NSD-NRC-97-4989, Requests Responses to NRC Request for Addl Info & Updates of Status for Several Areas Being Reviewed by Ecgb & Emeb for AP600
| ML20134Q175 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 02/19/1997 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD-NRC-97-4989, NUDOCS 9702260304 | |
| Download: ML20134Q175 (44) | |
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J Westinghouse Energy Systems Ba 355 Pittsburgh PennsyNania 15230-0355 Electric Corporation NSD-NRC-97-4989 l DCP/NRC0743 f Docket No.: STN-52-003
! February 19,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 TO: T.R. QUAY
SUBJECT:
RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION AND OPEN ITEMS ASSOCIATED WITH SSAR CHAPTER 3 l
Dear Mr. Quay:
In a letter dated February 7,1997, the NRC provided additional requests for additional information '!
- and updates of status for several areas being reviewed by ECGB and EMEB for AP600. Attachment
! I to this letter provides information and responses to a number of these items. Many of the items are
! resolved and do not require a response. The responses to other items will be provided later. The l responses are grouped by the enclosures of the NRC letter. Also attached are markups of SSAR
! revisions that will resolve a number of these items. These changes will be included in Revision 11 of .
the SSAR.
The resolution of the items addressed in the attachment will permit the NRC staff to provide input to the FSER for a number of the subsections.
If you have any questions please contact D. A. Lindgren at (412) 374-4856.
A Brian A. McIntyre, Manager ;
Advanced Plant Safety and Licensing jml Attachments j W
'I i- cc: D. Jackson, NRC (w/ attachments)
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. - i Attachment I to NSD-NRC-97-4989 NRC letter Enclosure 2 i
- 3. Open Item 3.2.2-1 (OITS 564) - Classification of Emergency Core Cooling System (ECCS)
Actinn W In a letter to Westinghouse dated August 20,1996, this open item was reported by the staff as being i
resolved. However, before this issue is considered resolved, the staff needs the following information
- l. and/or clarifications in the SSAR:
- a. The staff has identified the components and systems listed below as part of ECCS systems that are ;
classified as AP600 Class C (ASME Class 3):
In-containment refueling water storage tank (SSAR Fig. 6.3-2)
- Accumulator (SSAR Fig. 6.3-1)
Accumalator injection piping to discharge check valve V-028 (SSA'R Fig. 6.3-1) l Containment recirculating piping and valves to in-containment refueling water storage tank (IRWST) injection check valve V-122 (SSAR Fig. 6.3-1) *
- Piping from 1st,2nd & 3rd stage automatic depressurization valves (ADV) to the IRWST, !
including depressurization spargers (SSAR Fig. 5.1-5 & 6.3-2) :
Westinghouse is requested to verify in the SSAR Subsection 3.2.2.5, that all of the above components and systems and any other Class 3 ECCS not listed above are included in the commitment to random radiography for all ECCS.
commitment. Since this commitment is not stated in either Table 3.2-3 or applicable P& ids, how can the staff be assured that it will be implemented on all AP600 plants? -
Westinohnuse Resnonse
- a. Information will be added to SSAR subsection 3.2.2.5 to list the portions of systems to which the augmented weld inspection applies. The IRWST is not fabricated as a free standing tank but is formed using portions of containment internals structural modules. Reference to the requirements in 3.8.3.6.2 for inspection of the structural modules that form the IRWST will be included in subsection 3.2.2.5. A markup of these additions is attached.
- b. Design and fabrication requirements such as the need for these inspections are included in internal AP600 design documents. A reference to the requirements in subsection 3.2.2.5 will be included in SSAR subsection 6.3.2.3. A markup of this addition is attached.
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Attachment 1 to NSD-NRC-97-4989 ,
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- 7. RAI 210.219 (OITS 3509) - Table 3.2-3, Passive Cont. Cooling System !
Actinn N The response to this issue in the letter dated December 2,1996, is being evaluated by the staff.
1 Wentinghnume Resnnnte '
SSAR did not include all of the SSAR revision included in the December 2,1996 letter. These will be included in Revision 11 of the SSAR.
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- 9. RAI 210.221 (OITS 3512). Table 3.2-3, Reactor System Action W Revision 10 to the SSAR, Trble 3.2-3 provides acceptable responses to RAI 210.221a through d.
However, the response to 210.221e is not acceptable. This portion of the RAI requested the basis for the Core Barrel Nozzle to be Class D and non-seismic when the Core Barrel is Class B and Seismic Category I. In a letter dated December 2,1996, the response to this request states that the seismic classification of the nozzle would be changed to Category II, and the safety classification would remain as Class D because the nozzle does not provide core support and does not have to be safety-related. The staff's position is that the nozzle is an integral part of the core barrel (which is a safety-related component), and therefore should have the same safety and seismic classifications as the barrel. Table 3.2-3 should be revised to change the nozzle to be AP600 Class B and Seismic Category I. Therefore, OITS 3512 remains open.
Westinghnuse Remnonse The core barrel nozzle will be changed to have the same classification as the core barrel.
Westinghouse has determined that based on the criteria in subsection 3.2.2.5, the appropriate classification is Equipment Class C and the core support structures will be changed from Class B to Class C. Class C is a safety-related class to which 10 CFR Part 50 Appendix B applies and the seismic category does not change. Also the guide tube assemblies will be changed from Class D, seismic Category II to Class C, seismic Category I. A markup of these changes is attached.
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- 13. Open Item 3.6.2-1 (OITS 592) - Subcompartment Design I Action W The response to this issue in the letter from McIntyre to Quay dated October 23,1996, does not appear to contain the detailed information requested by the staff during the review meeting with Westinghouse on July 25 & 26,1995. As stated in the DSER, Section 3.6.2, page 3-94, the staff's !
position is that a minimum subcompartment pressure which bounds the effects of a high energy pipe i break (with consideration of leak-before-break (LBB) acceptance) must be determined. Specifically, the staff requests that for all subcompartments both inside and outside containment, SSAR Subsections 3.8.3.5 and 3.8.4.3.1.4 be revised to state that those compartments containing high energy piping are designed to the worst case of either the 5 psi load (the 7.5 psi load for the CVS room) or the double ended pipe rupture of the applicable high energy pipe.
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1 Attachment I to NSD-NRC-97-4989 Westinghnuse Rernonse The subcompartment design pressure bounds the effects of postulated breaks. The SSAR will be revised to specify that.
SSAR Revisions The sixth paragraph of subsection 3.8.3.5 will be revised as follows:
The determination of pressure and temperature loads due to pipe breaks is described in subsections 3.6.1 and 6.2.1.2. Subcompartments inside containment containing high energy piping are l i designed for a pressurization load of 5 psi. The pipe tunnel in the CVS room (room 11209, i i Figure 1.2-6) is designed for a pressurization load of 7.5 psi. These subcompartment design I
pressures bound the pressurization effects due to postulated breaks in high energy pipe. The l design for the effects of postulated pipe breaks is performed as desenbed in subsection 3.6.2.
Determination of pressure loads resulting from actuation of the automatic depressurization system is described in subsection 3.8.3.4.3.
In subsection 3.8.43.1.4 the paragraph under P, will be revised as follows:
The main steam isolation valve (MSIV) and steam generator blowdown valve compartments i l are designed for a pressurization load of 5 psi. The subcompartment design pressure bounds !
I the pressurization effects due to postulated breaks in high energy pipe. Determination of I l subcompartment pressure loads is discussed in subsection 6.2.1.2.
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- 15. RAI 210.40 (OITS 3702) - Break Exclusion in Steam Generator (SG) Blowdown, Startup FW, ;
and Chemical and Volume Control System (CVS) Lines Action W l I
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In a letter from McIntyre to Quay dated October 23,1996, and in the OITS 3702 report, )
Westinghouse states that additional information on the startup line, including the isometric drawmgs i will be provided during a forthcoming meeting with the staff. This issue will be discussed during the next meeting or a telephone conference.
In addition, Revision 10 to SSAR Subsection 3.6.2.1.1.4 added portions of the Chemical and Volume Control System (CVS) to the list of break exclusion areas. These new areas include makeup piping from containment to the anchors upstream of the outside isolation valve and downstream of the inside isolation valve, including branch connections. Revision 10 did not revise SSAR Figure 3E-5 to identify these areas. Therefore, the staff requests more information relative to the exact location of the anchors, the length of piping from the inside and outside isolation valves to each anchor, and the location and lengths of all applicable branch lines.
Westinghouse Resnonse Westinghouse is revising the break exclusion area for the CVS makeup line to be from the outside containment isolation valve to the inside containment isolation valve. Westinghouse has provided
- copies of the isometric drawings for this line for review in the Westinghouse Rockville hcensmg j
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Attachment 1 to NSD-NRC-97-4989
- ' office. This change will add two additional break locations to be considered in the evaluation of high .
energy breaks. The fluid in this line is cold and a break will not result in pressurization. A copy of l the SSAR markup showing changes for the revised break exclusion area are attached. l 1
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! 16. Open Item 3.6.2.3-1 (OITS 595) - Break Locations and Stress Summary
, Action N
! 1 In Revision 10 to the SSAR contains a significant revision to Subsection 3.6.2.5 which provides additional information on the pipe break hazard analysis. The staff's preliminary evaluation of this i submittal resulted in the following request.
l As discussed under Open Item 3.6.2.3-5 (OITS 599) below, Westinghouse has submitted a l
- revision to SSAR Subsection 3.6.1.3.2 which refers to the pipe rupture hazards analysis. In j j addition to the information in Revision 10, (1) add a reference in SSAR Subsection 3.6.2.5, to the ;
j new information in Subsection 3.6.1.3.2 that is applicable to the hazards analysis, and (2) in Subsection 3.6.4.1, state that the as-built reconciliation of the hazards analysis will be in l accordance with the criteria in SSAR Subsections 3.6.2.5 and 3.6.1.3.2.
i Westinehouse Resnonse
$ (1) A reference to the criteria in subsection 3.6.1.3.2 will be added to the paragraph under Essential -
l Target Evaluation in subsection 3.6.2.5.
