NSD-NRC-98-5634, Forwards Formal Transmittal of Correspondence Previously Sent Informally Over Period 980106-0227.Index of Attached Matl Also Encl
| ML20217E507 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 03/25/1998 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD-NRC-98-5634, NUDOCS 9803310061 | |
| Download: ML20217E507 (206) | |
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3 Westinghouse Energy Systems pil3yp Pennsylvania 15230-0355 Electric Corporation DCP/NRCl313 NSD-NRC-98-5634 1 Docket No.: 52-003 f
- March 25,1998 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T. R. QUAY
SUBJECT:
INFORMAL CORRESPONDENCE
Dear Mr. Quay:
Please Nnd enclosed a formal transmittal of correspondence we have previously sent to you informally.
This informal correspondence was sent over the period January 6,1998 though February 27,1998.
Attachment I provides the index of the attached material as you have requested.
/
A Brian A. McIntyre, Manager Advanced Plant Safety and Licensing jml Attachment linclosure te: N. J. Liparulo, Westinghouse (w/o Attachment, Enclosure)
J. W., Roe, NRC/DRPM (w/o Enclosure)
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-A Attachment I to Westinghouse Letter DCP/NRC1313 DATE ADDRESSEE DESCRIPTION 1/9/98 Iluffman Dran response to open item 440.783.
1/9/98 Sebrosky Spectra for open item 220.114F.
1/6/98 Quay Questioning if 410.432 and 410.433 are truly duplicates.
1/9/98- Quay Open item status charts.
1/13/98 Scaletti Open item responses for 410.418F,410.422F,410.423F, 410.435F.
1/13/98 Iluffman/Landry NOTRUMP documentation.
1/14/98 iluffman/Lois LBLOCA methodology limitation limits.
1/13/98 Scaletti Chapter 9 open items we have not answered yet.
1/14/98 Quay Open item status charts.
1/14/98 Scaletti Open item 410.427F response.
1/14/98 liuffman Draft responses to open items on shutdown ERGS. Please provide comments.
1/14/98 Iluffman . Proposed response to informal question from Ed Throm.
1/19/98 Quay Two items.
1/19/98 Quay Open item status charts.
1/19/98 Sebrosky Table correction to close open item 220.122.
1/22/98 Scaletti Corrections to chapter 8 from SSAR scrub.
1/22/98 Quay Open item status charts.
1/23/98 Quay Change to Westinghouse Electric Company.
1/26/98 Kenyon SSAR revision in response to open item 471.23.
1/26/98 Sebrosky Proposed SSAR changes from tclephone call.
1/26/98 Scaletti Response tr. open items (DCP/NRCl227). l 1/29/98 Quay Open item status charts.
2/2/98 Scaletti SSAR markup to close open item 280.33F.
2/5/98 Scaletti SSAR markup to close 460.28F. Will be in Revision 21 unless we hear from you. I 2/5/98 Quay Open item status charts.
2/6/98 11uffman Proposed changes to CVS material.
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DATE ADDRESSEE DESCRIPTION 2/5/98 Sebrosky Request for status of open items for chapter 19.
2/4/98 Sebrosky/Palla Proposed changes to insights table to resolve open item 720.439F.
2/13/98 Quay Open item status charts.
2/13/98 Quay /11uffman NRC staff ACRS testimony from 2/4/98.
2/13/98 Sebrosky PRA markup to resolve open item 720.434F.
2/17/98 Iluffman Draft of l A to address RG 1.97.
2/19/98 11uffman Suggested changes to address inconsistency in purge valve closure time.
2/23/98 Iluffman SSAR change to 6.2.2.2.4 to address shield building annulus drains.
2/19/98 Quay Open item status charts.
2/17/98 Sebrosky Markup to address open item 720.434F.
2/19/98 iluffman Paper on safety valve performance.
2/25/98 Sebrosky Notification that the stairs in the CVS room will be steel and liner in room under RPV will be painted carbon steel.
2/25/98 Sebrosky Revision to table 3.8.4-1.
2/25/98 Sebrosky Shutdown PRA question responses.
2/25/98 Sebrosky Markup to PRA insight table to address Marie's item.
2/26/98 Sebrosky SASR changes to resolve remaining structural issues.
2/26/98 Sebrosky Suggested changes for 2 of the open items (220.125,220.130).
2/27/98 Sebrosky Return of load combinations 6 to table 3.8.4-1 and 3.8.4-2.
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FAX COVER SHEET Recipient information Sender Information Name: D. Lindgren Date:
'2. /z 7 /9 8 l To: Joe Sebrosky Location: WEC E 330 Phone: Phone: (412) 374-4858
'(301) 415-1132 i Facimle: Facimle: Bet (412) 3744887 (310) 415-2002 Company: U.S.NRC WIN 284-4887 ,
location: Rockville, MD Cover + Pages 1+3 The following pages are being sent from the Westinghouse Energy Center, Monroeville, PA.
If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374-6529.
comments: -
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- 3. Design of Structures, Components, Equipment, and Systems Table 3.8.4-1 ,
l LOAD COMBINATIONS AND LOAD FACTORS FOR SEISMIC CATEGORY I STEEL STRUCTURES Load Combination and Factors i Combination No. I 2 3 4 5 6 7 8 9 Load description l Dead D 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 ,
l Lipid F 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 l Live L 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 l Earth pressure H 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Normal reaction Ro 1.0 1.0 1.0 1.0 1.0 1.0 Normal thermal To 1.0 1.0 1.0 1.0 Wind W l.0 1.0 Safe shutdown Es 1.0 1.0 earthquake Tornado Wt 1.0 l Accident pressure Pa 1.0 1.0 1.0 l Accident thermal Ta 1.0 1.0 1.0 l Accident thermal R, 1.0 1.0 1.0 reactions l Accident pipe reactions Yr 1.0 1.0 l l Jet impingement Yj l.0 1.0 l Pipe impact Ym 1.0 1.0 Stress Limit Coefficient (3)*(3) 1.0 1.0 1.6 1.6 1.6 1.6 1.7 1.5 1.5 (except for compression) l (for compression) 1.0 1.0 1.4 1.4 1.4 1.4 1.6 1.3 1.3 Entss:
- 1. Allowabis stress limits coefficients are applied to the basic stress allowables of AISI or AISC. The coefficiemes for AISC-N690 are supplemented by the requirements identified in subsection 3.8.4.5.
- 2. Where any load reduces the effects of other loads, the coefficient for that load is taken as zero unless it can be demonstrated that the load is always present or occurs simultaneously with the other loads.
- 3. In no instance does the allowable stress exceed 0.7F, in axial tension nor 0.7F, times the ratio of the plastic to elastic section modulus for tension plus bending'.
4 Loads due to maximum precipitation are evaluated using load combinanon 4 with the maximum precipitation in place of the tornado load.
Revision: 21
[ WBEtingh0088 3.8-93 Draft
, 3. Design of Structures, C:q r-ts, E_', m-:, and Systems w- .
Table 3.8.4-2 LOAD COMBINATIONS AND LOAD FACTORS FOR SEISMIC CATEGORY I CONCRETE STRUCI'URES Load Combination and Factors Combination No. I 2 3 4 5 6 7 8 9 Load Descriptios l Dead D 1.4 1.4 1.0 1.0 1.0 1.0 1.0 1.05 1.05 ..
l Liquid F l.4 1.4 1.0 1.0 1.0 1.0 1.0 1.05 1.05 l Live L 1.7 1.7 1.0 1.0 1.0 1.0 1.0 1.3 1.3 Earth H 1.7 1.7 1.0 1.0 1.0 1.0 1.3 1.3 Normal reaction 1.7 1.7 1.0 1.0 1.3 1.3 Ro Normal thermal To 1.0 1.0 1.3 1.3 Wind W l.7 1.3 Safe shutdown Es 1.0 1.0 earthquake Tomado W: 1.0 l Accident pressure 1.5 1.25 1.0 P.
l Accident thermal 1.0 1.0 1.0 Ta l Accident thermal 1.0 1.0 1.0 R.
reactions l Accident pipe reactions Yr 1.0 1.0 l Jet impingement Yj l.0 1.0 Pipe impact Ym 1.0 1.0 l
0.21 5 :
- 1. Design for ==rhaaral loads is in accordance with ACI 349 Strength Design Method for all load combenedman. Design for combinations including thermal loads is described in subsecuon 3.8.3J.3A
- 2. Where any load reduces the effects of other loads, the corresponding coefficient for that load is taken as 0.9 if it can be demonstrased that the load is always present or occurs simultaneously with the other loads. Otherwise the coefficient for the load is takaa as zero.
- 3. Loads due to maximum precipitation are evaluased using load'combinanoa 4 with the maximum precipitation in place of the tornado load.
Revision: 21 Draft -
3.8-94 Y WBElkgfl0080
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@ Westinghouse FAX COVER SHEET Recipient Information -
Sender Information Date: Name: D. Lindgren To: Joe Sebrosky Location: WEC E 300 Phone: Phone: (412) 374 4856 (301) 415 1132 Facimile: Facimile: Bel: (412) 374 4887 (310) 415-2002 Company: U.S.NRC WIN 284 4887 Location: Rockville, MD Cover + Pages 1+ l The following pages are being sent from the Westinghouse Energy Center, Monroeville, PA.
If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374 6529.
comments:
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d l 220.125 , j Revise the final paragraph of subsection 3,8.4.7 as follows:
' There are no other in-seruce testing or inspection requirements for the seismic Category I shield building and auxiliary building. However, during the operation of the plant the condition of these structures should be monitored by the Combined License applicant to provide reasonable confidence that the structures are capable of fulfilling their intended functions. >
220.130 Add the following paragraph to subsection 3.8.5.4.3 If it is necessary to perform reanalyses of the basemat, such as for evaluation of a nonuniform site in accordance with subsection 2.5.4.5.3.1, the member forces at the end of construction will be calculated considering the effects of settlement during construction.
~ 'Ihese member forces will be included as dead loads in each of the post-construction load combinations.
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FAX COVER SHEET l Recipient Information Sender Information Date: 2 /.tg /pg Name: D. Undgren To: Joe Sebrosky Location: WEC E 330 Phone: Phone: (412) 374 4856 (301) 415-1132 Facimile: Facimle: Bet (412) 374 4887 (310) 415-2002 Company: U.S.NRC WIN 284 4887 Location: Rockville, MD Cover + Pages 1+ h The following pages are being sent from the Westinghouse Energy Center, Monroeville, PA.
If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374 6529.
comments:
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220,120 Revise b. of subsection 3G.6.1.2 as follows:
- b. Blaxial Compression (oi-a2)/I01a I + 8 /2 02a s 1.0 De acceptance criteria identified in a. and b. above are used in the design of the containment vessel documented in section 3.8.2. Acceptance by the NRC staff and consultants of the containment vessel design is based on the Design Cenification review which included an independent confirmatory analysis as discussed in the Final Safety
' Evaluation Report. Where additional design evaluations are performed by the Combined License applicant, one of the following criteria shall be satisfied for biaxial compression to assure that the conclusions from the staff independent analysis remain valid.
c.1 og / c2a $ 1.0.-
This assumes conservatively that the smaller principal stress is equal in magnitude to the larger principal stress. De buckling capacity of a spherical shell is greater for unequal stresses than it is for equal stresses.
c.2 When c.1 is not satisfied, acceptability of unequal biaxial stresses may be demonstrated by nonlinear buckling analyses of a ponion of the containment vessel considering the effect of axisymmetric imperfections and plasticity. Where in-plane shear stresses are small, the analyses may use BOSOR-5 as described for the buckling evaluation close to the base in subsection 3.8.2.4.1.1. De analyses shall demonstrate a factor of safety in accordance with subsection 3G.2.
220.125 Revise the final paragraph of subsection 3.8.4.7 as follows:
There are no other in-service testing = 'ag: ; :quirements for the seismic Category I l shield building and auxiliary building. r,wuetwee-However, during the operation of the plant the condition of these structures should be monitored by the Combined License applicant to provide reasonable confidence that the structures are capable of fulfilling their intended functions.
220.128 Revise the second paragraph of 3.8.4.4.1 as follows:
The ductility criteria of ACI 318, Chapters 12 and 21, are considered in detailing, placing, anchoring, and splicing of the reinforcing steel. Chapter 21 of ACI 318 contains special requirements for design and construction for which the design forces, related to canhquake motions, have been determined on the basis of energy dissipation in the non-linear range of response. De special requirements are intended to provide a stmetural system with
. adequate details to accommodate non-linear response and displacement reversals without critical loss of strength. De nuclear island structures are designed for the safe shutdown u . _ _ . . .
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.a l earthquake wi^;ce: cr:di: f= ==gy dizip :!ca within the linear range of response. To provide additional margin, the special reinforcing details of Chapter 21 are applied to critical structural elements pro'viding major seismic load resist.mce.
. Sections 21.2 through 21.5 of Chapter 21 of ACI 318-95 are applicable to frame members resisting earthquake effects as shown in Table R21.2.1 of the commentary.
These requirements were considered in detailing of the basemat reinforcement as described in subsection 3.8.5. They are also applied in the detailing of the reinforcement in walls and floors for out of plane loads when the bending moments or out-of-plane shear forces due to the safe shutdown earthquake are significant relative to the design strength. Specifically they are applied to structural elements if the member force due to the safe shutdown earthquake is greater than 50 percent of the member design strength.
- Sections 21.2 and 21.6 of Chapter 21 of ACI 318-95 are applicable to walls, diaphragms, and trusses resisting earthquake effects as shown in Table R21.2.1 of the commentary. These requirements are applied in the detailing of reinforcement in the walls and floors of the auxiliary building and in the shield building cylindrical wall and roof. Paragraphs 21.6.2.3 and 21.6.6 impose special requirernents when the compressive stress in the concrete exceeds 0.2 ef ' based on factored forces which are determined on the basis of energy dissipation and are substantially below those corresponding to linear clastic response. The compressive stress level of 0.2 fc ' would be close to f'e based on linear elastic response. Since the safe shutdown earthquake member forces are based on linear elastic response, the 0.2cf ' limit is replaced by a 1.0 fc' limit when determining if the special requirements are applicable.
He pcc=ic= =: cpplied ic dm=:: $ : = prince .cinf====: ::=i!: ::::= : dc :
y!:!d = cc = :: ecmp==iv: ::mn:: dev: 6 eccc::: :::=gt ><h= 6: =f 'hutdown
- =hqud: !cci : in=:= d by a f ::= cf !.67 (=:ic cf 1: =ismi: == gin = view
!:v:! :=dqud: :c $c =f: du:dcwn :=iquie). He :: :::um! om=:: include:
in p!=: bbvi= cf i.:::-i= =d =::d= ::.!!: cf he nu&= i !=d
. . p!=: kbvi= cf cylindd=! x !! cf d!:!d building including 1 :c!cm= be:::=
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- cc: cf p!=: khvi= cf nu&= is!=d b=:m :
220.I30 Add the following paragraph to subsection 3.8.5.4.3 I If it is necessary to perform reanalysis or redesign of the basemat, such as for evaluation i of a nonuniform site in accordance with subsection 2.5.4.5.3.1, the member forces at the j end of construction will be calculated considering the effects of settlement during construction. Rese member forces will be included as dead loads in each o the post-construction load combinations.
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W lWestinghopse FAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION DATE: 2 -2 T -48 NAME: C_ , 14aacs TO: LOCATION: ENERGY CEN'TEh -
be 9hmsky EAST PHONE: FACSIMILE: PHONE: office: t//y'57967 ? t l COMPANY: Facsimde: win: 284-4887 GMC outside. (412)374-4887 I
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Cowf + Pages 1+ 1 ~2 The tesowing pages are being sent from the Westinghouse Energy Center, East Tower, Monroevtle, PA, W any problems occur during thle tronomission, please sail:
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Table 720.434F-1 Modifications to iP600 PRA based Insights DISPOSITION INSIGHT Certified Design Matenal l'
ADS provides a safety related means of depressurizing the RCS.
lb.
The following are some important aspects of ADS as represented in the PRA:
Certified Design f
ADS has four stages. Each stage is arranged into two Materialseparate groups o valves and lines.
Stages 1,2, and 3 discharge from the top of the pressuriter to the l
IRWST Stage 4 discharges from the hot leg to the RCS loop compartment.
Certified Design Each stage 1. 2, and 3 line contains two motor-operated Material valves (MOVs).
Certified Design Each stage 4 line contains an MOV valve and a squib valve. Material SSAR 6.3.2 !
The valve arrangement and posidoning for each stage is designed to reduce spurious actuation of ADS.
Stage 1,2, and 3 MOVs are normally closed and have separate controls.
Each stage 4 squib valve has redundant, series controllers.
Stage 4 is blocked from opening at high RCS pressures.
Cenified Design I
ne ADS valves are automatically and manually dactuated ia Material via the protection and safety monitoring system (PMS), and manually actuate v the diverse actuation system (DAS). y Certified Design De ADS valves are powered from Class IE de power. Material SSAR 6.3.7 The ADS valve positions are indicated and alarmed in the control room.
SSAR 3.9.6 Stage 1. 2. and 3 valves are stroke tested every 6 =nt cold shutdown.
Stage 4 squib valve actuators are tested every 2 years for 20% of the valves.
PRA If RNS is lost during reduced inventory conditiotis i with the reactor Attachment 54B coolant system open, a vent path through the ADS h by4th stage s j required to preclude the occurrencecf of surge A g line flooding and t ere not affect gravity injection. ldM ihe e pedgq OLp.6%c h, d % f Y ~2 ag , tLa. WS c c'binM hf /
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e 7/J FAX TO JOE SEBROSKY (NRC) 2/25/98 loc, Per our telecon this afternoon with Bob Palla, please find attached the following information I agreed to fax to you:
- 1. 2 pages (" Draft") of the shutdown PRA Attachment 54B showing the Level 2 focused shutdown large release frequency is dominated by failure to flood the reactor cavity -
contributing 82%. (This covers item 4a of the NRC 2/19/98 letter).
- 2. Markup of SD Evaluation Report page 6.1-3, item j. (This covers item 4b of the NRC 2/19/98 letter).
- 3. I added item 8 of NRC 2/19/98 letter into OITS. It's OITS #6623. See attached OITS printout.
The markup of the SD Evaluation Report will be incorporated into Rev. 3 of the report.
[
Cindy Haag Advanced Plant Safety & Licensing Westinghouse cc: Mike Corletti (item 2 above)
AP600 Informal Correspondance File!
oe 9 54B. Surge Line Floodi:g Effect en Low P:w:r tad Shutdzwn Risk AssessmInt T*RE'h
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-a 1 54B3.23 Level 2 PRA Quantificatior$ Results l
l The quantification of the containment event tree paths are presented on the tree figures.
I Summaries of the baseline and focused PRA quantification results and dominant sequences I are presented in Tables 54B-21 and 54B 22, respectively. Summaries of the baseline and I focused PRA diffusion flame sensitivities and dominant sequences are presented in l Tables 54B-23 and 54B-24, respectively.
I l 54B33 Shutdown Level 2 PRA Conclusion l
I The baseline PRA shutdown large release frequency is 1.5E-08 per reactor-year. The large I release frequency is dominated by reactor vessel failure due to the failure to flood the reactor i cavity which contributes 66 percent. Containment bypass due to steam generator tube rupture, I both as an initiating event and induced by high RCS pressure and temperature accounts for I approximately 24 percent of the large release. Ten percent of the large release frequency is I containment isolation failure. All other containment failure modes contribute negligibly to I the large release frequency. The assumption that a diffusion flame at the IRWST vents fails I the containment doubles the large release frequency. The diffusion flame failure sequence I accounts for more than 50 percent of the large release frequency in this case, however, its I frequency is on the order of 10-8 per rea: tor-year.
1 l The focused PRA shutdown large release frequency is 3.3E-07 per reactor-year.
I release frequency is dominated by reactor vessel failure due to failure to flood the reactor,.J l cavity which contributes 82 percentrEIrly hydrogen detonation contributes 12 percent of the I large release frequency. Late detonation contributes approximately 4.6 percent of the large i release frequency. 'Ihe other containment failure mode together account for approximately 1 1 percent of the large release. By assuming a diffusion flame at the IRWST vents fails the l containment, the large release frequency increases by 20 percent. The frequency of the I diffusion flame failure is 7.3E-08 per reactor-year and doesnt cause the large irlease i frequency to exceed the goal of IE-06 per reactor year.
I l 54B.4 References l
l 54B 1 "AP600 Shutdown Evaluation Report," WCAP-14837, Revision 0, March 1997.
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! y ahntdowrr, a,significant percentage of the severe accidents at shutdown result in dry reactor cavities in which the vessel fails and the ex-vessel debris cannot be cooled.
Iherefore, accident management strategies should consider a timely means of floodin j 7 ,
the reactor cavity; this would reduce the potential basemat penetration frequency l
which, in turn, would significantly reduce the overall large-release frequency.
I The overall CDF and large-release frequency at shutdown for the AP600 has been shown to l l be small as indicated in Chapter 54 of the PRA.
I
' l 6.1.1 References I l 6.1-1 AP600 Probabilistic Risk Assessment.
. l 6.1-2 Letter, Westinghouse to' NRC, DCP/NRC0702, Submittal of AP600 Emergency Response l Guidelines, Revision 2, January 10,1997.
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@ West nghouse FAX COVER SHEET Recipient Information Sender information Date: g/7f Name: D. Lindgren To: Joe hebrosky Location: WEC E 330 Phone: (301) 415-1132 Phone: (412) 374 4856 Facimile: Facimile: Bel: (412) 374 4887 (310) 415-2002 Company: U.S.NRC WIN 284-4887 Location: Rockville, MD Cover + Pegen 1+ l The following pages are being sent from the Westinghouse Energy Center, Monroeville, PA.
If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374-6529. .
Comments:
p g eg Q \$ A h f dlSto ^J O F M Lg 5, g. y.j D psz>As ss ws *uesn " paaur rpz crnss on a e. emes ~r.s iH THE TH/Rh ~SOLL.ET crP J,9, y, g , z
{
is sie I
l 3. Design er Structures, Components, Equipment, cnd Systems Table 3.8.4-1 LOAD COMBINATIONS AND LOAD FACTORS FOR SEISMIC CATEGORY I STEEL STRUCTURES Load Combination and Facton Combination No. I 2 3 4 5 6 7 8 9 Load description Dead D 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Liquid F 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Live L 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Earth pressure H 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Normal reaction Ro 1.0 1.0 1.0 1.0 1.0 1.0 Normal thermal To 1.0 1.0 1.0 1.0 Wind W l.0 1.0 Safe shutdown Es 1.0 1.0 earthquake Tornado Wt 1.0 Accident pressure Pa 1.0 1.0 Accident thermal Ta 1.0 1.0 Accident thermal Ra 1.0 1.0 reactions Accident pipe reactions Yr 1.0 Jet impingement Yj l.0 Pipe impact Ym 1.0 Stress Limit (3)W Coefficient
, 1.0 1.0 1.6 1.6 1.6 1.7 1.5 1.5 l (except for compression) l (For compression) 1.0 1.0 1.4 1.4 1.4 1.6 1.3 1.3 htm
- 1. Allowable stress limits coefficients are applied to the basic stress allowables of AISI or AISC. The coefficients for AISC N690 are supplemented by the requirements identified in subsection 3.8.4.5.
- 2. Where any load reduces the effects of other loads, the coefficient for that load is taken as zero l unless it can be demonstrated that the load is always present or occurs simultaneously with the other loads.
- 3. In no instance does the allowable stress exceed 0.7F u in axial tension nor 0.7F, times the ratio of the plastic to elastic section modulus for tension plus bending.
- 4. Loads due to maximum precipitation are evaluated using load combination 4 with the maximum precipitation in place of the tornado load.
1 Revision: 21 Draft 3.8-92 [ W85tiflgh00S8
@ Westingh6use .
FAX COVER SHEET Recipient Information Sender Information Date: 2/25/98 Name: D. Lindgren To: Joe Sebrosky Location: WEC E 330 Phone: Phone: (412) 374-4856 (301) 415-1132 Facimle: Facimle: Bel: (412) 374 4887 (310) 415-2002 Company: U.S.NRC WIN 284 4887 Location: Rockville, MD Cover + Pages 1+0 The following pages are being sent from the Westinghouse Energy Center, Monroevi!!o, PA.
If there are any problems during this transmission, please call the sender or WIN 284 6529 or Bell (412) 374-6529.
Comments:
Joe, This is to advise you of a couple of minor changes we have recently made to the construction of the containment internals. I do not think that there is any impact on your safety evaluation.
There is only a minor effect on the SSAR.
The stairs in the CVS room will be steel rather tnan concrete. This shows up in SSAR figures 3.8.3-7 and 3.7.212.
The liner in the room under the reactor vessel will be carbon steel instead of stainless steel.
For decontamination it will be painted with epoxy. The specific liner material is not included in the SSAR. I do not remember if we have referred to the type of materialin discussions.
Please give Bill Huffman a copy.
Don Lindgren l be o sTWFS
- 2. m acu t.
11 CM T.sa n
l' M Westinghouse FAX COVER SHEET W
RECIPIENT INFORMATION SENDER INFORMATION 1
OATE: A/l 7 /W NAME: Su e_. Fa(to l
(
TO: LOCATION: ENERGY CENTER -
0;Il Nuf[gan EAST PHONE: FACSIMILE: PHONE: omce:
COMPANY: FM win: 284 4887 MdC outside: (412)374-4887 LOCATION:
C - . p e. i. L 1he following pages are being sont from the Westinghouse Energy Center, East Tower, Morwoeville, PA. N any problems occur during thle transmission, please call:
c:29 WIN: 2SN( Bev ) or Outside: (412)374-6529 COMMENTS:
6,' II, L saSeA v
/
vaUe.
I aues' w e- te)ereeed.
b i
e Sealing of High-Pressure Steam Safety Valves t Br R. E ADAMS 8 Ann J. L CORCORANs A f" tal lavatisatism of eenling with high.
- pessouse steman safety valves has shows that peer eenHag valve le of vitalisterest to the demaner. %s e le a resalt of self.laduend growth et tiny laitial lenke. in making this say as smaB as posible sonstitutes a mapalha Espanelen of the lenklag steam osele leant arene of the factor of boiler design. De there saa be me wide dismenes be
- valm esat, enesing sentrasties of the samtlas aurinese la tween set and operating pressen.
Eny fastere contribeto le leakage in veins. Ansong the a mammar which Imeresses the else of the leak. A new dwiga et valw seat was i. i, i ' , - "--j- thin mest important is distortasa of the esating erfases enused by Sesie senting sudness. N seelias essete of tbe leaking outaide imensaeas laeloding instanation strains, pipias, etenam were mialmimed by providing better beat transfer and aanbient temperatees, p.c ' t on oesdoor5"'" _. _
freen the high.teenperature steam. N mew deelga re. Fathertheathis,anyminate anseats of dirt or foreign mate.
suited la "'we lamprowwment la eenllas, and riallodging on the seats, together with damage enmed by grit service tests how shown esamuent perferenames. in b stensa, een also be neuene of safety-valve leakage, Another serious probless la comanetion with anlety-valve lenk-Irracovertow .
age le that when leakage starts, for any reassa, sentiamed op.
~
erstion of b possume vessel at b normal pressare win result HE two primary requartments of steam anfety valves are I* "*d***3 3**I*8' d O' "3" D3' 3**d' ** **""* 'I **
(1) that b valves open without fall and release steam seating narfaces and to inerenses la leakass to tne point e15ciently when the popping presume is ranched, and (3) es that the valves close at a slightly lower premme and prevent O' "3"- must be taken est of earnee la order to meendinen leakage of stansa M all presumes under b eloning presses. A8 Oe 8te*8B prumaru han increased la es pad 26 yens
%e methods of popping and closing the amfety valves ass well the probium of safety-valve lenhage has beenme ' "y" understood. Howenr the problem of leakage of steam safety severs.
His was particularly true as pnamures rose above the valves has long been recognised by designers as one of b anast 600 to 90(Mb class to 11100 lb and above. It we dificult problems which by face.
experience of amiety-valve esses that valve daagne which could he essential elements of a safety valve, in so far as sealing is y b nh @ at 600 and 90(Mb pmenuse proved te be concerned, ass the feather and the sent bushing. De sent a ==lanble proWan at 1500 pai.
bombing is simply a hollow tube connected to the boiler, with a cardully Enished sent on the top surfeos, while the feather is his pmblem had enstad for many years with cely mymg degrees of aneeems in improving aanhas la various plaats throusb-essentially a cour with a similar sent held asamst the asst bushing out the country.
Maar difment tenheisimes of aest taishlag, by pressare from a spring. Men the steam presses os b changes in densa, and b like, were tried out, wi bottom of the feather esoseds the force of the spring, the design impmumet in the standasd of tightama of the nlus at 1800 of the valve is such that the feather pops up and steam is released i until the pnsure drops to a p.d.r " vahne. At 8*8*- that point psi, but without auking any duided changs( '
the feather is fomed down by the spring, and the valve le assia cloemd. n--- of this, it was desdad the a fundamental ressank tamtigaties of the ennen et niety.nin Inkage was amantial, i s ne renews for the daf5eelty of asahng with safety valves, as nis would not to a stady of the past esorts compend with other types of valves nach as abutof, gate, and ing the problema. Instand it would be an appmash fusa a An-globe valves, or the like, is that with these latter types of valves, damental basis by somaans entire a tammendees lead can be impoemd on the mating parts of b who would not be nabject to pastknowledg <
valves by simply inerendag the elesing fores on b operating W gretriction.
handle of the valve. In dieses oo'nparissa to this, on safety valves the senting surfaces are held together only by b diferense I"*'"** " " ' ' ~
in loadaag between the set lead of the spring and b operating presumo of the steam. dppareame Uesd. Aaserdingly, an espernamatal
4 his difmestial load is quite anaD, bensuse it is naturally investigaties to stady the amahng of safety valves wa desarable te' operate a bouer at a presume as eless to the sensa- A standard 8-in. saisty valve daagned for operatism at a mussauswahlevierkingpressenas pondble. Sineethemaximma of 1800 psiwas used as the test valve. -
allowable working pressee is the setting of the safety valva, Two typen of anatorialsfor the featherand the east hashing wess the gap bes=== the est and operating pnmes of the esisty of intenst in this in,-airaal== One one a duritts enshine smde of senialem stal, kasen as Type "F," essimising abset
' 3188ab Ehslassr. Batesbo Memertel lassieste, Calumbus. 0hi** 14 per esat ehromium and 9.13 per est enshen. Ha ether y g type, haeus as SteWeed trim, had matlag suminass of Stske
'lan. Mem.MME.
Cenergesad by the power weidad este an enstamitie sesimismetsel been.
Mesmes, Washnessa. D. C., April 13-84, legg,a-" er Theand presamled FIS.1 showsata the sortagdrawing af the sessen gassanter do.
sabannels u.am.m.d j ~h advessed in papee ese to be signed and used let lebenstery samung tests wet ta valse; e hailst osasisted singh of a stest ph espped e4 Ga leser et the Gastser.Ass Ptouredens es,, mama 38m. gS-4se. er their enehme and an e.e end, with a senaded s.im, anage at es top sur esammtesa o tan valve. %s hegar had a empe4 ef sheet 35 36 af water liar 4
1138 "1RAN8AC1' IONS OF THE ASME .
SAFETY M W NOVEMBElt,1950 TEST GAdg proved by the better anish on the s .
Charedmatics of Serhp Fish Niireyes and & Sisesi NEGLE[ "--
Ms
, CARE 0+k IU dillerent fmm esaling with nitrogen.y steam vatyt
, is FT. e ist. L0esees0LY STEEL PtPE, y8 ,
popping point, stana leakage was found to be a gfg 0 "" of 4 per cent below the popping po
- f. y MAC $PncED WATER LtytL AT[ r-obts.med when testing with nitrogen. mate 1500 L g$. 600*F. It was then reshand that
=,,, HEATING ELEMENTS TO CONSISTOF 8 CotLS saling as obtained with nitrogen was undoubtedly
- E '# #
WATER LEvtL AT)em AT C Lat gig *F..
4 Mnmg the diferemens whichs eam and tuung with nitrogen.
enstad b
" with steam and sealing with nitrogen was no ge i = i Rtpuct wat space taats with successively increaang and decreanng
- 0. 4 PER TURN his is abown graphically in Fig. 2. M for upper nitro-curve, p e THERuocouPLES gun, shows pressure increased.that the rate of leakage gradually e increasj TO CHECK totLEA t A # the popping premure was reached.Only very moderat g _ .J On redeems the pressure he. I v - - -
only aughtly greater than when the .
DiaanAs or,LAacaAtoav Tser Bou.ma ,
NITROGCN TEST "
which was added before attaching the valve.
by eight separste electric reestance eleme.nta woimd around theHeat was supplied y
elementa could be connected ing several diferent combinationsouts to provide the desired degree of heating. ; l placed at vanous positions on the boiler to insure d 8
that no sectionshrmocouples were atlafsetory for the aork, and pressure M could be controlled mai I by proper manipulation ments. of the connections ofE the beating ele Tscs TecArupe !
boiler, but early testa indicated that popping was not eenentialm to the invmtigation of sealing,
/'
/
valve coul j
slowly, the power being shut off na soon sa rapid leakage st so that the valve did not umaally pop. ,
STEAM TEST 0-l
_tanting,
^
& 1Jy than they norma valree would in leaked more rapidly and to lower pressuresWith this methodC of w.,= g when a valve closes after popptng,lt ticees erable with conadThis is partly baranaa, E i
impact, which sametain obtaining tightness. E ,
as b popping pressure; for mostg8 teste it ranged between f aboutThe 1100 and 1250 pm.
