NSD-NRC-98-5569, Forwards W Responses to FSER Open Items on AP600.Summary of Encl Responses,Fser Open Item Number,Associated Oits Number & Status to Be Designated in W Status Column of Oits, Included in Table 1

Forwards W Responses to FSER Open Items on AP600.Summary of Encl Responses,Fser Open Item Number,Associated Oits Number & Status to Be Designated in W Status Column of Oits, Included in Table 1

ML20202F161
Person / Time
Site:	05200003
Issue date:	02/12/1998
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD-NRC-98-5569, NUDOCS 9802190140
Download: ML20202F161 (11)

Text

.

, a Westinthouse Ene3y Systems Bm 355 Pinstugh Pennsylvana 15230 0355 Electric Corporation DCP/NRCl255 NSD-NRC-98-5569 Docket No.: 52-003

. February 12,1998 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATFENTION: T. R. QUAY

SUBJECT:

AP600 RESPONSE TO FSFA JPEN ITEhtS

Dear Mr. O iay:

Enclosed with this letter are the Westinghouse responses to FSER open items on the AP600. A summary ot .he enclosed responses is provided in Tablo 1. Included in the table is the FSER open item number, the associated OITS number, and the status to be de ignated in the Westinghouse status column of OITS The NRC should review the enclosures and inform Westinghouse of the status to be designated in the "NRC Status" column of OITS.

Please contact me oi. (412) 374-4334 if you have any questions concerning this transmittal.

-t M W Brian A. McIntyre, Manager

/.dvanced Plant Safety and Licensing jml Enclosure cc: W. C. Iluffman, NRC (Enclosure)

J. E. Lyons. NRC (Enclosure)

T. J. Kenyon, NRC (Enclosure)  !

J. M. Sebrosky, NRC (Enclosure) '

D. C. Scaletti, NRC (Enclosure)

N. J.1.iparuto, Westinghouse (w/o Enclosure) }kD)-] t g 9902190140 900212" PDR ADOCK 05200003 E PDR

,; i 1

[ DCP/NRCl255

- NSD-NRC-98-5569 February 12,1998 i

- Table 1 List of FSER Open Items included in Letter DCP/NRC1255 FSER Open item OITS Number We-tingbouse status in OITS 440,786F (R1) .- 6368 Confirm W 720.463 (RI) 6489 Confirm W 230.147F (Rl) 6506 Con irm W-c.

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NRC FSER OPEN ITEM 4 Question 440.786F AITS 6368) REVISION 1 TO RESPONSE TS 5.6.5 Core Operating Limits Report (COLR)

Specification 5.6.5, in addition to requiring that the core operating limits be established and documented in the COLR for the parameters listed, lists the approved analytical methods used to determine the core operating limits. WCAP 12472-P A, " BEACON - Core Monitoring and Operations Support Systera," is listed as the approved method used for TS 3.2.5 for monitoring compliance with the core operating limits specified in the COLR. The BEACON system has been accepted by NRC for performing continuous on-line core monitoring and operations support functions for Westinghouse PWRs subject to conditions desenbed in the staff safety evaluation report and accompany ahnical evaluation report for the acceptance of WCAP !2472.P A. Though the BEACON system is potentially suitable for other reactors, it has been examined in the topical report and by the staff only for current standard Westinghouse systems. For applications of BEACON to plants or core designs that differ sufficiently to have a significant impact on the WCAP 12472 P-A data base, the generic uncertainty components may require :eevaluation in order to ensure that the assumptions made in the BEACON uncertainty analysis remain valid, and assure that the power peaking uncertainties for enthalpy rise and heat flux provide 95 percent probability upper tolerance limits at the 95% confidence level. Acceptance for these applications would require further review and approval. SSAR Subsection 4.3.1.6.2 indicates that AP600 uses fixed in-core detectors, in-core thermocouples, and loop temperature measurements, his is different from the instrumentation data base described in WCAP-12472 P-A, which use, among other ir..;trumentation, movable incore detectors, ne applicant does not provide documentation for the use of fixed in-core detectors in the OPDMS for AP600 for NRC review and appro.al. 'IS .6.5 contains a " REVIEWER'S NOTE" stating that additional power distribution control and surveillance methodologies (for MSHIM and OPDMS monitoring) are currently under development and will be added upon NRC approval. This is an open item.