(2) A reference to the criteria in subsection 3.6.1.3.2 and 3.6.2.5 will be added to subsection 3.6.4.1.
I i A markup is attached.
- 18. Open Item 3.6.2.3-5 (OITS 599) - Separating Structures j Action W i
In Revision 10 to SSAR Subsection 3.6.1.3.2, information was added which provides a basis for
- resolving this issue as a part of the pipe rupture hazards analysis. Based on a preliminary review of
- this submittal, the staff has no further requests for information except to repeat the request in this open item to delete the exception to the standard review plan (SRP) Section 3.6.2 BTP MEB 3-1, j Section B.1.c.(4) in WCAP-13054, Revision 2. !
4 l Eprinohnuse Response j' WCAP-13054 will be revised to remove the exception to criteria B.1.c.(4). A markup of the change j
is attached.
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Attachment I to NSD-NRC-97-4989
- 23. Open Item 3.9.3.3-2 (OITS 793) - Anchor Bolts for Pipe Supports ;
, Action W '
s In a letter from McIntyre to Quay dated October 23,1996, Westinghouse responded to this item l$y
) referencing Revision 9 to SSAR Subsection 3.9.3.4. Revisions 9 and 10 contain no change to this portion of Subsection 3.9.3.4. It still commits only to the baseplate flexibility requirements ofIE Bulletin 79-02 and is silent on the factors of safety for concrete expansion anchor bohs. Since the factor of safety issue is being evaluated by the staff under DSER Open Item 3.8.4.2-2, Subsection 3.9.3.4 should contain a reference to the applicable portion of SSAR Subsection 3.8.4 for information
- relative to these factors of safety.
Westinohnuse Resnonse
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Supplemental requirements for fastening anchor bolts to concrete are provided in subsection 3.8.4.5.1.
reference to these requirements will be added to subsection 3.9.3.4. A markup of this addition is
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- 24. Open Item 3.10-1 (OITS 813) - Use of Seismic Experience Data
- Action W i
l Revision 10 to SSAR Section 3.10.6 states that the COL applicant, as a part of the Combined License
] application, will identify equipment qualified based on experience and include detatis of the
! methodology and the corresponding experience data. This agrees with the staff's request on this item, and is acceptable. However, the exception to SRP 3.10 in Revision 2 to WCAP-13054 contains statements which either need to be deleted or clarified. The first two sentences imply that IEEE 344-1987 is acceptable relative to the use of experience data. Regulatory Guide (RG) 1.100, Revision 2 states that this method of qualification in IEEE 344-1987 will be evaluated by the staff on a case-by-
, case basis. It appears to the staff that the exception in the WCAP is relative to RG 1.100, Revision 2.
l These two sentences should be revised to reflect the position in RG 1.100, Rev. 2. In addition, the i discussion relative to Generic Issue A-46 is not applicable to new plants. The staff's position is that
! A-46 is only used for verification of equipment in operating plants, and is not acceptable for i qualification of equipment in advanced light water reactors (ALWRs). This discussion should either be deleted or revised.
! . Westinohnuse Resnonse The discussion in WCAP-13054 on the Criteria 1 for SRP 3.10 will be revised to clarify that the j exception is to the revision on Regulatory Guide 1.100 and IEEE 344 and that the AP600 is in conformance with Regulatory Guide 1.100, Revision 2. The requirement that the combined license anplicant identify use of experience based data for equipment qualification and the methodology used i
is . cluded in subsection 3.10.6. A draft markup of the exception in WCAP-13054 is attached.
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Attachment 1 to NSD-NRC-97-4989 NRC Letter Enclosure 5
- 1. DSER# 3.9.2.3-2 (783) - Flow-induced vibration prediction analysis Actinn W In SSAR Revision 10, the first paragraph of Subsection 3.9.2.3 indicates that the flow-induced vibration assessment is documented in WCAP-14761, which is also included in the reference list in SSAR Section 3.9.9. This is acceptable. The WCAP-14761 is a replacement of previous report MIO1-GER-001, which was submitted by Westinghouse and reviewed by the staff and found acceptable.
However, the reactor internals of the first AP600 plant is designated as the prototype as defined in SRP 3.9.2 and RG 1.20 for vibration assessment of AP600 reactor internals. Information of vibration assessment from reference plants, which include H. B. Robinson, DOEL 3 and 4, etc. may only be used in vibration prediction analysis for the prototype and should not be confused with the prototype. l The wording in SSAR Sections 3.9.2.3 and 3.9.2.4 should be revised to avoid confusion between the
" prototype" and the " reference plants."
Westinghnuse Resnonse Subsections 3.9.2.3 and 3.9.2.4 will be revised to delete the portion that stated that prototype and reference were equivalent. A markup of the changes is attached.
- 3. DSER# 3.9.5-1, RAI 210.226, (OITS 3517)- 20% damping value for fuel assemblies Resolved Information provided in Westinghouse letter NSD-NRC-97-4933, dated 1/8/97, indicates that the damping value is justified by testing and is consistent with evaluations for Westinghouse-designed fuel in operating nuclear power plants. This is acceptable. However, Westinghouse needs to provide a suitable reference in the SSAR.
Westinghnuse Rennonse A paragraph will be added to subsection 3.9.2.6 to identify the damping for the fuel assemblies and reference WCAP-8236. A markup of the additions is attached.
NRC Letter Enclosure 1
Action W Revise the SSAR to reflect correct reference of OM Standards, OMa-1988 or the 1990 Edition of the OM Codes. From the January 31,1997, telephone conference, Westinghouse will send a letter requesting an exemption and revise the SSAR accordingly.
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Attachment 1 to NSD-NRC-97-4989 Westinohnuse Response
- AP600 used the 1990 Edition of the Code as the baseline for preparation of the inservice testing plan.
Letter NSD-NRC-97-4986, dated February 14,1997 requesting an exemption from the requirements in 10 CFR 50.55a for an earlier version of the Code has be sent to the NRC. The fourth paragraph of 4
subsection 3.9.6 will be revised to remove the reference to ANSI in the name of the Code. A markup l of the changes is a+tached.
) 3. RAI 210.228 - SSAR 3.9.6 (IST)
. Action W i J
. The main feedwater (FW) check valves, SGS-V058A/B appear to have a safety function to close based
} on SSAR Subsection 10.4.7.1.1. Revision 10 still indicates that these valves have a safety-related
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function, but they are not included in the ISTP (IST Program). 'In the January 31,1997, telephone i
conference, Westinghouse stated that these valves do not perform a safety function and will move the
- FW check valve description from SSAR Subsection 10.4.7.1.1 to SSAR Subsection 10.4/1.1.2.
Westinohnuse Resnonse
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a j These valves do not perform a safety function. The description of the feedwater check valve function is located in an incorrect location and will be inoved from SSAR Subsection 10.4.7.1.1 to SSAR Subsection 10.4.7.1.2. A markup of the changes is attached.
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) 5. RAI 210.230 - Safety Relief Valve Test - SSAR 3.9.6 (IST) i Actu.;n.E 4
J The SSAP. (Rev.10) was revised to include "5 years and 20% in 2 years. For Class 2/3, the SSAR i should '>e revised for consistency to include "10 years and 20% in 4 years." From the January 31,
- 1997, telephone conference, Westinghouse will revise the SSAR.
l i Westinohnuse Resnonse j The column in Table 3.9-16 will be revised to include 10 years and 20% in 4 years for the j appropriate valves. A markup of the changes is attached.
Actinn W '
i The discussion of Issue 87 in SSAR Section 1.9 should be revised to state tret valves built to Section i
III are required to be tested in accordance with the ASME Code. Revision 10 still states that these i j valves may be tested in accordance with the OM Code. From the January 31,1997, telephone j
! conference, Westinghouse will revise the SSAR. ;
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,' j Attachment 1 to NSD-NRC-97-4989 '
Westinohnuse Remnonse The second paragraph of the response for Issue 87 in Section 1.9 will be revised to indicate that compliance with the ASME OM is required rather than being an option. Also, the response will be revised to make the response consistent with the position in subsection 3.9.6.2.2 on testing of valves.
A markup of the changes is attached.
NRC latter Enclosure 6
- 2. OITS Item No. 801 Action W
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Note: This is only a portion of this item. The balance of the question will be answered with items '
related to valve qualification.
Westinghouse had added SSAR Subsection 3.9.6.2.3 to address valve disassembly and inspection.
The disassembly and inspection program must also be addressed in Section 3.9.8, as the COL will have to develop this program.
Westinohnuse Remnonse Specific reference to the valve disassembly and inspection program will be added to subsection 3.9.8.4. A markup of the SSAR changes is attached.
- 3. OITS Item No. 805 Actinn W Valves RNS-V002A/B are CIVs and are Type C tested per SSAR Table 6.2.3-1. These valves should '
be leak tested in the ISTP, Table 3.9-16. In the January 31,1997, telephone conference, Westinghouse will revise SSAR Table 6.2.3-1.
1 Westinohnuse Remnonse Because of the function and lay out of the normal residual heat removal system these valves are not subject to a containment leak test. The notes for these valves in Table 6.2.3-1 will be revised to clarify this and be consistent with the information in Table 3.9-16. A markup of the changes is attached.
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- 4. OITS Item No. 807 Actinn W Per NRC comments on the testing deferral, revise N9te 4, Note 9, Note 11, and Note 21 in SSAR Table 3.9-16. In the January 31,1997, telephone conference, Westinghouse stated they will provide additional justification on the use of solenoid operated valves for the head vent (Note 4); provide additional information on "sufficiently long" cold shutdown times (Note 9); delete RNS-PI V046 from xm 8
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- j Attachment 1 to NSD-NRC-97-4989 a
j Note 21 and the table (Note 21); provide additional information on the testing capabilities of VES-PL- !