To facilitate comparison among vanous gg
_ fi e testa, point. pressure data are repotted na per cent below pping
! & po i l_
I i
1 aspirator being used to pull the stamm through the jcondenserImak 1=aknee rates are reported as cubie feet of ste 8 ,, 8 pherie pressure per hour. am at atmoe- M popp,,opoh to 600 eu ft per hr could be measured.I.makageno.rates 2 ranging from about 6 Laassen-Panesvan Rauvione Wrva Braus the valves were made using tank nitrogue prommare gas, as a scoresAs part of the fu f == hap with nitrogen was determinedWith steam, byincrossed meseof a very diferest situation existed. As the high-uring the rate of A-" ut of known vohannewas of water freenfroesa a low value, leakage y n-at first gradm burette Probsw'=ery ooanected Temen. to the steam outlet of the valve.
slightly -
below the popping point, sue g indiented that -"mt of the rancesve parts increasedof the to na enormonely valThe highof results vales sambag skog
. If the testsline BCw within plant talesaaeas had little efect em osalhas untilit It was was f n =adaat to essen the valve to ;pop eend that enesidemble inquovemeest lae sneling with nitrog could be point C b preseme were decreased & line CB eenid m However,if at reatie.d br usina a W#
- 2 6mish on b amating =v laste mtraeed g lower mpe; esm,ad, b leakaan rema,ined quite high to a amm w shown by the Emo CL At point E, leakage
. s E. -
)
. ADAM 8, CORCORAN AFATING OF HIGHoFREBBURE IffEAM SAFETY VALVE 8 nas died down addenly to a normal vWue F, and if pnamare were in-creased at that point, !=h et would then retrace the line FB. son,E h difenace in behavior of the two game was striking, and it Soo. l j
was evident that if b reasons could be disconrod, s' solution to the problem of poor senhas could probably be dermed. /e',-
,,, - 7 ' N,' ,p j It was 6rst thought that & efect of b leaking steam might be to introduce thermal gradients in the valve body, canning it to
/f s g I distort and force the fenbr away from its true postion. Tem-
// \ g e i peratun measurements of the valve body did abow that tem.
t perature gradients enstad. However, when these were elimi- f
)I ,- \\
\
nated by beating the entire valve to the steam temperature, b ' '
improvement in sealing was negligible.
Temperasune of Folee Trin. In a further cEort to determine
,j , ,
E f
f
, i l g e l
& cause of rapid leakage with steam after leaking had once I started, temperatures of the vain trim wm sadaad during a g g
[ f leak test. In a test with Stellited valve trim, thermocouples were welded to the side of the feather, the location of b beads
\
l f
being indiestad in the diagram of Fig 3(a). Although the thermo.
couplee were not protected from the steam, good thermal contact
'sN { /
4/
1 between & thermocouple beads and the feather was assured aince & beads were spot welded to the feather. Hg. 4 shows \(*
{
graphically the temperature that were recorded during a seshng {
I I"t'
, _ $ TEAM OU_TLET I a 8UORE LEAKING STAATED,PRESSUREe Cg5 L81 0---LEAKING BADLY AT 1038 L85. / 50. IN.
LE AKING LESS AT 940 LBS./ $0. IN.
, &=-- AFTER AUOfSLE LEAKING CEASED AT 864 t.851511IN.
' ~
[ Tre. 4 Tanata4vvBas or Ov51.WB MvEEa Demme Laas Tapr
,, [ 'soo H i Ng ec ,
.gi o.
l , f 00* y' A '
CA no. Se (h/s) location or Tasamococyts Buse on Syst.uvs f
/ ,oo.,
o g-
}\ 3i \
l m- -
.g no. 36 (rvAs) Location or Turnmocovets Hotae in Tvra T l
g ,
l 4g Beenrwo '
,f A similar test mas made using the Type T valve tism. For this test, the thermocouples were spot. welded to the bottom of
.\
\\
a #
/,f
/
small holes cittlled in the seat bushing, so that the thermonples
/ ,)
r could be completely ahielded from the leaking steam. Iocations of the thermocouples are shownin Fig.3(6). The temperatures at pointa around the seat bushing are abown graphically q p y/ (
j
, In both cd these twta, the valve-trim temperatune were quite uniform up until the time leakage started. At that point, tem. STEAM XJTLIT peratures at several locations dropped considerably. When the -
- SEFOPE LEAKS 46 STANTQpMES$UREel100LSS./$4IN.
valve was leaking badly, severe temperature gradaents were ob. o .."LEAKil88 Ki 0 LES,/ 94 pt served in the valve trim, the temperature at one point of the e-.= AFTER AUDIOLE LEAF 0NG CEASED M CDS Las,/ $0. st.
feather being more than 300 F below the temperature at another point. As the boiler pnesure dropped and leakage became less, yie. 8 Taaesa4venas er Trvs T Besssse Drame Laas Taor b therenal gradients in the valve trim also decreased. When premmares dropped to b point that all leakage had esamed, the seat bushes were less than those noted on the feather for e valve-trian temperatures had returned to about their original number of reasoes. Leakage was oceederably less with the values, and the thermal gradients no longer easted. Type T trim than it was with the Stellited trim. h theres.
It woe thus apparent that the lealong steam was eachng couple beads on the Stellited feather were leented practisally be the seat serfaan. 'Ibe tasaperature difenmens emanared in the line with the stoaan Gow, med weeW be enoisd dready by eentest I
i
, gto -
TRANSACTIONS OF THE A8ME NOVEMBER,1900 with b steam. In b seat buebing, the beads were won abielded tore which plaet esponemos had shown to be ebarnetariads af frosa b steam, and were located essentially between the eccled area of the east and the hot body of the bushing, the body of the esalag could be explained on the beds of this theory Seahas was more ddlicult to obtain with higbar posseuse beenese bushing being malatained at a higher ta=r==a-se by contact the emoling efat from the expaaman of throttled stemm inanness with the bot steam over the atire internal dansaster. %us it aseses likely that the thermocouples on the fanabar measured directly with pressure, as ebown in Fig. 4. Similarly. enli !
i weses for esturated steam than for superheated steam beenese appeoniastely b temperstare of the leakag stamm, while the brooccupies is the sent bushing metaally mensand the tem- estorated steam ecols more when it expands la a throttling pseem he diference in eenhas between the tu." 7,. stainism peratures of the seat bushing at a Anite dastanes below the seat steel sad the 8tellited tria souM be explained by their diserenses sufses.
Caele'at gf neuerJef Steen.~ ne breal gradassia foundia b la thereaal espaamon, the -84=t of thermal ==r==a= for the 8teniu and for the austaitie been both being about 46 per sent vain trim were esused by the cooling of b steam as it espanded greater than for b Type T material.
through b vain sent When any gas is Glowed to expand PossiMs Methods /er Isiprerwir Seelear. On b beeis of this through a small ori6ee it espands in appmzieutely a thattJing theory of leakage, severni poemble monas of 8m% senhas process, and an imperfect gas will be cooled. De amount of cooM be M ': .' %ese are as follows:
ecoling for steam depende upon the laitial pressure and the degree of superbeat. Fig. 6 abows & degree of ecohng indneed as steam 1 Kaap laitial leaking low; if the amount of initial leaking of various qualities is allowed to espand ima diferent preamres N ,"I *MM.NN by the thmstling to awayhade presame. Satarated suam at 1000 poi cools about 250 F when p d-d to atmosphene pre. 2 Use a suting mhrial wie a low aa mW d w sure by a throttling process. Nitrogen, a far more perfect gas. ,, y ,
ewmia cools only about 6 F during throttling imm 1500 pm. d m% Q* tion d a mW soo dW est a 4 ' n--w M W W u e
=h==t of about 4 X 10" per des C, about 80 per east lem l l than that of the turbine-type semiala= It appeared that only a 5 s moisfunt m limited imprownet in sealing inight be espected from sah na satunArto straw - stuck.
eco 7 3 Use vain.eest materiale with high thermal conductivity, g
8
/
[ '
so that the thermal gradiente ladneed la the valve easts would more rodily equalise. Howent, most materials which han high bCO
, f thermal conductivity are not suited for use with hisbpressee l ; j steam; also, only a limited degree of improvement could be es-f , / saTunaTrositAnsl pated from eis attack.
4 Reduce thermal gradients in the valve seat by provulang a g( / /
s
/'" '
abort but pae between the seating surfaces and the hot stenan.
5 Decrease the rigidity of the vain aesta, so that less presem 1
diferential would be required to overcome any distortice of the p # j valve seat.
e 998 i ano'f New Dwipe of False aset It was soon appant that a change in design of the valve seat ofered the best means of =tnimimag
- "0" b characteristles r=Taa-kla for poor nenhag. Aceor:bagly, a o
o wo new type of valve aest was designed,in which the senting surfases icoo 1500 2000 2500 were undereut to provide relatively thin aesting surfasse. Fig. 7 emEssuftC-l. ass PER SCL st shows a drawing of b maddad valve seat. neue advantages yie, e coorune or Graias ano Nrvacean Wsaw Tamom.as to were expected of this design; thses are listed as follows:
Avvoarusase Paaesoma 1 Darinal gradients in the valve esat would be reduesd.
M ahansein */ Id'A*F. From this he cooling efect of the steam as it leaked would be eoanter.
feakage of steam amfety valves wasIf w,
_ , p, inform a tiny steamties a theory of acted by the hot steam on the beek of b thin ele tak exista at some point on the valve seat, the local area near wmW pmvWe a high mh d but Wm ud my W & I 4
that point is cooled, and the metal tends to continct This eco. on b mdag sufnem Md b ef= duly W
! traction causes deformation of the valve seat in two ways, Arut, 2 'I'h ein elmmu M da8at mily udw m @
by direct contraction i --- l -- e to the esating earfaces, and disanadal pm ud rudily hk my wwpass a eseood, by contractice in a directice parallel to the arenminnee g g ,g ,g of the seat. Both act to increase the else of the gap between the 3 The entire fores of the steam pressure would act to held l feether and b seat bushing at the local area of leakas, g;, ,gg, g ,g ,3,_ ,
g,,,,,, g , ,g _,
Thus, o.a. siman starts te iak areagh h vain out, the puhd est mling wund b M law 6 d b W siae of the gap inerenses rapidly until the rate of lenhage is es- 3,,,,,g,,,,,,,,,,
tresnely high. When preseme on the valve is redused, rapid A detailed drawing showing di====i == et the Amt sepms.
leakage does not step until the diferential forse on the feather mental valve seat of this design is shown la Fig. 8. He amat was la mesient to overeome the warpnes eneesd by the thermal dessmed as that whom the full lead of the spring was se the gradiente. See the sensen for the grist diferenes in sealing father, er during the impmet of enseing after a pop, the thin with leeressing pseases nad deerunang peseene with steam. edges weeld deform elastienDy, se that the load would be ennled nad the dismensela sealing between va== and antase is m- ,
plaimed. by the thicker portions of the vain trina. Howww, when steam !
4,ramas massem ne y and = .- ' premsee was sueh that the spring land ameeded the gas premme !
Many d the fee- by emir about a hundred poemde, e main land weend he i
l
1141 i ,
- ADAndS, CO3CORAN ==auM3 CF E105 PRISBURE STEAM SAFETY VALVE 8 1 N* $l f g.e ;; ,w -y gam,m 9
- V Yk h
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Fio. 7 Aseouste DaAwswo or Nsw.Damon Vatte SEAT carried by the thin portions, so that the sent would be quite Gezible in the range of pressure where poor sealing occure.
Although this design has been considered from the aspect of COO ,
a conicalmted valve, it should be equally appliesble to designs - m- __"= . i
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of Sat eented valves.
Im6eessary Seeling Tsass Win New Deeign of Foles Seat ,
?, t ry Sealing tens were made with b new design of valve seat, as well as wie a conventional valve seat of the Tne T material.
I /g X CONVENTIONAL v /*[g 'I iE I I
$co Prior to ti.e steam tests, both types of valte tnm were tested be
-5 EAT DESIGN with nitrogen to insure that no serious defecta existed in the seating surfseen.
f B
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'f E
h results of the sealing tests with steam are shown graphi-cally in Fig. 9. With both types of trim,steamleakaas was email I M" e
f at 6 ret and Goetuated considerably. The .. . . ~ values of 4 "
- i leakage with the conventional seat were reached af ter the pressure ge Isf Es
- a y! a 8
had dropped to between 1 per cent and 5 per east below the E /.
/ I e (
popping point. '111s simply indicated tho' 6 nite time required for the thermal gradaente and defortantion of the valve trim te { J o j e mN e ,
dev.io, an., tie iniasiieaung si.,ted.
- LEauAGE C -- . - -
Charseteristics of the asaling testa are discussed in the following i i ig paragraphs. j Type T Tries. With nitrogen, this set of valve trim was , l ,
g NrinoGEN 8EW SEAT found to seal about as well as those which had been tested pre- DESIGN L W AGE.
viously. De leakage curve for the Type T trim is the basis for y judging the analing charactenstics of the meda$ed demsn. 3 New Sees Demps, h greatimprovesmentin eenhas with the g s L'Nh new type valve sent is ciently indiented by the data. When '
tested with nitrogna, sealing with this tria was about onaparable l oj m,.-
,/ # /
to that at the samventional amat. When tested with steam,lenk-aos with the new design at 4 per cent below the popping point wee as low te with the oceventional valve sent at 12 per oest below the popping point. It had been espected that initial leakass =vais be i.wer for the new design, th for the so.- is is o e a e a o vestional trim, owing to the deeibuity et the saatu suriness. %-ettow poppine Patssung Bewever, the tests with nitrogen indiente that initial lenhage with the new deden was probably equite - , "- to that el> Fae. 9 Saanase Test Data Wres Two Vaave4 MAT Danemen e
11 9 TRANIACTIONS OF THE ASME NOVEMBER,1980 tained with tbs Type T trian. malt sonned that the eseellent breakdown tests to detanmne the femahdity of the daign were eenhas with steam must have been due in a lary part to b carrwd out. ;
better heat conducties to the senting efarfaces. i One of these testa conasted of the continuous popping at 1200 '
Ple<ed Tests Fid ds #ce Dseys. Since such ===11-t results Ib of one of b sent ebaants over a period of time. He valw had been obtained in asahag tests with the new design of valve was blown rapidly and continuously for a total of 200 thnen and sent in the laboratory, it was sent to the plant for forther in. than was subjected to chatter or very rapid popping to deternune vestigation and d.. ' r t. whether ammt damage would occur from & type of opention.
M Erst step in 6 was an exhaustin test of b ve)w which At b and of b test, the valn sent showed no damage in so far had been tried out in & laboratory. ne vain was tested at as bnakdown of b senting element was concerned, although b 1200 pai stesa presma over a period of some three weeks, & saating surfaces abowed wear, and a slight amount of leakage valw being tasted penodamily by popping and then being left had developed.
under loading for a pened of thne.
A test for measunas & fatigue life of the sent element was At the end of the tesung, the valve having been sabjected norked out using a camoperated weight. loaded device which to a total of some 200 pope, the aesta were --" nere dmpped b amat for the distance of its normallift with a weight were no signs of seat deformadan or failun, although some wear of 200 lb, so b disk would stab b hashiny with M--kt.-
had occurred on b amating surfaces. force.
nie indicated that the iden -haA d in b undercus sent was With h markan=n, b east was subjected to a test of 80,000 of a sui 5clandy rugged nature that it would withstand canunuous eyelms to deternune whether or not fatigue of b seat element service blowing, and that the research results abound be do. would develop. .
veloped into a practical denga.
At the end of the test, the disk and east were still structurally P ';t of tAs TAsneedese Demp. An analysis of expected per(set and abowed no signs of fatigue failure.
6 eld conditione naturmily abowed that it would be desirable Serviss Raeidas FiA TAarenadese Demp. h 6 ret expert.
to incorporate the deman changes only in the feather,insemuch mental htmodase seats were put into service in late 1944 l as this part would o8er the best opportunity for r-N-t and initially resulted in a tightnem comparshle with that which when wear oecuned. I'ndercutting of the bushing could be done, had been r=li=ad experunantally during the remnanh develop-but siisht went might necessitate replacement of the bushing, ment phamaa. It was found - y to incorporate a de8seting which is a more dimeult taak. piece on the nose of b Thermodme comparable in contour to Accordingly, & development was aimed at developing a design the older solid. disk design in order to elinunate a high-pitched which mould give the required best conductiftty and Sexibility noise which originated, appaready due to the toessein the seat without embodying any fundamental objecdon in the way of Of the hrmaA -a which were put into aernes initially in 6sid maintenanee. late 1944, two were later removed for ava=== tion and unrol Further analyas of past experience showed that if the seat are sull in aernce at the original planL wsre made so thin as to be very Sexible at the lower edge, as had Following the initially sucesssful testing of the Hermodisc been niet proposed, there was more danger of potential damage design and the ebecking of these after about six months'aernee, to this sent from gris passing through the safety valve. This is early in 1946 a complete change 4ver of a large high-presnare based on experience, pardeularly with superbenter valves. from the utility was made, allof the hrmaa riving very good results (met that very hard particles in the steam actually pit and abrade as to tightness and operating charactenstica during this installa.
thz noses of the safety. valve feather diska. Itcould camly be seen tion.
th t a durable edge on the undereut sent would be desirable. . Since that time, sental handwi of the hrmoduce have been In addition to this factor, making the sent edges quite thin put into service with equally swu steults, at pressures tanging up would result in very close tolerances in manufacture and would to 2650 pe.
increase the cost of & design, With all of these considerations in mind, a large number of Acamowzanonourt variations in anglen, hknames, and depth, were tried out, 6aally his rosea ch was done in co-operadon with Manning, Maxwell
- rnv%g at a @kness of 0.012 to 0.015 in. at the bottom edge of & hfoore, Inc. hir pernuance to publish these results is tha seat, vith a slightly deeper reces than had been indicated in gready appreciated.
tha original design.
h work at Battelle was under the dinet supernoon of Plans Tsade of Thanno6ae Desyi. Having determined the Dr. H. W. Russell and Dr. R. W. Dayton, hit aanstance demiss which seemed to be most practical from the 6elo-sernee throughout the work was very helpful. The iden for maaA = tion viewpoint and from the manufacturing viewpoint, a series of of the valve.eest design was eeneeived by Dr. Dayton.
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' Westinghouse FAX COVER SHEET l W .
RECIPIENT INFORMATION SENDER INFORMATION q
L DATE: 8 - l 7 -98 NAME: C . Ihmq :
TO: LOCATION: ENERGY CENER -
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LOCATION:
Cover + Pages 1 + 2} = 2g The fogowing peces are being sent from the Westinghouse Energy Center, East Tower, Monroev0le, PA. If any problems occur during this transmission, please call:
WIN: ( Bev ) or Outside: (412)374 -6529 ,
COMMENTS:
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- liF .5 E y Under item le (PRHR):
Capability exists for the control room operator to identify a leak in the PRHR HX before it can degrade to a tube rupture. denng-awbseque .t 6:ign be.:i: .cc!&n: (DB.^.),
, The PRilR HX, in conjunction with the PCS, can provide core coolingfor an indefinite period of time.
l After the IRWST water reaches its saturation temperature, the process of steaming to the containment I
initiates. Condensation occurs on the containment vessel, and the condensate is collected in a safety.
related gutter arrangement which returns the condensate to the IRWST. The gutter norma!!y drains to the containment sump, but when the PRHR HX actuates, safety-related isolation valves in the gutter drain line shut and the gutter overflow returns directly to the IRWST. (disposition = SSAR 6.3.2.1.1)
Under item 6 (RNS):
Planned maintenance oi the RNS and its support systems (CCS and SWS) is performed et-power in ofodes I, 2, J. (disposition = SSAR 16.3)
- Undet item 10:
The operation of RNS and its support systems (CCS, SWS, main ac power and onsite power) is RTNSS-l important for shutdown decay heat removal during reduced RCS inventory operations. [SSAR 16.3 disposition will be added) asDt.h vasr4*.g l %.,,uiu,, af.9'!:&at-power condition 1r "-- -- ;= ;:.- L.., .- - s .. ., T ^ !.nw in 5e e l
Short term availability controls of the RN& ;:::":b (disposition = SSAR 16.3}
- Under item 13, add: &
l To preventflooding in a radiologically controlled area (RCA) in the aardiary buildingfrom propagating i to non radiologically control'edareas, the non RCAs are separatedfrom the RCAs by J and 3-foot walls l andfloor slabs. In addition, electricalpenetrations between RCAs and non RCAs in the auxiliary l building are located above the maximum flood level. (disposition = SSAR 3.4.1.2.2.2) l New item (#42):
l No safety related equipment is located outside the Nuclear Island. (disposition = SSAR 3.4.1)
"3"-2 W westineoun I j
l e
NRC FSER OPEN ITEM I i=
- 8. De AP600 !ow pressure systems which interface with the RCS are protected against interfacing systems LOCA , f 4
(ISLOCJbyJu;ombinatinn(multiple isolation valves, valve interlocking, increase in the piping pressure limits V (nikTpressute relief capability.) gg d k b lo 7h% 57-29, 5 . I E Response: This is an accurate statement. _PRA Table 59 29. item A spectfically discusses the elements which prevent interfacing system LOCA between the RNS and the RCS.
\ !
b g 9 Solid state switching devices and electro. mechanical relays resistant to relay chatter will be used in the AP600 % r</a
!&C system [ Use of these devices and relaysdi= :!Hnere minimizes the mechanical discontinuities E
i ,
associated with similar devices at operating reactors. (A -
l E Response: It is not understood why the stafs statement is an insightfrom the AP60) PRA. The stag would need to explain why this is an important insight of the PRA to justify its placement in the DCD.
The stafs statement is accurate, but is not explicitly stated in the SSAR or PRA.
x /
- 10. Here are no watertight doors used for finnd protection in the AP600 design.
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E Response: This is an accurate statement per SSAR subsection 3.4.1.L2.
./11. He AP600 design minimizes poter.:ial Gooding sources in safety related equipment areas, to the extent possible.
He design also minimizes the number of penetrations through enclosure or barrier walls below the probable maximum Good level. All flood barriers (e.g., walls, floors and penetrations) are designed to withstand the maximum anticipated hydrodynamic loads) well as water pressures generated by floods in adjoining area]s E Response: Excluding the ending phrase "as well as water pressures generated byfloods in adjoining areas."
the staf statement is supported by SSAR subsection 3.4.1.1.2. This is essentially item 23 of PRA f ,
Table 59 29. hf hl
/ 12. Drains are capable to remove Dow from an assumed break in a line up to 4" in diameter and include features,
)'yt, v, i
such as check valves and siphon breaks, that prevent backDow. (t l f )#
v E Respo
..: !=l=--"~c nc wnkr ed in the stafs statement is not supported by text in the AP600 SSAR.
b'
& Glarge enough to accommodate more than the design flow gnd by making the flow path as
. h he .
straight as oossible. Drain headers are at least 4 inches in diametes" Regarding the portion of the stafs statement on backflow prevention, see the last bulletfrom item 15 below. 3 dT i
- 13. Dere is no cable spreading room in the AP600 design. \ l E Response: This is an accurate statement. $ *
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W Westinghouse
- I
- 1 l
l I
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J
j NRC FSER OPEN ITEM .
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- , W Response: Westinghouse recommends the staf revise this statement to read "The passive containment cooling f system (PCS) cooling water not evaporatedfrom the vessel wall 0ows down to the bottom of the inner containment annulus. Two 100 percent drain openings, located in the side wall of the shield ,
) INS s .g building, are alsays open with screens provided to prevent entry ofsmall animals into the drains.*y f
s }p' Note that the specyic drain configuration has changed since what was modeled in PRA Revision ^
8, when the drains were located on thefloor of the annulus (see also response to FSER open item ])l Y? "
720.440F). Thus the staf's statement should be revised.
.r 7 The annulus f drains, Wn m m.am:!s pys. 6aaddco mm tne waii vi k 1 'd Luiii will have k -
the same (or higher) HCLPF value as the Shield Building. M cr$fs that the drain system will not fail
(#
( at lower acceleration levels causing water blocking of the PCS W{ air
( 9 Response: Refer to ite n 6 above regarding placement of the annulus drains. 7 16-h The COL applicant should develop and implement policies, procedures, and trainin to close containment penetrations during Modes 5 and 6 in accordance with TS 3.6.8.
g f
$e[ Nl Response: A COL item in SSAR subsectior 13.S.1 states the Combined License applicant will address plant procedures. A COL item in SSAR subsection 13.2.1 states the applicant will develop and
\
implement training programsfor plant psvsonnel 7hese items inherently include following the Technical Specifications. The COL items in SSAR chapter 13 cover the staf's statement.
Auxiliary Buildine
- 1. Separate ventilation systems are provided for each of the two pairs of safety-related equipment divisions 5 y. supporting redundant functions (i.e., divisions A&C and B&D). His prevents smoke, hot gases, and fire
,' suppressants originating in divisions A or C from propagating to divisions B and D.
E Response: The stafs statement is covered by item 20 of PRA Table 59-29. Note this is essentially a duplicate \
Gp ofitem 17 of" general & plant wide requirements."
- 2. He major rooms housing divisional cabling and equipment (the battery rooms, equipment rooms, !&C ,
rooms, and penetration rooms) are separated by 3-hour rated fire walls without openings. Here are no doors,
)
dampers, or seals in these walls. He rooms are served by separate ventilation subsystems. In 'er for a j fire to propagate from one divisional room to another, it must move past a 3-hour barrier (e.g., a oor) into A a common corridor and enter the other room through another 3-hour barrier (e.g., another coor).
,k P W Response: This is an accurate statement. It is essentially what is described in SSAR subsection 9A.J.J.
W Westinghouse 3
48
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O NRC FSER OPEN ITEM -
52 t="i l
l Ensive Core Cooling Systems (PXS) ne passive core cooling system (PXS) is composed of (1) the accumulator subsystem, (2) the core makeup tanks (CMTs) subsystem. (3) the in containmeet refueling water storage tank (IRWST) subsystem and (4) the passive residual heat removal (PRHR) subsystem. In addition, the automatic depressurization system (ADS), which is part of the reactor coolant system (RCS), also supports passive core e ions.
l W Response:
The stafs statement is covered b stem ) of PRA Table 59-29. l()
Accumulators De accumulators provide a safety-related means of safety injection of borated water to the RCS. He following are some important aspects of the accumulator subsystem as represented in the PRA:
There are two accumulators, each with an injection line to the reactor vessel / direct vessel injection (DVI) nozzle. Each injection line has two check valves in series.
' He reliability of the accumulator subsystem is important. The ( 0L will maintain the reliability of the accumulator subsystem.
Diversity between the accumulator check valves and the CMT check valves minimizes the potential for common cause failures.
l W Response: The stafs statement on accumulators is covert by item la of PRA Table $9-29._ L Core Makeup Tanks (CMTs) I ne CMTs provide safety related means of high-pressure safety injection of borated water to the RCS. De following are some important aspects of CMT subsystem as represented in the PRA:
There are two CMTs, each with an injection line to the reactor vessel /DVI nozzle. Each CMT has a normally open pressure balance line from an RCS cold leg. Each injection line is isolated with a parallel set of air.
operated valves (AOVs) which open on loss of Class IE de power, loss of air, or loss of the signal from the PMS. De injection line for each CMT also has two normally open check valves in series.
.The CMT AOVs are automatically and manually actuated from PMS and DAS and their positions are indicated and alarmed in the control room.
CMT level instrumentation provides an actuation signal to initiate automatic ADS and provides the actuation signal for the IRWST squib valves to open.
De CMTs are risk important for power conditions be Ause the levelindicators in the CMTs provide an open signal to ADS and to the IRWST squib valves as the CMTs empty, ne COL will maintain the reliability of the CMT subsystem. These AOVs are stroke tested quarterly.
g , 720.434F-16 5
NRC FSER OPEN ITEM ,
E# 3=
CMT is required by the Technical Specifications to be available from power conditions down through cold shutdown with RCS pressure boundary intact.
I E' Rerponse: i The staff's statement on CAfTs is con ed b>3 tem Ic of PRA Table 59 29.
In-Containment Refuelins Water Storane Tank (IRWST) l The IRWST subsystem provides a safety related means of performing (1) low-pressure safety injection following ADS actuation, (2) long term core cooling via containment recirculation, and (3) reactor vessel cooling through the Gooding of the reactor cavity by draining the IRWST into the containtnent. He following are some important aspects of the IRWST subsystem as represented in the PRA:
- IRWST subsystem has the following flowpaths:
Two (redundant) injection lines from IRWST to reactor vessel DVI nozzle. Each line is isolated with a parallel set of valves; each set with a check valve in series with a squib valve.
Two (redundant) recirculation lines from the containment to the IRWST injection line. Each recirculation line has two paths: one path contains a squib valve and a MOV, the other path contains a squib valve and a check valve.
ne two MOV/ squib valve lines also provide the capability to flood the reactor cavity.
There are screens for each IRWST injection line and recirculation line which
- Mcy are nou' by debris or other materah generated in the IDWRT nr containment OL Applicant will maintain
@ility of such screens. J
'MA. tuvfd e Explosive (squib) valves provide the pressure boundary and protect the check valves from any potential /m adverse impact of high differential pressures. [ Q gjf He Squib valves and MOVs are powered by Class IE de power and their positions are indicated and alarmed De nt in the control room. /g He squib valves and MOVs for injection and recirculation are automatically and manually actuated via PMS, and manually actuated via DAS.
+
ne squib valves and MOVs for reactor cavity flooding are manually actuated via PMS and DAS from the control room.
Diversity of the squib valves in the injection lines and recirculation lines minimizes the potential for common cause failure between injection and recirculation / reactor cavity flooding.
Automatic IRWST injection at shutdown conditions is provided using PMS low hot leg level logic.
7"7 W westinghouse
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=
- 59. PRA Results and Insights I
Table 59-29 (Sheet 5 of 16) l I AP600 PRA BASED INSIGHTS INSIGHT DISPOSITION l
I Id. (cont.) SSAR 3.9.6 i 1RWST injection and recirculation check valves are exercised at each I refueling. IRWST injection and recirculation squib valve actuators are I tested every 2 years for 20% of the valves. IRWST recirculation MOVs I are stroke. tested quarterly.
SSAR 16.2 ne reliability of the IRWST subsystem is important. HeCO will I
( IP-M T' a- 3 l maintain the reliability of the IRWST subsysteyinetudtn [!'
SSAR 16.1 1 IRWST injection and recirculation are required by Technical Specifications I
to be available from power conditions to refueling without the cavity I flooded.
CMa\ H e s'rt . 68 M Redston: 9 W
Westinghouse fh_ 59 211 April 11,1997
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Q NFIC FSER OPEN ITEM
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1RWST injection and recirculation check valves are exercised at each refueling. IRWST injection and recirculation squib valve actuators are tested every 2 years for 20 percent of the valves. IRWST recircult.: ion MOVs are stroke tested quarterly.
+
ne reliability of the IRWST subsystem is important. De COL will maintain the reliability of the IRWST subsystem.
IRWST injection and recirculation are required by Technical Specifications to be available from power conditions to refueling without the cavity flooded.
s W Response: The stafs statements above on IRWST is covered by item Id of PRA Table 59 29. except
- Westinghouse wishes to note thefollowing change should be made to shat is written above:
- 0 Second bullet remove the work " ensure"for reasons provided earlier in this document. 7
+
An accurate statement would read " .. recirculation line which prevents cloreine bv debris
~ Also note the COL item is covered by a higher level action of the COL wil! maintain the reliability of the IRWST subsystem ISSAR Section 17.4). ,[v -{2 D ll G4 ne IRWST provides a safety.related long term source of water during shutdown conditions. He following are some iditional important aspects of the IRWST subsystem as represented in the shutdown PRA.
[ "(U7 e ration is applican entering th provide administrative controls to e IR T wh' trol trash gene ould possi
-during shutdown e screens.
w bSY Y&
W Response: As stated in SSAR section 13.5, the Combined Dcense applicant is responsible for developing ,
administrative controls. The COL item in SSAR chapter 13 covers the stafs statement at a higher $.3. ' g level.
On low hot leg level, the PMS actuates the squib valves to open allowing gravity injection from the IRWST.
i s w Response: This statement is a duplicate of the 8th bullet on IRWST(see above). O Passive Residual Heat Removal (PRHR) System De PRHR provides a safety related means of performing the following functions: (1) removes core decay heat ;
during accidents, (2) allows adequate plant performance during transient (non-LOCA and non-ATWS) accidents
{
without ADS,(3) allows automatic termination of RCS leak during a SGTR accident without ADS, and (4) provides I core cooling and pressure control during the early phase of an A'IWS accident.
E Response: For item (2), recommend changing the word " allows" to "provides." Item (4) is ambiguous by using the words early phase of an A7WS. The phrase should read, " allows plant to ride out an !
