NSD-NRC-98-5519, Provides Westinghouse Responses to FSER Open Items on AP600. Summary of Encl Responses Provided in Table 1.NRC Should Review Encls & Inform Westinghouse of Status to Be Designated in NRC Status Column of Oits

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Provides Westinghouse Responses to FSER Open Items on AP600. Summary of Encl Responses Provided in Table 1.NRC Should Review Encls & Inform Westinghouse of Status to Be Designated in NRC Status Column of Oits
ML20198R116
Person / Time
Site: 05200003
Issue date: 01/12/1998
From: Mcintyre B
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To: Quay T
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD-NRC-98-5519, NUDOCS 9801230223
Download: ML20198R116 (24)
v • d • e


Text

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I I Energy Systems g,3f5g ,,,, ,333 Ir o oration DCP/NRCl211 NSD-NRC 98 5519 Docket No.: $2 003

. January 12,1998 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATIENTION: T.R. QUAY SUluECT: AP600 RESPONSE TO FSER OPEN ITL.b3

Dear Mr. Quay:

Enclosure i of this letter provides the Westinghouse responses to FSER open items on the AP600. A summary of the enclosed responses is provided in Table 1. Included in the table is the FSER open item number, the associated OITS number, and the status to be designated in the Westinghouse status ,

column of 01TS.

'lhe NRC should review the el closures and infonn Westinghouse of the status to be designated in the "NRC Status" column of OITS.

Please contact me on (412) 374 4334 if you have any questions concerning this tiansmittal. l dr[ '

lirian A. McIntyre, Manager Advanced Plant Safety and Licensing

,imi Enclosure t t

ec: W. C. IlulTman, NRC (Enclosure) y ., , ,

T. J. Kenyon, NRC (Enclosure)

J. M. Sebrosky, NRC (Enclosure)

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D. C. Scaletti. NRC (Enclosure)

N. J. Liparuto, Westinghouse (w/o Enclosure) su.ppMS

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1)CP/NRCl211 NSI)-NRC,98 5519 2 January 12,1998 Table 1 1.16t of FSER Open items included in Letter DCP/NRCl211 FSER Open item OITS Number Westinghouse status in OITS 280.351 6501 Conurm W 470.411 5970 Con 0rm W 480.1105 6374 Action N 480.1106 637$ Action N 480.1108 6377 Action N 480.1110 6379 Action N 650.1917 6321 Action N 720,4601 6486 Action N 720.4631 6489 Action N i

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e I:nclosure to Westinghouse Letter DCI'/NRCl211 January 12,1998 II \ k GN

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, NRC FSER OPEN ITEM Ouest4on 280.35F (OITS 6501)

Re:

Spurious Actuation . The spurious actuation of ADS due to hot shorts of control circuits of motor operated valves from a fire in the main control room, rernote shutdown workstation, DC equipment rooms, and Clus IE penetration tunns has not been adequately addressed in the SSAR by the applicant. Spurious actuation of ADS results in a venting of the RCS in conflict with the fire protection of safe shutdown capability acceptance criteria specified above for the AfWN). This item remains open.

Response

The design features of the AIWX) and the expected operator actions to prevent spurious actuation of ADS due to hot shorts of control circuits of motor operated valves from a fire in the main control room, remote shutdown worketion, DC equiprnent room, and Class IE penetration noms have been described in SSAR Revision 18.

An overview is provided in 9A.2.7.1 in the last paragraph of the section titled " Spurious Actuation of Equipment."

Spunous actuation from a main control nom nre is addressed in SS AR 9A.3.1.2.5.1 in paragraphs 2 and 3 of the section titled " Safe Shutdown Evaluation."

Spurious actuation from a temote shutdown workstation fire is addressed in SSAR 9A.3.1.2.5.2 in the second paragraph of the section titled " Safe Shutdown Etatuation."

Spurious actuation from a Division A DC Equipment Room nre is addressed in SSAR 9A.3.1.2.1.1 in paragraphs 2,3. 4. and 5 of the section titled " Safe Shutdown Evaluation."