V0008 A and B (Note 21). The staff will respond to the Westinghouse December 17,1996, letter
] regarding Note 11. '
Wantinehane, p cnonne i
) Note 4 will be revised to identify the potential for quarterly testing to cause valve leakage. Solenoid i
valves are chosen for this application because they are the best overall choice to meet the several i disparate design requirements. These valves have safety-related functions to transfer open and to
- transfer close. Air operated valves are not well suited to such an application because they are t
.normally capable of safety-related transfer in only one direction. To achieve safety-related transfer capability in two directions requires the use of a piston operator and a safety-related air supply.
! Motor-operated valves are also not well suited to this application because they are larger and heavier
, with an extended operator that makes them difficult to locate and support. Motor-operated valves are i
also less reliabic. Both air operated valves and motor-operated valves have packing which is subject to leakage.
i Note 9 will be revised to add a partial exercise testing of the accumulator check valves at longer cold l shutdowns. Full stroke exercise testing during refueling shutdowns is maintained.
j Note 11 will be revised to provide additional information on the test device used to open the !
recirculation check valve disks.
i' Reference to Note 21 will be deleted from the entry for valve RNS-PleV046. )'
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A markup of the changes is attached.
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- 5. OITS Item No. 809 l Action W i
i Delete the first sentence of the second paragraph in SSAR Subsection 3.9.6.2, and move the second
- sentence to the end of the paragraph. Westinghouse will revise the SSAR.
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] Westinohouse Recnonse This paragraph will be revised to clarify the requirements for testing of valves with RTNSS important !
- missions. A markup of the changes it ettached.
- 6. OITS Item No.1730 j Actinn W i
- Valve PCS-V014A is a normally closed stop check valve (P&ID Fig. 6.2.2.1 of the SSAR, Rev. 6
p and Rev. 9) with a safety function to open. However, no check exercise is specified and the valve is !
still identified as a Category B valve in Revision 10. Westinghouse will revise the SSAR. )
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- Attachment I to NSD-NRC-97-4989
- Wantinohnnea Rannonse i A check exercise test will be included in Table 3.9-16 for this valve. A markup of the change is attached.
. 7. OITS ltem No.1731 Actinn W
. Westinghouse has stated that valves RCS-PL-150VA-D have an active function to move to the open .
position. However, SSAR Fig. 5.1-5 identifies these valves as failed closed. The IST Table should )
be revised such that these valves are subject to a fail-safe test, or the P&ID should be revised.
) Westinghouse will include the testing in the SSAR. I 1
l Wactinohnuse Rannonse Table 3.9-16 will be revised to include a safety function of Active-to-Fail for these valves, j 1
- 13. RAI Q952-96 l Actinn N i
In a letter response dated May 13,1996, Westinghouse continues to state that the ADS valves will be j
-tested at conditions determined with input from type selection testing. The qualification testing of the .!
- prototypical ADS valves should be performed under design basis conditions. From the January 31,
- 1997, telephone conference, Westinghouse stated the response did not apply to ISTP. The staff will
- review the issue further for acceptability.
i Westinohnuse Rannonne 1 l j This question is in reference to a response to an RAI about ADS system testing. Westinghouse has !
i been very specific that these tests were not valve qualification testing. Since this RAI response was 1 not about. valve qualification, there is no need to revise it to address a valve qualification question.
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- 1. Iitroduction cnd General Description of Plant Issue 87 Failure of IIPCI Steam Line Without isolation Discussion:
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Generic Safety Issue 87 addresses the uncertainty regarding the operability of the motor-operated isolation valves for the steam supply lines of the high-pressure coolant injection (HPCI) system in boiling water reactors following a postulated break in the supply line. A ,
break in the line could lead to high flow or high differential pressure that may inhibit closure I of the isolation valve. These valves typically cannot be tested in-situ for the design flow rates l
and pressures. Although the AP600 does not have a high-pressure coolant injection system, i it does have isolation valves designed to close against high flow or high pressure differential I in the event of a postulated pipe break.
The issue of the operability of motor-operated valves has received considerable attention since Generic Safety Issue 87 was initiated. The NRC provided guidance for inservice testing of l motor-operated, safety-related valves in Generic Letter 89-10. SECY-93-087 identifies the proposed position on inservice testing of safety-related valves for advance light water reactors.
The guidance in these documents recommends that safety-related valves be tested under full flow under actual plant conditions where practical. EPRI has a program to demonstrate operation of motor-operated valves.
AP600 Response l
Safety-related valves must meet the requirements of ASME Code,Section III to provide pressure boundary integrity. Valves and valve operators are sized to provide operation under l a full range of design basis flow and pressure drop conditions. For the AP600, safety-related motor-operated valve designs are subject to qualification testing to demonstrate the capability of the valve to open, close, and seat against maximum pressure differential and flow. The requirements for this testing are based on ANSI B16.41, " Functional Qualification Requirements for Power Operated Active Valve Assemblies for Nuclear Power Plants." See subsection 5.4.8 for an outline of AP600 valve requirements.
fAI The in-service testing program for safety-related valves is discussed in subsection 3.9.6. 2jei, EV l "'here prac:ical, mMotor-operated vaTv'es =d50!mM";'.Je T6 be opelability tested as I outlined in subsection 3.9.6.2.2. = der fu!! ficv = der ac:=1 p!=: ecadi:i=:;. Subsection 1 3.9.6.2.2 includes a discussion of the factors to be considered to determine which valves and I the test conditions to used for oberability testing of power-operated valves. Sufficient flow I is provided to fully open check valves during testing unless the maximum accident flows are I not sufficient to fully open the ch_cc)Lyalvey fthe valves builLIILASME Code, Section 11 a may beysted in compliance with the requirements found in the ASME code, "CodeTor peration and Maintenance of Nuclear Power Plants." For additional information on inservice testing of safety-related valves, see subsection 3.9.6.
Revision: 11 February 28,1997 1.9-62 [ W8Stiflgt100S8
- 3. Desig2 of Structures, Compone ts., Eq:1pment and Systems Handle spent fuel, the failure of which could result in fuel damage such that significant quantities of radioactive material could be released from the fuel and results in offsite doses greater than normal limits (for example, new and spent fuel racks, the bridge, and the hoist)
- Maintain spent fuel sub-critical
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Monitor radioactive effluent to confirm that release rates or total releases are within limits established for normal operations and transient operation
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Monitor variables to indicate status of Class A, B or C structures, systems, and components required for post-accident mitigation Provide for functions defined in Class B where structures, systems, and components, or portions thereof are not within the scope of the ASME Code,Section III, Class 2.
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Provide provisions for connecting temporary equipment to extend the use of safety related systems. See subsection 1.9.5 for a discussion of actions required for an extended loss of onsite and offsite ac power sources.
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'Ihe.' components and portions of systems that provide emergency core cooling functions and 7.2.2-/
are required to have radiography of a random sample of welds during construction include the following:
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A. cetimulators
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Injection ~ piping frorn the accumulators to the reactor coolant'sy' stem is~olation ch'eck valves in the direct vessel injection line Piping from the in-containment refueling water storage tank (IRWST) and recirculation screens to the reactor coolant system isolation check valves in the direct vessel injection l
line Piping from the Stage 1, 2, and 3 automatic depressurization system . valves'to'the IRWST including the spargers.. i The ~IRWST.is fonned from portions of structural modules that are elements of the containment inernal structures. The inspection requirements for the welds in these stmetural modules are provided in Subsection 3.8.3.6.2.
3.2.2.6 Equipment Class D Class D is nonsafety-related with some additional requirements on procurement, inspection or monitoring.
Revision: 11 ow,niivno2n.nii42:297 Draft,1997 3.2-8 T Westinghouse
- 3. Desig2 of Struct:res, Components, Equipment, cnd Syst:ms Table 3.2-3 (Sheet 31 of 61)
AP600 CLASSIFICATION OF MECHANICAL AND FLUID SYSTEMS, COMPONENTS, AND EQUIPMENT j
i Tag Number Description AP600 Seismic Principal Co:n- Comments Class Category struction Code Reactor System (Continued)
RXS-MI-01 Reactor Upper Internals C I ASME III, CS RXS-MI-02 Reactor Lower Internals Cs I ASME III, CS i RXS-MI-10 Non-Threaded Fasteners D NS ASME III, CS RXS-MI-11 Threaded Structural Fasteners hI ASME HI, CS RXS-MI-20 RXS-MI-21 Lower Core Support Plate Secondary Core Support hD I
II ASME III, CS ASME III, CS i RXS-MI-22 Vortex Suppression Plate D II ASME III, CS RXS-MI-23 Radial Reflector Assembly D II ASME III, CS RXS-MI-24 Radial Supports [4] CB I ASME IH, CS RXS-MI-25 Core Barrel CB I ASME HI, CS 6
RXS-MI-26 RXS-MI-27 Core Barrel Nozzle Head and Vessel Pins h D I-H II ASME HI, CS ASME HI, CS I
[ RXS-MI-28 Lower Support Plate Fuel Alignment Pins h I ASME HI, CS RXS-MI-29 Core Barrel Hold Down Spring h I ASME HI, CS RXS-MI-50 Upper Support CB I ASME HI, CS RXS-MI-5I Upper Core Plate CB- I ASME III, CS RXS-MI-52 Support Columns [38] @ I ASME III, CS RXS-MI-53 I-H ANSI B31.1 Guide Tube Assemblies [61] @
RXS-MI-54 Upper Support Plate Fuel Alignment Pins h I ASME HI, CS RXS-MI-55 RXS-MI-56 Upper Core Plate Inserts Safety Injection Deflector hID II ASME HI, CS ANSI B31.1 RXS-MI-57 Irradiation Specimen Guide D II ANSI B31.1 Tubes RXS-MI-58 Head Cooling Nozzles D II ANSI B31.1 RXS-MV-10 Reactor Integrated Head C I AISC-690 Packag:
RXS-MV-10A Integrated Head Package C I ASME-NF Shroud RXS-MV-10B Integrated Head Package C I ASME-NF Seismic Support Plate Revision: 11 Draft,1997 3.2-50 3 West lrighouse
==
- 3. Desigh of Structures, Components, Equipmest, and Systems l
l 1
=
For evaluation of spray wetting, flooding, and subcompartment pressurization effects, longitudinal enicks (with crack flow areas of I square foot) are postulated in the main steem and main feedwater piping. The dynamic effects of pipe whip and jet irapingement ce not evaluated for these cracks. Locations having the greatest effect on essential equipment are chosen.