A7WS event without rod insertion." 1 \
WO C n 7 l k % a% i g .g9 720.434F-18 W-Westinghouse 8
NRC FSER OPEN ITEM *
- Es ei The following important aspects of the PRHR design and operation features are incorporated in the PRA models:
+
PRHR is actuated by opening redundant parallel air-operated valves (AOVs). Dese AOVs are designed to fail open on loss of Clus lE power, loss of air, or loss of signal from the protection and safety monitoring system (PMS).
+ The PRHR AOVs are automatically actuated by two redundant and diverse I&C systems: (1) the safety-related protection and safety monitoring system (PMS) and (2) the nonsafety-related diverse actuation system (DAS). The PRHR can also be actuated manually from the control room using either PMS or DAS.
Diversity of the PRHR AOVs from the AOVs in the core makeup tanks (CMTs) minimizes the probability for common cause failure of both PRHR and CMT AOVs.
-+ ne positions of the inlet and outlet PRHR valves are indicated and alarmed in the MCR.
' E Response: The stafs above statements on PRHR are covered PRA Table 59-29.
+ ne PRHR AOVs and isolation MOV are tested quarterly. The PRHR HX is flow tested at shutdown.
E Response: It is true the PRHR AOVs are tested quarterly, per IST(SSAR subsection 3.9.6). As stated in the PRA and SSAR, the MOV is closed to test the AOVs, so indirectly, the MOV is also tested; i
& however, the MOVis not spectped as such per ISTand the PRA. The words "and isolation MOV" should be removedfrom the stafs statement to be technically accurate. It is accurate to say the h w 0.f PRHR HX is pow tested (as is stated by item le in PRA Table 59 29), but it is misleading to say '
it is tested at shutdown. The HX is pow tested at shutdown, but not every time the plant is shutdown. Per Technical Specspcation, the PRHR HX is pow tested every 10 years. It is not an 9
insightfrom the PRA to include this level ofdetail(thepow testfrequency). The recommendation is the stufs bullet above be changed to what is provided by item le in PRA Table 59 29. 0.G y
NO
+
Use of the PRHR heat exchanger (HX) for long-term cooling causes the IRWST water to heat up, resul ' g in inventory loss through evaporation. To ensure successful long-term cooling by the PRHR HX, the evaporated IRWST inventory must return to the IRWST after condensed on the containment liner and collected in the IRWST gutter system. De IRWST gutter system, which directs the water to the containment sump during normal plant operation, is automatically re-aligned to direct the water back to the IRWST during an accident. De following design features ensure proper re-alignment of the gutter system valves to direct water to the IRWST during accidents:
the IRWST gutter and its isolation valves are safety grade
- the valves that re-direct the flow are designed to fail-safe on loss of compressed air, loss of Class IE DC power, or loss of the PMS signal.
- the isolation vPves are actuated automatically by PMS and DAS.
W Westinghouse
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NRC FSEFI OPEN ITEM .
E Response: The stafs statement should be reworded asfollows, to be technically accurate. Note the statement below is consistent with SSAR subsection 6121.1. y "The PRHR HX, in conjunction with the PCS, can provide core cooling for an indefinite period of time. After the IRWST water reaches its saturation temperature, the process of \
)p steaming to the containment initiates. Condensation occurs on the steel containment vessel, and the condensate is collected in a safety related gutter arrangement which V f returns the condensate to the IRWST. The gutter normally drains to the containment Y sump, but when the PRHR HX actuates, safety related isolation valves in the gutter drain
\l. line shut and the gutter overflow returns directly to the IRWST. The following design gs features provide proper re alignmentfo the gutter system valves to direct water to t IRWST:" ,
i
'the stafs three sub-bullets above are accurate, except change the word " safety-grade" to " safety- l
^
related" and " fail safe" to " fail closed."
. Use of the PRHR HX for long-term cooling will result in steaming to the containment. The steam will normally condense on the containment shell and return to the IRWST via the gutter system. If the condensate does not return to the IRWST, the IRWST volume is sufficient for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of PRHR operation.
Connections to the IRWST are provided from the spent fuel system (SFS) and chemical and volume control system (CVS) to extend PRHR operation. A safety-related makeup connection is also provided from outside )
the containment through the normal residual heat removal system (RNS) to the IRWST. )
4 E Response: This is an accurate statement. - coe-Ht cb lL kfk YW *
- Capability exists in the control room to identify a leak in the PRHR HX which could degrade to a tube rupture under the stress conditions, such as RCS pressure increase and temperature gradients inside the HX '
tube walls, likely to occur during a postulated accident requiring PRHR operation.
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E Response: Recommend the stafs statement stop after the words " tube rupture". By continuing with the . $l< i specifics of tying this to a transient, it deminishes the leak tightness capability. Note the statement W gfI j will be consistent with PRA Tahir 59 29 item Is by ending the sentence as recommended. Also O note the operator guid$e is provided via Technical Specification 3.4.8. Y Technical Specifications require P" :: Smailable, with RCS boundary intact, from power condh .
MCL down through cold shutdown.-
Qwhich couid degradntr a tu ce is providedjor operator acuon wnen a leak is detected in the PRHR rupture cunng normal power operauon conditions or nder stress M %V d, conditions, such as RCS pressure i rease and temperature gradients inside the HX tube walls,li ly to occur g N, l during a postulated accident r iring PRHR operation. W I
' $ Response: The first sentenc is an accurate statement. The second sentence is essent' a repeat of the previous bullet. Recommend the second sentence be deleted.
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-r Table 59 29 (Sheet 6 of 16) l I
I AP600 PRA BASED [NSIGHTS DISPOSITION INSIGHT I
SSAR 6.3.1 &
I le. Passive residual heat removal (PRHR) provides a safety-related means of6.3.3 I performing the following functions:
l . Removes core decay heat during accidents Allows automatic termination of RCS leak during a steam generator tube I rupture (SGTR) without ADS.
l
'Ihe following are some important aspects of the PRHR subsystem as represented l in the PRA:
SSAR 6.3.2 PRHR is actuated by opening redundant parallel air-operated valves. These 1
l air operated valves open on loss of Class IE power, loss of air, or loss of I se signal from PMS.
Certified Design The PRHR air-operated valves are automatically actuated and manuallyMaterial l
actuated from the control room by either PMS or DAS.
I SSAR 6.3.2 l
Diversity of the PRHR air-operated valves from the CMT ait-operated valves minimizes the probability for common cause failure of both PRHR and CMT I
I air operated valves.
SSAR 6.3.1 &
l Long term cooling of PRHR will result in steaming to the containment. 'Ihe system drawings steam will normally condense on the containment shell and return to the I
IRWST. If the steam condensation does not return to the IRWST, the IRWST volume is sufficient for at least 72 hows of PRHR operation.
l Connections are provided to IRWST from the spent fuel system (SFS) and I
chemical and volume control system (CVS) to extend PRHR operation. A I
l safety related makeup connection is also provided from outside the containment through the normal residual heat removal system (RNS) to the I
l Capability exist 4for th control room ope tot to identify a leak in the16.1 PRHR HX before it can degrade to a tube rupture during a subsequent 1
I design basis accident (DBA).
SSAR 6.3.7 I
The positions of the inlet and outlet PRHR valves are indicated and alarmed I in the control room.
SSAR 3.9.6 1
PRHR air operated valves are stroke tested quarterly. The PRHR HX is flow tested to detect system performance degradation.
I SSAR 16.1 PRHR is required by the Technical Specifications to be available from l I
l power conditions down through cold shutdown with RCS pressure boundary J 1 intact. \
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De PRHRfs stem provides a safhy-r ated means of removing decay heat following loss of Abu
'R N 5 h
cooling dunng safe / cold shutdown wrtfrthe RCS intact. 4e
.W Response: Change the nords " shutdown cooling" to "RNS cooling" This is an accurate statement and is MA S1Qq covered by Technical Specification bases 3.5.5.. &Q, Automatic Deorenprization System (ADS) DIS 0*
ADS provides a safety-related means of depressurizing the RCS. The following are some important aspects of ADS as represented in the PRA:
ADS has four stages. Each stage is arranged into two separate groups of valves and lines. Stages 1,2, and 3 discharge from the top of the pressurizer to the IRWST. Stage 4 discharges from the hot leg to the RCS loop compartment.
+
Each stage 1,2, and 3 line contains two MOVs in series. Each stage 4 line contains an MOV valve and a squib valve in series.
De valve arrangement and positioning for each stage is designed to reduce spurious actuation of ADS.
Stage 1,2, and 3 MOVs are normally closed and have separate controls.
Each stage 4 squib valve has redundant, series controllers.
Stage 4 is blocked from opening at high RCS pressures.
- The ADS valves are automatically and manually actuated via the protection and safety monitoring system (PMS), and manually actuated via the diverse actuation system (DAS).
- De ADS valves are powered from Class IE de power and their positions are indicated and alarmed in the control room.
Stage 1,2, and 3 valves are stroke-tested every 6 months. Note: Westinghouse has indicated that this requirement may chan;c as a result of an NRC review. Stage 4 squib valve actuators are tested every 2 years for 20 percent of the valves.
ADS is required by the Technical Specifications to be available from power conditions down through I refueling without the cavity flooded.
- Depressurization of the RCS through ADS minimizes the potential for high-pressure melt ejection events.
Procedures will be provided for use of the ADS for depressurization of the RCS during a severe accident.
mm-21 T Westinghouse
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W Response: The stafs above statements on ADS are covered by item Ib of PRA Table 59-29. Notefor the 6th q bullet, as a result of NRC review, the stage 1, 2, and 3 valves are now stroke tested every cold
.9 f shutdown. With the number of cold shutdowns and refuelings assumed in the shutdown PRA, the test frequency is equivalent to being tested esery 6 months. PRA Table 59-29 will be revised 1 appropriately. Notefor the 9th bullet, the wording "during a severe accident" should be changed to "after core uncovery."
. Fire-induced hot shorts, especially in I&C copper cables from the protecdon logic cabinets to the squib valve operators, could cause detonation of a squib valve. His risk importaat concern should be addressed by -
appropriate power and control cable separation and routing and by the incorporation of features and requirements in the detailed design of ADS cabling. -
l 150 W Response: 2" =:: v:- ~ the words of the stafs statement be changed to read as describe S$bsection 9A.2. 1 pecifica , Spurious "
actuation of squib valves is prevented by th s j
\ , of a sauib valvo enntroller'circui which reauires multiole hot shorts for actuatiore physical f separation from thefire zone."of potential Note as stated in hot shortPRA the ternalfire locations, analysis, it isand orovisions conservatively modeledfor operator in the PRA analysis that one hot short can ause spurious ADS squib valve actua whereas, per design, multiple hot shorts are required. (f y , & , )Mut gj A.
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He first, second, and third stage valves, connected to the top of the pressurizer, provide a vent p l preclude pressurization of the RCS during shutdown conditionsif derny heat removal is los)-Cuc ivud- 7g s age ADS valve is required to open if gravity injection is actuated during cold shutdown and refueling with g )
p- the RCS is open to preclude surge line flooding. On low-low hot leg level (empty hot leg), the PMS signals the ADS 4th stage squibs to open to preclude surge line flooding.
vy gg ' yM lE Response: This is an accurate statement. A statement will be added to PRA Table 59 29./
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Normal Residual Heat Removal System (RNS)
)
I The normal residual heat removal system (RNS) provides the following nonsafety-related means of core cooling during accidents: (1) RCS recirculation at shutdown conditions,(2) low pressure pumped injection from the IRWST, and (3)long term pumped recirculation from the containment sump. Such RNS functions provide defense in-depth in mitigating accidents, in addition to that provided by the passive safety related systems.
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W Response: nis is an accurate statement. The statement is cove item 6 of PRA Table 59-29.h The following are some important aspects of RNS as represented in the PRA:
De RNS has redundant pumps, powered by separate non-Class lE buses with backup connections from the diesel generators, and redundant heat exchangers.
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& Response: nis is an accurate statement and is covered item 6 of PRA Table 59 29. o Ya l 1
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De RNS provides safety-related means for (1) containment isolation at the penetration of the RNS lines,(2)
RCS isolation at the RNS suction and discharge lines, and (3) IRWST and containment sump inventory
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makeup.
E Response: ficeptforpojuD), the above is an accurate statement and covered by item 6 of PHA Table 59-
- 29. Item (3) is incorrect. RNS does not provide a safety-related means. but rather a defense-in f depth function ofIRWST and containment sump inventory makeup. \.
- Th igned trom tne conuoi mum ;e F& m in core cooling functions (SSAR ,
Ernergency Response Guidelines (ERGS) are provided for aligning the RNS from the control room for RLs
, injection and recirculation.
W Response: This is an accurate statement. h. l$f,, y h Recirculation from the containment sump is actuated automatically by a low IRWST level signal or manually from the control room, if automatic actuation fails.
j m E Response: This statement is misicading as worded. It should read "PXS recirculation valves are_
automatically actuated _. .* It is believed the stagwas intending to mean the IRWST recirculation valves rather than an RNS recirculation (i.e., pumps stop, start) as could be interpreted by the @"
WI statement. Note that sf RNS is operating, the RNS pumps will continue to operate and provide containment recirculation. ware c, q y For long-term recirculation operation, the RNS pumps tc.ke suction from onfy one of the two sump (gcid)'
recirculation lines. Unrestric:ed flow through both parallel paths (one containing an MOV and a squib valve in series, the other containing a check valve and a squib valve in series)is required for success of the sump recirculation function when both RNS pumps are running. If one of the two parallel paths fails to open, g*
operator action (in the control room through PMS) is required to manually throttle the RNS discharge MOV (V0ll) to prevent pump cavitation. [ ERGS). I I
- 9 W Response
- nis is an accurate statement per the PRA.
. With the umps aligned either to the IRWST or the containment sump, the pumps' net positive suction head (NPSH) is adequate to prevent pump cavitation and failure even when the IRWST or sump inventory h ]
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is saturated. !
l
>p W Response: Change NRHR to RNS. nis above is an accurate statement.
i e
ne RNS containment isolation and RCS pressure boundary valves are safety related. He MOVs are powered by Class IE de power, m
JW Response: nis is consiste with item 6 of PRA Table 59-2% Q1 ,
W westinghouse
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l Table 59 29 (Sheet 13 of 16) l I AP600 PRA BASED LNSIGHTS INSIGHT DISPOSITION I
l 6. De noanal residual heat removal system (RNS) provides a safety-related means Certified Designj Material l of performing the following functions:
l - Containment isolation for the RNS lines that penetrate the containment I - Isolation of the reactor coolant system at the RNS tuction and discharge l lines l
! AMakeup of containment inventory. l L n3-4fr m , Pet-at, Red SSAR 5.4.7 )
l RNS'provides a nonsafery related means of core cooling through: '
1 - RCS recirculation at shutdown conditions l - Low pressure pumped injection from the IRWST and long term pumped j I recirculation from the containment. (
The RNS has redundant pumps and heat exchangers. The pumps are powered SSAR 5.4.7 &
l 8.3 1 by non-Class IE power with backup connections from the diesel generators.
SSAR 5.4.7 i RNS is manually aligned from the control room to perform its core cooling I functions. The performance of the RNS is indicated in the control room.
Cerufied Design I The RNS containment isolation and pressure boundary valves are safety-related.
Material l The motor-operated valves are powered by Class IE de power.
The containment isolauon valves in the RNS piping automatically close via SSAR 7.3.1 l
l PMS with a high radiation signal.
De RNS containment isolation MOVs are automatically and manually actuated SSAR 7.3.1 I
I via PMS.
f SSAR 5.4.7.2.2 i I interfacing system loss of coolant accident (LOCA) between the RNS and the I RCS is prevented by:
I - Each RNS line is isolated by at least three valves.
1 - The RNS equipment ouuide containment is capable of withstanding the 1 operating pressure of the RCS.
1 - 7he RCS isolation valves are interlocked to prevent their opening at RCS I pressures above its design pressure.
Certified Design I CCS provides cooling to the RNS heat exchanger.
Material I
i Planned maintenance of the RNS is performed at-power.
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The containmen olation valves in the RNS piping close automatically via PMS th a high radiation signal.
Wettingm~ n@c;inic= +ounAre att accident enndWn: Nilmse Mdthe containment radiation level is low the point that would cause the RNS MOVs to automatically cI6se.
WRespoaq{, &r us is consistent with item 6 of PRA Table 59 29. The second sentence tends to lead ihfjt.gjenience beyond an insightfrom the PRA. However, if the staff explains why it considers this an insight, then Westinghouse recommends the second sentence be reworded to read: The actuation setpoint was established consistent with a DBA non mechanistic source term associated with a large LOC *" i
- The following AP600 design features ccntribute to the low likelihood of interfacing system LOCAs through I the NRHR system:
p W'f -
1 The portion of the RNS outside containment is capable of withstanding the operating pressure of the p-RCS. \
A relief valve located in the common RNS discharge line outside containment provides protection against excess pressure.
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- Each RNS line is isolated by at least three valves y w eet ' b k He pressure in the RNS pump suction line is continuously indicated and alarmed in the main control ,
room. g( /
c ne pump suction isolation valves connecting the RNS pumps to the RCS hot leg are interlocked w
- h RCS pressure so that they cannot be opened until the RCS pressure is less than 450 psig. His preve overpressurization of the RCS when the RNS is aligned for shutdown cooling. M (y 0
The two remotely operated MOVs connecting the suction and discharge headers, respectively, to th IRWST are interlocked with the isolation valves connecting the RNS pumps to the hot leg. His prevents inadvertent opening of any of these two MOVs when the RNS is aligned for shutdown cooling A@
and potential diversion and draining of reactor coolant system.
De power to the four isolation MOVs connecting the RNS pumps to the RCS hot leg is admir:istratively blocked at their motor conttol centers during normal power operation. [ COL].
y e The operability of the RNS is tested, via connections to the IRWST, immediately before its alignment p c( .
to the RCS hot leg, for shutdown cooling,t: m = iei L.e e r.c eay cFa m.ax.' eatves in the n eairt-lines-13SAR, COL, FicEidresF---t mm24 W westinghouse 17
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h 2 Response: Westinghouse has the following commentsfor the stqff's above statement:
Change fR system to "Rab cX Second sub tullet u a hu au...w , but notfa d into the PRA and is not a key to providing a low likelihood ofinterfacing systems LOCA. Thus, Westinghouse does not see this as an important statement to include as an insight.
Last sub bullet: It is true that the system is tested; however, it is done to test operabilQ gtj+
of the system, not solely to minimize potentialfor interfacing syetems LOCA or to detect ad),
an open valve in ths drain lines. However, the testing does have this end result efect. n e t* g.
The words should be revised appropriately.
He IRWST suction isolation valve (V nd the RCS pressure boundary isolation valves (V001 A, V00k M V002A and V002B) are qualified f r DBA. onditions.
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- 3 g Response: It is not understood why the staf's statement is an insightfrom the PRA.
6 c.@d b, 3
The reliability of the IRWST suction isolation valve (V023) to open on demand (for RNS injection during .
,p , y power operation and for IRWST gravityiniection via the RNS hot leg connection during shutdown operatien) utc/d is importantQCOt~will ensure high reliabilitMCOL, D-RAP). "
Y,NY % b b 'r Y}L if h W Response: This item is acceptable and is covered by SSAR section 17.4 fRAP). M ,J, .
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An alternative gravity injection path is provided through RNS V-023 during cold shutdown and refueling conditions with the RCS open. He COL applicant should have policies that maximize the availability of this l P d'*fk valve and procedures to open this valve during cold shutdown and refueling operations when the RCS is open.
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W Response: The ERGS cover the operation of the valve. In addition, as stated in SSAR section 13.5, it is the Q responsibility of the Combined license applicant to develop procedures.
ne COL applicant will maintain RNS and its support systems (CCS and SWS) during power operation.
W Response: To be accurate and consistent whAR see 16.3 2, item 2.2), change the statement to read: " Planned maintenance afecting the RNS cooling function and_ its support yhguld be pedormed in Modes I, 2, 3 when the RNS is not no&ating."
He COL applicant will have administrative controls to maximize the likelihood that RNS valve V 023 Elh be able to open if needed during Mode 5 when the RCS is open, and PRHR ca W Response: As stated in SSAR section 13.5, it is the responsibility of the Combined License applicant to ($5 develop administrative procedures.
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%M f 1 nadvertent opening of RNS vahe V024 results in a draindo n of RCS inventory to the IRWST and requires gravity injection from the IRWST, e COL applicant *" hm administrative controls to ensure that gg -/
inadvertent opening of this valve is unlikel . 5 d":ica, S; CO!reppb %uM edealethisenorinthe ^'Q
-boman.seliability analysis /humanfactorifinjiheeringintegrationimplementation plan. Ed,,
W Response:
mwd SP.
As stated inhR sec1 ion.11 L.it is the responsibility of the Combined Ucense applicant to develop administrative procedures.
- De RNS is an important " defense-in-depth" system for accidents initiated while the plant is at power or at mid-loop during shutdown. De availability control of the RNS and its support systems (CCW, SWS and diesel generators) is covered in SSAR Section 16.3. [RTNSS]. [(Q g
W Response: Ths reason RNS is important while the plant is at power is not because it is important per the PRA V results or importance listings, but rather because it provides marginfor long term cooling T&H uncertainty. Otherwise, the stafs statement is accurate.
q
. P Startup Feedwater System (SFW) (,
ne SFW' system provides a nonsafety related means of delivering feedwater to the steam generators (SGs) whe the m'ain feedwater pumps are unavailable during an transient. His capability provides an alternate core cooling mechanism to the PRHR heat exchanger for non LOCA and SGTR accidents which minimizes the PRHR challenge l rate. He reliability of the SFW system will be maintained by the COL Applicant [D-RAP).
he W Response: The stafs statement is essentially taken directlyfrom the SS note the words should read startupfeedwater system pJL"lR.,t.
Table .4 (RAD be accurate.
rationale provided in this table W h
for why the startup feedwater pumps are included is based on the Expert Panel, not PRA.
Therefore, it is not clear why the stafs statement is considered an insightfrom the PRA. y N Instrumentation and Control (I&C)
- O ne following three I&C systems are credited in the PRA for providing monitoring and control functions during ,
accidents: (1) the safety-related Protection and Safety Monitoring System (PMS),(2) the nonsafety-related Diverse M Actuation System (DAS), and (3) the nonsafety-related Plant Control System (PLS).
{
De PMS provides a safety related means of performing the following functions:
Automatic and manual reactor trip.
=
Automatic and manual actuation of engineered safety features (ESF).
+
Monitor the safety-related functions during and following an accident as requiredhtRegulatory Guide 1.97.
m W Response: The stafs statements on PMS are cove d by item 2 of _PRA Table 59 29.) O m m 26 W Westinghouse M ,
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= g The DAS provides a nonsafety-related means of performing the following functions:
- Automatic and manual reactor trip.
. Automatic and manual actuation of selected engineered safety features.
. Provides control room indication for monitoring of selected safety-related functions.
E Response: The stafs statements on DAS are cove d by item 3 of PKA Table 59- &
ne PLS provides a nonsafety-related means of performing the followililrf-A=.
Automatic and manual control of nonsafety-related systems, including " defense-in-depth" systems (e g., RNS).
3
. Provides control room indication for monitoring overall plant and nonsafety related system performance.
W Response: SSAR subsection 7.1.1 support the stafs statements on PLS: however, on thefirst bullet, the word 's
" systems" should be changed to " functions."
he following are some important aspects of PMS as represented in the PRA:
. He PMS has four (redundant) divisions of reactor trip and ESF actuation and automatically produces a reactor trip or ESF initiation upon an attempt to bypass more than two channels of a function that uses 2-out-of-4 logic.
. He PMS has redundant divisions of safety-related post-accident parameter display.
. Each PMS division is powered from its respective Class IE de division.
. He PMS provides fixed position controls in the control room.
. He reliability of the PMS is ensured by redundancy and functional diversity within each division:
- The reactor trip functions are divided into two functionally diverse subsystems.
- He ESF functions are processed by two microprocessor based subsystems that are functionally identical in both hardware and software.
. Separate input channels are provided for the reactor trip and the ESF actuation functions, with the exception of sensors which may be shared.
. Sensor redundancy and diversity contribute to the reliability of PMS. Four sensors normally monitor variables used for an ESF actuation. Different type sensors, or same type sensors in different environment, minimize common cause failures.
720.434F-27 M@m c2O
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, sus u.!a W a; Continuous automatic PMS system monitoring and failure detection / alarm is provided.
- PMS equipment is designed to accommodate a loss of the normal heating, venulation, and air conditioning (11VAC). PMS equipment is protected by the passive heat sinks upon failure or degradation of the active liVAC.
- De PMS software is designed, tested, and maintained to be reliable under a controlled verification and validation program written in accordance with IEEE 7 4.3.2 (1993) that has been endorsed by Regulatory Guide 1.152. Elements that contribute to a reliable software design include:
- A formalized development, modification, and acceptance process in accordance with an approved software QA plan (paraphrased from IEEE standard, Section 5.3, " Quality")
- A verification and validation program prepared to confirm the design implemented will function as required (IEEE standard, Section 5.3.4, " Verification and Validation")
- Equipment qualification testing performed to demonstrate that the system will function as required in the environment it is intended to be installed in (IEEE standard, Section 5.4," Equipment Qualification")
- Design for system integrity (performing its intended safety function) when subjected to all conditions, external or internal, that have significant potential for defeating the safety function (abnormal conditions and events) (IEEE standard, Section 5.5. " System Integrity")
- Software configuration management process (IEEE standard, Section 5.3.5, " Software Configuration Management"). % 6 1Y Response: The stafs above statements on. PMS are covere by item 2 ofPRA Table 59-29. exceptfor the 7th bullet. Westinghouse does not claim specifically written as the-thrrtifntence of the stafs g .
7th bullet. Rathsr,Qnctional diversity minimizes the common causefailure among sensors. g ne following are some important aspects of DAS as represented in the PRA:
Diversity is assumed in the PRA that eliminates the potential for common cause failures between PMS and DAS. De DAS automatic actuation signals are generated in a functionally diverse manner from the PMS signals. Diversity between the DAS and PMS is achieved by the use of different architecture, different hardware implementations, and different software.
DAS provides control room displays and fixed position controls to allow the operators to take manual actions.
DAS actuates using 2-out-of 2 logic. Actuation signals are output to the loads in the form of normally de-energized, energize.to-actuate signals. De normally de-energized output state, along with the dual 2-out-of.2 redundancy, reduces the probability of inadvenent actuation.
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NRC FSER OPEN ITEM .
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_ m a ne actuation devices of DAS and PMS are capable of independent operation that is not affected by the operation of the other. De DAS is designed to actuate components only in a manner that initiates the safety function.
. Capability is orovided for on line testing and calibration of the DAS channels, including sensors.
. De DAS manual initiation functions are implemented in a manner that bypasses the signal processing equipment of the DAS automatic logic. This eliminates the potential for common cause failures between automatic and manual DAS functions.
. De DAS reacts tri,n knetinn it imntemented thmuch a trip of the control rods via the motor generator (M.
G) set which6eparate and diverse from the reactor trip bhne 3 COL will maintain the reliability of the M-G set breakers [D-RAP]. \v d e pgp(> l,,c/d;, -ft/6 SP PR
. h' DAS is an important " defense in-depth" system. The availability of DAS, with respect to both its reactor trip v and ESF actuation functions, will be controlled. [RTNSS]. The COL will maintain its reliability (D-RAP).
E Response: ne stafs above statements on DAS are covered em 3 ofPRA Table 59-2 , xceptfor the 5th bullet, which is supported by SSAR subsection 7.7.1.1IN.
De following are some important aspects of PLS as represented in the PRA:
C@f j,W ,
. PLS has redundancy to minimize plant transients.
. PLS provides capability for both automatic control and manual control.
. Redundant signal selectors provide PLS with the ability to obtain inputs from the integrated protection cabinets in the PMS. De signal selector function maintains the independence of the PLS and PMS. De signal selectors select those protection system t.igncis that represent the actual status of the plant and reject erroneous signals.
. PLS control functions are distributed across multiple distributed controllers so that single failures within a controller do not degrade the performance of control functions performej.bnum o llers.
m W Response: The stafs statements on PLS are cove y item 4 of PRA Table 59-29. &
Onsite Power ne onsite power system consists of the main ac power system and the de power system. De main ac power system is a non-Class IE system. De de power system consists of two independent systems: the Class IE de system and the non Class IE de system. _
P E Response: ne stafs statement is covered item Sa of PRA Table 59-29.
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- 59. PRA Results and Insights
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- I Table 59-29 (Sheet 9 of 16) l I AP600 PRA BASED INSIGHTS i INSIGHT DISPOSITION De diverse actuation system (DAS) provides a nonsafety-related means of Certified Design I 3.
Material I performing the following functions:
l - Initiates automade and manual reactor trip I - Automatic and manual actuadon of selected engineered safety features.
I Diversity is assumed in the PRA that climinates the pot:ndal for common cause I failures between PMS and DAS.
De DAS automatic actuation signals are generated in a functionally diverse Certified Design 1 -
manner from the PMS signals. Diversity between DAS and PMS is Material I
I achieved by the use of different architecture, different hardware l implementadons, and different software.
DAS provides control room displays and fixed position controls to allow the SSAR 7.7.1 1 I
! operators to take manual actions.
j i
DAS actuates using 2-out of 2 logic. Actuation signals are output to the loads SSAR 7.7.1.11 l
I in the form of normally de-energized, energize to-1ctuate signals. The normally I de-energired output state, along with the dual 2 out of 2 redundancy, reduces i the probability of inadvertent actuation.
De actuation devices of DAS and PMS are capable of independent operation SSAR 7.7.1.11 I
l that is not affected by the operation of the other. He DAS is designed to l actuate components only in a manner that initiates the safety function.
he DAS reactor trip function is to trip the control rods via the motor generator SSAR 7.7.1.11 I
I set.
I in th'e PRA it is assumed the following eliminates the potential for common I cause failures between automatic and manual DAS functions.
DAS manual initiation functions are implemented in a manner that bypam Certified Design 1 -
the signal processing equipment of the DAS automatic logic. Material I
l ;
The COL will maintain the reliability of the DASjkchMN Af6 Sd SSAR 16.2 o
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ir g The main ac power system is a non-Class IE system comprised of a normal, preferred, and standby power system:
It distributes power to the reactor, turbine. and balance of plant auxiliary electrical loads for startup, normal operation, and normal / emergency shutdown.
W Response: The stofs statement is covered by i _m Sa of PRA Tghlt,12:29. N The Class IE de and uninterrupdble power supply (UPSh-(L vi es reliable power for the safety-related equipment required for the plant instrumentation, control, monitoring, and other vital functions needed for shutdown of the plant.
W Response: y item $b of PRA TNie 59-h 4 flN The stafs statement is cover ne non-Class IE de and UPS system (EDS) cons m J e upply and distribution equipment that provide de and uninterruptible ac power to nonsafety-rel W Response: The stafs statement is covered y itemn.
$c of PRA Table 59-29.
O < O' He following are some important aspects of the main t puwei system as represented in the PRA:
Re arrangement of the buses permits feeding functionally redundant pumps or groups ofloads from separate buses and enhances the plant operational reliability.
During power generation mode, the turbine generator normally supplies electric power to the plant auxiliary loads through the unit auxiliary transformers. During plant startup, shutdown, and maintenance, the main ac l
power is provided by the preferred power supply from the high-voltage switchyard. De onsite standby power >
system powered by the two onsite standby diesel generators supplies power to selected loads in the event of loss of normal and preferred ac power supplies.
Two onsite standby diesel generator units, each furnished with its own support subsystems, provide power to the selected plant nonsafety-related ac loads.
On loss of power to a 4160 V diesel-backed bus, the associated diesel generator automatically starts and produces ac power. He normal source circuit breaker and bus load circuit breakers are opened, and the generator is connected to the bus. Each generator has an automatic load sequencer to enable controlled loading on the associated buses.
W Response: The stafs statements on main ac power are cov ed by item $a of PRA Table 59 2f D
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=- n De following are some important aspects of the Class IE de and UPS system (IDS) as represented in the PRA:
1 There are four independent Class lE 125 V de divisions. Divisions A and D cach consists of one battery bank, one switchboard, and one battery charger. Divisions B and C are each composed of two battery banks, two switchboards, and two battery chargers, ne first battery bank in the four divisions is designated as the 24-hour battery bank. He second battery bank in Divisions B and C is designated as the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> battery bank.
De 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery banks provide power to the loads required for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an event of less of all ac power sources concurrent with a design basis accident. De 72-hour battery banks provide I power to those loads requiring power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the same event-Battery chargers are connected to de switchboard buses. De input ac power for the Class IE de battery chargers is supplied from non-Class IE 480 V ac diesel-generator-backed motor control centers.
1 ne 24 hout and 72-hour battery banks are housed in ventilated rooms apart from chargers and distnbution equipment.
Each of the four divisions of de systems are electrically isolated and physically separated to prevent an event from causing the loss of more than one division.
Reliability of the Class IE batteries is important. The COL will maintain the reliability of the equipment.
}y Response: llte stafs statements on Class 1E de power are cover d by item $b of PRA Table 39 29.
}
The following are some important aspects of the non-Class lE de and UPS systern as represented in the PRA:
+
ne non Class IE de and UPS system consists of two subsystems representing two separate power supply trains.
EDS load groups 1,2, and 3 provide 125 V de power to the associated inverter units that supply the ac power to the non Class IE uninterruptible power supply ac system.