Spurious actuation from a Division B DC Equipment Room flic is addressed in SSAR 9A.3.1.2.2.1 in paragraphs 2. 3,4, and 5 of the section titled " Safe Shutdown Evaluation."

Spurious actuation from a Division C DC Equipment Room fire is addressed in SSAR 9A.3.1.2.3.1 in paragraphs 2. i,4, and 5 of the section titled " Safe Shutdown Evaluation."

Spurious actuation from a Division D DC Equipment Room Gre is addressed in SSAR 9A 3.1.2.4.1 in paragraphs 2,3,4, and 5 of the section titled " Safe Shutdown Evaluation."

Spunous actuation from a Division A Penetration Room fire is addressed in SSAR 9A.3.l.2.1.2 in the second paragraph of the section titled " Safe Shutdown Evaluation."

Spurious actuation from a Division B Instrumentation and Control / penetration Room nre is addressed in SS AR 9A.3.1.2.2.1 in paragraphs 2,3,4, and 5 of the section titled ' Safe Shutdown Evaluation."

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. NRC FSER OPEN ITEM Spurious actuation from a Division C Insuumentation and Control / Penetration Room fire is addressed in SS AR 9A.3.1.2.3.1 in paragraphs 2. 3. 4. and 5 of the section titled " Safe Shutdown Evaluation."

J Spurious actuation from a Division D instrumentation and Control / Penetration Room nre is addressed in SSAR 9A.3.1.2.4.1 in paragraphs 2. 3. 4. and 5 of the section titled

Most of the SSAR Sections refereaced above include a description of expected operator actions following detection of a Ilrc. These operator actions are assumed to be completed while the fire is relatively small and has not spread i

to the multiple fault kications requised for spurious ADS actuation.- 1he time.line analysis to demonstrate this assumption will be performed by the Combined License applicant as identified by the following SSAR Sections:

Secilon lH.6.. The stafung levels and qualifications of plant personnel including operations, maintenance, engineering. instrumentation and control technicians.etc. will be addressed by the Combined License applicant.

Sections 13.3 and 13.5.. Emergency planning and plant procedures are the responsibility of the Combined License applicant.

Section 9.5.l.H and Table 9.51.. The fire protec4an program implementation and the actions to be taken by individuals discovering a fire, the control room ope. itor, and the fire brigade will be addressed by the Combined License applicant.

To clarify what is expected from the Combined License applicant the following will be added to SS AR Section 9.5.l.N:

"The Combined License applicant will provide an analysis that demonstrates that operator actions w hich prevent the potential for spurious ADS actuation as a result of a fire can be accomplished within 30 minutes following  ;

detection of the fire."

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l NRC FSER OPEN ITEM SSAR Revision:

SSAR Section 9.5.1.8 will be revised as shown below:

9.5.1.8 Combined Lleense Information The Combined L.icense applicant will address qualification requirements for individuals responsible for development of the fire protection program, training of firenghting personnel, administrative procedures and controls governing the fire protection program during plant operation, and flic protection system maintenance.

The Combined License applicant will provide site specille nre protection analysis information for the yard area, the administration building, and for other outlying buildings consistent with Appendix 9A.

The Combined License applicant will address BTP CMEB 9.51 issues identified in Table 9.5.11 by the acronym "WA."

The Combined License applicant will address updating the list of NFPA exceptions after design certification, if necessary.

1he Combined license applicant willprovide an analysis that demostrates that opetutor actiom wKch prevent l the potertialfor spwious ADS actuation as a result ofafre can be accouplished within 30 ne'nutesfollowing l detection ofIhefire.

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280.35F 3

e NRC FSER OPEN ITEM Question 470.41F (OITS . 5970)

Meterology, October 7,1997 NRC letter in Section 2.3 of the AP600/SSAR. Westinghouse specifies bounding conditions and assumptions related to regional climatology, local meteorology, onsite meteo rological measurements program, shon-term (accident) diffusion estimates c/Q, and long term diffusion esti mates c/Q. The term c/Q is the relative atmospheric concen tration, c (Ci/m3), of radiological releases at the receptor point in terms of the rate of release, Q (Cihecond), from the point of release.