Guard pipe assemblies for high-energy piping in the containment annulus region between 5
the containment shell and shield building that are part of the containment boundary are designed according to the rules of Class MC, subsection NE, of the ASME Code. The ;
following requirements also apply. The design pressure and temperature are equal to or greater than the maximum operating pressure and temperature of the enclosed process pipe under normal plant conditions. Level C service limits of the ASME Code, l Section III, Paragraph NE-3221(c), are not exceeded by the loadings associated with containment design pressure and temperature in combination with a safe shutdown earthquake. The guard pipe assemblies are subjected to a pressure test performed at the maximum operating pressure of the enclosed process pipe. i l
Areas of system piping where no breaks, except as noted in subsections 3.6.1.3 and 3.6.1.2.2, are postulated are as follows:
l The main steam piping, from the containment penetration flued head outboard weld, to the upstream weld of the auxiliary building anchor downstream of the main steam i isolation valves, including the main steam safety valves and the connecting branch piping The main feedwater piping l from the containment penetration [16[dRE@iitb'ojifdj(eld]
to the auxiliary building anchor upstream of the isolation valve -including r 4 ranch connections-
=
The startup feedwater piping from the containment penetration to the auxiliary building anchor upstream of the isolation valve inekuling branch connections The steam generator blowdown piping from the containment to auxiliary building anchor i downstream of the isolation valve gc,. N
= '
The chemical and volume control system makeup piping from the containment to or the&nd= p:tr::= cf the outboard isolation valve !ne! din;; b=nd ====$=. g h The chemical and volume control system makeup piping from the containment t g %:i= de r=tr== cf!he inboard isolation valve ! c!:d!:;; b=nd ====1!:=.- k All other fluid system containment penetrations are for moderate-energy systems or for pipe of 1-inch nominal diameter or smaller. See subsection 6.2.3 for a discussion of containment penetrations.
Revision: 10 Y W85tiligh00Se 3.6-17 December 20,1996
,- , en
- 3. Design of Structures, Componerts, Equipment, and Systems Essential Target Evaluation } M To complete the essential target evaluation jet parameters, volumetric area of affected compartments, plant layout, and separating structures are considered. Parameters that determine the shape of the jet and the magnitude of the jet and thmst loads include pressure, ,
temperature, and friction losses between the break and the reservoir. The volumetric area affected is determined by considering jet shape and loads at the postulated location of the breaks. Where an initial evaluation of essential targets indicated adverse effects, layout may be changed to relocate the target or postulated break. If necessary, the location of whip of 1 restraints and jet shields is established to protect essential systems and components. Essential 7.4 2.3-/
l equipmentytected hy_ pipe _yzlip restraints orjet shields is listed in Table 3.6-3]lhe criteria l l or the break location postulated for evaluation of separating stmetures is outlined in I subsection 3.6.1.3.2. ~
Verification of the Pipe Break Hazard Analysis )
The ASME Code, Section 111 requires that each plant have a Design Report for the piping system that includes as-built information. Included in the Design Reports are the loads and loading combinations used in the analysis. Where mechanistic pipe bmak requirements are used to eliminate the evaluation of dynamic effects of pipe rupture in ASME Code,Section III, Class.1,2, and 3 piping system, the basis for the exclusion is documented in the Design Report.
As-built reconciliation of the pipe break hazard analysis is addressed by the Combined License applicant.
3.6.2.6 Evaluation of Flooding Effects from Pipe Failures) @ _
The effect of flooding due to high and moderate energy pipe failures on essential systems and components is described in Section 3.4.
3.6.2.7 Evaluation of Spray Effects from High- and Moderate-Energy Through-Wall Cracks) b Essential systems and components are evaluated for the potential effects of spray from high-and moderate-energy through-wall cracks. Spray effects are assumed to be limited to the compartment where the pipe failure occurs. The spray is assumed to wet unprotected components in the compartment. It is further assumed the spray does not damage non-electrical passive components, including piping, ducts, valve bodies, or mechanical components of valve operators. Spray may cause failure of electrical components not designed to withstand wetting. Components protected by NEMA 4 or NEMA 12 enclosures are not affected by spray effects.
The safe shutdown components inside containment are subject to wetting from design basis events inside containment. These conditions bound the effects of spray from moderate energy cracks. Sensitive components are qualified for this environment as described in Section 3.11.
J Revision: 10 December 20,1996 3.6-30 W Westingh0use
- 3. Design of Structures, Components, Equipment, and Systems 3.6.3.4 Documentation of Leak before-Break Evahations The leak-before-break evaluation is used to support the elimination of dynamic effects of pipe breaks from the loading conditions for the piping analysis. An evaluation ofleak-before-break using the as-built configuration of the piping system and suppons is required as part of the Design Report of the as-built configuration required to meet ASME Code requirements.
Appendix 3B contains a discussion of the bounding analysis methods for the leak-before-break evaluation.
The analysis methods, criteria, and loads used for evaluation of stress in piping systems are ontlined in subsections 3.7.3 and 3.9.3. The seismic input bounds the soil design profiles
' outlined in subsection 3.7.1.4 and Appendicies 2A and 2B. The evaluation also bound soil profiles qualified using site specific evaluations as outlined in subsection 2.5.4.5.5 3.6.4 Combined License Information )du 3.6.4.1 Pipe Break Hazard Analysis
} Tw@ ,
Combined License applicants referencing the AP600 cenified design will address as built L d 2.f -/
l reconciliation of the pipe b_reak hazagrdin i
accordance witli~1he criteria outIg
@ ions 3.6.1.3.2 and 3.6p N
3.6.4.2 Leak-before-Break Evaluation } DiEDL47 Combined License applicants referencing the AP600 cenified design will address:
- 1) verification that the as-built stresses, diameter, wall thickness, material, welding process, pressure, and temperature in the piping excluded from consideration of the dynamic effects of pipe break are bounded by the leak-before-break bounding analysis; 2) a review of the Certified Material Test Reports or Certifications from the Material Manufacturer to verify that the ASME Code,Section III strength and Charpy toughness requirements are satisfied; and
- 3) complete the leak-before-break evaluation by comparing the results of the final piping stress analysis with the bounding analysis curves documented in Appendix 3B.
3.6.5 References
} , g):-
- 1. NUREG/CR-2913, "Two-Phase Jet Loads," January 1983.
March 1977.
- 3. ASME/ ANSI-B31.1, Code for Power Piping,1989 Addenda to 1989 Edition.
- 4. ANSI /ANS-58.2-1988, " Design Bases for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture."
- 5. Moody, F. J., Fluid Reaction and Impingement Loads, paper presented at the ASCE Specialty Conference, Chicago, December 1973.
Revision: 10 December 20,1996 3.6-38 [ Westiflgh0tlSe
1 .
a h,s ' -
O' l .- a? ..
- 3. Design of Structures, Components, Equipment, and Systems Table 3.6-3 ROOMS WITH HIGH ENERGY PIPE BREAYS AND IOTENTIAL ESSEN'DAL TARGET INIERACHON Elevation Room IIigh Energy Essential Equipment Protected by Whip Restraints or Numben* Break Sourte Jet Shields >
66'-6" None 82'-6" f120J RCS Press.,Spra RCSsADS 'viilvesiN004KlN0Q4CAV0_{4AE014C ETefminal.End 11204 None 11209 None 96*-6" 11204 None 11209 None '
100'-0" and 107'-2" 11209 SGS E16sid6pH CVS hiaksupl CVS Isid6KCVSJIfdi6ffei[56j513 and Pipe chase Piping - Terminal SGS steam generator blowdown pipmg )
Fnd
.11 (CVS Makeup fp LoM Ji!" 3A-) I CVS Makeup valve V091 ,, g i Pi Piping - Terminal No DD R
11300 None 11301 None 11303/ RCS hiaEshp RGS-and-SGS sg")]6w' d6Whishdji~drais Piping, 11304 Piping - RCS pressurizer pressure and level instmmentation[shd Intermediate Break Pressurizer support steel 117'-6" 11400 SGS Stirt;Ui l h6h6 Raceways =d =b!= for Dimic= NC =d E/D
^
Fesdvpater Piping -
Terminal end 11401 None RCS ADS]alfc(V004&V004C,2014X}N0J4C ars yrbi'edtsdfMrs~i breik 166atal liiijsbiis:11403 11402 Steam Generator supports are protected fmm a break located in room 11400 i
Revision: 10 December 20,1996 3.6-48 W Westinghouse l
l
m"n"
- 3. Design of Structures, Components, Equipment, and Systems .
I1403 RCS Pf4SpFaj? None Tchhinal;Eid 11'403 RCSTitd6fG REaisfor DiviiU6ns A/C'andB/D hiterdsliifEBieak RNIfpis5Sp5y) RCSiADS3a!WsiV004AlN(XMC,W014A,7.V014C In:tErnreatigtig
. 135'-3" None R Q P R Sjirs9] RCS-fDSifsivEsO6w' is]Iciplitf6dii~ssphditMiel 11503, Teisisal.Tij 160'-6" and 153'-0" 11601 SGS $6iiLTI) RCS head vent piping
@Bf26 Piping - SGS;Tiv'eCuistnii6Esiitronpipihi Terminal end SGSMilil rRa%teiripidi, TEnnishl%Ed 11602 SGSjii6 None SGS]Geliisthiisestiti6n'pipirij Feediateffipin'gi Teddi5alEnd 11603 RCS BDS;Stig6J RCS piping and ADS valves 002B,003B,012B, & Ol3B Piping - Teminal Raceways and cables for Divisions A/C and B/D Eid 11703 RCS A' DSJStigEl1 RCS piping and ADS valves 002A,003A,012A, & Ol3A Piping - Teminal Raceways and cables for Division B/D X/C
&- ~ #gl 12244 CVS Makeup CVS Makeup valve V090 fG~OL1 N g p,qd Piping - Terminal goJ,3 End
- See Figures 1.2-1 through 1.2-8,1.2-10, and 12-11 for room numbers l
l l
Revision: 10 W Westingh00Se 3.6-49 December 20,1996
,. .. F7 ;
f 3. Desigm of Structures, Compone:ts, Equipment, and Systems
, L s .
i a Heat exchangers
=
l Filters
- t Passive valves -
Dynamic analysis without testing is used to qualify heavy machinery too large to be tested.