The onsite standby diesel generator-backed 480 V ac distribution system provides the normal ac power to the battery chargers.
The batteries are sized to supply the system loads for a period of at least two hours after loss of all ac power j sources. '
W Response: The stafs statements on non Class 1E dc power are cove d by item Se of RA Table 39-2 ht.CW
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- E N .t Comoonent Cooline Water System (CCS)
The component cooling water system (CCS)is a nonsafety-related system that removes heat from variou; components and transfers the heat to the service water system. The following are some important aspects of the CCS as repre-sented in the PRA:
a ne CCS is arranged into two trains. Each train includes one pump and one heat exchanger.
- During normal operation, one CCS purap is operating. He standby pump is aligned to automatically start in case of a failure of the operating CCS pump.
+ ne CCS pumps are automatically loaded on the standby diesel generator in the event of a loss of normal ac power. The CCS therefore, continues to provide cooling of required components if normal ac power is lost. 7-
}X Response: The stafs statements on CCS are cove d by item 7 of PRA Table 59-29. ON Service Water System (SWS) ne service water system (SWS)is a nonsafety-related system that transfers heat from the component cooling water heat exchangers to the atmosphere. He following are some important aspects of the SWS as represented in the PRA:
+ ne SWS is arranged into two trains. Each train includes one pump, one strainer, and one cooling tower cell.
- During normal operation, one SWS train of equipment is operating. De standby train is aligned to automatically start in case of a failure of the operating SWS pump.
. De SWS pumps and cooling tower fans are automatically loaded onto their associated diesel bus in the event of a loss of normal ac power. Both pumps and cooling tower fans automatically start after power from the diesel generator is available.
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}Y Response: The stafs statements on SWS are cov ed by item 8ff PRA Table $9 29.
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M..i.s*EE Chemical and Volume Control System (CVS) Tp JA Ae6 f ne chemical and volume control system (CVS) provides a safety related means to terminate inadvertent RCS boron dilution. In addition, the CVS provides a nonsafety related means to (1) provide makeup water to the RCS during g normal plant operation,(2) provide boration following a failure of reactor trip,(3) provide coolant to the pressuriz Q auxiliary spray line, (4) safety related portions of the CVS provide inadvertent bo_ron diliition prm-crinn andWry
-N-A PG S CVS pim,Jgolation of normal CVS letdown during shutdown operation on low hot gh(,sd:y 3.x leg level. _
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W Response: The stafs above statement on CVS is covered by item 9 of PRA Table S9-29 with supportfrom f& >
SSAR subsection 9.36. Note the second sentence begins by discussing nonsafety related means, but items (4) and (3) state safety-related portions. It could be a confusing sentence. Also note, item (4) is a repeat of theprst sentence.
De following are some important aspects of CVS as represented in the PRA:
- He CVS has two makeup pumps and each pump is capable of providing normal makeup.
W Response: This statement is cover d by item 9 of PRA ble 6 One CVS pump is configured to operate on demand while the other CVS pump is in standby. He operation by -[o of these pumps will alternate periodicallyp x.JJ,). - I M < Ny .
(Ntr W Response: The stafs statement is accurate per PRA assumptions. The prst sentence is true. The second 1 sentence's monthly statement is an assumption of the PRA: however, good operating practices would callfor the COL to periodically alternate the pumps.
- On low hot leg level, the safehlated PMS signals4hated safe CVS AOVs to close automatically I to isolate letdown during Mode 4 (Pe, RNS is in operation), b 5, and Mode 6 (with the upper internals in place and the refueling cavity less than half full) as required by AP600 TS. M IN
& J W Response: Only two of the AOVs are safety-related, the third is nonsafery-related. Exceptfor this error, th above statement is true per the ESF Technical Specipcation.
,y ff D (4)0 Re safe @ elated PMS boron dilution signal automatically re aligns CVS pump suction to the boric acid tan his same signal also closes the two safety-related CVS demineralized water supply valves. His signal actuates on any reactor trip signal, source range flux multiplication signal, low input voltage to the Class IE ,
DC power system battery chargers, or a safety injection signal. 7 W Response: This is an accurate statement.
hl The COL applicant will maintain procedures to respond to low hot leg level alarms.
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E Response: The shutdown ERGS cover the procedure to respond to low hot leg level a rms.
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FAX COVER SHEET Recipient Information Sender Information Date: 2/23/98 Name: D. Lindgren To: Bill Huffman Location: WEC E 330 Phone: Phone: (412) 374 4856 (301) 415-1141 Facimile:
(310) 415-2002 Facimile: Bet (412) 374 4887 Company: U.S.NRC WIN 284-4887 Location: Rockville, MD Cover + Pages 1+0 This page is being sent from the Westinghouse Energy Center, Monroe'/ille, PA. If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374-6529.
Comments:
In response to a request by the NRC Staff, a description of the shield building annulus drains will be included in the SSAR as shown below.
Revise the fourth paragraph of subsection 6.2.2.2.4 as follows:
The cooling water not evaporated from the vessel wall flows down to the bottom of I the inner containment annulus into Heer-annulus drains. The redundant Heer I annulus drains route the excess water out of the upper annulus. The annulus I drains are located in the shield building wall slightly above the floor level to I minimize the potential for clogging of the drains by debris. The drains are I horizontal or have a slight slope to promote drainage. The drains knes-are always open (without isolation valves) and each is sized to accept maximum passive i containment cooling system flow. The outside ends of the drains are located I above catch basins or other storm drain collectors.
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FAX COVER SHEET l Recipient Information Sender Information Date: z/j 9/96 Name: D. Lindgren To: Bill Huffman Location: WEC E-330 Phone: Phone: (412) 374 4856 (301) 415-1141 Facimile: FacimDe: Bet (412) 374-4887 )
(310) 415-2002 Company: U.S.NRC WIN 2844887 Location: Rockville, MD Cover + Pages 1+ [
The following pages are being sent from the Westinghouse Energy Center, Monroeville, PA.
If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374-6529.
Comments:
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We have discovered an inconsistency in the SSAR changes proposed for our responses to Items 640.167F and 480.1135F and the closure times for containment isolation valves including the containment purge valves and the main steam and main feedwater isolation valves.
To clarify the require closure times the following changes are suggested.
The second paragraph of 6.2.3.4.1 will be changed as follows:
De safety related function of containment boundary integrity is verified by an integrated -
leakage rate test. The integrated leakage rate is verified to be less than L, as defined in Table 6.5.31. De integrated containment leakage rate system is utilized to measure the containment leak rate for determination of the integrated leakage rate. The containment l isolation valves are verified to close within the time specified in Table 6.3.2-1. 20 = 6 f= $: =:d:xx: p;g; ='d;; v6= xd $^ xxxi f= :.!! r :: d :c;;;
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Revise the last two sentences of subsection 6.2.1.5.3 as follows:
Two4016-inch diameter flow paths were provided in the containment model to simulate containment purge lines. Valves within these lines were closed 5-20 seconds after 8 psig was reached.
In Table 6.2.31 the closure times for the VES valves V003, V004, V009, and V010 will be changed to 20 seconds. The current time is 5 seconds. De times for the MSIVs and MFIVs will remain 5 seconds.
The response for 480.ll35F revised a note to indicate that the industry standard closure is less than or equal to 60 seconds.
SSAR markups are attached.
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- 6. Engineered Safsty Fectures I e
a relative humidity of 99 percent. An air annulus temperature of 0 F is assumed and a linear temperature profile between 0*F and 90 F is used in the containment shell, which separates the annulus from the containment volume.
6.2.1.5.3 Other Parameters Containment parameters, such as containment volume and passive heat sinks, are biased to predict a conservative low containment backpressure. The containment volume used in the calculation is conservatively set to 1.05 times the cold volume. Passive heat sink surface areas were approximately doubled from the heat sinks, presented in Tables 6.2.1.1-5 and 6.2.1.1-6. Material properties were biased high (density, conductivity, and heat capacity) as indicated in CSB 6-1 (Reference 8). To further minimize containment pressure, containment I purge was assumed to be in operation at time zero. Two4016-inch diameter flow paths were provided in the containment model to simulate containment purge lines. Valves within these I lines were closed 6-20 seconds after 8 psig was reached.
6.2.1.6 Testing and Inspection This section describes the functional testing of the containment vessel. Testing and in-service inspection of the containment vessel are described in subsection 3.8.2.6. Isolation testing and leak testing are described in subsection 6.2.5. Testing and inspection are consistent with regulatory requirements and guidelines.
l The valves of the passive containment cooling system are stroke tested periodically.
Subsection 6.2.2 provides a description of testing and inspection.
De baffle between the containment vessel and the shield building is equipped with removable panels and clear observation panels to allow for inspection of the containment surface. See subsection 3.8.2 for the requirements for in-service inspection of the steel containment vessel.
Subsection 6.2.2 provides a description of testing and inspection to be performed.
Testing is not required on any subcompartment vent or on the collection of condensation from the containment shell. The collection of condensate from the containment shell and its use in leakage detection are discussed in subsection 5.2.5.
6.2.1.7 Instrumentation Requirements Instmmentation is provided to monitor the conditions inside the containment and to actuate the appropriate engineered safety features, should those conditions exceed the predetermined levels. He instruments measure the containment pressure, containment atmosphere radioxtivity, and containment hydrogen concentration. Instrumentation to monitor reactor coolant system leakage into containment is described in subsection 5.2.5.
The containment pressure is measured by four ir,Jependent pressure transmitters. The signals )
are fed into the engineered safety features actuation system, as described in subsection 7.3.1.
Upon detection of high pressure inside the containment, the appropriate safety actuation Revision: 21 Draft,1998 6.2 20 W W85tingh00S8 m
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- 6. Engineered Safety Features l
. One locked-closed isolation valve inside and one automatic isolation valve outside containment. (A simple check valve is not used as the automatic isolation valve outside containment.)
- One automatic isolation valve inside and one automatic isolation valve outside containment. (A simple check valve is not used as the automatic isolation valve outside containment).
l Isolation valves outside containment are located as close to the containment as practical.
Upon loss of actuating power, air-operated automatic isolation valves fail closed.
E. Each line penetrating the containment that is neither pan of the reactor coolant pressure boundary nor connected directly to the containment atmosphere, and that satisfies the ,
I requirements of a closed system, has at least one containment isolation valve. This containment isolation valve is either automatic, locked-closed, or capable of remote-manual operation. The valve is outside the containment and located as close to the containment as practical. A simple check valve is not used as the automatic isolation valve. This design is in compliance with General Design Criterion 57.
F. The containment isolation system is designed according to seismic Category I l I
requirements as specified in Section 3.2. The components (and supporting structures) of any system, equipment, or structure that are non-seismic and whose collapse could result in loss of a required function of the containment isolation system through either impact or resultant flooding are evaluated to confirm that they will not collapse when subjected to seismic loading resulting from a safe shutdown earthquake.
Air-operated isolation valves fail in the closed position upon loss of air or power.
Containment isolation system valves required to be operated after a design basis accident or safe shutdown earthquake are powered by the Class lE de electric power system.
6.2.3.4 Tests and Inspections 6.2.3.4.1 Preoperational Testing Preoperational testing is described in Chapter 14. The containment isolation system is testable through th: operational sequence that is postulated to take place following an accident, i including operation of applicable portions of the protection system and the transfer between normal and standby power sources.
He safety related function of containment beundary integrity is verified by an integrated leakage rate test. De integrated leakage rate is verified to be less than L, as defined in Table 6.5.3-1. The integrated containment leakage rate system is utilized to measure the containment leak rate for determination of the integrated leakage rate. De containment I isolation valves art verified to close within the time specified in Table 6.3.2-1. 20 secent fer-theen::!n r: purge bela:!ca vcJves =d 60 secone for ;!! c:hcr centainm=: isclation vakr
. Revision: 21 Draft,1998
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- 6. Engineered Safety Features 4
Table 6.2.3-1 (Sl>
Containment Mechanical Penetri Containment Penetraties System Line Flow Closed Sys IRC Valve / Hatch identincation SSAR Subsection s
l VFS Corn. air fiber supply in No VFS-NV003 9.4.7 l VFS-PL.V004 I Cont. air fdter cahause Out No VFS-PbV010 9.4.7 l VFS-PbV009 VFS PbV008 VWS Fan Coolers out Om No VWS-%V006 9.2.7 VWS-NV082 Fan coolers in in No VWS-Nv058 9.2.7 VWS-PLV062 WlJ Reactor coolant drain tank gas Out No WLS-%V068 II.2 W13-LV067 Normal cont. surnp Out No WLS-NV057 11.2 WLS.6V055 SPARE N/A No P40 6.25 SPARE N/A No P41 6.2.5 SPARE N/A No P42 6.2.5 CNS Main equipment hasch N/A No CNS MY YOI 6.2.5 Map- hasch N/A No CNS-M Y-YO2 6.2.5 Penennel hasch N/A No CNS MY.YO3 6.2.5 Parsonnel hasch N/A No CNS-MY YO4 6.2.5
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s.." a r;tions and Isolation Valves DRAFT Isaiseles Device Test Fe.ies Cwre N S.A Signal Times Typel & Note Medhun Direction CoC T. HR.DAS 2 C5 Air Forward C-O C T. HR DAS 20&sec Forward em f \8 CoC T,HR DAS 20h C.5 Air Forward C4C T HR,DAS 20 bus Forward CCC N/A N/ Forward Dec T std. CJ.4.5 Air Forward 04C T std.
04C T std. C,3,4,5 Air Forward 04C N/A rid.
CCC T std. C Air Forward C-C4 T std.
CCC T.DAS std. C Air Forward C-C-C T,DAS std.
C C-C N/A N/A B Air Forward C-C 4 N/A N/A B Air Forward C-C-C N/A N/A B Air Forward C-C-C None N/A B Air Forward C-C-C None N/A B Air Forward CC4 None N/A B Air Forward CCC Nons N/A B Air Forward Revision: 21 Draft,1998 6.2 189
]
Table 6.2.3-1 (Sheet 4 of 4)
Containment Mechanical Penetrations and Isolation Valves Closure Time:
Required valve closure stroke time I std: Industry standard for valve type (5 60 seconds) f j
N/A: Not Applicable i in SSAR Section 6.2.3.1.1 Test: 'Ihese fields refer to the penetration testing requirements a giv:n penetration Type: Requind test type Sure A: Integrated Leak Rate Test P&ID or figure B: Local Leak Rate Test - penetration C: Local Leak Rate Test -- fluid systems Note: See notes below Medium: Test fluid on valve seat Direction: Pressurization direction Forward: High pressure on containment side Reverse: High pressure on outboard side lation C pressurizer level 50, Appendix L tup feedwater, blowdown and sampling piping from the steam generators to the containment penetration, is considered an extension of the cor itainment atmosphere during post accident conditions. During Type A tests, the secondary side of the steam generators is vented to the atmos Type A test in order to maintain stable containment atmospheric conditions.
\ test ta facilitate testing. Their leak rates are measured separately, iction isolation valves is not vented during local leak rate test to retain double isolation of RCS at elevated pressure. Valve is flooded during ve sine] the test pressure tends to unscat the valve disc, whereas containment pressure would tend to seat the disc.
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5' M Westinghese FAX COVER SHEET W '
l RECIPIENT INFORMATION SENDER INFORMATION DATE: A/I?/9y NAME: $v& Fach TO: LOCATION: ENERGY CENTER .
bill NuNwq EAST PHONE: FACSIMILE: PHONE: Omce. '/OA t COMPANY: FacsinwW. win: 284 4887 A)RC- outside: (412)374-4887 LOCATION:
Cover + Pages 1+ 3, The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. N any problems occur during this transmission, please call:
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[ 1. Introduction end G:ner:1 Description cf P12t Criteria Referenced , AP600 Section Criteria Position Clarification / Summary Descdption of Exceptions Reg. Guide 1.94, Rev.1,4/76 - Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants General ANSI N45.2.5-1974 N/A Not applicable to AP600 design certification.
This is the responsibility of the Combined Li-cense applicant. See Section 17.4 for the Combined License information item.
Reg. Guide 1.95, Rev.1,1/77 Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release General N/A The AP600 does not have onsite chlorine sources.
Therefore, these guidelines are not applicable to the AP600. Offsite chlorine sources are site-specific and are the Combined License applicant's responsibility.
Reg. Guide 1.%, Rev.1,6/76 Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants General N/A Applies to boiling water reactors only.
Reg. Guide 1.97, Rev. 3,5/83 Instrumentation for Light Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident General ANS-4.5- 1980 Conforms The variables to be monitored are selected ac-cording to usage and need in the plant Emergen-cy Response Guidelines. They are assigned design and qualification Category 1,2, or 3 and I classified as Type A, B, C, D, or E. Due to l AP600 specific design features A: disc;=ed h Sec::r .5, the selection of some plant-specific variables and their classifications and categories I are different from then those of this regulatory I guide. For example, the use of the passive I residual heat removal system as the safety I grade heat sink allows steam generator wide I range level to be category 2, not category 1 as I specified in Regulatory Guide 1.97.
The AP600 has no Type A variables. See Section 7.5 for additional information.
Since Category 3 instrumentation is not part of a safety-related system, it is not qualified to pro-vide information when exposed to a post-accident adverse environment. Category 3 instrumentation is subject to servicing, testing, and calibration DRAFT
[ W8Silfigh0USe 1 A-43
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) i 1. litroduction cnd General Description cf Pl=t Criteria Referenced '
AP600 )
Section Criteria Position Clarification / Summary Description of Exceptions programs that are specified to maintain their l capability. However, these programs are not in accordance with Regulatory Guide 1.118. which l applies to safety-related systems.
C.1 2 Conforms Reg. Guide 1.98, Rev. O,3/76. Assumptions Used for Evaluating the Potential Radiological Consequences of a Radioactive Offgas System Failure in a lloiling Water Reactor General N/A Applies to boiling water reactors only.
Reg. Guide 1.99,(Task ME 305-4), Rev. 2,5/88 - Radiation Embrittlement of Reactor Vessel Materials C.! Conforms C.2 Conforms C.3 Conforms Reg. Guide 1.100,(Task EE 108-5), Rev. 2,6/88. Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants General IEEE Std. 344-1987 Conforms Reg. Guide 1.101, Rev. 2,10/81 Emergency Planning and Preparedness for Nuclear Power Reactors General NUREG-0654, Conforms Emergency planning is the responsibility of the FEMA REP-1 Combined License applicant. See Section 13.3 for the Combined License information on emergency planning. RG 1.101 (Revision 2) references NUREG-0654/ FEM A-REP-1 and item
!!.H. " Emergency Facilities and Equipment" of NUREG-0654/ FEMA-REP-1 is applicable to the technical support center (TSC). operations support center (OSC), and the emergency operations facility (EOF) in the AP600 design. Designing the EOF, including specification of its location, in accordance with the AP600 human factors engineering program is the responsibility of the Combined License applicant. See section 18.2.6 for the Combined License information on designing the EOF. 'Ihe AP600 design conforms with the design criteria of item II.H that pertain to the TSC and the OSC.
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RECIPIENT INFORMATION SENDER INFORMATlON DATE: ,)//.3/98 NAME: O . Mao %
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U PHONE: Office: QI2 3 7y-ya7 7 COMPANY: Facsimile' win: 284 4887 DS N R.C- outside: (412)374 4887 LOCATION:
Cover + Pages 1 + J J:t 5 23 The foRowing pages are being sent from the Westinghouse Energy Center, East Tower, Monroevule, PA. If any problems occur during this transmlesion, please sail:
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I COMMENTS:
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NRC Staff Insights of the AP600 Level 1 PRA and Westinghouse Feedback -
OsagIlllA_JdaDhwide requirejngnis .
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- 1. WEewrt+mamtamg list of risk important systems, structuses and components (SSCs in the D RAP. [hsjw'b.
-ST/fd 2 Response: The risk.important SSCs within the scope of D RAP are provided in SSAR Table 17.41. There / 7. '/'
is no additionalaction required by Westinghouse to maintain this list after Final Design Approval.
Westinghouse does not agree that this item is an insight of the Leve ! PRA, rather th PRA results are usedfor identifying the risk imponant SSCs in D RAP , ,
- 2. The COL Applicant should perform a seismic walkdown to ensure that the as-built plant conforms to the assumptions in the AP600 PRA based seismic margins analysis and to assure that seismic spatial systems
,. interactions do not exist. Details of the seismic walkdown will be developed by the COL applicant. .
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. W Response: As provided in the response to FSER open items 720.451F through 720.433F, the seismic margin Combined License applicant action item will be changed in AP600 PRA Revision i1, subsection if 59.10.6 to read as follows:
Y The Combined License applicant referencing the AP600 certified design should perform a seismic walkdown to confirm that the as-built plant conforms to the design used as the basisfor the seismic margin evaluation and that seismic spatial systems interactions do not exist. Details of the seismic walkdown will be developed by the Combined License applicant.
- 3. WEC will maintain a list of the SSC HCLPF values used in the AP600 Seismic Margins Assessment in the D-RAP. The COL Applicant should compare the as-built SSC HCLPFs to those assumed in the AP600 seismic margins analysis (SMA). Deviations from the HCLPF values or assumptions in the SMA should be evaluated by the COL Applicant to determine if any vulnerabilities have been introduced. MM E Response: The HCLPF values usedfor the AP6')O seismic margin analysis are provided in AP600 PRA Table 551. The SSCs captured by the D RAP process using the results of the seismic margin analysis as the rationalefor inclusion, are provided in SSAR Table 17.41. There is no additional action required by Westinghouse to maintain this list after Final Design Approval. Westinghouse does not agree that "WEC will maintain a list of the SSC HCLPF values"is an insight of the Level i PRA.
As provided in the response to FSER open items 720.451F through 720.453F, the following Combined License applicant action item will be included in AP600 PRA Revision iI, subsection 59.10.6:
The Combined License applicant referencing the AP600 certified design should compare the as built SSC HCLPFs to those assumed in the AP600 seismic margin evaluation. Deviationsfrom the HCLPF
\
Q values or assumptions in the seismic margin evaluation should be evaluated by the Combined License applicant to determine y unacceptable vulnerabilities have been introduced.
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4 e COL Applicant will maintain an operation reliability assurance process based on the system reliability information derived from the PRA and other sources. The COL Applicant should incorporate the list of nsk-important SSCs, as presented in the SSAR section on D-RAP, in its D-RAP and operation reliability assurance process.
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W Response: There is a Combined License applicant 0 RAP action within SSAR subsection 17.4.8 that reads SStig the ' Combined license applicant is responsiblefor performing the tasks necessary to maintain the j y, f reliability of risk-sienificant SSCs." In addition. SSAR subsection 17.4.7.3 states the "C6E' applicant will need to establish PRA importance measures, the expert panel process, and other deterministic methods to determine the site specific list ofSSCs under the scope of RAP.* These two COL action items address the stafs insight statements.
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@k 4 important operator actions from the F'# PRA, as y presented in Chapter 18 of the SSAR on human factors engineering 3in developing and implementing procedures, ef5 All training and other human reliability related programs,,y,.y ,h pe+t , f6
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W Response: In the AP600 PRA, credit is taken for various tasks to be performed in the control room team of trained operators. These tasks are rule based andproceduralized. The tasks refer to the
& g completion of a nell defined mission by a team of trained operatorsfollowing procedures. As stated in SSAR section 18.10. operator training is the responsibility of the COL Westinghouse W p //
input to the COL is provided in WCAP 146SS. PRA Table 59 29, item 11, also reflects what is written in SSAR chapter 18. Westinghouse believes what is provided in SSAR section 18.10, and q 24 1
_ hogt is captured in PRA Table 59 29, addresses the stafs insight statement.
g 6 ' dmikddesignAgelthhkht 9bekhpdhe i inn the final desinrt .nformation and
-TI'et specifi~cinformation. s dee_med necess
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he COL Applicant shoul u i i PRKihIcTuifmg the fire'
,MIT;hd-arralfse th at-powcLand shutdown operation @ on site-specific information, the COL ~
] Applicant should also re evaluate the qualitative screening of exteiiiEwentif any site specific suscepti,bj ' ,
. c' are found, the applicable external event should be included in the updated PRA. ,
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W Response: There is a COL item provided in PRA subsection 59.10.6 that reads the " Combined Licens f' applicant referencing the AP600 certified design will venfy the as builtplant is consistent with the h
design used as the basisfor the baseline AP600 PRA." It is the COL's responsibility to describe
'g,f how this will be done and whether any ponions of the baseline PRA need to be updated.
- 7. No safety-related equipment is located outside the Nuclear Island.
W Response: This is an accurate statement. Inc l de d M mw'W f
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- 8. De AP600 low pressure systems which interface with the RCS are protected against interfacing systems LOCA ,f (ISLOCALby.arombinatica(multiple isolation vals es, valve interlocking, increase in the piping pressure limits V g andife{ure relief capability.,) .
g g d 4t, W im (c c %% SP- M, W Response: This is an accurate statement, ,PRA Tan 59 29. item 6. specifically discusses the elements which prevent interfacing system LOCA betseen the RNS and the RCS.
g 9 Solid state switching devices and electro-mechanical relays resistant to relay chatter will be used in the AP600 *.cfd I&C system;I Use of these devices and relays either eliminates or minimizes the mechanical discontinuities y slale' i
f associated with similar devices at operating reactors. @
\Y E Response: It is not understood why the stafs statement is an insightfrom the AP600 PRA. The staf would need to explain why this is an important insight of the PRA tojustify its placement in the DCD.
The stafs statement is accurate, but is not explicitly stated in the SSAR or PRA.
> 10' There are no watertight doors used for flood orotection in the AP600 design.
(6c .p E Response: This is an accurate statement per SSAR subsection 3.4.1.1.2.
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./11. He AP600 design minimizes potential Gooding sources in safety related equipment areas, to the extent possible.
The design also minimizes the number of penetrations through enclosure or barrier walls below the probable maximum flood level. All Good barriers (e.g., walls, floors and penetrations) are designed to withstand the maximum anticipated hydrodynamic loads ~4s well as water pressures generated by floods in adjoining area}
E Response: Excluding the ending phrase "as well as water pressures generated byfloods in adjoining 3ar the staf statement is supported by SSAR subsection 3.4.1.1.2. This is essentially item 23 of PRA 1 g
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Table S9 29.
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- 12. Drains are capable to remove now from an assumed break in a line up to 4" in diameter and include features, t such as check valves and siphon breaks, that prevent backflow. (t
' - W Respo The l=-+" m wn d in the stafs statement is not supported by text in the AP600 SSAR. .
SSAR dsection 9.3.5.1.2 es read *Pluning of the drain headers is minimi:ed by designing ^ e
. Q 'Y&.Jtaree enough to accommodate more than the design 11ow and by making the flow cath .'s$gas straight as cossible. Drain headers are at least 4 inches in diameter," Regarding the portion of
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the stafs statement on backflow presention, see the last bulletfrom item 15 below.
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- 13. There is no cable spreading room in the AP600 design. ky
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/l'd. The separatio/
h- A ngf equipment and cabling associated with dif erent divisions of safety related equipment as well as the separation of safety-related from nonsafety-related eq spment, rmm16es the lifelihood that a fire or flood bl '
would affect more than one safety-related system or train. MP [' f E Response: This is an accurate statement. PRA R Table $9-29, item 13, provides the same information. /* ,
i 15, ne following minimize the probability for dre or flood propagation from one area to another and helps limit risk from internal fires and floods: %
Fire barriers are sealed and flood barriers are watertight.
& Response: This statement isfrom PRA Table 59-29, item 14, but is isissing the words "to the extent possible" after the word sealed. L Each fire door is alarmed in the control room.
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The COL Applicant will ensure the reliable performance of fire barriers through appro riate inspe ion and maintenance of doors, dampers, and penetration seals. Also, all water tight penetrations will be I
maintained with high reliability during power operation to prevent the propagation of water from one d area to the next.
h h esponse: The stafs statement appears to be concentrating on a COL item for inspection and maintenance y jq of fire barriers and maintenance of reliable water tight penetrations. Westinghouse is not y specsfying the COL items to this level because it is the COL's responsibility to describe how this will be done. Rather, Westinghouse includes a COL item provided in SSAR subsection 9.5.1.8 that reads the " Combined License applicant will address qual:fication requirements for individuals responsible for development of the fire protection program, training offire fighting personnel, administrative procedures and controls governing the fire protection program during plant operation, andfire protection system maintenance " In addition, as stated in SSAR Table 9.5.1 1, items 29, it is the COL's responsibility for " establishing administrative controls to maintain the Q performance of thefire protection system and personnel."
De COL Applicant will ensure the availability of proper fire fighting equipment in all plant areas, emf--
.-espt'clally an me most risk sig(11fidAnt1TrnTr KI'--
H)e onse: SSAR Table 9.5.1 1, items 4, 30. and 32, cover this stagstatement. Note that it is not appropriate
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to ac.'d the phrase "and especially m the most risk-sigmficantfire areas" because Table 9.5.11
,f cover.: .s!! fire areas. There is no need to limit this to the most risk significantfire areas.
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. The COL Applicant will maintain an adequately staffed, well trained, and well prepared fire brigade.
V h. esponse: SSAR Table 9.5.11, items 4 and 30 through 34, cover this staf statement.
When a fire door, fire barrier penetration, or flood barrier penetration must be open to allow specific maintenance (e.g., during plant shutdown), appropriate compensatory measures will be taken to mini.
g mize risk. Risk during shutdown is minimited by appropriate outage management, administrative controls, procedures, and operator knowledge of plant configuration. In panicular, this will require
\ , configuration control of fire /Dood barriers to ensure the integrity of fire and flood barriers between areas containing equipment performing redundant safe shutdown functions.
}y Response: The intentions of what is described in the stafs statement is covered by' good plant operating ;
Df practices. It is covered in a higher level by SSAR Table 9.5.1 1, items 4 and 29.
/
Drains include features, such as check valves and siphon breaks, that prevent backflow.
fptd%r.' }y Response: Assumption m, as written in PRA Chapter 56. reads "forfloor drains, appropriate precautions such as check valves, back flow preventors, and sighon break.s are assumed to prevent back flow and ff 1 any potentialflooding." g' y 9,3,;)
9
- 16. Fire detection and suppression capability as wc!! as flooding control features and sump level indication are g4 provided in the AP600 design. Appropriate compensatory measures will be taken by the COL Applicant to $0,g maintain adequate detection and suppressioti capability during maintenance activities, gt
}! Response: Per SSAR section 13.5, the Combined License applicant is responsible for developing the plant procedures. The stafs statement is part of good plant practices, and should be addressed by the f'
applicable procedures which the COL will develop.
- 17. In addition to the MCR which has its own dedicated ventilation system, separate ventilation systems are provided for each of the two i p
h divisions A&C andBAD)$
of smoke from a non-saf pairs urthermore, theof safety plant related ventilation equipment systems divisions include features to prevent supportinnedundant propagation related area to a safety related area or between safety-related areas supported _by fu wQifferent divisions. The COL holder must ensure the reliable performance of such smoke propagation prevention features.
S Response: E.xcluding the COL statement, the staff's statement is covered by_ item 20 of PRA Table 59-29.
Regarding the . COL statement, this level of detail is not included wiihin Westinghouse COL items 1 ofSSAR 9.5. ;
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W Response:
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,5S_ A_Rytion 13.5.1 (as revisedper the response to FSER open item 440.763F. Westinghouse letter DCP/NRC)]98, dated December 22, 1997) includes the following statement: WCAP 14837 provides input to the Combined License applicantfor the development ofplant specific refueling plans." This means the maintenance guidelines, as well as other guidelines specofied within the WCAP, should be considered by the Combined License applicant when they develop the plant procedures. This SSAR COL item covers at a h stafs statement.
- 9. The COL applicant should control transient combustibley 3 g a 2 Response: The intentions of what is described in the stab's statem t is cos r d in a higher level by SSAR dane 9.5.1 TNtem 4d.
- Ye- h/.5 f.
Main Control Room (MCR) and Remote Shutdown Workstation (RSW)
- 1. De automatic function of the AP600 actuation systems (i.e., PMS and DAS)is not affected by a fire in either the MCR or the RSW. His ensures an independent, automatic means, to reach safe shutdown even 3
e when a fire occurs in the MCR or the RSW (manual actuation is not needed unless the automatic actuation M fails). Also, even though a fire in the MCR may defeat manual actuation of equipment from the MCR, it will not affect the manual operation from the RSW. Dis is because the I&C cabinets are located in fire areas outside the MCR and the RSW.
E Response: The stafs statement is covered by. item 19 of PRA Table 59 29.
. Redundancy in MCR operations, in terms of both monitoring and manual control of safe shutdown equipment, g is provided within the MCR itself. This provides an alternative means for mitigating certain MCR fires
>h before deciding to evacuate the MCR and use the RSW, E Response: The stafs statement is covered by item 17 of PRA Table 59-29.
- 3. If MCR evacuation is necessary, the RSW provides complete redundan:;y in terms of control for all safe shutdown functions.
E Response: This statement is paraphrasedfrom SSAR section 7.4.3.1.1. The stafs statement is covered by item 12 of PRA Table 59 29. (fS adQ &ge,, bsgl2 4 De MCR has its own dedicated ventilation system and is pressurized. His eliminates the possibility of smoke, hot gases, and fire suppressants, originated in areas outside the MCR, to migrate via the ventilation system to the control room.