Westinghouse states that site specific meteorology information would be provided by the COL applicant. Further, Westinghouse notes that, if the site specific meteorology parameters exceed the bounding c/Q values in AP600/SSAR Section 2.3.4 and Table 21, the COL applicant will address how the radiologi-cal conse-quences resulting from the design basis accidents (DIIAs) continue to meet the dose reference values given in 10 CFR 50.34 and control room operator dose limits given in GDC 19 using site 4pecific c/Q values. The staff has not accepted the short term (accident) bounding c/Q values proposed in AP600/SSAR Section 2.3.4 and Table 21 pending completion of the radiological consequence assessments resulting from postulated design basis accidents (DIIAs) his is FSER Open issue 2.31.

Response

he site boundary atmospherk dispersion factor for the AP600 reference site will not be changed from the values currently reponed in the SSAR. The design basis containment leak rate is being reduced from 0.12 percent per day to 0.1 percent per day and this design change, together with the resolution of open item 470.44F, assure that the nutiological consequences of the design basis ac41ents do not exceed regulatory guidelines.

New X/Q values for the AP600 main control room have been calculated using the ARCON96 model.

He calculated X/Q values were then further adjusted upward to provide additional margin to accommodate a greater range of potential plant sites. Attached are revised SSAR subsection 15A.3.3 which provides a discussion of the approach used to determine the main control room X/Q values and revised SSAR Table 15A 5 which lists the new set of main control room X/Q values.

SSAR Revision: See attached markup.

470.41F 1

15. Acident Anal)sb gap fraction is conservatively increased by 20 percent to address this uncertJunty. The resulting gap fraction of 3.6 percent is used for the fuel handling accident, the rod ejecuon accident, and the iocked reactor coolant pump rotor accidero.

15 A.3.2 Nuclide Parameters The radiological consequence anajyses consider radioacuse decay of the subject nuclides pnor to their release, but no additional decay is assumed aher the acuvity is released to the environment. Table 15A-4 lists the decay ccastants for the nuclides of concem.

Table 15A 4 also lists the dose convenion facion for calculacon of the CEDE doses due to mhalation of iodines and other nuclides, dose consersion facton for calculauon of the acute dose due to immenion in a cloud of noble gases. and average gamma disintegration energies for determination of acute dose due to immersion in a cloud of nongaseous activity. These dose conversion facton are from EPA Federal Guidance Report No.11 (Reference 3) and the aserage gamma disintegrauon energies are from ICRP Pubhcauon 38 (Reference 4).

15 A.3.3 Atmospheric Dispersion Factors Sut,section 2.3.4 lists the short term atmospheric dispersion facton (x/Q) for the reference site.

Table 15A 5 reiterates these x/Q values,and-inc!* $c WQ vic; for ic rr.c.ir, cc=cl-h- musem e j>

l The locations of the potential release points for a Loss-of Coolant Accident and their I relationship to the main control room air intake and the penonnel access door are shown in i Figure 15A 1. Figure 15A 2 shows the locauons of the potential release points assochted I with other postulated accidents relat;ve to the possible paths for air entry into the mam 1 control room.

15A,4 References I. Murphy, K. G., Campe, K. M., " Nuclear Pow er Plant Coctrol Room Ventilation System Design for Meeting Ge' ! I Criterion 19," paper presented at the 13th AEC Air Cleaning Conference.

2. Soffer, L., et al., " Accident Source Terms for Light Water Nuclear Power Plants,"

NUREG 1465. February 1995.

3. EPA Federal Guidance Report No. I1. " Limiting Values of Radionuclide Intake and Air Concentration and Dose Convenion Factors for Inhalauon, Submersion, and Ingesuon,"

EPA 520/188-020, September 1988.