Response: REVISJON 1 The purpose of the " Reviewer's Note" in TS 5.6.5 is to recognize that adOonal NRC review and approval of BEACON application methodology with fixed incore detectom will be completed prior to generation of a plant specific TS 5.6.5. Westinghouse intends to seek NRC review and approval of an addendum to the BEACON topical report (WCAP 12472-P-A) covering the use of AP600 fixed incore detectors in conjunction with use of similar detectors at operating plants. Discussion of urmertainty components relative to use of BEACON methodology with fixed incore detectors is proviued in SSAR i subsection 4.3.2.2.7. The attached markup of TS 5.6.5 provides a revision to the " Reviewer's Note" which clarifies the commitment and provides the basis for closing this item for AP600 Design

' Certification. A statement has also been added to SSAR subsection 4.3.4 which commits the i Cembined License applicant to referencing an approved addendum to WCAP-12472-P-A covering

! fixed incore detectors.

I SSAR Revision: See attached markups.

440.786F(R1) 1

' nepui m y ne w o cw ncs 5.6 e

. 5.6 Reporting Requirements t

5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) l 2c. NUREG 0800. Standard Rev1 w Plan. U.S. Nuclear Regulatory Comnission, Section 4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.31.

Westinghouse Constant Axial Offset Control (CAOC),

Rev. 2. July 1981.

l (Methodology for Specification 3.2.3 Axial Flux Difference (Constant Axial Offset Control).)

3. WCAP 10216 P A, Revision 1

• Relaxation of Constant Axial Offset control FQ Surveillance Technical l Specification " February 1994 (Westinghouse l Proprietary).

(Methodology for Specifications 3.2.2 Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.1 -

Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FQ Methodology).)

4. WCAP 12945 P, Volumes 1 5, ' Westinghouse Code Qualification Docunent for Best Estimate Loss of Coolant Accident Analysis,' June 1992 ,'une 1993.

(Methodology for Specification 3.2.1 Heat Flux Hot Channel Factor.)

6. RAP 14807, *NOTRUMP Final Validation for AP600,"

R.L. Fittante ct al., January 1997 (Westinghouse Proprietary).

(Methodology for Specification 3.2.1. Heat Flux Hot Channel Factor.)

g 6. WCAP 12472 P A, ' BEACON Core Monitoring and Operations Su port Systen,* August 1994 and Addendum

1. May 1996 (Astinghouse Proprietary).

(HethMology for Specification 3.2.5 OPOMS -

Monitored Power Distribution Parameters.)

.................... REVIEWER'S NOTE -

Additional power distribution control and surveillance methodologies (for MSHIM and OPONS sonitoring) are currently under development and will be added upon NRC approv 1.

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the fuel is used for X Y calculations. In the early stages of desige a two-dimensional few group diffusion code TORTISE, which is an updated version of TUR'LE (see WCAP 721.1.

A Reference 36) was userl.

Spatial few group calculations are carried out to determine the entical boron concentrations and power distnbutions. The moderator coefficient is evalua:ed by varying the inlet temperature in the same kind of calculations as those used for power distnbution and reac-uvity predictions.

Validation of the reactivity walculations is associated with validation of the group constants themselves, as discussed in subsection 4.3.3.2. Valida: ion of the Doppler calculations is associated with the fuel temperature validation discussed in cubsection 4.3.3.1, Validation of the moderator coefficient calculations is obtained by comparison with plant measurements at hot zero power conditions, similar to that shown in Table 4.3 9.

Axial calculations are used to determine differential control rod worth curves (reactivity versus rod insertion) and to demonstrate load follow capability. Group constants are obtained from the three-dimensional nodal model by flux volume weighing on an axial slicewise basis.