W Response: The stafs statement is covered by item 20 of PRA Table 59 29. Note it is recommended that the stafs wording of " eliminates" be changed to " prevents".
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- 5. He MCR and the RSW are in separate Gre and Good areas. Rey have separate and independent ventilation k(. systems.
_W Response: The staffs statement is cosered by items 18 and 20 of PRA Table 59-29.
AP600 MCR Gre ignition frequency is limited as a result of the use of low-voltage, low-current equipment o and 6ber optic cables.
W Response: The stafs statement is covered by item 16 of PRA Table 59-29.
Cpntainment/ Shield Buildina
- 1. Containment isolation functions are protected from the impact of internal fires and floods by redundant containment isolation valves in each line which are located in separate fire and flood areas and,if powered, C are served by different power and control divisions. Always, one isolation component in a given line is located inside containment, while the other is located outside containment, and the containment wall is a fire / flood barrier.
W Response: The staffs statement is covered by item 22 of PRA Table 59-29.
- 2. Although the containment is a single fire area, redundant divisions are generally separated by continuous structural or fire barricts without penetrations and by labyrinth passageways. In a few situations, the divisions are separated by large open spaces without intervening combustibles.
t ty sponse: Westinghouse recommends the stafs wording of this insight repect what is written in SSAR Q '-) - subsection 9A.3.1.1, specifically: The containment / shield building comprises onepre area which
- U is separated intofire :ones. "These :ones are based on the establishment ofboundaries (structures V or distance) that inhibit pre propagation from :one to :one. Complete pre barrier separation cannot be provided inside containment because of the need to maintain thefree exchange of gases ,
f for purposes such as passive containment cooling " -
1
- 3. Here are wo compartments inside containment (PXS-A and PXS B) containing safe shutdown .
equipment other than containment isolation valves that are floodable (i.e., below the maximum flood height). .
Each of these two compartments contains redundant and essentially identical equipment (one accumulator with associated isolation valves as well as isolation valves for one CMT one IRWST injection line and one gj containment recirculation line). Dese two compartments are physically separated by-i or 3.ioot Watts anA v9P
- 5 de to ensure that a flood in one compartment does not propagate to the o er. Drain lincs from the yd PXS A and PXS-B compartments to the reactor vessel cavity and steam generato compartment are protected y
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frorn backflow by redundant backflow preventers. ("
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d W Response: Westinghouse recommends she staff remose the word "only"in theprst sentence. The cavity also
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)y has source range detectors. Itnot ispropagate correcttothat thehowever, PXS Westinghouse A and PXS B compar b such that a flood in one compartment does the other;
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('y O *, recommends the staf remove the specifics that the compartments are separated "by 2 or 3 foot f
walls andpoor slabs." It appears the staffinadsertently used the words regarding 2 and 3. foot walls andpoor slabs that appear in SSAR subsection 3.4.1.2.2.2 which pertains to the auxiliary y building separation of RCA and nonRCA areas. Once these recommendations are implemented, W the stafs statemesa is fully supported by SSAR subsection 3.4.1.2.2.1.M$g6 IT1W
- 4. Containment isolation valves located below the maximum flood height inside containment or in the Auxiliary lluilding are norma!!y closed and are designed to fait closp..-.w.
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r H Response: The stafs statement is not technically accurate. The va re not designed tofait closed when submerged. West ghauss-recomm sta ange the wording of their statement to read f consistently with SAR subsection 3.4.L2.2; Specifically, the SSAR reads "There are fdy_r.
automatically actuated containment isolation valves inside containment subject toflooding. These four normally closed cpntainment isolation val _ves . wouldawLfaiLopea a.L_asesult of the /. .pf
.f compartment flooding. Also, there is a redundantJermilv closed ennon *n~nr isolation valve \ t*% 'y located outside containment in series with each of these valves." )
- 5. He fragility of valve rooms, labeled 11206/11207, where the passive core cooling system valves are concentrated is an important fa: tor in the AP600 capability to withstand earthquakes. He capacity of the as-built SSCs to meet the HCLPF values assumed in the AP600 PRA will be checked by a seismic walkdown. i(). d.
E Response: It is not understood what the stag means by the fragility of valve rooms 11206 and 11207 is an importantfactor in the capability ofAP600 to withstand earthquakes. The HCLPF valuefor these rooms is 0.96g (per PRA Table $$-1). The HCLPF valuefor these valve rooms is not the limiting fr gg.
HCLPF elementfor the nuclear island. Westinghouse recommends thefirst sentence of the stafs O.v sA statement be removed. ./
co b The stafs statement regarding a seismic wa'kdown is already addressed under item 2 of " general p- 5
& plant wide requirements." ,
- 6. The passive containment cooling system (PCS) cooling water not evaporated from the vessel wall flows down .
to the bottom of the inner containment annulus into floor drains. The redundant floor drains route the excess a water to storm drains. He drain lines are always open (without isolation valves) and each is sized to accept maximum PCS flow, ne interface with the storm drain system is an open connection such that any blockage in the storm drains would result in the annulus drains overflowing the connection, draining the annulus independently of the storm drain system.
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NRC FSER OPEN ITEM .
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E Response: Westinghouse recommends the staf revise this statement to read "The passivugetainment cooling b0, f system (PCS) cooling water not evaporatedfrom the vessel wallflows down to the bottom of the inner containment annulus. Two 100 percent drain openings. located in the side wall of the shield h
y building, are always open with screens pros ided to prevent entry of small animals into the drains.
f
'p ' Note that the specfic drain conpguration has changed sinceIwh L 8, when the drains were located on thefloor of the annulus (see also response to FSER open item pl 720.440F). Thus the staff's statement should be revised.
t 7 ie annulus door drains, which are essentially pipes embedded into the wall of the Shield Building, will have g .
the same (or higher) HCLPF value as the Shield Building. This ensures that the drain system will not fail
[ at lower acceleration levels causing water blocking of the PCS air bafDe.
- rE Response
- Refer to item 6 above regarding placement of the annulus drains.
[ [h The COL applicant should develop and implement policies, procedures, and trainin o close containment
.; penetrations during Modes 5 and 6 in accordance with TS 3.6.8.
A COL item in SSAR subsection 13.5I states the Combined License applicant will address plant f [f $ Response: procedures. A COL item in SSAR subsection 13 2.1 states the applicant will develop and implement training programs for plant personnel These items inherently include following the Technical Specepcations. The COL items in SSAR chapter 13 cover the staff's statement.
Apxiliary Building 9
J 1, Separate ventilation systems are provided for each of the two pairs of safety-related equipment divisions
,[ supporting redundant functions (i.e., divisions A&C and B&D). His prevents smoke, hot gases, and Gre
- ,' suppressants originating in divisions A or C from propagating to divisions B and D.
'd
\ E Response: The stafs statement is covered by item 20 of PRA Table 59-29. Note this is essentially a duplicate
~
Gp ofitem 17 of " general & plant wide requirements."
- 2. De major rooms housing divisional cabling and equipment (the battery rooms, equipment rooms, !&C rooms, and penetration rooms) are separated by 3-hour rated fire walls without openings. Dere are no doors, d.
dampers, or seals in these walls. The rooms are served by separate ventilation subsystems. In order for a l i fire to propagate from one divisional room to another,it must move past a 3-hour barrier (e g., a door) into d a common corridor and enter the other room through another 3. hour barrier (e g., another door).
E Response: This is an accurate statement. It is essentia!!y shot is described in SSAR subsection 9A.3.1.
7* * "
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. NRC FSER OPEN NEM
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- 3. A two foot concrete floor (barrier) protects important safety related !&C equipment as well as the main )(o i control room and the remote shutdown panel. located in the north end of the Auxiliary Building, from @
y potential debris produced by a postulated sei. nically-induced structural collapse of the adjacent Turbine
[f, d
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.\ Building and propagated through the access bay separating the two buildings, W Response: To be an accurate statement, the stafs wording should be changed asfollows: (1) change "a two. J foot concrete floor (barrier)" to "An access bay"; and f 2) delete the ending words "andpropagated t,,
through the access bay separating the two buildings." By changing these words, the statement is now consistent with PRA subsection _5Qeg?-
- 4. 'Ihere are no r,onnuswens to sources of " unlimited quantity of water in the Auxiliary Building. /pl$fN f W Response: It is not understood what is the definition of " unlimited quantity of water" or the purpose o is N tatement. Uponfurther understanding of this statement, it may be accurate to state there are o fmally open connecQ
$5. To ensure that a flooding in a radiologically controlled area (RCA) in the Auxiliary Building does not propagate to non RCAs (where all safety related equipment except for some containment isolation valves is p located), the non RCAs are separated from the RCAs by 2 and 3-foot walls and Ocor slabs. In addition, electrical penetrations between RCAs and non RCAs in the Auxiliary Building are located above the maximum Good level.
W Response: As it is not appropriate to use the word " ensure" since its interpretation is su ective.
CA U
Westinghouse recommends the stafs statement be reworded to read "To preventfooding in in the auxiliary buildingfrom propagating to . " and to remove the st'itement in parentheses. The G #
statement will then be consistent with SSAR subsection 3.4.1.2.2.2.
sfD IS[-
' 6. The two 72-hour rated Class IE division B aad C batteries are located above the maximum flood height in the Auxiliary Building considering all possible flooding sources (including propagation from sources located l outside the Auxiliary Building). 5 psf'.* M NN W Response: It is not clear why the stafincludes this statement as an important insightfrom the PRA. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Class 1E batteries are usedfor safe shutdown operation; the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> batteries are usedfor functions such as post accident sampling.
- 7. Flood water propagated from the Turbine Building to the Auxiliary Building valve / piping penetration room at grade level (the only Auxiliary Building area that interfaces with the Turbine Building)is directed to drains g and to outside through access doors. This, combined with the presence of water tight walls and floor of the valve / penetration room, limits the maximum flood height in the valve / piping penetration room (to about 36 y\ inches) and ensures that the flooding does not propagate beyond this area.
2 Response: Change the words " ensures that thefooding does not propagate ... ~ to "preventsfloodingfrom If propagating beyond this area." The statement is then accurate per SSAR subsection 3.4.1.2.2.2 e
~
Auxiliary Building Level 3 non RCA discussion. p
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- 8. The mechanical and electrical equipment in the Auxiliary Building are separated to prevent propagation of C"A leaks from the piping and mechanical areas to the Class lE electrical and Class IE I&C equipment rooms. J- '
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W Response: By resising the wording to read "... the piping and mechanic equipment eas", the stafs statement becomes consistent with SSAR subsection 3.4.l_.2.2.2.
Turbine Beilding 930 1. No safety related eqaipment is located in the turbine building. Der is a 5our fire barrier wall between g the turbine building and the safety related areas of the Nuclear Island. g6 E Response: This is an accurate statement, per SSAR subsection 3.4.1.2
{0 (Note there was not item 2 or 3 in Attachment 2 of NRC's November 7,1997 letter.]
. 4.! Connections to sources of "large" quantity of water are located in the Turbine Building. They are the service water system (SWS) which interfaces with the component cooling water system (CCS) and the circulating water system (CWS) which interfaces with the turbine building closed cooling system (TCS) and the condenser. Features that minimize flood propagation to other buildings are:
Flow from any postulated ruptures above grade level (elevation 100' 0")in the Turbine Building flows j down to grade level via floor grating and stairwells. His grating in the floors also prevents any significant propagation of water to the Auxiliary or Annex Buildings via flow under the doors. .
j\ - A relief panet in the Turbine Building west wall at grade level directs the water outside the building to the yard and limits the maximum flood level in the Turbine Building to less than 6 inches. Flooding p propagation to areas of the adjacent Auxiliary and Annex Buildings, via flow under doors or backflow
- ef
)
through the drains, is possible but is bounded by a postulated break in those areas. q 1
W Response: Information in,,SSAR subsection 3.4.1.2.2.3 supports the stafs statement once the word " Annex" t,
is removedfrom the two suo ouiiers, ,
k D $ P- (/ i Annex Building p rM
/ 1. There is no safety related equipment located in the Annex Building. t k 2 Response: This is an accurate statement, per SSAR subsection 3.4.1.2.2.3. 0
- 2. Flood water in the Annex Building grade level is directed by the sloped floor to drains and to the yard area d
- k through the fr t door of the Annex Building.
WQ6'0 i N W Response: Remove the word " front"from the statement, and then it becomes an accurate statement, per SSAR subsection _t 4.1.2.2.3. -
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- 3. Flow from postulated ruptures above grade levelin the Annex Building is directed by floor drains to the
[
l Annex Building sump which discharges to the Turbine Building drain tank. Alternate paths include flows to the Turbine Building via flow under access doors and down to grade level via stairwells and elevator shaft.
.\ p -
, to 1 l W Response: Remove the word "any"from the statement, and then it becomes consistent with SSAR subsection 3.4.1.22.3. _
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4 The floors of the Annex Building are sloped away from the access doors to the Auxiliary Building in the .
vicinity of the access doors to prevent migration of flood water to the non-radiologically controlled areas of th dd-the Nuclear Island where all safety related equipment, except for some containment isolation valves, is
\ locatedpTAAC).+
2 Response: This is an accurate statement per SSAR subsection 3.4.1.2.2.L cDo SP.
- 5. "Ihere are no connections to sources of " unlimited" quantity of water in the Annes Building. p ::.5S M h E Response: It is not understood what is the definition of " unlimited quantity of water" or the pu pose of this statement. t), , j - l Beactor Coolant System l
- 1. To prevent overdraining, the RCS hot and cold legs are vertically offset which permits draining of the steam generators for nozzle dam insettion with a hot leg level much higher than traditional designs. un> scu. .; -
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m.mm.J,j 00 ;r :- L .J ... de- !"? Le'o-l , p S L%Wg W Response: This is an accurate statement per Sfsubsection 5.4.6.2. Although the second sentence may be m,Qd ,
an insight of the Shutdown Evaluation Report, it is not understood why this is.an important insight from the PRA.
- 2. 40 ivwu a LJ .o ia; hei kg dumw.se == a step nozzle connection between the RCS hot leg #1 and the RHR suction line is used. L ai; r.na .; e 20 lrch sch:d6 4'Jpipe appcoximately1 r feetto M
i o.
W Response: Although this may be stated within the Shutdown Evaluation Report it is not understood why this detail ofinformation is an important insight of the PRA. For example, the schedule of the piping is not important in calculating the failure probability. Ilowever, sf the staf explains why this is important as a PRA insight, then please revise the openin ' t j',4 .. ..p-nce it appears to be missing some words. , be consisttRLwi.th R subitct in" U 7 ' IJMorringhny /
r ccmmnulc 'k c-nroner r- lower the RCS hot leg level at yhich a vortex occurs in the A
RNS suction I tep noge . "
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- 3. Should vortexing occur,ic u,uimum air entrainment into the pump suction ^~ , +imonuuiy gr- .M, y S no g o.icc ih.n 5 pe A M ,
'T' W Response: Although this may be stated uishin the Shutdown Evaluation Report, it is not understood why this information is an important insight from the PRA. However. if the staff explains why this is important as a PRA insight, then please revise the sentence to read " . RNS pump suction .. ".
\)f .
4 Here are two safety-related RCS hot leg level channeh, one located in each hot leg. These level instruments
[ ., g are independent and do not share instrument lines. Rese level indicators are provided primarily to monitor 'L g RCS level during midloop operations. One level tap is at the bottom of the hot leg, and the other tap is on
,\'( the top of the hot leg as close to the steam generator as possible.
N W Response: Although this may be stated within the Shutdown Evaluation Report, it is not understood why this information is an important insight from the PRA.
- 5. Wide range pressurizer level indication (cold calibrated)is provided that can measure RCS level to the bottom Ich hen of the hot legs. % "pner 1-e 1 'ap h cme d to an ^.DS-valve 4aleMe;' hic Sc tcp cf ic 4
.V g, y murirer. ne1% Lei tap is connuicd to thee"r ^hh hotleg-Dis nonha. f wid 37 level indication can be used as an alternative way W used oftomonitoring identify level inconsistencies in the safety related hot leg level instrumentation. g W Response: Although this may be stated within the Shutdown Evaluation Report, it is not understood why this information is an important insightfrom the PRA.
- 6. The RNS pump suction line is sloped continuously upward from the pump to the reactor coolant system hot leg with no local high points. His design eliminates potential problems in refilling the pump suction line if a RNS pump is stopped when cavitating due to excessive air entrainment. His self-venting suction line allows the RNS pumps to be immediately restarted once an adequate level in the hot leg is re established.
W Response: This is an accurate statement per SSAR subsection $,4121 %d A.
- 7. De COL applicant should have procedures and policies to maximize the availability of the nohafe@ la wide range pressurizer level indication (cold calibrated) during RCS draining operations durin cold shutdown. De operators shall be trained to use this indication to identify inconsistencies in the safe lated hot leg level instrumentation to prevent RCS overdraining.
W Response: SSAR section 13.S provides the committment that the Combined License applicant is responsible for developing procedures. The COL items reported in szetion 13.5 provide the committment at a higher level than described in the stafs statement above.
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_ e IRWST injection and recirculation check valves are exercised at each refueling. IRWST injection and recirculation squib valve actuators are tested every 2 years for 20 percent of the valves. IRWST recirculation MOVs are stroke-tested quarterly.
The reliability of the IRWST subsystem is important. He COL will maintain the reliability of the IRWST subsystem.
IRWST injection and recirculation are required by Technical Specifications to be available from power conditions to refueling without the cavity flooded.
s W Response: The staff's statements above on IRWST is covered by item Id of PRA Table 39 29 except Westinghouse wishes to note thefollowing change should be made to what is wrinen above:
Second bullet - remove the work " ensure"for reasons provided earlier in this document.
An accurate statement would read * . recirculation line which prevents clogging by debris
. ." Also note the COL item is covered by a higherlevelaction of the COL willmaintain the reliabdity of the IRWST subsystem (SSAR Section 17.4).
l The IRWST provides a safety-related long term source of water during shutdown conditions. He following are some
\.g ditional important aspects of the IRWST subsystem as represented in the shutdown PRA.
(U e applic provide administrative controls t ntrol trash g cd-during shutdown g / ro RCS-a ens.
W Response: As stated in SSAR section 13.5, the Combined License applicant is responsible for developing b administrative controls. The COL item in SSAR chapter 13 covers the staf's statement at a higher $. 3 ? q level.
On low hot leg level, the PMS actuates the squib valves to open allowing gravity injection from the IRWST.
'E Response: This statement is a duplicate of the 8th bullet on IRWST(see above).
Passive Residual Heat Removal (PRHR) System l
The PRHR provides a safety-related means of performing the following functions: (1) removes core decay heat during accidents, (2) allows adequate plant performance during transient (non-LOCA and non ATWS) accidents without ADS,(3) allows automatic termination of RCS leak during a SGTR accident without ADS, and (4) provides core cooling and pressure control during the early phase of an ATWS accident.
W Response: For item (2), recommend changing the word " allows" to "provides." Item (4) is ambiguous by using the words early phase of an ATWS. The phrase should read. " allows plant to ride out an ATWS event without rod insertion."
W westinghouse *"8
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NRC FSER OPEN ITEM h
2 Response: The stafs above statements on ADS are covered by item ib of PRA Table 59-29. Notefor the 6th bullet, as a result of NRC review, the stage 1, 2, and 3 valves are now stroke tested erery cold Q
f shutdown. With the number of cold shutdowns and refuelings assumed in the shutdown PRA, the test frequency is equivalent to being tested esery 6 months. PRA Table 59 29 nill be revised appropriately. Notefor the 9th bullet, the wording "during a severe accident"should be changed to "after core uncovery."
+ Fire induced hot si. orts, especially in I&C copper cables from the protection logic cabinets to the squib valve operators, could cause detonation of a squib valve. This risk important concern should be addressed by appropriate power and control cable separation and routing and by the incorporation of features and requirements in the detailed design of ADS cabling.
f W Response: Westinghouse recommends the words of the stafs statement be changed to read as described in I SSAR subsection 9A.2.7.1, specifically, " Spurious actuation of squib valves is prevented by the use Q
f of a squib valve controller circuit which requires multiple hot shorts for actuation, physical f separation of potential hot short locations, and provisions for operator action to remove power from thefire zone." Note as stated in the internalfire PRA analysis, it is conservatively modeled in the PRA analysis that one hot short can cause spurious ADS squib valve actuation, whereas, per design, multiple hot shorts are required.
gj A .9 g f
. The first, second, and third stage valves, connected to the top of the pressurizer, provide a sent path to 7, f, preclude pressurization of the RCS during shutdown conditionsjf._ decay heat removal is lospOne-fomthJ s age ADS valve is required to open if gravity injection is actuated during cold shutdown and refueling with po -
the RCS is open to preclude surge line flooding. On low low hot leg level (empty hot leg), the PMS signals the ADS 4th stage squibs to open to preclude surge line flooding.
!W Response: This is an accurate statement. A statement will be added to PRA Table 59-29./
Normal Residual Heat Removal System (RNS)
The normal residual heat removal system (RNS) provides the following nonsafety-related means of core cooling during accidents: (1)"CS recirculation at shutdown conditions,(2) low pressure pumped injection from the IRWST, and (3) long term pumped recirculation from the containment sump. Such RNS functions provide defense in-depth in mitigating accidents, in addition to that provided by the passive safety related systems.
2 Response: This is an accurate statement. The statement is covered by item 6 of PRA Table 59 29. /
The following are some important aspects of RNS as represented in the PRA: ,
l
. The RNS has redundant pumps, powered by separate non Class IE buses with backup connections from the diesel generators, and redundant heat exchangers.
W Response: This is an accurate statement and is covered by item 6 of PRA Table $9 29. / _
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NRC FSER OPEN ITEM -
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De containment isolation valves in the RNS piping close automatically via PMS with a high radiation signal.
Westinghouse analyses indicate that under all accident conditions but large LOCAs, the containment radiation level is well below the point that would cause the RNS MOVs to automatically close.
LV Response: i The first serlitBClis CQuLstent with item 6 of PRA Table 59 29. The second sentence tends to lead beyond an insightfrom the PRA. However, if the staf explains why it considers this an insight, then Westinghouse recommends the second sentence be reworded to read: The actuation setpoint was established consistent with a DBA non mechanistic source term associated with a large LOCA. ~
- De following AP600 design features contribute to the low likelihood of interfacing system LOCAs through the NRHR system:
De portion of the RNS outside containment is capable of withstanding the operating pressure of the RCS.
A relief valve located in the common PJ4S discharge line outside containment provides protection against excess pressure.
- Each RNS line is isolated by at least three valves.
De pressure in the RNS pump suction line is continuously indicated and alarmed in the main control room.
De pump suction isolation valves connecting the RNS pumps to the RCS hot leg are interlocked with RCS pressure so that they cannot be opened until the RCS pressure is less than 450 psig. This prevents overpressurization of the RCS when the RNS is aligned for shutdown cooling.
ne two remotely operated MOVs connecting the suction and discharge headers, respectively, to the IRWST are interlocked with the isolation valves connectig the RNS pumps to the hot leg. His prevents inadvertent opening of any of these two MOVs when the RNS is aligned for shutdown cooling and potential diversion and draining of reactor coolant system.
De power to the four isolation MOVs connecting the RNS pumps to the RCS hot leg is administratively blocked at their motor control centers during normal power operation. [ COL).
De operability of the RNS is tested, via connections to the IRWST,immediately before its alignment pod .
to the RCS hot leg, for shutdown coolingpe : nx 0,.; Jim ec sc, ;.g q.c.. ; nan! valvet in the
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W Response: Westinghouse has thefollowing commentsfor the staf]'s above statement:
- Change "NRHR system" to "RNS".
Second sub-bullet is a true statement, but notfactored into the PRA and is not a key to providing a low likelihood ofinterfacing systems LOCA. Thus, Westinghouse does not see this as an important statement to include as an insight.
- Last sub-bullet: It is true that the system is tested; however, it is done to test operabilQ vt.N='-
of the system, not solely to minimize potentialfor interfacing syetems LOCA or to detect W) an open valve in the drain lines. However, the testing does have this end result efect. a cN g, The words should be revised appropriately.
. De IRWST suction isolation valve (V023) and the RCS pressure boundary isolation valves (V001 A, V001 B, V002A and V002B) are qualified for DBA conditions.
H Response: It is not understood why the staf's statement is an insight from the PRA.
k.h a ne reliability of the IRWST suction isolation valve (V023) to open on demand (for RNS injection during .
g ,y power operation and for IRWSTgravity injection via the RNS hot leg connection during shutdown operation)
- is importantQtOL~will ensure high reliabilily?[ COL, D-RAP]. htc/d v.i w ,r - S p M,,,J, c ,g s
-O Nk E Response: This item is acceptable and is covered by SSAR section 17.4 (RAP).
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- An alternative gravity injection path is provided through RNS V-023 during cold shutdown and refueling *fb
( conditions with the RCS open. He COL applicant should have policies that maximize the availability of this 3 "P valve and procedures to open this valve during cold shutdown and refueling operations when the RCS is open. ,
H Response: The ERGS cover the operation of the valve. In addition, as stated in SSAR section 13.5, it is the Q responsibility of the Combined License applicant to develop procedures.
E Response: To be accurate and consistent whAR sec 16.3 2, item 2.2), change the
- statement to read: " Planned maintenance affecting the RNS cooling function and its suppgrt l yshould be performed in Modes 1._2, 3 when the RNS is not nylly.nperating."
+ De COL applicant will have administrative controls to maximize the likelihood that RNS valve V-023 wiil~N be able to open if needed during Mode 5 when the RCS is open, and PRHR cannot be used for core cooling. p 4 W Response: As stated in SSAR section 13.5, it is the responsibility of the Combined License applicant to [$
develop administrative procedures. M .
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720.434F-25
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. nadvertent opening of RNS valve V024 results in a draindo n of RCS inventory to the IRWST and gg g requires gravity injection from the IRWST, e COL applicant *" hra administrative controls to ensure that inadvertent opening of this valve is unlikel . & df&n, i COL ; pine ^~M 4valuamhis erros inahe
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% man.seliability-analysis /humanVsV6rifTrigTnliering integrationimplementation plan. NW, W Response:
mwd SP.
As stated inhR seltlanJLL.Jt is the responsibility of the Combined hcense applicant to develop administrative procedures.
The RNS is an important " defense-in depth" system for accidents initiated while the plant is at power or at mid loop during shutdown. He availability control of the RNS and its support systems (CCW, SWS and diesel generators) is covered in SSAR Section 16.3. [RTNSS].
W Response: The reason RNS is important while the plant is at power is not because it is important per the PRA results or importance listings, but rather because it provides marginfor long term cooling T&H uncertainty. Otherwise, the stafs statement is accurate.
Startuo Feedwater System (SFW) ne SFW system provides a nonsafety-related means of delivering feedwater to the steam generators (SGs) when the main feedwater pumps are unavailable during an transient. His capability provides an alternate core cooling mechanism to the PRHR heat exchanger for non-LOCA and SGTR accidents which minimizes the PRHR challenge rate. He reliability of the SFW system will be maintained by the COL Applicant [D RAP].
W Response: The stafs statement is essentially taken directlyfrom the SSAR Table 17.4 (RAP). To be accurate, note the words should read startupfeedwater system cumos De rationale provided in this table for why the startup feedwater pumps are included is based on the Expert Panel, not PRA.
nerefore, it is not clear why the stafs statement is considered an insightfrom the PRA.
Instrumentation and Control (I&C) ne following three !&C systems are credited in the PRA for providing monitoring and control functions during accidents: (1) the safety-related Protection and Safety Monitoring System (PMS),(2) the nonsafety-related Diverse Actuation System (DAS), and (3) the nonsafety-related Plant Control System (PLS).
He PMS provides a safety-related means of performing the following functions:
- Automatic and manual reactor trip.
- Automatic and manual actuation of engineered safety features (ESF).
Monitor the safety-related functions during and following an accident as required by Regulatory Guide 1.97.
W Response: ne stafs statements on PMS are covered by item 2 of PRA Table 59-29.
720.434F-26
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- .E; Cbemical and Volume Control System (CVS)
M MAed5f '
1 y The chemical and volume control system (CVS) provides a safety-related means to terminate inadvertent RCS boron
\3 dilution. In addition, the CVS provides a nonsafety related means to (1) provide makeup water to the RCS during normal plant operation,(2) provide boration following a failure of reactor trip,(3) provide coolant to the pressuriz
@9 auxiliary spray line, (4) safety related portions of the CVS provide inadvertent _b_o.lon dit"'ina prntactinn andh s
. p hg , leg s Wlevel f
- d a~'i m #6: C'/S Mgolation i of normal CVS letdown during shutdown operation on low hot a- __ .
- f W Response: The stafs above statement on CVS is covered by item 9 of PRA Table 59-29 with supportfrom qh S5AR tubsection 9.3.6. Note the second sentence begins by discussing nonsafety-related means, jf but items (4) and ($) state safety-related portions. It could be a confusing sentence. Also note, item (4) is a repeat of the first sentence.
De following are some important aspects of CVS as represented in the PRA:
He CVS has two makeup pumps and each pump is capable of providing normal makeup.
W Response: This statement is covered by item 9 of PRA Table 59-29 One CVS pump is configured to operate on demand while the other CVS pump is in standby. De operation of these pumps will alternate periodically (monthly).
W Response: The stafs statement is accurate per PRA assumptions. The first sentence is true. The second sentence's monthly statement is an assumption of the PRA: however, good operating practices would callfor the COL to periodically alternate the pumps.
On low hot leg level, the safehlated PMS signals h safe lated CVS AOVs to close automatically g o' to isolate letdown during Mode 4 (when RNS is in operation), h e 5, and Mode 6 (with the upper internals in place and the refueling cavity less than half full) as required by AP600 TS.
E Response:
& J Only two of the AOVs are safety-related, the third is nonsafety-related. Exceptfor this error, th /
above statement is true per the ESF Technical Specification.
D (4)0 pr De safe lated PMS boron dilution signal automatically re aligns CVS pump suction to the boric acid tan .
His sam @e signal also closes the two safety-related CVS demineralized actuates on any reactor trip signal, source range flux multiplication signal, low input voltage to the Class IE ,
DC power system battery chargers, or a safety injection signal.
7 W Response: This is an accurate statement. p<hb He COL applicant will maintain procedures to respond to low hot leg level alarms. 1- .
H Response: The shutdown ERGS cover the procedure to respond to low hot leg level rms.
W Westinghouse 720.434F-33
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- 59. PRA Results and Insights M 'ge Y,# J h1
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[ Qbyfpy' l Table 59 29 (Sheet 15 of 16)
I AP600 PRA BASED INSIGHTS l
l
[ INSIGHT DISPOSITION t er po i sty the COL ASARAth l 12. Sufficient instrumentati and control is provided at the remote shutdown l
workstation to bring th plant to safe shutdown conditions in ase the control SSAR ,M '
db pg u v5 F room .emust be evacuat . M N ^* kNo % #4 . y #g u<g}u if p
- D 13.
Separatio,nyln W wanthe equippent and cabling among the [SARdivisions 3.4.1.1.2,of safety-related l
- [.n M e uipment puam and e probab separati n of safety related from nonsafety that a fire or flood would affect m- d-related 9.5.1.2.1.1 g
& 9A equipment j I stated-systent a Mih!/.M.s }. dwd j,jL
= :d., -
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l 14 J 8 The following minimizes the probability for fire and flood propagation from one area to another and helps limit risk from internal fires I '
l -
r 's K Fire barriers are sealed, to the extent possible? and flood arriers are I watertight. SSAR 3.4.1.1.2 &
I Each fire door is alarmed and monitored in the control room. 9.5.1.2.1.1 l -
Requirements for fire barrier and maintenance will be implemented in l SSAR 9.5.1.8 Combined License applicant programs.
I I -
When a fire door, fire barrier penetration, or flood barrier penetration must l SSAR 9.5.1.8 I
be open to allow specific maintenance activities, additional compensatory l
measures are expected to be taken. Control of compensatory measures is a Combined 1.icense applicant item.
I 15.
I Fire detection and suppression capability is provided in the design. Flooding SSAR 3.4.1, l
control features and sump level indication are provided in the design. 9.5.1.2.1.2, & ;
Compensatory measures are expected to be taken to maintain the detection and 9.5.1.8 I
suppression capability to allow specific maintenance activities.
I 16.
AP600 main control room fire ignition frequency is limited as a result of the use l SSAR 7.1.2 &
of low voltage, low current equipment and fiber optic cables.
7.1.3 l 17. i Redundancy $' control room operations is provided within the control room l SSAR 9.5.1.2.1.1 itself for fires in which control room evacuation is not required.
I 18.
The remots shutdown workstation provides redundancy of control and SSAR 7.3 & 9,5 l
I monitoring for safe shutdown functions in the event that main control room evacuation is required.
I l
I The remote shutdown workstation is in a fire and flood area separate from SSAR the 7.1.2, 1 main control room.
l 7.4.3.1.1.&
9A.3.1.2.5 l
T M gh0084 g Revision: 9 mdzat April 11,1997 59 221
- W hc w Iwim i
i Brian A. McIntyre,01:27 PM 2/1388, NRC Staff ACRS testimony frosi j t-Date: Fri,13 Feb 199313:27:42'-0500 To: TRQ@NRC. GOV, wch@nrc. gov From: " Brian A. McIntyre" <mcintyba@wesmail.com>
Subject:
NRC Staff ACRS testimony from 2/4/98 Cc: meintyba@wesmail.com Ted and Bill, It seems that the staff did say on the record that the WGOTHIC problems were not resolvable. What did I miss????