For active equipment, it is verified that deformations due to seismic loadings do not cause !
binding of moving pans to the extent that the component cannot perform its required safety l function.
Dynamic Testing Dynamic testing is used for components with mechanisms that must change position in order to perform the required safety function. Section 3.10 discusses the seismic qualification of electrical equipment and combinations of valves and valve operators. Such components include the following:
Electric motor valve operators
= Valve position sensors
=
Similar appunenances for other active valves Combinations of Analysis with Testing Combinations of aaalysis, static testing, and dynamic testing are used for seismic qualification of complex valves. Section 3.10 discusses the requirements for these combinations for equipment, which includes the following:
Main steam and main feedwater isolation valves j Other active valves i 3.9.2.3 Dynamic Response Analysis of Reactor Internals under Operational Flow Transients and Steady State Conditions The vibration characteristics and behavior due to flow-induced excitation are complex and not readily ascertained by analytical means alone. Assessment of vibrational response is
- done using a combination of analysis and testing. Comparisons of results obtained from i reference plant vibration measurement programs have been used to confirm the validity of OT scale model tests and other prediction methods as well to confirm the adequacy of reference ? '/.3, p 2. i plant internals regarding flow induced vibration [h S: fc!l: wing ..- .... .. .
"nf:x=; p!= " i: a;; =!=: :: 1: :xrc. prc:c:yp: = ==d =d if;;d in S:c;d d R: c::: ...] '
P:= 39.2 =d R:g;!::cj 0;i !.20 f= -ib.;.: ca ==nm::: cf r x:= i;;c=!;. The j
flow-induced vibration assessment is documented in WCAP-14761 (Reference 18).
Reactor components are excited by flowing coolant, which caases oscillatory pressures on the surfaces. 'Ihe integration of these pressures over the applied area provides the forcing functions to be used in the dynamic analysis of the structures. In view of the complexities of the geometries and the random character of the pressure oscillations, a closed form Revision: 11 .wmumane !
Draft,1997 3.9-34 Y W8Stkighouse
- -m
. . . )
- 3. Design of Structures, Components, Eq:lpment, cud Systems The reactor coolant canned motor pumps of the AP600, have the same rotational speed and the same number of impeller blades as in previous plants. Therefore, a significant change I in vibration is not expected. The forcing function frequencies are similar to previous plants. l For calculation of pump induced pulsations acting on the AP600 reactor internals, the j pulsation level at the pumps is taken to be the same as the level previous shaft seal pumps.
Since the horsepower of the AP600 pumps is lower than in shaft seal pumps, the shaft seal !
pulsation is a conservative analysis basis for the AP600 3.9.2.4 Pre-operational Flow-Induced Vibration Testing of Reactor Internals The pre-operational vibration test program for the reactor intemals of the AP600 conducted on the first AP600 is consistent with the guidelines of Regulatory Guide 1.20 for a comprehensive vibration assessment program. Design features that have not previously been tested in the reference plants or subsequent testing are tested to verify the vibration analysis.
Conformance with Regulatory Guide 1.20 is summarized in Section 1.9.1.
The program is directed toward confirming the long-term, steady-state vibration response of the reactor internals for operating conditions. The three aspects of this evaluation are the following: a prediction of the vibrations of the reactor internals, a preoperational vibration test program of the internals of the first plant, and a correlation of the analysis and test results.
With respect to the reactor internals preoperational test program, the first AP600 plant m ed in Regulatory Guide 1.20. The reactor vessel intemals are classified as prototype as def' AP600 reactor vessel internals do not represent a first-of-a-kind or unique design based on the arrangement, design, size, or operating conditions. The units referenced in the subsec-tion 3.9.2.3 as supporting the AP600 reactor vessel internals design features and configuration have successfully completed vibration assessment programs including vibration measurement programs. These units have subsequently demonstrated extendM satisfactory inservice operation.
g j A The[ge:c:y;M same size operatmg conditions Structural .
as thedifferences AP600eferencehlant include modifications resulting from the use of 17x17 fuel, the removal of the '.hermal shield and the change to the inverted top hat upper internals support assembly. These design changes were incorporated into the Doel 3 and Doel 4 reactor internals as well as the AP600.
The effects of these design evolutions from the reference plant were shown by instrumented preoperational testing at the Doel 3 (upper internals) and Doel 4 (lower internals) plants.
The vibrational responses of the AP600 reactor internals are characterized by the Doel 3 and 4 vibration measurement programs.
The pre-operational test program of the first AP600 plant includes a limited vibration measurement program and a pre- and post-hot functional inspection program. This program satisfies the guidelines for a Regulatory Guide 1.20 Prototype Category plant. The AP600 reactor internals design does not require supplemental testing including component vibration Revision: 11 chnll030% Ril.020797 Draft,1997 3.9-38 [ W85tiligt10US8
,. , imm
- 3. Design of Structures, Components, Equipment, and Systems considering the maximum stresses for each condition and combining them with square root of the sum of the squares method.
The system seismic analysis of the reactor vessel and its internals is either performed by a response spectrum analysis method or by a time-history integration method. Both of these analysis techniques are consistent with guidelines in the Standard Review Plan.
For cenain systems or components, when time dependeni seismic response is desired, the nonlinear time history analysis is used. The seismic time-history analysis technique is essentially the same as that for the pipe rupture analysis, except that in seismic analysis time history accelerations are used as the forcing function. The seismic response is combined with the pipe rupture response, as outlined in subsection 3.9.3, in order to obtain the maximum stresses and deflections.
Reactor internals components are within acceptable stress and deflection limits for the postulated pipe rupture combined with the safe shutdown eanhquake condition.
3.9.2.5.3 Control Rod Insertion Dudng full power plant operation, rod cluster control assemblies and the corresponding drive rod assemblies are held at a fully withdrawn position by their respective control rod drive mechanisms. During certain accident conditions, such as small break loss of coolant accident or a safe shutdown canhquake condition or both, control assemblies are assumed to drop to their fully inserted position. The guide tubes are evaluated to demonstrate the function of the control rods for a break size of 144 inches and smaller.
No credit for the function of the control rods is assumed for large breaks in the safety analyses outlined in Chapter 15. However, for break sizes consistent with use of the leak-before-break criteria, the design of the guide tubes permits control rod insertion at each control rod position.
3.9.2.6 Correlation of Reactor Internals Vibration Tests with the Analytical Results The results of dynamic analysis of reactor internals have been compared to the results of preoperational testing in reference plants. This comparison verifies that the analytical model used provides appropriate results.
~
of
\
The damping for fuel assemblies of 20% for the evaluation of the response during a safe b't.6 &
shutdown earthquake is for the fundamental mode. This damping value is supponed by test results presented in WCAP-8236 (Reference 21).
The preoperational vibration test program for the reactor vessel internals of the AP600 conducted on the first plant, conforms to the intent of the guidelines in Regulatory Guide 1.20 for a comprehensive vibration assessment program. This program includes a correlation of the analysis and test results. This comparison provides additional verification for the analytical model.
Revision: 11 ossarrvino3oon aii 020797 Draft,1997 3.9-44 3 Westinghouse
, . ' mw i
- 3. Design of Structures, Compone:ts, Equipment, and Systems "
Use of baseplates with concrete expansion anchors is minimized in the AP600. Concrete expansion anchors may be used for pipe cuppons. For these pipe support baseplate designs, the baseplate flexibility requirements of IE Bulletin 79-02, Revision 2, dated November 8, g
1979 are met by accountingfor the baseplate flexibility in the calculation of anchor bolt 3 , 9, y, g Supplemental reqmrements for fastening anhits to concrete are outlined in) subsection 3.8.4.5.1.
4 Friction forces induced by the pipe on the suppon must be considered in the analysis of .
sliding type supports, such as guides or box supports, when the resultant unrestrained '
thermal motion is greater than 1/16 inch. The friction force is equal to the coefficient of friction times the pipe load, and acts in the direction of pipe movement. A coefficient of I friction of 0.35 for steel-on-steel sliding surfaces shall be used. If a self-lubricated bearing '
plate is used, a 0.15 coefficient of friction shall be used. The pipe load from which the !
friction force is developed includes only deadweight and thermal loads. The friction force l can not be greater than the product of the pipe movement and the stiffness of the pipe support in the direction of movement.
Small gaps are provided for frame type suppons built around the pipe. These gaps allow for radial thermal expansion of the pipe as well as allowing for pipe rotation. The minimum gap (total of opposing sides) between the pipe and the suppon is equal to the diametral expansion of the pipe due to temperature and pressure. The maximum gap is equal to the diametral expansion of the pipe due to temperature and pressure plus 1/8 inch.
For standard component pipe suppons, the manufacturer's functional limitations for ,
example, travel limits aad sway angles, should be followed. This criterion is applicable to 1 limit stops, snubbers, rods, hangers and sway struts. Snubber settings should be chosen such that pipe movement occurs over the mid range of the snubber travel. Some margin should be provided between the expected pipe movement and the maximum or minimum snubber-stroke to accommodate construction tolerance.
3.9.3.4.1 ASME Code Class 1 Component Supports The load combinations and allowable stresses for ASME Code Class I component suppons are given in Tables 3.9-8 and 3.9-9.