Radial bucklings are determined by varying parameters in the buckling model while forcing the one-dimensional model to reproduce the axial characteristics (axial offset, midplane power) of the three-dimensional model.

Validation of the spatial codes for calculating power di:tributions involves the use of in. core and ex-core detectors and is discussed in subsection 4.3.2.2.7, As discussed in subsection 4.3.3.2, calculation measurement compansons have been made to operating reactor data measured during startup tests and during normal power operation.

'lhese compansons include a variety of core geometries and fuel loadF patterns, and incorporate a reasonable extreme range of fuel enrichment, bumable absort>er leading, and cycle bumup. Qualification data identified in Reference 40 indicate small mean and standard deviations rel.tive to measurement which are equal to or less than those found in previous reviews of similar or parallel approved methodologies. For the reload designs the spatial codes described above, other NRC approved codes, or bota are used.

4.3.4 Combined License Information This section contains no requirement for additional information to be provided in support of the combined license. Combined License applicants referencing the AP600 certified design will address changes to the reference design of the fuel, burnable absorber rods, rod cluster control assemblies, or initial core design from that presented in the SSAR. Minor changes to design feart ms and design parameters within the bounds of safety analysis and changes covered by NRC-approved topical reports may be made with NRC staff review.

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Question: 720.463F (OITS #6489) REVISION I TO RESPONSE I

Footnote 3 to Table 54 7 indicates that a separate fault tree was not constructed for containment isolation during loss of ofhite power (LOOP) events since LOOP does not dominate the shutdown CDP. The response to RAI 720.306 indicates that approximately 20 percent of the shutdown CDF involves LOOP. He revised surge line flooding analyses submitted by letter dated October 8,1997 indicates a somewhat lower contribution from LOOP in the base case, but LOOP dominates the focused PRA shutdown CDP. Please discuss the impact that LOOP would have on containment isolation and containment closure capability, and justify why separate containment isolanon fault trees or a bounding treatment of containment isolation is not needed for LOOP events.

Response

Baseline Shutdown PRA The contaironent isolation model is dependent upon Class IE de and UPS power system (IDS) in order to power motor-operated valves (MOVs) and the protection and safety monitoring system (PMS). He IDS can utilize power from three sources; offsite power through the battery chargers and regulating transformers, the diesel generators again through the battery chargers and regulating transformers, and the Class IE batteries. De current containment isolation model credits all three of these power tources; a ~ ial loss of of ' site power case would orily credit the -

diesels and the Class IE batteries as power sources.

Tne containment isolation model is dominated by common cause and random failures of the containment isolation I valves. Support sys. ems account for a very small fraction of the containment isolation functional unreliability.

However, using the containment isolation simplification noted in Table 54 7. cutsets involving power failures may be missed in the Level 2 assessment. Specifically, loss of offsite power event cutsets in which offsite power restoration fails and the Class IE de power system fails dould result in containment isolation failure, ne baseline shutdown PRA quantification in the response to FSER Open item 720.431F (mark-up of attachment 54B) indicates that the loss of offsite power events contribute approximately six percent of the total baseline shutdown core damage frequency (CDF) Of that six percent, about 25 percent have successfully recovered offsite power (basic event SUC RIS), in which case the current containment isolation model is correct. The remainder of the loss of offsite poiver cutsets are dominated by cutsets involving failure to restore offsite power and failure of components belonging to front line systems, such as the normal residual heat removal system (RNS), the automatic --

depressurization system (ADS), and the gravity injection and recirculation functions of the passive ccre cooling system (PXS). Class IE de power system failures account for only about four percent cf the loss of offsite power event core damage frequency.

Based on the above facts, the increase in large release frequency due to the impact of this simplification is approximately:

Increase in LRF = 6%

• 75%

• 4% = - 0.2% of current CDF which is judged to be negligible.

g 720.463F(R1)-1

o NRC FSER OPEN ITEM

= me iPI l!!