Part of the transcript from the 2/4/98 Advanced Plant ACRS subcommitte meeting MR. HUFFMAN: This is Bill Huffman, one of the P600 PM's. I have some slides that I could present. 'Ihey reiterate the open issues on GOTHIC, NOTRUMP and the testing program that Brian McIntyre discussed with you. We have no quarrels with the identified open issues. I guess my only point would be that -- that some of these are -- are not obviously resolvable, especially in the area of GOTHIC, and the staffis looking at them very hard, taking in advisement and in consideration what discussions have taken place at the ACRS thermohydraulic subcommittee meetings.
MR. McINTYRE: Bill, when you were talking, you thought the GOTHIC issues were not resolvable?
MR. HUFFMAN: I did not say that --
MR.Mc!NTYRE: Okay.
MR. HUFFM AN: - and I don't believe that's on the record.
MR.McINTYRE: Okay. l Brian A. McIntyre Bell 412.374.4334 WIN 284.4334
~ ~ ~
[IEintedk"B'rEn~nXMcIntyref <mciniyina_@ wesmall.com> ~1_1 J
J
); .
1
- FAX TO NRC February 4,1998 g/cc.
To: Bob Palla (NRC) ed 4 j der #4p d-%
f g
Joe Sebrosky (NRC)
From: Cindy Haag (Westinghouse)
Subject:
Proposed changes to AP600 PRA insights table to resolve FSER OI 720.439F.
Bob.
Per our telephone discussion yesterday (2/3/98), you requested that the following items be added or modified to the PRA insights table (Table 59 29). You also stated that these were the only items you saw as needing to be re. addressed in order to resolve this insights open item. The items we discussed as summarized below along
, with the proposed Westinghouse disposition of each. Also attached are the segments of the PRA insights table which cover the three items. I've included in the table all the pieces weNe promised to include for PRA Rev.11 (i.e., those items from 720.439F), and then hand-marked the proposed resolution modifications. Hope this is of assistance to you. Please review this material and let Joe Sebrosky (301-4151132) and me (Joe will reply your message to me) know if this information will resolve the 01. Once we have agreement on resolution, then I will issue a revision to the open item response to include the acceptable information. Thank you.
Regards, Cynthia Haag (4l2 374 4277)
Insight items NRC Requested:
- 1. Containment spray . modify insight #36 to state AP600 has a nonsafety.related containment spray system, and direct the disposition to the SSAR. 'Ihe rest of what was written in #36 was acceptab;e.
W Proposed Resolution: A sentence will be added to insight #36 with a disposition to SSAR subsection 6.5.2.
- 2. Reactor vessel insulation . need disposition to be included in the FSER OI response under "PRA Revision" discussion.
W Proposed Resolution: The disposition will be added as ~SSAR 5.3.5, and Cern) fed Design Material". Some of the information is within the SSAR and some is within ETAAC, thus the reason to mention both within the disposition column.
- 3. Protection from diffusion flames (related to 720.444F which requested an ITAAC) . Some diffusion flame vent information was added to SSAR and ITAAC, please include disposition in PRA insights table.
W Proposed Resolanion: Per the revised response to 720.444F (W lener DCP/NRCl240, dated 1/3068), the third item ofinsight #31 is now covered by SSAR 6.2.4.5 and Certified Design Material (includes ITAAC). The disposition column in the insights table will be noted as "Certyled Design Material"
. l
8 h
i Mop ch pryC'Ab ' f- M AP600 PRA BASED INSIGHTS INSIGHT DISPOSITION i 26. The reflective reactor vessel insulation provides an engineered flow path to PRA Chapter 39, 7
allow the ingression of water and venting of steam for externally cooling the @ $$AR vessel in the event of a severe accident involving core relocation to the lower plenum.
g, 3, Q
) CecW A h he reflective insulation panels and support members can withstand precsure gyggi b
,t M differential loading due to the IVR boiling phenomena.
PRV/ insulation panel clearances, water entrance and steam exit flow areas, and loss coefficients are based on scale test data from the ULPU facility.
Water inlets and steam vents are provided at the entrance and exit of the insulation boundary.
I No coatings are applied to the outside surface of the reactor vessel which will 4 inhibit the wettability of the surface.
l Reactor vessel insulation is an important SSC. De COL will maintain the
/P reliability of the insulation.
4 31. The containment layout prevents the formation of diffusion flames that can SSAR 1.2, General h challenge the integrity of the containment shell.
Arrangement
(
i Vents from compartments where hydrogen releases can be postulated Drawings away from the containment wall and penetrations or are hatched and locked i closed.
b IRWST vents near the containment wall are turned to direct releases away Qficcl j
( from the containment shell, t;g<;p Q h j l'
eb 36. APheo ha o. nens -rtOd MainM sp 554R (,.T c4 spwm.
Containment spray is not credited !: the PRA. Failure of the nonsafety related PRA Chapter 43 h containment spray does not prevent the plat.t achieving the safety goals.
g P
he COL will develop and implement severe accident management guidance PRA Chapter 59 h
for operemos of the nonsafety related containment spray system using the (59.10.6) j' suggested framework provided in WCAP 13914.
)
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February 5.1998 .
To: ' Joe Sebrosky, NRC .
cc: Ted Quay. NRC ~
Srian McIntyre, Westinghouse
Subject:
' Request NRC Status of Chapter 19 Open Items and Closure Path Joe.
I've taken an OITS printout (as of 2/4/98 pm) of Chapter 19 open items and created the attached table. The table does not list those items that W and NRC both agree are Resolved or Action W (based on 1/22 mtg or telecons).
Please let me know the NRC status for those items on the table listed as " Action W" and the status for those
' items listed as " Action N." Specifically, I'm trying to determine how the staff's review is progressing and how closure is coming with these open items since the Jan. 22 meeting, and to determine if there are any problem areas. Remember as discussed with you and Teo, Westinghouse does not plan to provide PRA Revision 11 until we hear from the staff on large issues such as shutdown PRA and PRA insights. It should be noted that the staff has markups of new/ revised information going into PRA Revision i1.
Please inform Westinghouse of the Chapter 19 open item closure status.
Regards, Cynthia Haag Westinghouse I
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@ West'nghbuse .
FAX COVER SHEET Recipient Information Sender Information Date: g/g/pg Name: D. Lindgren To: Bill Huffman Location: WEC E 330 Phone: Phone: (412) 374-4858 (301) 415-1141 Facimile: Facimile: Bet (412) 374 4887 (310) 415-2002 Company: U.S.NRC WIN 284 4887 Location: Rockville, MD Cover + Pages 1+ l The following pages are being sent from the Westinghouse Energy Center, Monroeville, PA.
If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374-6529.
comments:
S 1 L, t. ,
TLfASE S62 ip YHFSE eNM& 6S wiu SYS*^ U E 'r@ M) E d o g sTlaius oa c. v s mum I
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Revise the first paragraph of subsection 5.2.3.1 as' follows.
Table 5.2-1 lists material specifications used for the principal pressure-retaining applications in Class 1 primary components and reactor coolant system piping. Material specifications with grades, classes or types are included for the reactor vessel components, steam generator components, reactor coolant pump, pressurizer, core makeup tank, and the passive residual heat removal heat exchanger. Table 5.2-1 lists the application of nickel-chremium-iron alloys in the reactor coolant pressure boundary. De use of nickel-chromium-iron alloy in the reactor coolant pmssure boundary is I'mited to Alloy 690.
Alloy 600 may be usu. for cladding or buttering. Steam generator tubes use Alloy 690 in the thermally treated form. Nickel-chromium-iron alloys are used where corrosion resistance of the alloy is an important consideration and where the use of nickel-chromium-iron alloy is the choice because of the coefficient of thermal expansion.
Subsection 5.4.3 defines reactor coolant piping. See subsection 4.5.2 for material specifications used for the core support structures and reactor internals. See appropriate sections for intemals of other components. Engineered safeguards features materials are included in subsection 6.1.1. De nonsafety-related portion of the chemical and volume I' control system inside containment in contact with reactor coolant is constructed of or clad 1.l with corrosion resistant material such as Type 304 or Type 316 stainless steel or material I . with equivalent corrosion resistance. that-e-ne materials are compatible with the reactor coolant pxx : 5:;9 j. De nonsafety-related portion of the chemical and volume control system is not required to conform the process to requirements outlined below.
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r 1-Westinghoose FAX COVER SHEET W
RECIPIENT INFORMATION SENDER INFORMATION DATE: fes4a d $~ /99.8 NAME: L@g TO: LOCATION: ENERGY CENTER -
EAST Duno Scmvrrt PHONE: FACSIMILE: PHONE: Omce: #tt-s n/-sz9o COMPANY: Feceimile: win: 284 4887 Lt s Mec- outside: (412)374 4887 LOCATION:
Cover + Peges 1+/ )
l The fogowing pages are being sent from the Weetinghouse Energy Center, Esot Tower, 1 Monroevsle, PA. If any problems occur during this tronomleelen, please cell:
WIN: 344125 ( Bey ) er Outside:(412)374-6529 COMMENTS:
D.ve TL. aLLJ e.Lo s k/d ena to< G L; s h ai c. - . - - an o,_ r% vt.o. 2as to m 400). 23 m.// 4 ,,,,,,.ou ,,4 rrA4 /2-o - z. I a Jan w < L., &~ faa, A-n
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- 11. Radioactive Wast) M: nag: ment 2 g
I the auxiliary building provide more than 2 years of spent resin storage at the expected rate l cf more than 6 months of storage at the maximum generation rate.
l l The dry solid radwaste includes 1246 cubic feet per year of compactible and non-compactible I waste packed into about 13 boxes (90 cubic feet each) and ten drums per year. Drums are i used for higher activity compactible and non-compactible wastes. Compactible waste includes i HVAC exhaust filter, ground sheets, boot covers, hair nets, etc. Non-compactible waste I includes about 60 cubic feet per year of dry activated carbon and other solids such as broken i tools and wood. Solid mixed wastes will occupy 7.5 cubic feet per year (one drum). De low I activity spent filter cartridges may be compacted to fill about 0.26 drums per year (2 ft'/ year)
I and are stored in the packaged waste storage room. Compaction is performed by mobile I equipment or is performed offsite. High activity filter cartridges fill three drums per year I (22.5 cubic feet per year) and are ston:d in portable processing or storage casks in the rail car I of the auxiliary building.
l The total volume of solid radwaste to be stored in the radwaste building packaged waste I . torage room is 1256 cubic fett per year at the expected rate and 1922 cubic feet per year at I the maximum rate. His includes the dry wastes packaged in drums or steel boxes, the mixed I solid chemical wastes and the lower activity filter cartridges. De quantities of liquid I radwaste stored in the packaged waste storage room of the radwaste building consists of 20 l cubic feet of chemical waste and 17 cubic feet of mixed liquid waste. The useful storage l volume in the packaged waste storage room is approximately 3900 cubic feet (10 feet deep, l 30 feet long, and 13 feet high he packaged waste storage room provides storage for about i 3 years at the expected rate f generation and more than 2 years at the maximum rate of I generation. One four-dru containment pallet provides more than 8 months of storage I capacity for the liquid m' ed wastes and the volume reduced liquid chemical wastes at the I
expected rate of genera on and more Jsseemjh than 4 months owet% en at the IIo maxi urp rg& *9 *Ll,, f us; =
s b tesder t m k.
I A conservative estimate of solid wet waste includes blowdown,m, aterial based on continuous operation of the steam generator blowdown purification system, with leakage from the primary I to secondary cycles. He volume of radioactively contaminated material from this source is estimated to be 540 & fm per year. Provisions for processing and disposal of radioactive i steam generator W W o usins and membranes are described in subsection 10.4.8. Note I that, although incl.g we for conservatism, this volume of contaminated resin will be I removed from the phc.. within the contaminated electrodeionization unit and not stored as I wet waste.
De condensate polishing system includes mixed bed ion exchanger vessels for purification of the condensate as described in subsection 10.4.6. Should the resins become radioactive, the resins are transferred from the condensate polishing vessel directly to a temporary processing unit or to the temporary processing unit via the spent resin tank. The processing unit, located outside of the turbine building, dewaters and processes the resins as required for offsite disposal. Radioactive condensate polishing resin will have very low activity. It will
. be disposed in containers as permitted by DOT regulations. After packaging, the resins may be stored in the radwaste building. Based on a typical condensate polishing system operation of 30 days per refueling cycle with leakage from the primary system to the secondary system, Revision: 13 3 Westinghous8 11.4-5 May 30,1997
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- TX CONFIRNATION REPORT ** AS OF FEB 5 '98 12:21 PAGE.01 AP600 DESIGN CERT DATE TIME T0/FROM MODE M1H/SEC PGS STATUS 01 2/ 5 12:19 3014152300 G3--S 01"26 02 OK
e- ,'
t FAX COVER SHEET Westinghoose W !
RECfPIENT INFORMATION SENDER INFORMATION DATE: FFa. o2.1998 NAME: Lgm TO: LOCATION: ENERGY CENTER -
Ec 4tein EAST D ett o PHONE: FACSIMILE: PHONE: Omco. Msz- 37V-rz po COMPANY: Feceimde: win: 284-4887 Lt 5 Alec. outsede: (412)374-4887 3 LOCATION: f I
{
Cmer + Pages 1 + S~ f Rf'5L6th The fogowing pages are being sent from the Westinghouse Energy Center, East Tower, Monroewme, pA. N eny problems occur during thle tronomleelen, pieseo cell:
I WIN: 344125 ( Bev ) or Outelde: (412)374 -6529 COMMENTS:
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Costrel of R=rhactive Materials As secsed in the fire prottriion analysis, rusterials that collect or contain radioactivity, such as spent ion exchange resins and filters, are protected acd stored in accordance with BTP C M E B 9.5-1.
9.5.1.2.1.2 Fire Detection and Alarm Sy*==
Fire detection s A alarm systems are provided where r'ximred by the fire protection analysis, I in accordance with BTP ChEB 9.5-1, NFPA 803, NFPA 804, and NFPA 72 (Reference 2).
Plie detectors respond to smoke, flame, beat, or the products of combustion. The installation of fue detectors is 'in accordance with NFPA 72 (Reference 2) and the mm#s-turer's recoaunendations. The selection and installation of fire desactors is also based on consideranon of the type of hazard, combustible loading, the type of cornbustion products, and detector response characteristics. The types of detectors used in each fire area are identified in the fue protection analysis.
The fue detection system provides audible and visual alarms and system trouble annunciation in the main control room and the security central alarm station. Annunciation carcuits connecdng zone, main, and remote annunciadon panels are electrically supervised.
Each fue detection, indicating and alarm unit is provided with reliable ac electrical power from the non-Class IE unintenuptible power supply system. This system is described in subsection 83.2.1.2.
93.1.2.13 Mre Water Supply System The fue water supply system is designed in accordance with BTP CMEB 9.51 and the applicable NFPA standards. 6 q rd 4ke- camMnme,& Sprg Fire waser is supplied from two separate fresh water storage ranh. 'Ihe primary fue w sislam .
tank is ded2cated to the fue protection system. The secondary fire waser tank serves the ra water systern but contains e "" m'- M water for use by the fue prosecuon system i There are two 100-percent capacity fire pumps. 'Ihe lead pump is electric motor <iriven and the secondary pump it diesel engine-driven. A motor 4 riven jockey pump is used to keep the fire water system full of water and pressurized, as required. For additional information regarding the fire water tanks and pumps, see subsection 9.5.1.2.3.
The fue water tanks are permanently connected to the fue pumps suction piping and are ananged so that the pumps can take suction from either or both tanks. Piping between the fus water sources and the fire pumps is in accordance with NFPA 20 (Reference 2). A failure in one taak or its piping cannot cause both tanks to drain.
i i
1 Revision: 18 )
T W WhWloung 9.5-7 December 5,1997 i i
JAN 26 '98 17819 8 *
,, r.u o c. t. . s u m .s sua m, a c, so , a . a s i. o . v u r .w vo
, 9. Antnfary Systems l'
Fire protection water is distributed by an utv.ktground yard main loop, designed in acconiance with NFPA 24 (Reference 2). The yard main includes a building interior header that distributes wiser to suppression systems within the main plant buildings. Indicator valves provide secionalimd control and permit isolation of pomons of the yard main for maintenance or repair. An indicator vahe also separates the irdividual fire pump connections te the main.
Sprinuer and nandpipe systems are supplied by connections from the yard main. where plant I
areas, other than the containment and oudying buildings, are protected by both spinkler systems and standpipe systems, tbc connections from the yard main are arranged so that a single active failure or a crack in a moderate energy line cannot impair both systems.
Manual valves for sectiembred control of the yard main or for shutoff of the war.er supply to suppression systems are electncally supervised iflocated above gmund and administratively contmiled if l<v ared underground.
Hydrants are provided on the yard main in accordance with NFPA 24 (Reference 2), at intervals of up to about 250 feet. They provide bose strearn protection for every part of each buildmg and two hose streams for every part of the interior of each building not covered by standpipe protection, exclading certain ternote areas of the shield building. The lateral to each hydrut is controlled by an isolation vahe.
IIose houses are in sa:oniance with NFPA 24 (Reference 2). They are located at intenals of not more than 1000 feet along the yard main.
Outdoor fire water piping and water suppression systems located in unheated areas of the plaat are potected frtxn freezing.
A permanent connection between the fire protection system and the component cooling water Cp f' hr$s Teb$h W -md &* 'k uL h4 Cernkamrd' w.h.Y*ro.a.i)a 9.5.1.2.1.4 At yarepression Systems 7 = am m sm .
Automatic fire suppression systems are in accordance with Bl? CMEB 9.5-1 and the applicable NFPA standards, with consideration of the unique aspects of each application, including building characteristics, matenals of construction, environmental conditions, fire area contents, and adjacent structures.
Fixed automatic fire suppression systems are provided based on the results of the fire yw-ciics analysis.
The erlemne of automatic suppression systems for each plant area ts based on the guidance 1
of NFPA 803 and NFPA 804 (Reference 2). Water systems are p:cferred, but the use of automatic water suppression systems for firefighting in radianon areas is minimized because of the possible spread of contarnination. Haloo and carbon dioxide fixed flooding systems att not used.
1 Revision: 18 i i
hbor $,1997 9.5-8 y Westhgh0058 JAN 26 '98 178tl 3014152300 PAGE.003
,. re i t. L s v A - 4 .u s u v son o ao io.2v n o . vv e r . v.a vo
- !!L""""**
- 6. F f - - : Safety Feenns 6.5 F1ssion Prodnet Remoyni and Control Systams 6.5.1 Eastseered Safety Feature (ESF) Mhar Systems his subsection is not applicable to the AP600 6.5.2 Cont *=+st Spray System i
In the event of a design basis LOCA there is an assurted core degradation that results in a i
signi6 cant telease of radioactivity to the contamment atmosphere. %is activity would consist I
of noble gases, particulares, and a small amount of elemental and organic iodine (as discussed I
in subsection 15.6.5.3, most of the iodine would be in the particulate form). He AP600 does i
i not include a safety.related containment spray system to remove airborne particulates or I
elememal iodme. Removal of altbome activity is by natural processes that do not depend on sprays (that is sedimentation, diffusiophoresis, and thermophoresis). %ese removal I
rnechanisms are discussed in Appendix 15B.
Much of the non-gaseous airborne activity would eventually be deposited la the containment i
I sump solution. Long term retention ofiodine in the containment sump following design basis accidents requires adjustment of the sump solution pH to 7.0 or above. ~Ihis pH adjustment is accomplished by the passiw con: cooling s ' ystem and is discussed in subsection 6J.2.1.4.
l l
In accordance with Referena 1, the Src protection system provides a nonsafety.related l l j l
containtnent spray timetion for accident management following a severe accident. This design j I
featme is not safety-related and is not credited in any accident analysis including the dose I
analysis provided in section 15.6.5. Dose reduction following a severe accident may be '
1 mba-d over the natural removal mechamsms via the consafety.related containment spray.
Subsection 15.6.5.3.2 provides additionaldiscussion of the namral removal mechanisms. The '
I I
following subsections provide a discussioo of the nonsafety related containment spray functio provided by the fire protection systern.
I I 6.5.2.1 System Description i
I
%c fire protection system provides a ncmsafety related containment spray function for severe I a I
acew t management. Subsection 9.5.1 provides a description of the fire protection system inclnding equipment and valves that support the containment spray function such as the fu' n I !
i potaps and fue main header. This section provides the descripdon of the portion of the fire protection system designed spect5cally to provide the containment spray function.
I I
i De source of water for the containment spray function is provided by asher the ,+ ~j ^--
i secondary fire protection system water tank.TMr;; = rl d irr q d'Z Iq. 1 i
Edher the motor driven or diesel dnves Sie protection systern pump rnay be used to deliver Ste woor to the containment spray bender. De flow path to containment is via the nonnal l I
I fue main header as shown in Figure 9.5.1 1, sheets I through 3. De containment spray flow path is from the fire main extension, through the 6te protection system !!ne that penemucs Revision: 17 UEMm'** 63 1 October 31,1997 3014t52300 PAGE.004
, ,r.u.s w ,1s w a a a 'o '**' " ' ' ' ' - '**
g . - Sa,et, a_
l enatamment, to the containtnent spray riser that connects to the fire pnxection system header I
inside contamment, nis riser supplies two ring headers located above the containment I polar crane.
I I 6.511.1 Valves l
I I
The containment spray flow path from the fire main header contains one normally open reannal valve (FPS-V048). one normally closed manual valve (FPS-V101), one lock cicsed I
manual cont =iam isolation valve outside containment (PCS-V050), a contamnent isoladoo I
check valve inside contamment (FPS V052), a norma!!y open manual isoladon valve in the I
spray riser (FPS-V700), and a normally closed remotely 4parated valve (FPS V701)
I downstream of the manualisolation valve in the gy A,. .[
l Contamment spray is initiated by first cl ng the passive containment cooling water system i
fire be*ier isolation reactor (PCS V005).
l by opening the remotely <perated valv, opening the rnanual valves outsxie l contamme I
e inside containment. The manual valves outside contamment are located in valve / piping penetrabon room 12306. The valves are located I
close to the entrance door such that radiation exposures to an indiddcal required to enter the I
toom and align the valves would nor *.xceed the prescribed post-accident dose limits di-d I in subsection 12.4.1.8.
1 I
Valve FPS-V701 is a fail + pen air-operated valve such that the conenmment spray flow path I
it can be opened following a loss of the nonsafery-related compressed air system. During i
shatdown operations, the fire protecdon system hender inside containment is prest.urued from I
the passive containment cooling water storage tank for fire protection and manual isolation i valve FPS-V700 is closed.
I i 6.5.2.2.2 Containement Spray Hender and Nonles I
%e contammert spray header consists of a single header that feeds two ring henders located I
above the containment polar eme. The containment spray ring beaders and spray nozzles are l i oriented to maxumze containment volurne coverage. A lower ring headeris locatec at plant '
I elevadon 235 feet, and contains 44 spray nozzles. An upper ring header is located at plant i elevation 250 feet, and contains 24 spray nozzles.
1 I
Be nozzles within the spray ring header are conventional containment spray nant** utilized I
la past Westinghouse pressurized water reactors. The spray nozzles are selected on the basis I
of drop size to provide adequate absorption of fission products from the containtnent I annosphere.
I I 6.511.3 Appaleable Codes and MRe=%
1 I I De containment spray function is not safety-related, and therefore the valves and piping in !
I the containment spray flow path are not required to be safety relaied for the containment spray I function. However, the conuunment isolazion piping and valves are safery-relaand (AP600 Revision: 17 l
Odober 31,1997 Westingh0088 6.5-2 JAN 26 '98 17t12 30141 MM N+M
,, t, x t< stL.sutoi w uu son to so io:At ru .vuz r .um vo 6 Engineered Safety fenaurus I
i Equipment t
Class B) to perform the safety.rdated ftmetion of cootainment The isolatioet.
I cf=**Nion of the remaining portions of the fire neader are noosafety related, and a I classd5ed as Class F as discussed in subsecuons 3.2.2.7 and 9.5.1. The con I header and valve, downstream of the manual isolation valve inside containme related and ela**WA as Class E.
I category D. The containment sprzy header is classified as Seismic I
i 6.5.2.1,4 Systems Operation i
i I During normal operacion. the fire protection system header inside containment is isola I the fire main he=W. the containment spray piping is therefore not pressurized du I operation. During plant shutdown rnodes. personnel access to containment is require I as such, the fire protection system standby header inside containrr.ent is pressuriz I water in the passive contenment cooling water storage tank. During these modes, man i (solation valve FPS-V700 is closed to Arther isolate the cor.tainment sway heade passive containment cooling water storage tari.
I free he. scee.dag Mrs I
Severe accident management guidelines provide the weMc4cnk I perator with guidance to initiate the I containmrsit spray feature of the fire protection syste . Operator action to open two manua I isolation valves outside of contamment followed by motely opening the cocemn I isolation valve within containment from either the m workstation will f aitiate the spray function. Coo control room or the remote shutdown i at spray may be termman*4 at any time by closing the remotely operused isolation valve wi I
manual valves in the containment spray flow pet. i outside containment.' containment, or by I Operation of the containment spray will have no effect on the a I vailability of the remainder of the fire 1
protection systern other than the loss ofinventory[due to the sprayed water.
ince the fire protection system operates in the active *=# l I
pr==ari=t ooce the remotely 1 mode. i e h xM 4,, kept full and I isotation valve is opermj the stem perf the comala-nt spmy Nnction.
eesane, sityeml*W h I
I 1,}4ee wn yank s s sseldedigone.n
. , . , dug M (e
{ smopersbon.
I When water pressure in the fue main begins to fall, due to a demand for water fmm I containment spray, the motor driven pump starts automadcally on a low pressure signa I the rnocor-driven pump fails to start, the dieselwiriven pump starts upon a lower pre I
alsnal The pump continnes to nm until it is stopped manually.
1 6.5.2.2 Design baluation l
I 6.512.1 (%=tal= mane Coverage I
i l
i he containment sprr/ nonles are the Lachler (SPRACO Company) spray nozzles or !
I equivalect, which provide a drop size discibunon which has been established by test I found suitable for fission product removat ne fire protection systern header provides I contain=ar spray oonle differential pressure of 40 psid, which fixes the drop site I discibution. The unass rneen drop size produced at this differennal pressure is conse assamed to be 1000 micreas.
Revidea: 17 6.5-3 October 31,1997 JAH 26 '90 17813 3014152300 PAGE.006
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T ( ,/ l Westingtiouse Energy System sa ns Electric Corporation *,,, "c"'r 4*an8 W m
- e# DCP/NRCl227
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,( Docket No.: 52-003 I
% January 23, 1998
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% S Document Control Desk U.S. Nuclear Regulatory Commission
%'p ka Washington, DC 20555
%, g& j p e y
S ATTENTION: T.R. QUAY o
SUBJECT:
AP600 RESPONSE TO FSER OPEN ITEMS
Dear Mr. Quay:
Enclosure 1 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 j item number, the associated OITS number, and the status to be designated 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 on (4 ') _74-4334 if you have any questions concerning this transmittal.
d~. / /Y' Brian A. McIntyre, Manager Advanced Plant Safety and Licensing .
jml Enclosure ee: W. C. Huffman, NRC (Enclosure)
T. J. Kenyon, NRC (Enclosure) ;
J. M. Sebrosky, NRC (Enclosure)
D. C. Scaletti, NRC (Enclosure)
N. J. Liparuto, Westinghouse (w/o Enclosure)
0 l
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NRC FSER OPEN ITEM Question 280.33F (OITS 6499) ,
l Fire Protection Water Supply I The fire protection water supply for the AP600 is not dedicated for fire protection purposes. The SSAR markup states that the fire protection water supply is used to supply the containment spray following a sesere accident. This additional demand on the fire protection water supply was not included in presious revisions of the SSAR submitted by the applicant for staff review. The existing fire protection water supply cannot meet the demand of both the fire protection system and containment spray simultaneously with either one fire pump or one water storage tank out of service.
This item remains open.
Response
SECY-97 044, " Policy and Key Technical issues Pertaining to the Westinghouse AP600 Standardized Passive Reactor Design" was issued by the staff on February 18,1997. The purpose of the SECY is to request Commission approval of a staff position requiring additional nonsafety-related system (s) that would achieve an appropriate balance between prevention and mitigation of severe accidents for the AP600. In desenbing an acceptable containment spray system design for the AP600, the staff states:
"The :taff beheves that the containment spray system could be similar to the ac-independent water addition system (AICWA) system in the Advanced Boshng Water Reactor (ABWR) design. The AICWA system consists ofpiping and manual valves connecting thepre protection system to the upper devuell containment spray ring header. The piping and manual valves are of a quahty to ensure containment isolation. The diesel-drivenprenater pumpfor the ABWR provides 1500 gpm flow at a differential pressure of 12S pounds per square anch (guage) (psig)."
"The AP600 pre protection inchedes a diesel-driven prewater supply system pump that is ratedfor
- 000 gym. A containment spray system supphed by such a system would considerably reduce the i imcertamty associated with severe accident aerosolpssion product removal. The spray coverage l area and the number ofspray ringh headers that could be supphed would have to be appropriately estabbshed consistent with this flow rate."
The Commission approval of this SECY in their Staff Requirement Memorandum of June 30.1997 pernuts the use of the AP600 fire protection system to enhance severe accident management. The use '
of containrr.ent spray during a severe accident does not limit the availability of the fire protection system during normal operations, or following an accident. Guidance for the use of containment spray for a sesere accident is provided by the Sescre Accident Management Guidelines which are the responsibility of the COL applicant. The decision to use the fire protection system for containment l spray will be based on the, set of circumstances that are present following the severe accident. For l csample, if a fire water storage tank is out of sersice, the decision to use containment spray is l weighed against the need for fire protection. and the availability of other water sources for either fire protection or containment spray.
2so.33M T Westinghouse
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E NRC FSER OPEN ITEM It is important to note that many operating plants also utilize their fire protection system in their Sesere Accident Management Guidelines as a means for containment spray or RCS cooling via the steam generators. Therefore, the use of the AP600 fire protection system for containment spray is consistent with its use in the severe accident management for current plants.
It should also be noted that the fire protection for safety related portions of the AP600 are seniced by the other fire water sources such as the passise containment cooling system water storage tank and and ancillary water storage tank. These tanks m aligned to provide fire protection of the safety-related portions of the plant, and do not rely on the availability of the fire water storage tanks that could be used for containment spray.
Finally, if the containment spray were in use, the fire protection system header would still be functional, even if one fire pump or fire water storage tank was out of service. The fire protection system header preferentially feeds the fire system header, and it would be available for fire protection during or after containment spray operation if necessary.
In discussions with Westinghouse, the NRC staff has indicated that it is their opinion that the AP600 is in siolation of Branch Technical Postion 9.5-1, Position C.6 b.10 which states that when a common tank is shared between fire water storage and the sanitary or service water system, a minimum fire wster storage tanks volume should be preserved for fire protection system by passive means.
Westinghouse agrees that such an approach is warranted for normally operating systems such as those specified in Position C.6 b 10 to avoid frequent challenges to the fire water storage insentory.
Howeser. the use of the fire water for containment spray is not comparable to service water,in that containment spray is only used to assist severe accident management, and is not required to meet the offsite dose limits following a sesere accident. Therefore Westinghouse does not beliese that AP600 is in uolation of BTB 9 51 Position C.6 b.10.
In conclusion, precedence exists for the use of the fire protection system for containment spray in sescre accidents Currently operating plants as well as the ABWR have connections to the fire protection sy stem for sescre accident management. In addition, the NRC Commission approsal of SECY 97 044 identified this as an acceptable system design for the AP600.
SSAR Resision:
None 280.33F 2 W-Westinghouse
l NRC FSER OPEN ITEM Question 280.37F (OITS 6502)
Spent Fuel Cooling The applicant has not provided adequate fire protection to ensure that spent fuel pool ecoling sy stem j (SFS) will be operable following a single fire. Automatic fire suppression has not been provided for SFS. The SFS cooling pumps are located in the same fire area as the rail bay and spent resin storage rooms. These areas are considered significant fire exposure hazards to the SFS and have not been provided with automatic suppression. This item remains open.
Response
The nonsafety-related SFS cooling pumps are n_pj located in the same fire area as the rail bay and spent resin storage rooms. The SFS cooling pumps are in fire area 1200 AF01 while the rail bay and the spent resin storage rooms are in fire area 1200 AF02.
Branch Technical Position 9.5-1 does not require automatic fire supression of the fuel pool area.
Specifically, Position 7 m specifies that protection for the spent fuel pool area should be provided by local hose stations and portable extinguishers. Automatic fire detection should be provided to alarm and annunciate in the control room and to alarm locally. The AP600 meets this criteria.
Unlike some operating plants, the AP600 SFS is not safety-related. Safety-related makeup to the spent fuel pool is provided to maintain spent fuel cooling in the unlikely event that the SFS becomes unavailable. The safety related makeup is unaffected by fires that could damage the SFS pumps. The SFS does provide defense in-depth by providing redundant pumps and heat exchangers and is designed to withstand a single active failure and maintain spent fuel pool cooling.