3.9.3.4.1.1 Class 1 Component Supports Models and Methods The static and dynamic structural analyses employ the matrix method and normal mode theory for the solution of lumped-parameter, multimass structural models. The equipment support structure models are dual-purpose, since they represent quantitatively the clastic restraints that the supports impose upon the component, and represent the individual support member stresses due to the forces imposed upon the supports by the component.
A description of the suppons for the reactor pressure vessel, steam generator, and l pressurizer is found in subsection 5.4.10. The supports are modeled using elements such j as beams, plates, and springs where applicable.
l Resision: 11 Y W85thlgt10USB 3.9-67 Draft,1997
- 3. Design of Struct:res, Compone:ts, Eq:lpment, cnd Systems He[instmment columns housing the in-core detector provide a protective path for the det& tors during instauation, reactor operation, and removal at refueling outages.
The guide tube assemblies sheath and guide the control rod drive shafts and control rods. ;
The guide tubes are fastened to the upper suppon and are restrained by pins in the upper l core plate for proper orientation and suppon. i The upper core support assembly is positioned in its proper orientation, with respect to the {
lower core support assembly, by flat-sided pins in the core barrel flange. Four equally i spaced flat-sided pins are located at an elevation in the core barrel where the upper core I plate is positioned. Four mating sets of insens are located in the upper core plate at the same positions. As the upper support assembly is lowered into the lower support assembly, the insens engage the flat-sided pins in the axial direction. Lateral displacement of the plate and of the upper support assembly is restricted by this design.
Fuel assembly locating pins protrude from the bottom of the upper core plate and engage the fuel assemblies as the upper assembly is lowered into place. This system of locating pins and guidance arrangement provides proper alignment of the lower core support assembly, the upper core suppon assembly, the fuel assemblies, and control rods.
The upper and lower core suppon assemblies are preloaded by a large circumferential spring, which rests between the upper barrel flange and the upper core support assembly.
This spring is compressed by installation of the reactor vessel head.
Vertical loads from weight, eanhquake acceleration, hydraulic loads, and fuel assembly preload are transmitted through the upper core plate via the suppon columns, to the upper support, and then into the reactor vessel head. Transverse loads from coolant cross-flow, eanhquake acceleration, and possible vibrations are distributed by the support columns to the upper support and upper core plate. The upper support plate is particularly stiff to minimize deflection.
3.9.5.1.3 Radial Reflector The radial reflector is between the lower core barrel and core, surrounding the core and forming the core cavity. The reflector is manufactured of solid rings of stainless steel with holes bored venically for water cooling. The stainless steel reflects fast neutrons back to the core regions. This results in lower neutron loss from the core and decreased fluence on the reactor pressure vessel. Each reflector ring is sized in height so that adjoining sections meet at a fuel grid elevation.
3.9.5.1.4 Reactor Internals Interface Arrangement Figure 3.9-8 shows the arrangement of reactor internals components shown in Figures 3.9-5 and 3.9-6 and their relative position in the reactor vessel. As shown in the figure, the lower reactor intemal (Figure 3.9-5) rests on the vessel ledge. The upper core suppon structure (Figure 3.9-6) also rests at the same location on the top of a large compression spring (hold Revision: 11 ow=vinom sii.oarm Draft,1997 3.9-84 T W85thigt100S6
iii-iii~
- 3. Design of Structures, Components, Eq:1pment, and Systems 3.9.6 Inservice Testing of Pumps and Valves Inservice testing of ASME Code,Section III, Class 1, 2, and 3 pumps and valves is performed in accordance with Section XI of the ASME Code and applicable addenda, as required by 10 CFR 50.55a(f), except where specific relief has been granted by the NRC in accordance with 10 CFR 50.55a(f). The Code includes requirements for leak tests and functional tests for active components.
The requirements for system pressure tests are defined in the ASME Code,Section XI, IWA-5000. These tests verify the pressure boundary integrity and are part of the inservice inspection program, not part of the inservice test program.
Testing requirements for components constructed to the ASME Code are in several parts of the ASME OM Code (Reference 2). The ASME OM Code used to develop the inservice ,
testing plan for the AP600 Design Certification is the 1990 Edition. The edition and l addenda to be used for the inservice testing program are administratively controlled by the Combined License applicant. l
%AI The specifi ASME Code requirements for functional testing of pumps are found in the 2/o.2t7 ASM '
M Code, Subsection ITSB. Thegefic ASME Code requirements for I
functional testing of valves are found in the ASMELANM QM Code, Subsection ISTC. The functional tests are required for pumps and valves that have an active safety-related function.
The AP600 inservice test plan does not include testing of pumps and valves in nonsafety-related systems unless they perform safety-related missions, such as containment isolation.
This is based on the AP600 implementation of the regulatory treatment of nonsafety-related systems (RTNSS) process (WCAP-13856, Reference 15). Fluid systems with RTNSS important missions are shown to be available by operation of the system.
The AP600 inservice t st plan includes periodic systems level tests and inspections that j demonstrate the capability of safety-related features to perform their safety-related functions such as passing flow or transferring heat. The test and inspection frequency is once every 10 years. Staggering of the tests of redundant components is not required. These tests may be performed in conjunction with inservice tests conducted to exercise check valves or to >
perform power-operated valve operability tests. Alternate means of performing these tests j and inspections that provide equivalent demonstration may be developed by the Combined License applicant in the inservice test program. Table 3.9-17 identifies the system inservice te ts.
A preservice test program, which identifies the required functional testing, is to be submitted to the NRC by the Combined License applicant prior to performing the tests and following the start of construction. The inservice test program, which identifies requirements for functional testing, is to be submitted to the NRC prior to the anticipated date of commercial operation by Combined License applicant. Table 3.9-16 identifies the components subject v to the preservice and the inservice test program. This table also identines the method, l extent, and frequency of preservice and inservice testing.
Revision: 11 !
[ W85tlngh00S8 3.9-89 Draft,1997 i
. wunsma
- 3. Desigm of Struct:res, Components, Equipme;t, and Systems 3.9.6.1 Inservice Testing of Pumps Safety-related pumps are subject to operational readiness testing. The only safety-related mission performed by an AP600 pump is the coast down of the reactor coolant pumps. As a result, the AP600 inservice test plan does not include any pumps.
The AP600 inservice test plan does not include testing of pumps in nonsafety-related systems unless they perform safety-related missions. Systems containing pumps with RTNSS important missions have the capability during operation to measure the flow rate, the pump head, and pump vibration to confirm availability of the pumps. These measurements may be made with temporary instruments or test devices. The AP600 l inservice test plan does not include testing of nonsafety-related pumps (including RTNSS l important pumps) because they do not perform safety-related missions.
3.9.6.2 Inservice Testing of Valves Safety-related valves are subject to operational readiness testing. Inservice testing of valves assesses operational readiness including actuating and position indicating systems. The valves that are subject to inservice testing include those valves that perform a specific function in shutting down the reactor to a safe shutdown condition, in maintaining a safe shutdown condition, or in mitigating the consequences of an accident. The AP600 safe shutdown condition includes conditions other than the cold shutdown mode. Safe shutdown conditions are discussed in subsection 7.4.1. In addition, pressure relief devices used for protecting systems or portions of systems that perform a function in shutting down the reactor to a safe shutdown condition, in maintaining a safe shutdown condition, or in mitigating the consequences of an accident, are subject to inservice testing.
bc AP600 in;e:vice :=: pha dcc nc: incbde :=:ing cf vd/= in ncn=fety =k:e g cy; enu un!=
i Q:rurnen::, er::=: : device p d =fe'" -b0 inservice test plan does not mciua hey pc:fer:r m...m nonsafety-related valves (including RTNSS important valves) because they do not perform safety-related missions. Valves tha ' dent {i sed as having RTNSS important missio~
have provisions to allow testing gad-but are notjincluded in the inservice test plankliiless,
@incervice testingjs_ identified _as_part of the regnhtn oversight required for RTNSS.
@his testing may use temporary instmments or test devices.
The valve test program is controlled administratively by the Combined License holder and is based on the plan outlined in this su 'on. Valves (including relief valves) subject to
[ inservice testing in accordance it Sectica X! he ASME Code are indicated in Table 3.9-16. This table includes the type of testi]ng to be performed a which the testing should be performed. The test program conforms to the requirements of ASME OM, Subsection ISTC, to the extent practical. The guidance in NRC Generic Letters, AEOD reports, and industry and utility guidelines (including NRC Generic Letter 89-04) is also considered in developing the test program. Inservice testing incorporates the use of nonintrusive techniques to periodically assess degradation and performance of selected valves.
Revision: 11 ownmo9.asi o20797 Draft,1997 3.9-90 W W85tiflgh0ljS8
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- 3. Desig; of Struct res, Components, Equipme t, and Systems '
n-3.9.8 Combined License Information i
3.9.8.1 Reactor Internals Vibration Response l
Information including predicted vibration response and allowable response will be provided prior to the preoperational vibration testing of the first AP600 consistent with the guidance of Regulatory Guide 1.20.
3.9.8.2 Design Specifications and Reports Combined License applicants referencing the AP600 design will have available for NRC audit the design specifications and design reports prepared for ASME Section III components.
3.9.8.3 Snubber Operability Testing Combined License applicants referencing the AP600 design will develop a program to verify operability of essential snubbers as outline in subsection 3.9.3.4.3.
3.9.8.4 Nonintrusive V.,lre Inservice Testing Combined License applicants referencing the AP600 design will develop an inservice test program in conformance with the valve inservice test requirements outlined in subsection 3.9.6 and Table 3.9-16. This nrogram will include provisions for nonintrusi - #U c k valve testing mejhedfand the program for valve disassembly and inspection outlined i in subsection 3.9.6.2.3. _
l 3.9.8.5 Feedwater Line Thermal Monitoring A monitoring program will be implemented by the Combined License holder at the first AP600 to record temperature distributions and thermal displacements of the feedwater line piping as outlined in subsection 3.9.3.1.2.