4. ICRP Publication 38, "Radionuclide 'Transformanons - Energy and Intensity of Emissions," 1983.

Revish.: 17 October 31.1997 15A-6 $ Westinghouse qw. r4ir - z

Insert in subsection 15A.3.3 De a'tmospheric dispersion factors (x/Q) to be applied to air entering the main control room following a design basis accident, were calculated at the HVAC intake and at the annex building '

entrance (which would be the air pathway to the main control room due to ingress / egress). De calculation of X/Q values was perfonned using the computer code ARCON96 (Reference 5), which provides a time-based building wake model for the determination of x/Q. De x/Q values were calculated using three separate meteorological data bases to encompass a range of potential plant sites. Bree existing power plant sites were used: a scacc.ast site, a river valley site, and a rolling.

hills site. Each site data base included five years of meteorological d ta.

To address the uncenainty regarding the actual orientation of the AP600 at a site, the calculation of ,

X/Q was performed with the plant orientation ranging through the 16 compass points (every 22.5 degrees).

De ARCON 96 code was run for all combinations of source and reception points for each of the 16 plant orientations at each of the three sites. Dese runs produced the 95* percentile X/Q values for the five post accident time periods of interest. Additional conservatism was added to the results for each site by selecting the maximum 95* percentile x/Q value from the 16 compass points and then Selecting the maximum of the three site values for application to the AP600 reference site. De AP600 reference site thus, for each time period, selectively combines the most conservative of the calculated values for the three sites and the 16 plant orientations.

Funhe, adjustment was made to the main control room x/Q values to add margin to the calculated values, his adjustment decreases the likelihood that a selected site would have higher x/Q values than identified in this application, while still meeting the dose guiJelines for the main control room, ne X/Q values are provided in Table 15A 5.

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4 Table 15A 5 (Sheet 2 of 2) l ATMOSPilERIC DISPERSION FACTORS (x/Q)  ;

FOR ACCIDENT DOSE ANALYSIS g Main control room o*ud p X/Q (s/m') al IIVAC Intake Ior t,a(n Melease Point Conta was $ team 1.ine Ceedenner Fuel H f uel Building l Ground If ane Stearn Wats Safety Valtes Air Remon) Area Reher Pat.,;;

} . 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 2=10' 39:10' 3 7ml0' l's10' =10' 3Ia10' 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 15=10' 30=10' 3 0= 10' 9 2m 10' 12x10' 19:10' L . 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 13:10' 27 10' 27 10 8 6 ts lV '0 10' 14:10'

. 4 days 84r10' 2 0=10' 21=10' 3610' 6 7ml0' 84 10' i . 30 days 48,10' t,3=10' l.5=10' les10' 37m10' 3 9:10' 8

X/Q (s/m ) a Control Room Door for Ea Release Point Containment Steam Line Ceedeanee Fuel Handling Fuel Building l Ground Release $teato Vents Safety Valv Alt Remoral Area Relwf Paret o.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 18:10' 3 3s10' 63 0' I!=10' I4 10' 1Ial0' f

' . 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 13s10' 24:10* =10' 9t:10' 88m10' 6 8:10d f i . 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> I1=10' 20=10' 10* 81:10' 7.1 x 10' 53:10' 4 days 7 3 = 10' 14:10* 2 32=10d 4 4= 10' 3 0:10' a 30 days 43m10' 88:10' l imi 2 *a l0* 23:10' 14:10' N

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l. The main control room x/Qs for ain control room habitability assess , nt are determined by:
  • Using the methodoto en NUREG/CR.50$$ (Reference 5) to determm the hourly average xlQs for vanous win speeds and stabil asses I
  • Using meteoro gical data from three different sites, the annua! aserage main c trol room xtQ is dete:Tnined.

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4 Determin' 'g the fifth penentile trsa conuol room 1/Qs from the same tinteorologi data.