SSAR Resision:
None l
1 1
l ll 280.37F-1 1
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@ Westinghouse FAX COVER SHEET Recipient Information Sender Information Date: , j'y /f g Name: D. Lindgren To: Joe Sebrosky Location: WEC E 330 Phone: Phone: (412) 374 4856 (301) 415-1132 Facimile: Facimile: Bel: (412) 374 4887 (310) 415-2002 Company: U.S.NRC WIN 284-4887 Location: Rockville, MD Cover + Pages 1+7 f The following pages are being sent from the Westinghouse Energy Center, Monroeville, PA.
If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374-6529.
comments:
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- 3. Design of Struct:res, Components, Eq 1pme t, rnd Systems guide the operator on a timely basis to determine if the level of canhquake ground motion requiring shutdown has been exceeded. He procedures will follow the guidance of EPRI Reports NP 5930 (Reference 1), TR-100082 (Reference 17), and NP-6695 (Reference 18), as modified by the NRC staff (Reference 32).
3.7.5.3 Seismic Interaction Review The seismic interaction review will be updated by the Combined License applicant. This review is performed in parallel with the seismic margin evaluation. The review is based on as-procured data, as well as the as-constructed condition.
3.7.5.4 Reconciliation of Seismic Analyses of Nuclear Island Structures The Combined License applicant will reconcile the seismic analyses described in subsection 3.7.2 for detail design changes such as those due to as-procured equipment information.
Deviations are acceptable based on an evaluation consistent with the methods and procedure of Section 3.7 provided the amplitude of the seismic floor response spectra including the effect due to these deviations, do not exceed the design basis floor response spectra by more than 10 percent. If it is necessary to update the soil structure interaction analyses, these analyses should be performed with site specific soil properties using seismic input defined by the response spectra given in Figures 3.7.1-1 and 3.7.12.
3.7.6 References
- 1. EPRI Report NP-5930 "A Criterion for Determining Exceedance of the Operating Basis Earthquake," July 1988.
- 2. Uniform Building Code,1991.
- 3. ASCE Standard 4-86, " Seismic Analysis of Safety-Related Nuclear Structures and Commentary," American Society of Civil Engineers, September 1986.
- 4. ASME B&PV Code, Code Case N-411.
- 5. H. B. Seed, and I. M. Idriss, " Soil Moduli and Damping Factors for Dynamic Response Analysis," Report No. EERC 70-14 Earthquake Engineering Research Center, University of Califomia, Berkeley,1970.
- 6. H. B. Seed, R. T. Wong, I. M. Idriss, and K. Tokimatsu, " Moduli and Damping Factors for Dynamic Analysis of Cohesionless Soils," Report No. UCB/EERC-8914, Eanhquake Engineering Research Center, University of California, Berkeley,1984.
- 7. Bechtel Corporation, " User's and Beoretical Manual for Computer Program BSAP (CE800)," Revision 12,1991.
Revision: 20
[ Westingh0US8 3.7-53 February 6,1998
l I
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, 3. Design ef Structures, Ccmponerts, Eq:1pmert, cnd Systems 3.83.5.6 Steel Form Modules The steel form modules consist of plate reinforced with angle stiffeners and tee sections as shown in Figure 3.83-16. The steel form modules are designed for concrete placement loads defined in subsection 3.83.3.2.
The steel form modules are designed as steel structures according to the requirements of AISC-N690. This code is applicable since the form modules are constructed entirely out of stmetural steel plates and shapes and the applied loads are resisted by the steel elements.
3.83.5.7 Design Summary Report A design summary report is prepared for containment internal structures documenting that the l structures meet the acceptance criteria specified in subsection 3.8.3.5.
Deviations from the design due to as-procured or as-built conditions are acceptable based on an evaluation consistent with the methods and procedures of Section 3.7 and 3.8 provided the following acceptance criteria are met.
. the structural design meets the acceptance criteria specified in Section 3.8
- the seismic floor response spectra meet the acceptance criteria specified in subsection 3.7.5.4 Depending on the extent of the deviations, the evaluation may range from documentation of an engineering judgement to performance of a revised analysis and design. j l
3.8.3.5.8 Design Summary of Critical Sections 3.8.3.5.8.1 Structural Wall Modules This subsection summarizes the design of the following critical sections:
- South west wall of the refueling cavity (4' 0" thick)
- South wall of west steam generator cavity (2' 6" thick)
- North east wall of in-containment refueling water storage tank (2' 6" thick) I i
The thicknesses and locations of these walls which are part of the boundary of the in-containment refueling water storage tank are shown in Table 3.83-3 and Figure 3.8.3-18.
They are the portions of the structural wall modules experiencing the largest demand. The stmetural configuration and typical details are shown in Figures 3.83-1,3.83-2. 3.8.3-8, 3.83-14,3.83-15, and 3.8317. The structural analyses are described in subsection 3.8.3.4 and summarized in Table 3.83-2. The design procedures are described in subsection 3.83.53.
Revision: 20
[ W85tingh0Use 3.8-37 February 6,1998
l 1
\
- 3. Design of Structures, Cemponents, Eq11pmert, cnd Systems The three walls extend from the floor of the in-containment refueling water storage tank at elevation 103' 0" to the operating floor at elevation 135' 3" ne south west wall is also a boundary of the refueling cavity and has stainless steel plate on both faces. The other walls have stainless steel on one face and carbon steel on the other. For each wall design j information is summarized in Tables 3.8.3-4, 3.8.3-5 and 3.8.3-6 at three locations. Results I are shown at the middle of the wall (mid span at mid height), at the base of the wall at its mid point (mid span at base) and at the base of the wall at the end experiencing greater demand (corner at base). The first pan of each table shows the member forces due to individual loading. The lower part of the table shows governing load combinations. The steel plate thickricss required to resist mechanical loads is shown at the bottom of the table as well as the thickness provided. The maximum principal stress for the load combination including thermal is also tabulated. If this value exceeds the yield stress at temperature, a supplemental evaluation is performed as described in subsection 3.8.3.5.3.4; for these cases the maximum stress intensity range is shown together with the allowable stress intensity range which is twice the yield stress at temperature.
I 3.8.3.5.8.2 In Containment Refueling Water Storage Tank Steel Wall l
The in-containment refueling water storage tank steel wall is tne circular boundary of the in-containment refueling water storage tank. The structural configuration and typical details are shown in sheet 3 of Figure 3.8.3-8. The structural analyses are described in subsection 3.8.3.4 and summarized in Table 3.8.3-2. The design procedures are described in subsection 3.8.3.5.3. The steel wall extends from the floor of the in-containment refueling water storage tank at elevation 103' 0" to the operating floor at elevation 135' 3" The wall is a 5/8" thick stainless steel plate. It has internal vertical stainless steel T-section columns spaced 4'-
8" apan and external hoop carbon steel (L-section) angles spaced 18" to 24" apan. The wall is fixed to the adjacent modules and ficor except for the top of columns which are free to slide radially and to rotate around the hoop direction.
The structural evaluation is performed separately for the central and end regions. The central region envelopes results for the wall except for the last four columns at each end. The end region envelopes results for the four columns at each end. The wall is evaluated as venical and horizontal beams. The vertical beams comprise the T-section columns plus the effective width of the plate. The horizontal beams comprise the L-section angles plus the effective width of the plate. The evaluations are summarized in Table 3.8.3-7. Design loads and load combinations are shown on sheet 1. Sheet 2 shows the ratio of the design stresses to the l allowable stresses. When thermal effects result in stresses above yield, the evaluation is in accordance with the supplemental criteria as described in subsection 3.8.3.5.3.4.
3.8.3.5.8.3 Column Supporting Operating Floor l
This subsection ;ummarizes the design of the most heavily loaded column in the containment internal structures. De column extends from elevation 107'-2" to the underside of the operating floor at elevation 135'-3" In addition to supporting the operating floor, it also supports a steel grating Door at elevation 118'-0" l
Revision: 20 February 6,1998 3.8-38 W Westinghouse
F-4 i 3. Desigm of Structures Components, Equipment, rnd Systems -
He load combinations in Table 3.8.4-1 were used to assess the adequacy of the column. For load combination 1 in the table, the interaction factor due to biaxial bending and axial load is 0.38. For load combination 6 without thermal loads, the interaction factor is 0.42 and with thermal loads the interaction factor is 0.61. Since the interaction factors are less than 1, the column is adequate for all the applied loads.
3.8.3.6 Materials, Quality Control, and Special Construction Techniques Subsection 3.8.4.6 describes the materials and quality control program used in the construction of the containment internal structures. The structural steel modules are constructed using A36 plates and shapes. Nitronic 33 (American Society for Testing and Materials 240, designation S24000, Type XM-29) stainless steel plates are used on the surfaces of the modules in contact with water during normal operation or refueling. The structural wall and floor modules are fabricated and erected in accordance with AISC-N690. Loads during fabrication and erection due to handling and shipping are considered as normal loads as described in subsection 3.8.4.3.1.1. Packaging, shipping, receiving, storage and handling of structural modules are in accordance with NQA-2, Pan 2.2 (formerly ANSI /ASME N45.2.2 as specified in AISC N690).
3.8.3.6.1 Fabrication, Erection, and Construction of Structural Modules Modular construction techniques are used extensively in the containment internal structures (Figure 3.8.3 1). Subassemblies, sized for commercial rail shipment, are assembled offsite and l
transported to the site. Onsite fabrication consists of combining the subassemblies in j structural modules, which are then installed in the plant. A typical modular construction i I
technique is described in the following paragraphs for Module MI, which is the main structural module in the containment internal structures.
The M1 module is a multicompartmented structure which, in its final form, comprises the central walls of the containment internal structures. The vertical walls of the module house the refueling cavity, the reactor vessel compartment, and the two steam generator compartments. The module (Figure 3.8.3-14) is in the form of a "T" and is approximately 50 feet long, 65 feet wide and 60 feet high. The module is assembled from about 40 l prefabricated wall sections called structural submodules (Figure 3.8.3-15). The submodules are designed for railroad transport from the fabricator's shop to the plant site with sizes up to 12 feet by 12 feet by 80 feet long, weighing up to 80 tons. A typical submodule weighs between 9 and 11 tons. The submodules are assembled outside the nuclear island with full penetration welds between the faceplates of adjacent subunits. The completed M1 module is lifted to its final location within the containment vessel by the heavy lift construction crane.
Following placement of the M1 module within the containment building, the hollow wall structures are filled with concrete, forming a portion of the structural walls of the containment intemal structures.
Tolerances for fabrication, assembly and erection of the structural modules conform to the requirements of section 4 of ACI-ll7, sections 3.3 and 3.4 of AWS DI.1, and sections Ql.23 and Ql.25 of AISC-N690.
Revision: 20
[ Westingh00Se 3.8-39 February 6,1998
pp q 9 3. Design of Structures, Cempone;ts, Equipmert, end Systems l
. Anchors are designed wherever possible with sufficient depth of embedment and side cover such that the steel anchor yields prior to failure of the concrete.
- The effect of concrete cracking is considered for fasteners located within the tensile zone of supporting concrete.
3.8.4.5.2 Supplemental Requirements for Steel Structures Supplemental requirements for use of AISC-N690 are as follows:
. In Section Ql.0.2, the def'mition of secondary stress applies to stresses developed by temperature loading only.
- In Section Ql.3, where the structural effects of differential settlement are present, they are included with the dead load, D.
. In Table Ql.5.7.1, the stress limit coefficients for compression are as follows:
1.3 instead of 1.5 in load combinations 2,5, and 6.
1.4 instead of 1.6 in load combinations 7, 8, and 9.
1.6 instead of 1.7 in load combination i1.
- In Section Ql.5.8, for constrained members (rotation and/or displacement constraint such that a thermal load causes significant stresses), supporting safety-related structures, systems, or components, the stresses under load combinations 9,10, and 11 are limited to those allowed in Table Ql.5.7.1 as modified above.
- Sections Ql.24 and Ql.25.10 are supplemented as follows:
Shop painting is in accordance with Section M of the Manual of Steel Construction, Load and Resistance Factor Design, First Edition. Exposed areas after installation are field painted in accordance with the applicable portion of Chapter M of the Manual of i Steel Construction, Load and Resistance Factor Design, First Edition. See subsection 6.1.2.1 for additional description of the protective coatings.
3.8.4.5.3 Design Summary Report A design summary report is prepared for seismic Category I structures documenting that the structures meet the acceptance criteria specified in subsection 3.8A.5.
Deviations from the design due to as-procured or as-built conditions are acceptable based on )
an evaluation consistent with the methods and procedures of Section 3.7 and 3.8 provided the )
following acceptance criteria are met. )
l a the structural design meets the acceptance criteria specified in Section 3.8 l
- the seismic floor response spectra meet the acceptance criteria specified in subsection 3.7.5.4 Revision: 20 February 6,1998 3.8-52 3 Westinghouse
ll r y
- 3. Design cf Structures, Componezts, Eq:1pme:t, r_nd Systems l
Depending on the extent of the deviations, the evaluation may range from documentation of an engineering judgement to performance of a revised analysis and design. He results of the evaluation will be documented in an as-built summary report by the Combined License applicant.
3.8.4.5.4 Design Summary of Critical Sections The design of representative critical elements of the following structures is described in Appendix 3H.
- Passive containment cooling system water storage tank e Shield building roof to cylinder connection
. Shield building to auxiliary building connection at elevation 180'
. South wall of auxiliary building (column line 1)
. Interior wall of auxiliary building (column line 7.3)
. West wall of main control room in auxiliary building (column line L),
e!cvation 117'-6" to elevation 153'-0"
. North wall of auxiliary building (column line 11 between Q and P), elevation 117'-6' to elevation 153'-0"
. Floor slab in north end of auxiliary building at elevation 135'-3" it cluding:
- 9 inch concrete slab on metal deck
- 24 inch reinforced concrete slab
- 24 inch finned floor above the main control room
. Spent fuel pool structural module 3.8.4.6 Materials, Quality Control, and Special Construction Techniques This subsection contains information relating to the materials, quality control program, and special construction techniques used in the construction of the other seismic Category I structures, as well as the containment intemal structures.
3.8.4.6.1 Materials 3.8.4.6.1.1 Concrete The compressive strength of concrete used in the seismic Category I structures and containment internal structures is fc =4000 psi. The test age of concrete containing pozzolan is 90 days. He test age of concrete without pozzolan is the normal 28 days. Concrete is batched and placed according to Reference 6 Reference 7, and ACI-349.
Portland cement confoims to Reference 8. Type II, with the sum of tricalcium silicate and tricalcium aluminate limited to no more than 58 percent. It is also limited to no more than 0.60 percent by weight of alkalies calculated as Na2O plus 0.658 K20. Certified copies of mill test reports showing that the chemical composition and physical properties conform to the specification are obtained for each cement delivery.
Revision: 20
[ WeMingh00Se 3.8-53 February 6,1998
~.
IEE EE <
- 3. Design cf Structures, Ccmponerts, Eq';ipment, cnd Systems F 1 !
section. The governing scenario is the case with a delay in the auxiliary building construction for the soft soil site with alternating layers of sand and clay. The delay is postulated to occur just prior to the stage where the auxiliary building walls are constructed. De evaluation used the following steps:
i a Member forces in the basemat just prior to the stage where the auxiliary building walls i are constructed were obtained from the construction settlement analyses. Bearing pressures were also obtained at the same stage and at the end of construction.
- Member forces at the end of construction were obtained by adding the change in member force that occurs after the construction of the auxiliary building walls has started. He increase in design member forces was obtained from the increase in bearing pressure after construction of the auxiliary building walls using the results of the analyses of the fully constructed nuclear island descdbed in subsection 3.8.5.4.1.
- The bearing pressure and member force due to the safe shutdown earthquake were obtained from the analyses described in subsection 3.8.5.4.1. The member force was added to the dead load member force at the end of construction and compared against the basemat design capacity.
The member forces, including the locked-in forces, during various stages of plant construction are within the design capacity for the five critical locations. De evaluation demonstrated that the member forces including locked in forces calculated by clastic analyses remain within the capacity of the section.
3.8.5.4.4 Design Summary Report A design summary report is prepared for the basemat documenting that the structures meet the acceptance criteria specified in subsection 3.8.5.5.
Deviations from the design due to as-procured or as-built conditions are acceptable based on an evaluation consistent with the methods and procedures of Section 3.7 and 3.8 provided the following acceptance criteria are met.
a the structural design meets the acceptance criteria specified in Section 3.8 a the seismic floor response spectra meet the acceptance criteria specified in subsection 3.7.5.4 Depending on the extent of the deviations, the evaluation may range from documentation of an engineering judgement to performance of a revised analysis and design. He results of the evaluation will be documented in an as-built summary report by the Combined License applicant.
l Revision: 20 W Westinghouse 3.8-63 February 6,1998 j l
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!E E 7 .4 3. Design of Structures, Crmponents, Eq ipment, rnd Systems l 3.8.5.4.5 Design Summary of Critical Sections The basemat design meets the acceptance criteria specified in subsection 3.8.4.5. Two critical portions of the basemat are identified below together with a summary of their design. The boundaries are defined by the walls and column lines which are shown in Figure 3.7.2-12 (sheet I of 12). Table 3.8.5-3 shows the reinforcement required and the reinforcement provided for the critical sections.
Dasemat between the shield buildine and exterior wall (line 11) and column lines K and L.
His portion of the basemat is designed as a one way slab spanning a distance of 23' 6" between the walls on column lines K and L. The slab is continuous with the adjacent slabs to the east and west. The critical loading is the bearing pressure on the underside of the slab due to dead and seismic loads. This establishes the demand for the top flexural reinforcement at mid span and for the bottom Gexural and shear reinforcement at the walls. The basemat is designed for the bearing pressures and membrane forces from the analyses on uniform soil springs described in subsection 3.8.5.4.1. He design moments and shears are increased by 20 percent to accommodate the nonuniform sites defined in subsection 2.5.4.5. Negative moments are redistributed as permitted by ACI 349.
The top and bottom reinforcement in the east west direction of span are equal. The reinforcement provided is shown in sheets 1, 2 and 5 of Figure 3.8.5-3. Typical reinforcement details showing use of headed reinforcement for shear reinforcement are shown in Figure 3 H.5-3.
Ilasemat between column lines 1 and 2 and column lines K-2 and N This portion of the basemat is designed as a one way slab spanning a distance of 22' 0" between the walls on column lines I and 2. The slab is continuous with the adjacent slabs to the north and with the exterior wall to the south. The critical loading is the bearing pressure on the underside of the slab due to dead and seismic loads. This establishes the demand for the top Gexural reinforcement at mid span and for the bottom nexural and shear reinforcement at wall 2. The basemat is designed for the bearing pressures and membrane forces from the analyses on uniform soil springs described in subsection 3.8.5.4.1. The design moments and shears are increased by 20 percent to accommodate the nonuniform sites defined in subsection 2.5.4.5. The reinforcement provided is shown in sheets 1, 2 and 5 of Figure 3.8.5-3. Typical reinforcement details showing use of headed reinforcement for shear reinforcement is shown in Figure 3H.5-3.
Deviations from the design due to as-procured or as-built conditions are acceptable based on an evaluation consistent with the methods and procedures of Section 3.7 and 3.8 provided the following acceptance criteria are met.
- Re structural design meets the acceptance criteria specified in Section 3.8 Revision: 20 February 6,1998 3.8-64 3 Westinghouse 1
- 3. Design of Structures, Components, Eqtipment, cnd Syst:ms 3.8.5.7 In Service Testing and Inspection Requirements There are no in-service testing or inspection requirements for the nuclear island structures I foundation.
The need for foundation settlement monitoring is site-specific and is the responsibility of the Combined License applicant (see subsection 2.5.4.5.1 1 ).
3.8.5.8 Construction Inspection Construction inspection is conducted to verify the concrete wall thickness and quantity of concrete reinforcement. The construction inspection includes concrete wall thickness and reinforcement expressed in units of in2 /ft (linear length) equivalent when compared to standard reinforcement bar sections. Inspections will be measured at applicable sections excluding designed openings or penetrations. Inspections will confirm that each section provides the minimum required reinforcement and concrete thickness as shown in ;
Table 3.8.5-3. The minimum required reinforcement and concrete thickness represents the !
required minimum values to meet the design basis loads. Table 3.8.5-3 also indicates the reinforcement provided which may exceed the required minimum reinforcement for the following reasons:
. Structural margin
. Ease of construction
. Use of standardized reinforcement sizes and spacing 3.8.6 Combined License Information 3.8.6.1 Containment Vessel Design Adjacent to Large Penetrations The final design of containment vessel elements (reinforcement) adjacent to concentrated l masses (penetrations)is completed by the Combined License applicant and documented in the I ASME Code design report.
l 3.8.6.2 Passive Containemt Cooling System Water Storage Tank Examination The Combined License applicant should examine the structures supporting the passive containment cooling storage tank on the shield building roof during initial tank filling as described in subsection 3.8.4.7.
l 3.8.6.3 As-Built Summary Report The Combined License applicant will evaluate deviations from the design due to as-procured or as-built conditions and will summarize the results of the evaluation in an as-built summary report as described in subsections 3.8.3.5.7. 3.8.4.5.3 and 3.8.5.4.4.
Revision: 20 February 6,1998 3.8-68 3 Westingh00S8
Westinghobse FAX COVER SHEET W
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1 i l l] NRC REQUEST FOR ADDITIONAL INFORMATION
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minutes to one hour, depending upon radioactivity concentrations. He only time the detectors will not be operating under statistically accurate conditions will be the time following a filter change or a -
system shutdown for maintenance. Sample transport times are minimized by locating the detectors as 1 close as practicable to the process sample point.
The response time for the in-line detectors is less than ten seconds. Rese detectors are provided with dynamic background radiation compensation.
Combining the minimum detectable levels shown in the table above with the detector response times discussed above, it has been shown that each monitor is sensitive enough to detect 10 DAC-hours.
SSAR Revision:
Those airborne radiation monitors which monitor plant areas which may be occupied by plant personnel will be capable of detecting 10 DAC-hours. The specific radiation monitors which are included in thes category are identified in Table 11.5-1.
In Table 11.51, " Radiation Monitor Detector Parameters", add the following:
Add "(Note 5)" in the " Service column for:
Containment Atmosphere Gas '
Containment Atmosphere N"/F" Fuel Handling Area Exhaust Auxiliary Building Exhaust Annex Building Exhaust Main Control Room Supply Air Duct (Par;iculate)- both entries MCR Supply Air Duct (Iodine)- both entries MCR Supply Air Duct (Gas)- both entries Containment Air Filtration Exhaust H.P. & Hot Machine Shop Exhaust Radwaste Building Exhaust 471.25-4 3 Westinghouse Rev.2
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NRC REQUEST FOR ADDITIONAL INFCRMATION
.7
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"S. Monitor is sensitive enough to detect 10 Derived Air Concentration (DAC)-hours."
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- TX CONFIRMATION REPORT ** AS OF- JAN 26 '98 12:39 PAGE.01
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- PITTSBURGH, PA 152'10-0355 ADVANCED PLANT SAFETY AND LICENSING FAX NO: 412 374-4M7 (WDi 284)
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- NUCLEAR REGULATORY COMMIS810N
( I wAssiwarow, o.c. .a=== m "%gyg January 21, 1998
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- JAN 23 1998 L.3 < g MEMORANDUM TO: Theodore R. Quay, Director -
0f 3 Standardization Project Directorate Division of Reactor Program Management s( 'g .j g ;ggg Office of Nuclear Reactor Regulation g Brian A. Mc Intyre FROM: William Huffman, Project Mana r Standardization Project Direct e Division of Reactor Program Management Office of Nuclear Reactor Regulat*
SUBJECT- " ' " ^
- EETING WITH WESTINGHOUSE ELECTRIC CORPORATI S DATE AND TIME: January 22,1998 - 8:30 a.m. - 4:30 p.m. ggp LOCATION: U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Room 0-9 A18 (8:30 am - 12:30 pm)
Room 0-3 B4 (1 pm - 4:30 pm)
PURPOSE: To discuss and resolve issues related to the review of the Westinghouse AP600 design certification application. Issues from AP600 SSAR Chap-ters 6,15, and 19.
PARTICIPANTS *: NRC WESTINGHOUSE J. Lyons, et al. C. Haag J. Sobrosky J.Scobel W. Huffman M. Corietti B. McIntyre, et al.
Docket No.52-003 cc: See next page
" Meetings between the NRC technical staff and applicants or licensees are open for interested members of the public, petitioners, intervenors, or other parties to attend as observers pursuant to " Commission Policy Statement on Staff Meeting Open to the Public," 59 Federal Reaister 48344, 9/20/94. However, portions of this meeting may be closed to protect Westinghouse proprietary information. Members of the public who wish to attend should contact me at (301) 415 1141.
i h6 Westl@088 FAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION DATE: or/2 z [9g NAME: L (J A ,1j r e n TO: LOCATION: ENERGY CENTER .
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COMPANY: Facsimile: win: 284 4887
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- TX CONFIR$ATION REPORT ** AS OF JAN 19 '98 13:13 PAGE.01 AP600 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 1/19 13:13 #23:NRC G3--S 00*29 01 OK
3 R-i WESTINGHOUSE ELECTRIC CORPORATION O PO 80X 355 PITTSBURGH, PA 15230 0355 ADVANCED PLANT SAFETY AND LICENSING rm FAX NO: 412474-4487 (W94 284)
CONFIRMATION NO: 412474 4237 DATE: /I / NE
't TO: So OM LOCATON:
FAX NUMBER:
FRoM: Brian A. McIntyre PHONE: WIN: 284-4334 i BELL: 412-374-4334 NUMBER OF PAGES (INCLUDING COVER SHEET):
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% UNITED STATES
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NUCLEAR REGULATORY COMMISSIONp t:fAsHINSTcN, D.C. 200864001 p'
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- January 7, 1998 - <
!aN i 21998 asBrian A. Mc intyrec Mr. Nicholas J. Uparulo, Manager Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230
SUBJECT:
OPEN ITEMS IN THE AP600 DESIGN CERTIFICATION REVIEW Deat Mr. Upsrulo:
As a result of the Nuclear Regulatory Commission staff's continuing review of the AP600 design certification application, the Quality Assurance and Maintenance Branch has prepared the enclose status report on the AP600 Initial Test Program. This report identifies nine open items and one confirmatory item that require resolution by Westinghouse before the staff can complete its review. If you have any questions regarding this request, please contact me at (301) 415 3145.
Sincerely, o
ws'~D Jo Wilson, Senior Policy Analyst tandardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation i
Docket No. 52 003 j (jf
Enclosure:
As stated cc w/o enci: See next page t
V,, fb Jo .
Y
V y Evtn if the commission gives st:ff the go-ahead for rulemaking, a rule would not likely be fint!ized until i)
., next year. And until a new regulition is issued. Pietrangelo noted, the industry will continue to be " walking on l eggshells."-Jenny Well, Miami and Washington (Jenny _weilemgh.com)
RISK-INFORMED TECH SPECS ABANDONED IN AP600 AS REVIEW SCHEDULE SLIPS NRC staff review of the Westinghouse AP600 passive safety, advanced, light-water reactor design is running about six months behind schedule and may fall further behind.
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One of the causes and one of the casualties of the schedule slippage was an attempt to use risk informed /\
technical specifications for the future plant relative to allowed outage times and surveillance times of essential equipment. , 6 "Unfortunately, as these risk informed technical specifications were developed by Westinghouse, the staff determined that the applicant was not providing sufficientjustification for the proposed technical specification Q%
values," staff said in a paper to commissioners, Secy 97 298, released last week. ( g "At that point the staff realized that, even if additional justification were provided for these values, the staff Q could not complete the review" on schedule and asked Westinghouse to use the standard technical specifications N instead.
The attempt to develop risk-informed TS was one cause of the delay in the review schedule, staff said. The latest schedule calls for a final design approval to be issued by September, rather than March of this year, but only g if certain conditions are met. One of them has already been missed. S Staff said the cause for the existing delays was mostly Westinghouse's fault, for not meeting its own proposed %
schedule of submittals, and partly, NRC staff's fault.
Staff admits reassigning AP600 reviewers to other higher priority work, such as the shutdown rule, which the commission later rejected, the steam generator rule, which never survived a cost-benefit analysis, and "other operating reactor issues." i In addition to the risk-informed TS, certain unique attributes of the AP600, or Westinghouse approaches to its licensing, that were different from those of the two evolutionary designs, previously approved, also caused delays, h
staff said, A ko 4
)
Such issues include the inspections, tests, analyses, and acceptance criteria (Itaac) for the design; the initial test program; the regulatory treatment of nonsafety related systems; code documentation and qualification; systems reliability of hydrogen mitigation systems; the need for a containment spray system (INRC,7 July '97,1); ([ 3 base mat design; fire protection program; spent fuel pool cooling system; and adverse system interactions.
Staff said it was in the process of clarifying its positions on fire protection and the spent fuel pool design and
%8 would forward a policy paper to commissioners in the future.
Staff said the new current schedule will be challenging. One of the conditions for meeting it was that certain
" docketed draft information" be submitted by Westinghouse no later than January 5. NRC AP600 Project j Manager Tom Kenyon said January 14 that the January 5 date had not, in fact, been met. However, Kenyon said a the time may be made up elsewhere and it did not necessarily mean the new schedule was already outdated.
-David Stellfox, Washington (dstel@mh.com)
FIRE WATCH: THERMO LAG ORDERS SOON? RIVER BEND ' HOT SHORTS' EXIT SET ]
I Commissioners are reviewing a staff proposal to issue confirmatory orders to licensees who have not yet resolved their Thermo-Lag fire barrier problems.
Licensees with Thermo-Lag have been out of compliance for almost six years, since the agency issued Bulletin 92-01 in June 1992 declaring ~Ihermo Lag fire barriers inoperable. But problems with the barriers surfaced as early as 1989 when a fire test done for Entergy's River Bend plant failed.
NRC has always maintained that the fire barrier deficiencies were of low safety significance, but the noncompliance with NRC requirements as a result of the "Ihermo-lag barriers was widespread. Licensees with only a little Thermo-Lag have already replaced it, but some licensees with larger quantities or with 'thermo-lag )
in inaccessible places have not done so yet. Schedules for doing so extend beyond 2000, or eight years after the i product was declared ineffective. l Licensees have implemented actions to compensate for the defective barners, mostly the use of fire watches.
But the history of using fire watches is replete with falsified logs and watches sleeping on the job.
NRC staffers have been talking about using confirmatory orders to close out the problem for a year or more.
NRC fire protection engineer Steven West said a paper proposing the orders be issued is currently pending l before commissioners.
Meanwhile, Entergy's River Bend may finally get its exit meeting on NRC's first fire protection functional inspection (FPFI), which occurred at River Bend last summer.
"
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Machd is 4ke p%oseA ssAR chanac b O
sec.kon (.2.i.3.2.4 desu.Wna whv 4h AP6oc ma.ss and energ <e\ ease caf cula.b as do not auun a. single (navut. .
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I for a given time period is determined from the difference between 'the energy required to raise the temperature of the incoming flow to saturation and the sum of the decay heat, core stored I energy, reactor coolant system metal energy and SG mass and metal energy release rates. The energy release rate for the available break flow is determined from a comparison of the total energy available release rate and the energy release rate assuming that the break flow is 100-percent satura:ed steam. Saturated steam releases maximize the calculated containment pressurization.
Rep \ca e. wA 6.2.1.3.2.4 Single Failure Analysis , do ckd g ,
ho single failure is assumed in the mass and energy release calculations. De safety in ]
system for the AP600 is passive, as opposed to active pumped safety injection systems for a ,
conventional PWR, As a result, there is no single failure postulated for the mass and energy release analysis ne effects of a single failure are taken into account in the containment analysis of subvetion 6.2.1.1.
6.2.1.3.2.5 Metal Water Reaction Conmient with 10 CFR 50. Appendix K criteria, the energy release associated with the zirconium. water exothermic reaction has been considered. The LOCA peak cladding temperature analysis, presented in Chapter 15, that demonstrates compliance with the Appendis K cntena demonstrates that no appreciable level of zirconium oxidation occurs.
Thn lesel of rexuon has been bounded in the containment mass and energy release analysis by mcorporstmg the heat of reaction fmm 1 percent of the zirconium surrounding the fuel.
This exceeds the level predicted by the LOCA analysis and results in additional conservatism in - t mass and energy release calculations.
6.2.1.3.2.6 Energy Insentories insentones of the amount of mass and energy released to containment during a postulated I LOCA are presided in summary Tables 6.2.1.3-2 through 6.2.1.3 7.
6.2.1.3.2.7 Additional Information Required for Confirmatory Analysis System parameters and hydraulic characteristics needed to perform confirmatory analysis are I provided in Table 6.2.1.3 8 and Figures 6.2.1.3-1 through 6.2.1.3-4.
6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Rupture Inside Containment Steam line ruptures occurring inside a reactor containment structure may result in significant releases of high-energy fluid to the contaipment environment, possibly resulting in high containment temperatures and pressures. De quantitative nature of the releases following a steam line rupture is dependent upon the configuration of the plant steam system, the containment design as well as the plant operating conditions and the size of the rupture. His section describes the methods used in determining the containment responses to a variety of
. Revision: 13 g Westinghouse 6.21I May 30,1997
i.