I Focused Shutdown PP.A The focused PRA assumes failure of all non-class IE systems, structures and components following the initiating e /ent. "iherefore, the focused PRA does not credit offsite power in any Level 1 or Level 2 fault tree or event tree model; the note is only applicable to the baseline shutdown PRA.

SSAR Revison: In response to NRC staff request in a 2/3/98 telecon, the SSAR will be revised as shows. on the attached markup regarding equipmer.t hatch power supplies, s PRA Revision: None.

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, 3. Design of Structures, Components, Equipme:t, ard Spicms The containment ves:el includes the shell, hoop stiffeners and crane girder, equipn ent hatches, personnel airlocks, penetration assemblies, and miscellaneous appurtenances and attachments, he polar crane is designed for handling the reactor vessel head during normal refueling. The crane girder and wheel assemblies are designed to support a special trolley to be installed in the eve.nt of steam generator replacement.

He contamment vessel supports most of the containment air baffle as described in subsection 3.8.4. The air baffle is arranged to permit inspection of the exterior surface of the -

containment vessel. Steel plates are welded to the dome as part of the water distribution system, described in subsection 6.2.2. The polar crane system is described in subsection 9.1.5.

3.8.2.1.2 Containment Vessel Support ne bottom head is embedded in concrete, with concrete up to elevaaon 100' on the outside and approximately clevation 108' on the inside, ne containment vessel is sssumed as an independent, free. stand! _ ,,tructure above elevation 100' He thickness of the lower head is the same as that of the upper head. Here is no reduction in shell thickness even though credit could be taken for the concrete encasement of the lower head.

Vertical and lateral loads on the containment vessel and internal structures are transferred to the ba emat below the vessel by friction and bearing. Seals are provided at the top of the concrete on the inside and outside of the vessel to prevent moisture between the vessel and concrete. A typical cross section design of the seal is pre ented in Figure 3 8.2-8. sheets 1 3.8.2.1.3 and 2.

Equipment Hatches yf/ p f gYlg

/ V Two equipment hatches are provided/One is at the operating floor (eievation 135'-3"). He hatch has an inside diameter of 22 fiet, to pennit repla. ament of a steam generator. He other is at floor elevation 107'-2" to per'mit grade-level access into the containment, with an inside diameter of 16 feet, ne hatches; shown in Figure 3.8.2 2, consist of a cylindrical, sleeve with a pressure seated dished head bolted on the inside of the vessel. He containment intemal

' pressure acts on the convex face of the dished head and the head is in compression. De flanged joint has double O ring or gum-drop seals with a pressurized for leak testing the seals. Et 4. .g 70wC/Mda der %n annular space that m Mf+^WMHnZf % ,uXqi L&.a ~&4 >~-e / u cL 3.8.2.1.4 Pyenngeg y  % ( gg .g g n ie M are a aMnt to ach t liatches.

Figure 3.8.2-3 shows the typical arrangement. Each personnel airlock has about a 10-foot extemal diameter to accommodate a door opening of w!dth 3 feet 6 inches and height 6 feet 8 inches. He airlocks are long enough to provide a clear distance of 8 feet, which is not impaired by the swing of the doors within the lock. The airlocks extend radially out from the containment vessel through the shield building. They are supported by the containment vessel.

Revision: 11 February 28,1997 3.8 2 T W96tingh0038 o

i NRC FSER OPEN ITEM 55 W

enm l Open Item 230.147F (OITS #6506) Response Revision 1 j Westinghouse states in its submittal that the isolation valves between the RCS and the CVS are tested a per the Inservice Testing Program. One of the safety functions of these valves is to establish a leaktight barrier essuming ti.at integdty of the downstream piping is lost. Westinghouse states that ex:rcise testing of these valves minimizes the probability of unexpected gross failure and gross leakage of the valves. He staff considers the currently proposed IST program for these valves is insufncient considering that the CVS SSCs are classified as non ASME Section III. He valves should be leak 3 tested at full RCS differential pressure across the valves (simulating the isolation of a downstmri

} piping break). The leak testing and acceptance cdteria should be as currently required by ASTM L Section XI(OM 10) which allows 0.5D gpm or 5 gpm (whichever is less), where D is the nominal valve dbmeter expressed in inches During c.iscussions between Westinghouse and NRC staff after submittal of Revision 0 of this response the staff requirements were clarified. These requirements included leak testing of all three kI isolation vahes on each line between the RCS and CVS. Clarification of the leak test requirements in

( Note 32 of Table 3.916 was also requested. Verification requirements for valves that have the characteristic of decreasing leakage with pressure must also be defined.