00t#ning time averaged x/Qs for other than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> using loganthmic approximation (scussed in Regulatoy

( ide 1.1d5 (Reference 6). m j Revisiool 17 >

October 31,1997 15A.16 (Y Westinghouse go.4tf-5

lasert in Sheet 2 ef Table 15A.5 8 I X/Q (s/m ) at HVAC Intake for the Identined Release Points )

Con tainment Secondary Side Fuel Handling Fuel Building)

Release Points (3) Release Points") Arcad) Relief Panel' 0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.0E 3 2.0E.2 2.0E 3 3.0E-3 2 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.0E 3 1.8 E.2 1.3 r.3 2.0E.3 8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 5.0E 4 8.0E 3 8f r4 1.0E 3 1 - 4 days 5.0E 4 7.0E.3 8.0b t 1.0E 3 4 30 days 4.0E 4 6.0E 3 7.0L 1 9.0E-4 X/Q (s/m 3) at Co itrol Room Door for the Identined Release Pointg2)

Containment Secondary Side Fuel 11andling Fuel Building)

Release Points (3) Release oointsH) ArenU) Relief Panelt 0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.0E 3 2.5E.3 1.0E 3 1.0E.3 2 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6.0E 4 2.0E.3 6.0F 4 6.0E-4 8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.0E-4 1.0E.3 3.0F 4 3.0E 4 I 4 days 3.0E.4 9.0E 4 3.0E 4 3.0E-4 4 30 da> s 3.0E 4 8.0E 4 2.5E 4 2.5E 4 Moln

1. These dispersion factars are to be used 1) for the time period preceding the isolation of the main control room and actuation of the emergency habitability system,2) for the time after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the compressed air supply in the emergency habitability system would be exhausted and outside air would be drawn into the main control room, and 3) for the detenuination of control room doses when the non safety ventilation system is assumed to remain operable such that the emergency habitability system is not actuated.
2. These dispersion factors are to be used when the emergency habitabihty system is in operation and the only path for outside air to enter the main control room is that due to ingress / egress.
3. The listed values bound the dispersion factors for releases from the main equipment hateb, the staging area hatch, and the plant vent (elevated release). These dispersion factors would be used for evaluating the doses in the main control room for a loss of-coolant accident, for the centainment leakage of activity following a rod ejection accident, and for a fuel handling accident occurring inside the containment.
4. The listed salues bound the dispersion factors for, releases from the steam vents, the steam line safety & power operated relief valves, and the condenser air removal stack. These dispersion factors would be used for evaluating the doses in the main control room for a steam generator tube rupture, a main steam line break, a locked reactor coolant pump ro',or, and for a the secondary side release from a rod ejection accident. Additionally, these dispersion coefficients are conservatise for the small line break outside containment.

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5. The listed values bound the dispersion factors for releases from the fuel storage and handling area.

.These dispersica factors would be used for the fuel handling accident occurring outside  ;

. containment.-

6. The listed values bound the disperrion fectors for releases from the fuel storage area in the event that spent fuel boiling occurs and the fuel building rel'ai panel opens on high temperature. These dispersion factors are to be used for evaluating the imphet of releases associated with spent fuel

.. pool boiling.

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NRC REQUEST FOR ADDITIONAL $NFORMATION -

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. EAE Em Questioni 480.1105 (OITS 6374)

No information can be found on t..< drag induced forces acting on structure or piping submerged in the IRWST.-

Since the Al%00 has submerged piping within the IRWST, the drag forces acting on those structures would need to be addressed. Rose forces would include both the standard drag load associated with a structure within a constant >

- velocity fluid field and the acceleration drag load from forces associa .d with Guid acceleration. Please provide the methodology used to calculate the drag force on submerged piping and structural columns within the IRWST. Also, include the results of the analyses.  ;

Response

WCAP-13891. "AP600 Automatic Depressurization System Phase A Test Data Report " was intended to provide only the test results and therefore does not contain the results of analyses subsequently performed. De drag forces due to moving fluid and due to fluid acceleration have been considered for both the structural columns and submerged

piping as described below.