A d
1 Revised paragraph for SSAR 6.2.1.3.2.4 l l
The assumptions for the containment mass and energy release analysis are intended to maximize the calculated release. A single failure could reduce the flow rate of water to the RCS, but would not disable any passive core cooling system. For example, if one of the two parallel valves from the CMT were to fail to open, the injection flow rate would be reduced and, as a result, the break mass release rate would decrease. Therefore, to maximize the releases, the AP600 mass and energy release calculations do not 4 assume a single failure. The effects of a single failure are taken into account in the containment analysis of subsection 6.2.1.1.
A
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DRAFT _ l NRC FSER OPEN ITEM Question 440.798F (OITS - 6450)
Westinghouse needs to document how the critical safety functions are satisfied during shutdown. It is not clear that the ERG Shutdown Guidelines are comprehensive enough to mitigate the expected transients and accidents during shutdown. Westinghouse should explain how procedures SDG 1 to 6 are sufficient to mitigate the expected transients and accidents during shutdown conditions.
Response
The following discussion will be added to the Introduction section in the background document for the SDF-0.1, AP600 SHUTDOWN SAFETY STATUS TREE DURING SHUTDOWN:
The AP600 shutdown emergency response guidelines (ERGS) were developed using the same philosophy and methodology that was used for developing the at power ERGS for the AP600. Since there was no corresponding generic guidance for shutdown conditions for operating plants, the shutdown guidelines for AP600 are first of a kind guidance and focus on protecting the general public by monitoring and protecting the plant critical safety functions. The same barriers (fuel cladding, reactor coolant system and containment building) protect the public whether the plant is at power or is shutdown. Thus the critical safety functions which are used to monitor plant conditions during accidents for at-power conditions (operating modes 1,2,3, and 4) were used as the basis for developing a monitoring tool to detect challenges to the plant safety state for the remaining shutdown conditions (modes 5 and 6). Each symptom (question in the status tree) for each at-power critical safety function status tree and the underlying intent of that safety function was evaluated with respect to shutdown conditions. The resultant was a single status tree for shutdown operations during modes 5 and 6 that represents all six of the critical safety functions. Additional attention was given to the prevention of boiling in the reactor core and establishing containment closure early, if boiling in the reactor core cannot be prevented because these are the prime issues being addressed by the USNRC and operating plants for shutdown operations. The following thought process was used for aniving at the single shutdown status tree and the symptoms used to determine if there is a challenge to shutdown plant safety.
- l. Review of the at-power subcriticality status tree for shutdown conditions showed that since the reactor is already shutdown, only decay heat is generated. (Note that the check that the reactor is below 5 % power in the at power ERGS is to assure that only decay heat is generated by the reactor since safety systems are designed for decay heat removal only.) Thus, the only concern applicable during modes 5 and 6 is inadvertent criticality. A flux doubling alarm is provided for the AP600 design to identify a loss of shutdown margin which precedes an inadvertent criticality while nuclear flux is well below the point of adding heat into the system. An orange path is designated for this condition since prompt operator action should be initiated to reestablish shutdown margin and prevent the retum to criticality.
- 2. Review of the at power core cooling status tree showed that the core cooling safety function is applicable, but due to the initial energy levels, elevated core exit temperatures are not expected for 440.798F-1
~
DRAFT NRC FSER OPEN ITEM a long period of time. As long as water level is maintained above the reactor core, heat can be removed to prevent core heatup to temperatures that could possibly damage the core. Adequate water level to maintain core cooling can be determined by either pressurizer level (if the RCS is closed) or hot leg level (if the RCS is at reduced inventory operations). Since the primary system is at a low temperature (and saturation pressure) condition, the prime recovery strategy would be to restore water level over the reactor core by any means possible. No higher prioritization symptoms are necessary (such as core exit temperature symptoms that prioritize FR-C.1 and FR-C.2 in the at-power ERGS), since primary pressure at saturated conditions is low enough for all possible injection paths to physically inject. An orange path is designated for this condition since prompt operator action should be initiated. To prevent heat up and possible high saturation pressures, the core cooling safety function challenged was prioritized first on the shutdown status tree. Injection of borated water is desired to prevent and/or mitigate a simultaneous loss of shutdown margin as discussed in item I earlier.
- 3. Review of the at-power heat sink status tree showed that the main heat sink during shutdown conditions in modes 5 and 6 is the normal residual heat removal system (RNS). If the RNS is lost, prompt mitigating actions must be taken by the operator. An orange path is designated for this condition, since prompt operator action should be initiated to reestablish RNS cooling or provide attemate ways of removing core decay heat. This symptom is checked just after the shutdown core cooling status check, since primary system inventory is needed for the operation of the RNS pumps. It is also prioritized just behind the challenge to the core cooling function since the primary system will start to heat up on the loss of RNS, while more time is available to the operator for addressing a loss of shutdown margin and containment closure (containment closure actions are also in the guideline for addressing the loss of RNS). Loss of the ability for RNS to remove heat from the primary system is addressed as its own symptom by means of a check on RCS temperature being maintained within a small band when no heatup or cooldown is in progress.
- 4. Review of the at-power integrity status tree showed that the only challenge to primary system integrity during operating modes 5 and 6 is system overpressurization. Since the primary system is already at a low temperature, a significant rapid cooldown cannot occur. 'Ite cold overpressure limits are checked and if exceeded, prompt operator action is required.
- 5. Review of the at-power containment status tree showed that rapid containment overpressure during modes 5 and 6 cannot occur due to the initial low energy condition of the primary system. The main function for containment is to prevent radiation releases during an accident. During the shutdown conditions of modes 5 and 6, the containment building may be open to allow easy access for maintenance activities. The intent of the containment function can be maintained if containment closure is initiated upon indication of abnormal radiation in containment. Therefore only containment radiation is monitored on the shutdown ERG status tree. Abnormal radiation will require prompt operator action to establish containment closure.
440.798F-2 3 W85tingh0US8
i NRC FSER OPEN ITEM
. DRAFT e-
^
- 6. Review of the at-power inventory status tree showed that normal primary system inventory could vary during modes 5 and 6 depending on whether or not the plant is in reduced inventory operations. The inventory status tree for the at-power ERGS checks if pressurizer level is normal during operation is modes 1,2. 3 or 4. Normal water level for operation in modes 5 and 6 is either pressurizer level (if the RCS is closed) or hot leg level (if the RCS is at reduced inventory operations. Since departure from the normal primary inventory is check already in the shutdown status tree to verify adequate core cooling capability, no additional checks are made to specifically address the inventory function.
Since the plant is in a shutdown and low energy state in modes 5 or 6, the symptoms for determining a challenge to the plant safety state are different because they are tailored to the shutdown plant conditions. The priority between functions is also different. The top priority is given to core cooling (inventory on the core) since loss of inventory can result in core boiling in a short period of time since there is less water to absorb core heat and the subsequent loss of RNS due to inadequate water level in the reactor coolant system. Second priority is given to heat sink (loss of RNS) which will also result in a heatup of the primary water. Third priority is given to containment (abnormal radiation) so that prompt closure of containment will be achieved prior to any prolonged releases to the environment.
Fourth priority is given to suberiticality (loss of shutdown margin) since more time is available before a reactivity accident causes either boiling in the core or releases to containment than in the previous accidents. The fifth priority is given to integrity (cold overpressurization) since it is less likely to result in core boiling or quick radiation releases outside the containment. The sixth priority is given to the loss of heat sink due to support system failures (unexpected heatup of the RCS) since the expected heatup rate of the primary system would be small. All six of the functional challenges require prompt operator action to mitigate them.
SSAR Revision: None l
)
8N T Westinghouse 1
1 1
NRC FSER OPEN ITEM
- Question 440.799F (OITS . 6451)
The ERGS do not rely on containment sump level, hydrogen concentration or core exit temperature to monitor the integrity or performance of containment. In addition, the ERGS do not have guidance for initiating hydrogen igniters and cavity flooding system. Westinghouse should explain why monitoring containment radiation is sufficient to monitor the performance of the containment during shutdown and should explain why these other parameters are not required to be monitored.
l
Response
A rspid containment overpressure during modes 5 and 6 cannot occur due to the initial low energy condition of the primary system. During the shutdown conditions of modes 5 and 6, the containment building may be open to allow easy access for maintenance activities. Rus, the main objective for containment safety function during modes 5 and 6 is to prevent radiation releases during a challenge.
The intent of the containment function can be maintained if containment closure is initiated upon indication of abnormal radiation in containment. High containment radiation would be detected well before any core damage would occur or large quantities of hydrogen are released into the containment.
Bus adequate time is available for determining if additional actions are necessary, as stated in the guideline for a response to a high radiation in containment, SDG-3, RESPONSE TO HIGH CONTAINMENT RADIATION DURING SHUTDOWN. Therefore only containment radiation is monitored on the shutdown ERG status tree. Abnormal radiation will require prompt operator action to establish containment closure.
Use of the hydrogen igniters or cavity flooding is not specified in the shutdown ERGS. It is intended that these actions be covered under the Step 20 of SDG-1, Evaluate Long Term Plant Status, and under Step 4 of SDG-3 Determine if Additional Actions are Necessary.
SSAR Revision: None I
l om T Westinghouse i
j
J DRAFT - - !
NRC FSER OPEN ITEM Question 440.800F (OITS - 6452)
The standard safety function sequence applied in the AP600 ERGS for operating emergencies is not l followed for shutdown conditions. Westinghouse should justify the different sequence used for the j shut down conditions. !
I
Response
This item is addressed in the response to FSER Open Item 440.798F.
SSAR Revision: None l
I 44 .800F-1 W westinghouse
d NRC FSER OPEN UEM
. DRAFT -
a
~
a Question 440.801F (OITS - 6453)
I The shutdown safety status tree, SDF-0.1, does not include the following critical safety functions: j Subcriticality, Core Cooling, Heat Sink, Integrity, Containment and Inventory. Westinghouse should explain how these safety functions are monitored and maintained during shut down conditions for all transients.
Response
This item is addressed in the response to FSER Open Item 440.798F.
SSAR Revision: None l
440.801 F-1 W Westinghouse l
\
l
. I
o
- DRAFT =--
NRC FSER OPEN ITEM Question 440.802F (OITS 6454) 1 i
The shutdown safety status tree, SDF-0.1, implies that operator actions are to be taken in a panicular sequence for all shutdown alarms. For example, the " Nuclear Flux doubling alarm", has the operator verify whether the normal residual heat removal system (RNS) is in service or not before any action associated with the flux doubling alarm is taken. The staff does not understand the precedence for action in the shutdown safety status tree. Westinghouse should clearly explain how this status tree would be used to develop the plant specific shutdown emergency operating procedures.
Response
He priority of monitoring symptoms in the shutdown status tree is addressed in the response to FSER Open item 440.798F. The actual method of monitoring the status tree will be determined through the MMI process. Since the shutdown status tree was developed to apply at all times during shutdown conditions, is could be monitored by the plant computer and when a challenge occurs, the operator would be alerted to the challenge so that prompt action could be taken to mitigate the challenge.
SSAR Revision: None i
i 440.so2Fa W westinghouse
i
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r s le 1 NRC FSER OPEN ITEM 8 ;. i
- s. . 3- s.
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Question 410.427F (OITS 6402) g, j'%- h
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Re: %
RAI #: 410.427F In response to a question concerning compliance of the VBS with defense in depth criten .ghouse provided a response (OITS No. 2897) which stated that the system is located in the auxiliary buildiag md equipment will be procured with the appropriate environmental qualification and as with the other equipment of the Nuclear Island, it is protected from defined natural phenomena. Howeser, Westinghouse needs to be specific, and provide details describing what requirements of environmental qualification are met when procuring VBS equipment for protection against natural phenomena. Westinghouse needs to state in the SSAR Section 9.4.1 that VBS is classified as a defense in depth system.
Response
SSAR subsection 9.4.1.1.1 will be revised to indicate additional requirements for the system.
l l
SSAR Revision:
Subsection 9.4.1.1.1, revise to state:
Those portions of the nuclear island nonradioactive venulation system which penetrate the main control room envelope are safety-related and designed as seismic Category I to provide isolation of the main control room envelope from the surrounding areas and outside environment in the event of a design basis accident. Other functions of the system are nonsafety-related. HVAC equipment and ductwork whose failure could affect the l
operability of safety-related systems or components are designed to seismic Category 11 requirements. The l
l remaining portion of the system is nonsafety-related and nonseismic. The equipment is procured to meet the l environmental qualifications used in standard building practice.
The nuclear island nonradioactive ventilation system is designed to control the radiological habitability in the main control room within the guidelines presented in Standard Review Plan (SRP) 6.4 and NUREG 0696 l I
(Reference 1), if the system is operable and ac power is available.
l R::=!::.- 1:!=d :=:'i=:in = d!::ic ;y:::- inc: =f::y ::!:::d. Mc :=r,1: pPortions of the system l that provide the defense in depthfunction of filtration of main control room / technical support center air during conditions of abnormal airborne radioactivity are designed, constructed, and tested to conform with Generic Issue l B-36, as described in Section 1.9 and Regulatory Guide 1.140 (Reference 30), as described in Appendix 1 A, ASME N509 (Reference 2) and ASME N510 (Reference 3). "
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- TX CONFIRMATION REPORT ** AS OF JAN 14 '98 12:23 PAGE.01
- AP600 DESIGN CERT DATE TIME TO/FROM MODE MlH/SEC PGS STATUS 01 1/14 12:23 #23:NRC 63--S 00'28 O! OK 1
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ng h Westinghouse FAX COVER SHEET W
RECIPIENT INFORMATION SENDER INFORMATION DATE: / /r 5/9 J' NAME: % (f)rgg TO: LOCATION: ENERGY CENTER .
bado {c A L ETI~1 EAST PHONE: FACSIMILE: PHONE: Osco:M/2-3 7V-5 29o COMPANY: Facemie: win: 284-4887 O 3 VA/ C outside. (412)374 4887 f LOCATION:
I co . P. ,.y The fogowing pages are being sent from the Weetinghouse Energy Center, Esot Tower, j Monroeville, PA. If any probleme occur during thle tranemleelen, please call:
WIN: 344125 ( Bev ) or Outelde: (412)374 -6529 COMMENTS:
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- TX CONFIRMATION REPORT ** AS-OF JAN 13 '98 9:14 PAGE.01 I
- APG00 DESIGN CERT DATE TIME TO/FROM MODE MlH/SEC PGS STATUS 01 1/13 09:10 3014152300 G3--S 03'22 05 OK 4
.e
I d T -
Fanto, Susan V.
- i From: Novendstern, Earl H.
Sent: Wednesday, January 14,199811:53 AM To: 'Huffman, Bill (NRC)'; 'Lambrose Lois (NRC)'
Cc: Fanto, Susan V.; Nissley, Mitchell E.; Kemper, Bob (Notes); Novendstem, Earl H.
Lambrose and Bill, Per our telecon yesterday, we would propose words along the lines of the following suggested text. Comments?
Thanks.
EARL WCAP-14171, Revision 1, documents the application of the Westinghouse large break LOCA best-estimate methodology to the AP600. In developing that methodology some parameters were not included in the uncertainty analysis. Some of these parameters were bounded conservatively, while others were judged to have an insignificant impact on the low calculated PCT result and for simplicity were not considered. Among the latter set of parameters are (1) the AP600-unique passive safety i system (CMTs and PRHR) modeling, and (ANY OTHERS AS IDENTIFIED BY NRC).
A bounding assessment of the potential impact of the PRHR and CMT systema modeling on the calculated PCT was obtained by eliminating those systems from the large break transient calculation. As noted on page 2-33 of WCAP-14171, Revision 1, the blowdown and reflood phase PCT values for the WCOBRA/ TRAC case in which these systems were eliminated each differ by less than 10 deg F from the base case PCT result.
The AP600 large break LOCA 95% PCT values for blowdown and reflood are j reported in SSAR Table 15.6.5-9. The limiting PCT is that for the l blowdown, at 1676 deg F. In the event that either the blowdown or reflood phase PCT were to exceed 1725 deg F in a future AP600 best l estimate methodology large break LOCA analysis, the PCT increase of 50 l deg F would constitute a significant change in results in the 10CFR50.46 analysis parlance. In the event that a PCT increase beyond 1725 deg F did occur for AP600, Westinghouse must perform the following l to reestablish the WCAP-14171, Revision 1 methodology for the AP600 '
application: l
- 1. Repeat the global model matrix of calculations and the final 95%
uncertainty calculation.
- 2. Address the sensitivity to the above-listed parameters which are not now included in the AP600 uncertainty methodology. Repeat the study that identifies the PCT sensitivity to PRHR/CMT elimination, and add the blowdown and reflood PCT impacts as a bias to their respective 95%
PCT results. Likewise, for other parameters which have been judged to have no significant impact on calculated PCT for a 95% PCT of 1725 deg F or below, perform a study to assess their impact on the PCT result.
- 3. Perform an analysis of the maximum local oxidation using the Page 1 ;
e
't j' techniques approved *for 3/4 loop plant applications to show compliance '
with the applicable 10CFR50.46 criteria. A transient with PCT in cxcess of the 95% PCT value identified in Step 1 above, augmented by the biases identified in Step 2, would be used for the oxidation analysis. The core-wide oxidation analysis would also be performed using the methods approved for use in 3/4 loop plant applications.
i Sast East H. Novendstem Phone:(412) 3744790 Far (412) 3744011 ECE 426 B
======
Page 2
k i
l Fanto, Susan V.
- From: Novendstem, Earl H.
Sent: Tuesday, January 13,1998 3:51 PM To: 'Huffman, Bill (NRC)'
Cc: ' Ralph Landry (NRC)'; Fanto, Susan V.; Fittante, Randy (Notes); Young, Michael Y.; Gagnon, Andy (Notes); Novendstem, Earl H.
Subject:
NOTRUMP Documentation Bill + below is what we would propose doing for the AP600 version of NOTRUMP. Please let me know when we should talk (Wed am or next Tuesday best for me). Thanks.
=================================================================
Reference 1: WCAP 10079-P-A, August 1985.
Reference 2: WCAP 14807, Revision 3, November 1997.
The original NOTRUMP central numerics (documented in Reference 1) were modified for application to AP600. The modifications were performed in the following three areas: 1) change from net mass flow-based to net volumetric flow-based momentum equation,2) Implicit treatment of bubble rise and droplet fall, and 3) implicit treatment of gravitational head. Although these rnodifications are described in Reference 2 (in Sections 2.4,2.9, and 2.11, respectively), the descriptions may be difficult to follow because of the frequent referencing to the original equations in Reference 1. Thus, the task (which is currently underway in response to SER open item t 440.795F) is to document the changes to the NOTRUMP central numerics which were made for AP600 all in one place, from start to finish (i.e., to document the derivations from the starting differential equations to the resulting difference form in the code).
The aforementioned changes to the NOTRUMP numerics affect the energy and mass conservation equations for each interior fluid node (i.e., Equations 21 through 2-4 in Reference 1), and the momentum conservation equation for cach non-critical flow link (i.e., Equation 2-33 in Reference 1). Thus, til of these starting differential equations will be presented, both in their original and modified forms, including the necessary derivation details which were required to modify them. Next, the linearizations and derivations which were performed on these modified differential equations to place them in their finite-difference form to be solved will be i
presented. The final result of this will be a presentation of the modifications to the central matrix and its elements, analogous to what is documented in Appendix E in Reference 1.
l
[dd EEEEEEEEESSEEEEEES Earl H. Noverdstem Phone:(412) 374 4790 i Far (412) 3744011 j ECE 426 B Page 1 l
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& M y NRC FSER OPEN ITEM
- Co m pt Lti m a C c" LU 637 w(Ro u u-Pha'*
Phone % L- 3 W -r2 90 Fan e Fax n Question 410.418F (OITS 6393)
Re: {
RAI #: 410.418F in order to demonstrate compliance with Position C.2 of Regulatory Guide (RG) 1.29, Westinghouse must state in the applicable SSAR sections for each HVAC system (VAS, VBS, VCS, VFS, VHS, VRS, VTS, VXS, and VZS) that the equipment, ductwork, supports, and accessories are designed and constructed in accordance with the requirements of seismic Category II, or provide an alternative to preclude them from collapsing onto safety-related structures, systems, and components during an SSE.
Response
Subsections of SSAR section 9.4 will be revised to address seismic Category 11 requirements. For VBS see SSAR subsection 9.4.1.1.1, second paragraph.
SSAR Revision:
VAS ,
Subsection 9.4.3.1.1, revise to state: I "9.4.3.1.1 Safety Design Basis The radiologically controlled area ventilation system serves no safety-related function and therefore has no l nuclear safety design basis. System equipment and ductwork located in the nuclear island whosefailure could l affect the operability of safety-related systems or components are designed to seismic Category 11 l requirements. The remaining portion of the system is nonseismic."
VBS See SSAR subsection 9.4.1.1.1, second paragraph.
I VCS Subsection 9.4.6.1.1, revise to state:
"9.4.6.1.1 Safety Design Basis The containment recirculation cooling system serves no safety-related function and therefore has no nuclear safety design basis. 'Ihe contamment recirculation system is not required to mitigate the consequences of a l design basis accident or loss of coolant accident. System equipment and ductwork whosefailure could affect 410.418F-1 1
]
~4 NRC FSER OPEN ITEM k
the operability ofsafety related systems or components are designed to seismic Category 11 requirements. The remaining portion of the system is nonseismic."
VFS Subsection 9.4.7.1.1, revise to state:
"9.4.7.1.1 Safety Design Basis The containment air filtration system serves no safety-related function. other than containment isolation, and therefore has no nuclear safety design basis except for containment isolation. See subsection 6.2.3 for a description of the containment isolation system. System equipment and ductwork whosefailure could affect the operability ofsafety related systems or components are designedto seismic Category 11 requirements. The remaining portion of the system is nonseismic."
VHS Subsection 9.4.11.1.1, revise to state:
"9.4.11.1.1 Safety Design Basis The health physics and hot machine shop HVAC system serves no safety-related function and therefore has no l nuclear safety design basis. The system is nonseismic."
VRS Subsection 9.4.8.l.1, revise to state:
"9.4.8.1.1 Safety Design Basis The radwaste building HVAC system serves no safety-related function and therefore has no nuclear safety design l basis. The system is nonseismic."
VTS Subsection 9.4.9.1.1, revise to state:
"9.4.9.1.1 Safety Design Basis The turbine building ventilation system serves no safety-related function and therefore has no nuclear safety {
\ design basis. The system is nonseismic."
410.418F-2 W Westinghouse
en NRC FSER OPEN ITEM e4+
4 VXS Subsection 9.4.2.1.1, revise to state:
"9.4.2.1.1 Safety Design Basis The annex / auxiliary buildings nonradioactive HVAC system serves no safety-related function and therefore has no nuclear safety design basis. System equipment and ductwork located in the nuclear island whosefailure could affect the operability of safety-related systems or components are designed to seismic Category 11 requirements. The remaining portion of the system is nonseismic."
VZS Subsection 9.4.10.1.1, revise to state:
"9.4.10.1.1 Safety Design Basis The diesel generator building heating and ventilation system serves no safety-related function and therefore has l no nuclear safety design basis. The system is nonseismic."
410.418F-3 W Westinghouse
f NRC FSER OPEN ITEM anan=
- I Question 410.422F (OITS 6397)
Re:
RAI #: 410.422F Westinghouse needs to reference the RG 1.140-1979, Revision 1, in SSAR Section 9.4.13 and provide the same reference to all HVAC subsections including 9.4.1 and 9.4.7.
Response
SSAR subsection 9.4.13 will be revised to include Regulatory Guide 1.140 as a reference and this reference will be identified in each use of Regulatory Guide 1.140 in the text of Section 9.4. Regulatory Guide 1.140 states: "This guide applies only to atmosphere cleanup systems designed to collect airborne radioactive materials during normal plant operation, including anticipated operational occurrences, and addresses the atmosphere cleanup systems, including the various components and ductwork in the normal operating environment." It applies only to the systems described in SSAR subsections 9.4.1 and 9.4.7.
SSAR Revision:
Section 9.4, wherever used, (subsection 9.4.1, 3 places, subsection 9.4.7, 3 places, subsection 9.4.12 on COL l information. I place) replace; " . . 1.140. . " with: " . . 1.I40 (Reference 30) . ."
Subsection 9.4.13, add:
~30. " Design, Testing, and Maintenance Criteria for Normal Ventdation Exhaust Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants," Regulatory Guide (RG) 1.140-1979, Revision 1."
410.422F-1
e i
'A NRC FSER OPEN ITEM Question 410.423F (OITS 6398)
Re:
RAI #: 410.423F Westinghouse needs to revise SSAR Section 9.4.1.4 under " Abnormal Plant Operation," and state that a representative charcoal sample, used or new, is laboratory tested to verify a minimum charcoal efficiency of 90 percent in accordance with RG 1.140 and conforms to ASME N510-1989 for test procedures and test frequency.
Response
SSAR subsection 9.4.1.4 will be revised to state that a representative charcoal sample, used or new, is laboratory tested. l SSAR Revision:
Subsection 9.4.1.4, second paragraph. revise to state:
"The supplemental air filtration unit, HEPA tilters, and charcoal adsorbers are tested in place in accordance with ASME N510 to verify that these components do not exceed a maximum allowable bypass leakage rate. Used sSamples of charcoal adsorbent, used or new, are periodically tested to verify a minimum charcoal efficiency of 90 percent in accordance with Regulatory Guide 1.140 (Reference 30), except that test procedures and test frequency are conducted in accordance with ASME N510."
i l
410.423F-1
1 4
a NRC FSER OPEN ITEM , I I
. T i i i
I. l 1
Question 410.405F (OITS 6410) l Re: l 1
RAI #: 410.435F During nonnal operation, one of the redundant supply ABUs, return fans and battery room exhaust fans operate 1 continuously to maintain acceptable environmental conditions, maintain the Class lE electrical room emergency passive heat sink below its initial ambient air temperature, and prevent a hydrogen gas buildup in the Class IE battery rooms. The battery exhaust is vented directly to the turbine building vent to limit the hydrogen gas concentration to less than 2 percent by volume in accordance with RG l.128. Revision 1, " Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants " However, Westingh,use needs to reference the RG 1.128. Revision 1, in SSAR Sections 9.4.1 and 9.4.13. This is Open item 9.4.1-17.
Response
SSAR subsections 9.4.1.1.2 and 9.4.13 will be revised to include a reference to Regulatory Guide 1.128.
SSAR Revision:
Subsection 9.4.1.1.2, third portion, revise to state:
Class IE Electrical Rooms ,
I The nuclear island nonradioactive ventilation system provides the following specific functions:
- Exhausts air from the Class IE battery rooms to limit the concentration of hydrogen gas to less than 2 l percent by volume in accordance with Regulatory Guide 1.128 (Reference 31).
l
. Maintains the Class lE electrical room emergency passive cooling heat smk below its initial design ambient !
air temperature limit of 75'F
=
Provides smoke removal capability for the Class IE electrical equipment rooms and battery rooms" Subsection 9.4.13. References, add reference 31:
l "31. " Installation Design and Installation of Isrge Isad Storage Baneries for Nuclear Power Plants",
l Regulatory Guido 1.128, Revision I, October 1978."
g 410.435F-1
- =
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- Date: Tue,06 Jan 199810:21:03 -0500 To: TRQ@NRC. GOV From: " Brian A. McIntyre" <mcintyba@wesmail.com>
Subject:
Duplicate FSER OI Cc: meintyba@wcsmall.com,jwr@nre. gov Ted, In your letter of December 18,1997, the text for FSER open items 410.432 and 410.433 are identical as far as we can determine. There obviously is an error of some sort. Either there is an open item we do not have, or an open item was sent twice. I guess a third option is there is something we are missing in our reading of the open item..
Please provide clarification by the close of business on Wednesday, January 7 so that we can take the appropriate actions on our end to meet our goal of resolving all open items by January 31,1998.
Thanks for your assistance.
Brian A.McIntyre Bell 412.374.4334 WIN 284.4334 FAX Bell 412.374.4887 FAX WIN 284.4.887 l
l l
[, Printed for _" Brian A. McIntyre" <mcintyba@wesmai!.com> 11
.i 4
6
@ Westinghouse FAX COVER SHEET Recipient information Sender Information Date: ;[9 /9 g Name: D. Lindgren To: Joe Sebrosky Location: WEC E 330 Phone: Phone: (412) 374 4856 (301) 415-1132 Facimile: Facimile: Bet (412) 374-4887 (310) 415-2002 Company: U.S.NRC WIN 2844887 Location: Rockville, MD Cover + Pages 1+[
The following pages are being sent from the Westinghouse Energy Center, Monroeville, PA.
If there are any problems during this transmission, please call the sender or WIN 284-6529 or Bell (412) 374-6529.
comments:
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- Fax to NRC FSER Open item 220.114F ,
Following our telecon today we reviewed the vertical Floor Response Spectra at elevation 272'and compared them to those at elevation 307' We believe that the information already available to you is sufficient to close your comment # 4 of this issue without revising the SSAR.
The Model "A" enveloped spectra and the raw spectra for Model "B".were included in our letter of July 28.1997. The FRS for elevation 272' and 307'are attached. They show very little difference (2 to 3%) between the spectra for the two elevations. This is also shown in the revised spectra from the AP600 Floor Response Spectra design document which are also attached.
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0N Nd mM EAST x </o (g' PHONE: FACSIMILE: PHONE: Osco.
COMPANY: - Facsimile: wirt 284 4887 M% outside: (412)374 4887 LOCATION:
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The foRowing pages are being sent from the Westinghouse Energy Center, East Tower, Monroev0fe, PA, any problems occur during thle transmission, please call:
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NRC FSER OPEN ITEM ,
Question 440.783F (OITS #6365)
LCO 3.4.12 ADS Operating LCO 3.4.12 specifies that the ADS, including 10 flow paths, shall be operable during MODES 1 through 4 operation. If one flow path, or one stage 1 flow path and one stage 2 or stage 3 flow path is inoperable. Action A 1 requires restoration of these flow path within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. BASES 3.4.12 does not provide sufficient justification for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action completion time. In response to RAI 440.671 (Westinghouse letter, B. McIntyre to T. Quay, NSD NRC 97-5278, August 27,1997), Westinghouse provided justifications for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time. Westinghouse contended that the basis for 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time is consistent with the STS PORV,3.4.11. Action B, and consistent with two train ECCS systems that can perform their safety function without a single failure. The applicant also contends that the inoperable ADS flow path conditions (Condition A) have been assumed as single failures in the Chapter 15 LOCA analyses. The staff finds that the design bases and functional requirements are so different between the PORVs and the AP600 ADS valves that it is not prudent to draw an equivalence between them. In addition, a single failure assumption in the safety analyses is not to be used for a failure that has been found to exist. The applicant should provide additional justifications, which should be documented in the TS BASES, to justify the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time.
This is open item.
Response
The justification for allowing this condition to exist for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is that this time is long enough to fix many problems, the time is short enough that the probability of having a DBA is low and if a DBA did occur it can be mitigated as long as a single failure does not also occur. Note that with ADS valves INOPERABLE in accordance with LCO 3.4.12 A (one ADS stage 1 and one ADS stage 2/3 paths) or LCO 3.4.12 8 (one ADS stage 4 path), the AP600 can withstand any DBA as long as it does not suffer a single failure during the DBA. In addition, as mentioned in the background for this LCO, the PRA shows that adequate Core Cooling can be provided with the failure of two (or more) flow paths.
The following wording is proposed to clarify the justification of the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time:
"A Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is reasonable based on the capability of the remaining ADS valves to perform the required safety functions assumed in the safety analysis and the low probability of a DBA during this time period. This completion time is the same as is used for two train ECCS systems which are capable of performing their safety function without a single failure."
A revision to this technical specification BASES is attached.
440.783F-1
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NRC FSER OPEN ITEM ,
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- i SSAR Change:
A revision to LCO BASES 3.4.12 is attached.
ITAAC Change:
None 440.783F-2
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E NRC FSER OPEN ITEM
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BASES [LCO 3.4.12]
APPLICABILITY in MODES 1,2,3 and 4 the ADS must be OPERABLE to mitigate the potential consequences of any event which causes a reduction in the RCS inventory, such as a LOCA.
l The requirements for the ADS in MODES 5 and 6 are specified in LCO 3.4.13. " Automatic Depressurization System (ADS) - Shutdown, RCS Intact."
and LCO 3.4.14. " Automatic Depressurization System Shutdown, RCS Open."
ACTIONS &l If any one, or if two flow paths, consisting of one stage 1 and one stage 2 or 3 are determined to be inoperable, the remaining OPERABLE ADS flow paths are adequate to perform the required safety function. A flow path is inoperable if one or two of the ADS valves in the flow path are determined to be inoperacle. A Comp l6;;en T;me of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ;s acceptede a;nce ;he Oi'CilACLC AOC pe:he ;en m.:;;;;; COAs #:hout a s;ngle fe;lure. A Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is reasonable based on the capability of.
the remaining ADS valves to perform the required safety functions assumed in the safety analysis and the low probability of a DBA during this time period. This completion time is the same as is used for two train ECCS systems which are eteable of performing their safety function without a single failure.
B.1 and 8.2 If the Required Actions and associated Completion Times are not met or the requirements of LCO 3.4.12 are not met for reasons other than Condition A, the plant must be brought to MODE 5 where the probability and consequences on an event are minimized. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Timee are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner, without challenging plant systems.
g 440.783F-3
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