F l Response: Revision 1 The valves that isolate the CVS from the RCS are not PlVs because the CVS design pressure is not lower than the RCS design pressure. * 's actually higher (3100 psig vs 2485 psig). In addition, these isolation valves do not meet any of the criteria used to select PIVs in Technical Specificanon LCO 3.4.16 (PIV Integrity). As a result, these valves should not be included in Technical Specification LCO 3.4.16 and SSAR Table 3.9-18, which lists the PlVs that are subject to Technical Specification LCO 3.4.16.

- In order to support classification of the CVS purification loop located inside containment as nonsafety-related, leak testing will be provided on the fin::wo valves isolating the CVS from the RCS.

Regulatory Guide 1.26 states that portions of systems that are connected to the reactor coolant i boundary and are capable of beinCi solated from that boundary by two vaives, each of which is capable of automatic closure made be classified as Group C. A footnote to this requirement allows influent lines to be classified as Group D if they are capable of being isolated by an additional valve which has high leaktight integrity, ni :d::d: k & :c : requi= men: :c kd :::: cn =he in-the innuen:!!ac.

He AP600 CVS design uses a similar approach to this footnote for isolation of the class D portion of the CVS purification loop located inside containment. Three valves capable of automatic closure are located in safety-related piping between the RCS at 6 the class D portion of the CVS. Westinghot.se is aisc providing seismic analysis of the class D portion of the CVS as discussed in the response to T westinghouse 2mm

e s

NRC FSER OPEN ITEM N

l Open Item 230.146F which is not required by Regulatory Guide 1.26. "'n:!r.;;h:=: = quin: S:: :we i

vdce be kd :=::d =:::.d of cc: = ;;;ind by R:;;;h:crj 0;!d: 1.25. He requirement for leak testing is contained in the incervice Testing SSAR information (Table 3.916). He leak rate acceptance criteria is specified in a note to this table consistently with ASTM Section XI.

Verification that valves have the characteristic of decreasing leakqe with pressure will be provided with two tests at different test pressures. The note to describe leak testing of the valves will be

, revised to include the requirements for this verification .

SSAR Change: Dese changes use SSAR Revision 20 as the base. he SSAR changes in Revision 0 of this response were included SSAR Revision 20.

Revise Table 3.9-16 as follows: <

l For valve CVS-PL-0RO mder IST Notes add "32".

Revise Nota 32 as follows:

nese valves are subject to leak testing to support the nonsafety related classification of the CVS purification tebsystem inside containment. Rese valves are not included in the PIV integrity Techni-al Specification 3.4.16. He leakage through valves CVS-V001, andCVS ' 22, and CVS V090 will be tested separate!y with a leakage limit of 1.5 gpm for each valve heleakage through valves CVS-V080;-V081, V082, V084, and V085 will be testM at the sam. me as a

roup with a leakage limit of 1 ;s.m for the group. The leak tests will oc performeo at reduced RCS pressures. The observed leakage at lower pressures can be assumed to be the leakage at the maximum pressure as long a the valve leak. age is verified to diminish with increasing pressure differential. Verification that the valves have the characteristic of decreasing leakage with pressure may be provided with two tests at different test pressures. The test requirements including the minimum test pressure and the difference between the test pressures will be defined by the Cocnbined License applicant in the inservice test program.

Revise Figure 9.3/1 as follows:

Add a test connection for valve V080 as shown in the attached figure.

ITAAC Change:

None 230.147F(RI)-2 s

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