Acceleration drag due to hydrodynamic forces resulting from sparger operation have been quantified as part of the analysis for the passive residual heat removal heat exchanger. Rese forces are very small and have no impact on the structural analysis of components within the IRWST. The drag forces due to the movement / circulation of fluid in the IRWST have been determined based on the circulation velocity and fluid behavior observed in nutacrous Phase A and Bl testing blowdowns performed over the full range of expected AP600 conditions. For the submerged piping, the drag force' ve small compared to the dead weight, thermal, and seismic piping loads already considered, and therefore no specific analysis is required. Since the structural columns within the IRWST are closer to the sparger and are directly in the path of Duid movement induced by sparger operation, the drag forces can be significant. For these colamns tne drag forces have been considered by applying a conservative force of 7 psi on the entire projected area of these columns. The resultant applied shear force of 1512 lb. per foot of length conservatively accounts for both the drag forces and acceleration drag due to hydrodynamic forces resulting from sparger operation, and is combined with the other livi and dead loads in determining the column stresses.

SSAR Revision: None 9

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2 NRC REQUEST POR ADDIT;0NAL INFORMATION  !

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w-Question: 480.1106 (OITS 6375)

WCAP 13891. "AP600 Automatic Depressurization System Phase A Test Data Report," does not discuss the Guid structure interaction (FSI) effects at the tank walls. The issue is how the stiffness of the test facility tank versus the stiffness of the IRWST tank Door and walls affects the measured loads, if the IRWST wall is excited, the wall may transmit forces to other attached structures and piping. Please describe how the FSI is considered for the specific AP600 plant IRWST. how the test facility analysis considers FSI, and what were the calculated loads which will be imparted to the IRWST attached piping and structures.

Response

'The fluid structure interactions resulting from ADS sparger operation have been analyzed for the IRWST structure, and attached piping and structures. These analyses are outlined in Reference 480.1106-1, "AP600 IRWST Hydrodynamic Load Roadmap," in which specific documents are listed, and are discussed in SS AR section 3.9.3.4.2 These analyses consider te impact of the test tank stiffness and resonant frequency, conservatively add:.it the test tank resonant frequency response to the source loading used in the plant IRWST analysis. Also, these analyses show that the ADS indued FSI has no significant impact on the nuclear island equipment seismic evaluations, and that no gificant loads are transmitted to the passive residual heat removal heat exchanger or to auxiliary piping systems.

- Note that these analyses have already been thoroughly reviewed by the NRC Mechanical Branch as part of the AP600 design certification process.

References:

480.1106-1;- Westinghouse letter DCP/NRC 0727. "AP6001RWST Hydrodyumic Load Roadmap,"

February 3,1997.

SSAR Revision: None.

480.1106-1 T,i m Aingflouse

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NRC REQUEST FOR ADDITIONAL INFORMATION 7

Ouestion: 480.1108 (OITS 6377)

De IRWST attached piping and submerged struc*ures such as the support columns,

  • Integrated ilD Support" as shown on drawing 1030 P2 001 revision 7 could experience an acceleration and or drag force from the IRWST tank during a quencl:er discharge. Please provide a discussion on the magnitude for those accelerations and forces and explain how those accelerations ard forces were developed.

Response- ,

Please refer to the response for RAI 480.1105.

' SSAR Revision: None

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NRC REQUEST FOR ADDITIONAL INFORMATION

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. Question: 480.1110 (OITC6379)-

Ilased on the measured test forces, how were the AP600 design loads determined and with what level of margin.

Response

Hydrodynamic analyses were performed using measured test forces as decribed in SSAR subsection 3.8.3.4.2 and as discussed in the response to RAI 480.1106.

SSAR Revision: None.

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..- c w l NRC FSER OPEN ITEM Question 650.19F (OITS . 6321) l Issue 83: Control Room Habitability 1

As discussed in NUREG-0933, Issue 83 addressed the significant discrepancies found dunng a survey l of existing plant control rooms before 1983. Rese discrepancies included the inconsistencies between i the design, construction, and operation of the control room habitability systems and the descriptions in ]

the licensing-basis documentation. In addition, the staff determined that total system testing was ]

inadequate and that the control systems were not always tested in accordance with the plant TS. De following issues are related to issue 83:

(1) Issue B 36 on criteria for air filtration and adsorption units for atmospheric cleanup systems.

-(2) Issue B-66 on control room infiltration measurements, and (3) Issue III.D.3.4 also on control room habitability. Dese three issues are discussed elsewhere in Sections 20.2 and 20.4 of this report.