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l Revision: 11 osummo309n aii.02:297 l Draft,1997 3.9-102 T Westirighouse
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- 3. DesE of I'tructures, Components, Eq:1pment, r.nd Systems
- 14. NRC BULLETIN NO. 88-11: Pressurizer Surge Line Thermal Stratification, December 20,1988.
- 15. "AP600 Implementation of the Regulatory Treatment of Nonsafety-Related Systems P ocess," WCAP-13856 September 1993.
- 16. NRC IE Bulletin 79-13. " Cracking in Feedwater System Piping," June 25,1979 and Revisions 1 and 2, dated August 30,1979 and November 16,1979.
- 17. " Investigation of Feedwater Line Cracking in Pressurized Water Reactor Plants,"
(Proprietary) WCAP-9693 June 1980,
- 18. "AP600 Reactor Intemals Flow-Induced Vibration Assessment Program,"
WCAP-14761, March,1996.
- 19. " Functional Capability Criteria for Essential Mark II Piping," General Electric Company, NEDO-21985,78NED174, E. C. Rodabaugh, September,1978.
- 20. " Functional Capability of ASME Class 2/3 Stainless Steel Bends and Elbows," ASME 83- PVP-66, T. H. Liu, E. R. Johnson, K. C Chang.
er
- 21. " Safety' Analysis of the 17.x 17 Fuel Assembly for Combined Seismic and Loss-of 3,9, P/
Coolant Accident," WCAP-8236 (Proprietary), WCAP-8238 (non-proprietary) i Revision: 11 os. nim 3mn.nii42i397 Draft,1997 3.9-104 [ W85tiligh00S8
- 10. Steam and Pow:r Cony:rsion System 1
1 The portion of the feedwater system to be constructed in accordance with ASME Code,Section III, Class 2 requirements is provided with access to welds and removable insulation for inservice inspection, in accordance with ASME Code,Section XI. The portion of the feedwater system to be constructed in accordance with ASME Code, 1
Section III, Class 3 requirements is also designed and configured to accommodate inservice inspection in accordance with ASME Code,Section XI. (
The condensate and feedwater system classification is described in Section 3.2. The control functions and power supplies are described in Chapters 7 and 8.
For a main feedwater or main steam line break (MSLB) inside the containment, the condensate and feedwater system is designed to limit high energy fluid to the broken loop. High energy line break for piping not qualified for leak before break (LBB) criteria is discussed in subsection 3.6.3.
Double valve main feedwater isolation is provided via the main feedwater control valve (MFCV) and main feedwater isolation valve (MFIV). Valves fail closed on loss of actuating fluid. Both valves are designed to close automatically on main feedwater isolation signals, an appropriate engineered safety features (ESP) isolation signal, within the time established within the Technical Specification, Section 16.1.
y; A ch=k valve, "hich =:: cn mvmm pamure differai;!, ;; prcvided in Se m 7se.2rg feedver.::: !!nc :c =ch :;;== g= m:c be:vec= ic ME" =d $c containm=:
peneter.:i= Se ch=k valve r; daig =d :c ii:;;=d $c force; en==::=d "h=
elesm; af:= ; main feed .;;m 1: : mp:um. 3 =!va pmv=: b!cvedev: from marc
$= =c :::=m g= m:cr during feedline br=k sti!: $c appropria:: ESF :ig=! i;
- =
- =::d :c isc! :: =ing Sc MHV =d MFCV D=ing n==:1 = upm::caditic=, Sc f=c:icn cf $cm ch=k =!va i: :c pav=: mva:c Scv from i ;:== g= mp ,
"hmeva $c feedz := :y:::m i: nc: in opem:i= -
=
The MFCVs provide backup isolation to their respective containment isolation valves in order to terminate feedwater flow. The MFCVs are located in the auxiliary building in piping designed to ASME Code,Section III, Class 3 seismic Category I requirements.
These valves are components of the steam generator system (SGS). i
=
For a steam generator tube rupture event, positive and redundant isolation is provided for j the main feedwater system (MFIV and MFCV) with ESF isolation signals generated by <
the protection and safety monitoring system.
10.4.7.1.2 Power Generation Design Basis l
l The condensate and feedwater system provides a continuous feedwater supply to the two l
steam generators at the required pressures and temperatures for steady-state and )
anticipated transient conditions. I nwnuupoumun.2nm Revision: 11 W Westirighouse 10.4-21 Draft,1997
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- 10. Steam and Pow:r Conversion Syst m l
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Plant operation is possible at 100-percent power with one condensate pump out of i service, and at 70-percent power with one booster / main feedwater pump assembly out-of service.
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Plant operation is possible at greater than 70-percent power with one feedwater heater string out of service. l
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The feedwater and condensate pumps and pump control system are designed so that loss of one booster / main feedwater pump assembly or one condensate pump does not result in trip of the turbine-generator or reactor.
The pumps and other system components are designed so that the condensate, feedwater i booster and feedwater pumps are protected from running with very low net positive I suction heads without tripping on short transient low levels in a hotwell or deaerator tank.
The condenser hotwell is designed to store, at the normal operating water level, an amount of condensate equivalent to at least three minutes of full-load condensate system operating flow.
The system is able to accommodate ten-percent step or five-percent per minute ramp load changes without significant deviation from the programmed water levels in the steam generators or major effect on the feedwater system. ;
The system has the capability of accommodating the necessary changes in feedwater flow to the steam generators with the steam pressure increase resulting from a 100-percent ;
load rejection. l The booster / main fecdwater pumps are tripped simultaneously with the feedwater
)
isolation signal to close the main feedwater isolation valves. In addition, the same '
isolation signal closes the isolation valve in the cross connect line between the main feedwater pump discharge header and the startup feedwater pump discharge header. M/
va.2ze A check valve, which acts on reverse pres differential, is' provided in the' main .
feedwater line to each steam generator between the. MFIV aiid the containment I penetration. The checkLyalve is designed to withstand the forces encountered when closing after a main feedwater line rupture. The valves perform.no safety-related i function but will serve to prevent blowdown from morc than one steam generator during
{
feedline break while the appropriate' engineered safety features signal is generated to isolate using the MFIV and MFCV. : During normal or upset conditions, the function of these check valves is to prevent reverse flow from the steam generators whenever the l I feedwater system is not in operation. ,
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l Revision: 11 twmnwum4.,n.2nm Draft,1997 10.4-22 3 Westingh0USB l
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'W SRP Chapter 3 - DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS m.nr-inn mr P?!'
-w gp Criteria Referenced AP600 APcoo Section Criteria Position Comments / Summary of Exceptions B.I.c.(1)(a) Acceptable B.I.c.(1)(b) Acceptable B.I.c.(1)(c) Acceptable B.I.c.(2)(a) Acceptable B.I.c.(2)(b) ASME Code, NC-3653 Acceptable B.I.c.(3) Exception 'Ihe locations of postulated breaks in non-ASME pipe are based on ANSUANS-58.2-1988," Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture" October,1988. Stress value is calculated using equations in Section 104.8 of ANSUASME B31.1, Power Piping Code considering normal and upset plant conditions and compared to 0.8 (X+Y) where X and Y are the allowable stress values for Equations 12 and 13 of B31.1. OBE is not used to determine pipe break locations in the AP600. oy B.I.c.(4) -- c :
-'"h ;!::hhfc= 5.;;iis ac:;pi cd, di xpueens a a .w ec de;:gned for-postelated. 3
- 2-
--! ::d : d h:gh 2ncas uc.; :x.nmm. -
i B.I.c.(5) Acceptable B.I.d. Acceptable B.I.e.(1) ASME Code NB-3653 Acceptable D" m:ui35.-3. rt::b-080696 - '3 3-19 i
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SRP Chapter 3 - DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS r ti-Criteria Referenced AP600 AP6O3 Section Criteria Position Comments / Summary of Exceptions
- 2. ASME XI Exception He IWV Section of the ASME Section XI has been replaced by ANSI /ASME-OM Section IWV Part 10. He AP600 valve test program will meet the requirements of OM-10 and incorporate appropriate requirements from NRC Generic Letter 89-10. Generic Letter 89-04 will also be reviewed for applicable guidance. See SSAR Table 3.9-16 for a description of AP600 Inservice Test Requirements.
'3 .10 - I SRP 5 3.10 - Seismic and Dynamic Qualifiention of Mechanical and Electrical Equipment (Rev. 2,7/81)
SR? 3.so Ai- m itap:M5pc15 KtM. 4uroL I ' *'*> R G v. s u %,a
- 1. IEEE 344-1975 Exception ne AP600 references qualification standards IEEE 323-1974 and IEEE 344-1987 As N4 R.G.1.100, R 8 " 3 "'
65tidin IEEID44~1987, saTety relatedEquipment may be qualifieirbWd on new testing '
rv un m .-
' and/or ansivsk or based on properly documented past test and experience data '*-
-GDC+2,& ,
(Section v.0 ofIEEE 344-1987). He concept of using properly documented experience
/ data is cost effective as evidenced by its proposed use in the resolution of the A-46
- ~ ~ ~
-- ' problem (NUREG-1030). The choice ofqualification method is based upon many factors including practicality, complexity of the equipment, economics, and availability of !
' gre den AW 4ve o E / . /oe, f'EV/5p- 2. , hersb previous qualificStion and experience data. If experience data is used, the COL applicant
[ will identify the specific equipment and include details of the methodology and the Iva g M88 Aus Pn ais v5E or /#Et 399- corresponding experience data for each piece of equipment.
!?87 Structural integrity and pressure retaining capability will be demonstrated by analysis using appropriate design codes such as the codes issued by the American Institute of Steel Construction (AISC) and American Society of Mechanical Engineers Boiler and Pressure Vessel Code (Section III).