In Section 1.9.4 of the SSAR, Westinghouse states that habitability of the main control room during normal operation is provided by the non-safety related nuclear island nonradioactive ventilation system (VBS). In the event of a design basis accident involving a radiation release or a loss of all ac power event, the non-safety related nuclear island nonradioactive ventilation system is automatically terminated, the main control room pressure boundary is isolated, and the safety related main control room emergency hebitability system (VES) is actuated, ne safety related main control room emergency habitability system supplies breathable quality air for the main control room operators while the main control room is isolated. In the eveN of external smoke or radiation release, the non safety related nuclear island nonradioactive ventilauon system provides for a supplemental filtration mode of operation, as discussed in Section 9.4 of the SSAR. In the event of a Hi Hi radiation level, the soety related main control room emergency habitability system is actuated. in the unlikely event of a toxic chemical release, the safety-related main control room emergency habitability system has the capability to be manually actuated by the operators.

Further, a 6-hour supply of self contained portable breathing equipment is stored inside the main

- control room pressure boundary, in the DSER, the staff requested that Westinghouse address control room habitability and issue 83 for the AP600 design. His was designated as Open item 20.311. Open item 20.311 (650,19F) is still unresolved because (1) the staff has determined that the AP600 design does not meet the dose limits of G1X' 19 and (2) Westinghouse needs to state in Section 6.4.7 of the SSAR that the COL applicant referencing the AP600 design is responsible to verify that the as-built design, procedures, and training are consistent with the licensing basis documentation and the intent of Issue 83 (see COL Action item 6.4-2).

Derefore, Issue E3 is not resolved for the AP600 oesign.

650.19F-1

g 4H NRC FSE.i OPEN ITEM I

Response

FSER Open item 470.46 provided revised parameters for use in calculation of the control room doses.

With these revised inputs, the control room doses are projected to be within the GDC 19 limit of 5 rem TEDE.

FSER Open Item 410.367 provided a change to subsection 6.4.7 to address the COL responsibility for procedures and training consistent with the intent of Generic issue 83.

SSAR Revision: None.

650.19F 2 3 Westinghouse

4 NRC FSER OPEN ITEM

+ g I M

Ouestion: 720.460F (OITS #6486)

In the level 2 analysis provided by letter dated September 17, 1997, WEC indicates that the fault trees used to quantify the system nodes of the shutdown CETs were based on the fault tree linking results of the surge line flooding case. WEC needs to identify the fault trees used to quantify each system top event (e.g., by providing tables similar to Table 36-1), and indicate how each fault tree was modified from the at power model to reflect changes in equipment / system availability permitted during modes 5 and 6. The response should justify that for each system.related top event in each CET, the issumptions regarding equipment / system availability are consistent with the TSs or availability targets for the respective operating mode. Assumptions regarding DAS, isolation valves, cavity flooding valves, and igniters should be specifically addressed.

- Response:

It is essential to note that the fault tree top events used in the containment event tree (CET) quantification documented in the September 17,1997 letter were fault trees specifically modeled for the shutdown conditions; not at power fault trees that were modified for the new CET calculations, nese fault tree top events were previously used in the Level 2 assessment of shutdown events documented in the revision 6 PRA, June 1995. In the revision 8 PRA Table 54 7 and the success criteria tables contain the information regarding the special assumptions made for the shutdown modes. Also. Table 54-8 presents the system unavailability status information with referer.cc to the proper Technical Specification for that system.

The following containment event tree (CET) nodes use fault tree linking results for assigning the CET failure probabilities:

Node DP: Depressurization Node IS: Containment isolation Node IR: Reactor Cavity Flooding from IRWST Node IG: Hydrogen Igniters Tables 720.460-1 through 720.460-4 identify which of the accident classes use fault tree linking for the abose nodes and the fault tree top event name. De LP 3BL accident class is included in the following tables to cover the Les el 2 quantification reported in the rerponse to FSER Open item 720.43 tF (mark up of attachment 54B), in which new core damage sequences involving recirculation failures reside.

T westinghouse

t '

t- ,

NRC FSER OPEN REM t'

Table 720.460F.1 Depressurization (DP) Node Failure Probabilities Accident Class Fauure Probability LP1A (No CET; Large Release Assumed)(Note 1)

LP 3D 0 1.P 3BR 0 (Note 1)

LP 3BE o LP 3BL 0 LPCDP (No CET; Large Release Assumed)

Note 1: He LP I A and LP 3BR accident class frequencies are defined based on failure or success of ADS after high pressure core damage occurs. De fault tree that defines success or failure of ADS post core damage is ADTLTS. Accident class LP l A contains high pressure core damage events in which ADS also fails post core damage; accident class LP 3BR contains high pressure core damage events in which ADS is successful post core damage.

Table 720.460F.:

Containment isolation (IS) Node Fauure Probabilities Accident Class Failure Probability LP I A (No CET; Large Release Assumed)

LP 3D CIST LP 3BR CIST LP 3BE CIST LP-3BL CIST LPCBP (Ne CET; Large Release Assumed) 4 720.460F-2

NRC FSER OPEN ITEM s Table 720.460F.3 Reactor Casity Flooding from IRWST (IR) Node Failure Probabuities Accident Class Fauure Probability LP-1 A (No CET: Large Release Assumed)

LP-3D IWFS LP 3BR IWFS

  • IW2AB LP 3BE IWFS . , _ .

LP 3BL 0 LPCDP (No CET; Large Release Assumed)

Table 720.460F 4 Ilydrogen Igniter (IG) Node Fallure Probabilities Accident Class Failure Probabuity LP-I A (No CET; Large Release Assumed)

LP-3D VLHS LP 3BR VLHS LP 3BE VLHS LP 3BL VLHS LPCBP (No CET; Large Release Asst med)

As was meationed above, section 54 of the PRA contains the descriptions of these fault trees and the assumptions made regarding the shutdown modes of operation.

PRA Revision: None.

T westinghouse l

1

  • l l 42 a NRC FSER OPEN ITEM Ouestion: 720.463F (OITS #6489) -

Footnote 3 to Table 54 7 indicates that a separate fault tree was not constructed for containment isolation during loss of offsite power (LOOP) events since LOOP does not dominate the shutdown CDF. The response to RAI 720.306 -

indicates that approximately 20 percent of the shutdown CDF involves LOOP. The revised surge line flooding.

analyses submitted by letter dated October 8,1997 indicates a somewhat lower contribution from LOOP in the base e

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 isolation fault trees or a bounding treatment of containment isolation is not needed for LOOP events.

Response

Baseline Shutdown PRA De containment 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). The 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. The current containment isolation model credits all three of these power sources; a special loss of offsite power case would only credit the diesels and the Class IE batteries as power sources.

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

However, using the containment isolation simplification noted in Table $4 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 should result in containment isolation failure.

The baseline shutdown PRA quantification in the response to FSER Open Item 720.43tF (mark-up of attachment 548) indicates that the loss of offsite power events contribute approximately six percent of the total bascime

- 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. De remainder of the less of offsite power 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 injec4on and recirculation functions of the passive core cooling system (PXS). Class IE de power system failures account for only about four percent of the loss of offsite power esent core damage frequency.

Based on the abc= 'rm S increase in larp release frequency due to the impact of this simplification is approximately:

Increase in LRF = 6%

  • 75%
  • 4% = - Ol% of current CDF which is judged to be negligible.

+

720.463F 1

. - ~_ . . . . . . . -. . -. - .. --- .. . - . -. . . ....... - . ..-

-a

'O NRC F/ It OPEN ITEM -

IM i Focused Shutdown PRA ,

The focused PRA assume '-ilure of all non-class IE systems, structures and components following the initiating

, event. Herefore, the focuwe PRA does not credit offsite power in any Level I or Level 2 fault tree or event tree

' model; the note is only a, >1icable to the baseline shutdown PRA.

PRA Revision: None.

F J

a 4

s 720.463F 2 g

p w i ,'rp y g .w , y g- 9 - - - - w y w-9 y ,- p

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