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_ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . __ _ . - _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _________._____.___m. _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _
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- 3. Design of Strudores, Components, Equipment, and Systems f l
)
. Table 3.9-16 (She VALVE INSERVICE Valve Tag Safety-Related Nunnber _ Description (D Valve Type Missions Safel CVS-LV100 Makeup Line enneminment Isolation Relief Check Maintain Oose Active )
Transfer Oose rans=ir=u Transbr Open Safety Q CVS-LV136A Deminerahzed Water System Isolation Remote Maintain Oose Actived Transfer Oose Remote (
l CVS-LV136B Deminernhzed Water System Isolation Remote Maintain Oose Active-(
Transbr Oose Remote (
l DWS-LV244 Deminerahzed Water Supply Ca=*=i-aant Isolation - Outside Manual Maintain Oose Containn Safety R DWS-LV245 Demmernhzed Water Supply Caa*=iamant Isolation - Inside Check Maintain Oose Cantamn Safety R PPS-LV050 Phe Water Conrainment Supply Isolation Manual Maintain Oose Contaun Safetyq PPS-LV052 Fire Water Containment Supply Isolation -Inside Check Mamenin Oose Cnarama Safety R PCS-LV001A PCCWST Isolation Remote Maintain ^ pen Activey Transfer open Remote; PCS-LV001B PCCWSTIsolation U Remote Maminin Open Activeg Transfer Open Remote PCS-LV002A PCCWST Series Isolation Remote Mmmenin Open Active Transfer Open Remote ;
PCS-LV002B PCCWST Series Isolation Remote Maintain Open Active Transfer Open Remote ;
PCS-LV005 PCCWST Supply to Fire Protection Service Isolation Manual Maintain Oose Active Transfer Oose PCS-LV014A Post-72 Hour Water Source Isolation u Transfer Open Active Check PCS-LV023 PCS Recirculation Retum Isolation Manual Maintain Oose Active Transfer Oose PSS-LV008 canrainmaat Air Sample Caa*=3amaat Isolation IRC Remote Mamtam Close Active 4 Transfer Oose Contaisu Safety 2 Remote l l
t .
W Westingh0use N 4
1 L of 16)
LEQUIREMENTS ASME IST Functions (2) Category Inservice Testing Type and Frequency IST Notes AC Containment Isolation Leak Test /2 Years 23,27 s Isolation Check Exercise / Refueling Shutdown miled B Remote Position Indication, Exercise 4 Years APERTURE sition Exercise &ll Stroke /Quartedy CARD nled B Remote Position Indication, Exercise /2 Years Also Avanchie ott sition Exercise Full Stroke / Quarterly Apertura Crrd W Isolation A Containment Isolation Leak Test (See Notes) 27 Leakage a Isolation AC Containment Isolation Leak Test (See Notes) 27 Irakage u Isolation A Containment Isolation Leak Test (See Notes) 27 Leakage a Isolation AC Containment Isolation Leak Test (See Notes) 27 Ieakage Wied B Remote Position Indication, Exercise /2 Years sition Exercise Full Stroke /Quartedy Wied B Remote Position Indication, Exercise /2 Years sition Exercise Full Stroke / Quarterly B Remote Position Indication, Exercise /2 Years sition Exercise Full Stroke /Quartedy B Remote Position Indication, Exercise /2 Years sition Exercise Full Stroke / Quarterly B Exercise Full Stroke / Quarterly 13 B Wene Full Stro edy Refuel 6.g Sht: : 21
[ Check Exercise / Refueling' B Exercise Ril Stroke / Quarterly 13 tailed A Remote Position Indication, Exercise /2 Years 27 x Isolation Containment Isolation leak Test (See Notes)
Ieakage Exercise Full Stroke /Quartedy sition Revision: 11 9702260304 -Q "'*""'
3.9-135
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- 3. Design of Structures, Components, Equipment, cad Systems' !
I e Ethie 3.9-16 (Shei 1
l VALVE INSERVICE TEk Valve Tag Safety Related )
Number Description (I) Valve Type Missions Safe i PXS-LV016B Core Makeup Tank B Discharge Owck Ched Maimmin Open Active j j Transfer Open Remote ;
l Transfer Oose i PXS-LV017A Case Makeup Tank A Discharge Check Check Mainemin Open Active - )
Transfer Open Remoto Transfer Oose ;
l PXS-LV017B Core Makeup Tank B Discharge Check Check Mainemin Open Active Transfer Open Remote (
Transfer Oose l PXS-LV022A Accumulator A Pressure Relief +
Relief Mainemin Oose Active
. Transfer Open j i
Transfer Oose PXS-LV022B Accumulator B Pressure Relief _ Relief Mainemin Oose Active Trausfer Open Transfer Oose PXS-LV027A Accumulator A Discharge Isolation Remote Maintain Open Remote )
PXS-PL-V027B Accumulator B Discharge Isolation Remote Mannenin Open Remote ]
PXS-LV028A Accumulator A Discharge Check Check Maintain Oose Active Transfer Open RCS Pre Remote 1 PXS-LV028B Accumulator B Discharpe Check ' Queck Maintain Oose Amive Transfer Open RCS Pre Remote )
PXS-LV029A Aminanlator A Discharge Check , Check Maintain Oose Active Transfer Open RCS Pse i Remote )
PXS-LV029B Accumulator B Discharge Check Check Maintain Oose Active Transfer Open RCS Pse Remote 1 PXS-LV042 Nitrogen Supply Ona*=ia-at Isolation ORC Remote Maintain Oose Active-u Transfer Oose Contain Safety &
Remote 1 PXS-LV043 Nitrogen Supply Containment Isolation IRC Check Maintain Oose Active Transfer Oose Cnasaia Safety S Remote 1 W Westingh0use
5 of 16)
REQUIREMENTS ASME IST FunctionsO) Category Inservice Testing Type and Frequency IST Notes BC Remote Position Indication, Exercise /2 Years 10 sition Check ExerciseSefueling Shutdown ANST"ao '"
BC APERTURE Remote Position Indication, Exercise /2 Years sition Check Exercise / Refueling Shutdown 10
'g CARD DC Also Avntichte on Remote Position Indication, Exercise /2 Years 10 Aporture Card sition Cbeck Exercise / Refueling Shutdown BC Qass 2S Relief Valve Tests /10 Yeanhd 20 Fin 4:-Years)
BC Qass 2S Relief Valve Tests /10 Years c -
and 20%irf(Jear)s sition B Remote Position Indication, Exercise /2 Years I
sition B Remote Position Indication, Exercise /2 Years BC Remote Position Indication, Exercisch Years 9 tre Boundary Ceck Exercise / Refueling Shutdown pition BC Remote Position Indication, Exercisen Years 9 tre Boundary Qeck Exercise / Refueling Shutdown kition BC Remote Position Indication, ExerciseE Years 9 tre Boundary Check Exercise / Refueling Shutdown
- ition BC Remote Position Indication, Exercise /2 Yerts 9 tre Boundary Qcc
- Exercise / Refueling Shutdown ition kiled A Remote Position Indication, Exercise 4 Years 27 k Isolation Containment Isolation leak Test (See Notes) 1.catage Exercise Full Stroke / Quarterly ition AC Remote Position Indication, ExerciseA Yean 27
, Isolation t Containment Isolation leak Test (See Notes)
'Izakage Cbeck Exercise /Quanedy ition Revision: 11 9702260804 g "" h'";
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- 3. Design of Structures, Components, Equipment, and Systems Table 3.9-16 (Shee VALVE INSERVICE TEST Valve Tag Safety-Related Number Desaiption(I) Valve Type Missions Safet RCS-PbV004A Fourth Stage Au omatic Depressurization System Squib Maintain Open Active Maintain Oose RCS Pm Transfer Open Remote I RCS-PbV004B Fowth Stage Automatic Depressurization System Squib Maintain Open Active Maintain Oose RCS Pm Transfer Open Remote I RCS-PLV004C Fourth Stage Automatic Depressurization System Squib Maintain Open Active Maintain Oose RCS Pm Transfer Open Remote 1 RCS-PLV004D Fourth Stage Automatic Depressurization System Squib Maintain Open Active Maintain dose RCS Pm Transfer Open Remote 1 RCS-PLV005A Pressurizer Safety Valve Relief Maintain Oose Active Transfer Open RCS Pm Transfer Oose Remote 1 RCS-PLV005B Pressurizer Safety Valve Relief Maintain Oose Active Transfer Open RCS Pres Transfer Oose Remote 1 RCS-PbV010A Automatic Depressurization System Discharge Header A Vacuum Relief Transfer Open Active Relief RCS-PbV010B Autornatic Depressunzation System Discharge Header B Vacuum Relief Transfer Open Active Relief RCS-PLV011A First Stage Automatic Depressunzation System Isolation Remote Maintain Open Active Maintain Oose RCS Pre Transfer Open Remote 1 RCS-PLV011B First Stage Automatic Depressunzation System Isolation Remote Maintain Open Active Maintain Oose RCS Pre-Transfer Open Remote 1 RCS-PbV012A Second Stage Automatic Depressunzation System Isolation Remote Maintain Open Active Maintain Oose RCS Pm Transfer Open Remote !
RCS-PLV012B Second Stage Automatic Depressurization System Isolation Remote Maintain Open Active Maintain Oose RCS Pm Transfer Open Remote !
RCS-PLV013A Third Stage Automatic Depressunzation System Isolation Remote Maintain Open Active Maintain dose RCS Pm Transfer Open Remote 1 W Westingh0uSe
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B of 16) ;
REQUIREMENTS ASME IST Functions (2) Category Inservice Testing Type and Frequency IST Notes l D Remote Position Indication, Altemate/2 Years ne Boundary Charge Test Fire /20% in 2 Years 5 )'
sition l D Remote Position indiention, Altemate/2 Years 5 ANSTEC tue Boundary Charge Test Fire /20% in 2 Years L'" " APERTURE l D Remote Position Indication, Altemate/2 Years 5 CARD ,
ure Boundary Charge Test Fue/20% in 2 Years Sid " Also Available 0:t Aperture Card D Remote Position Indication, Altemate/2 Years 5 ,
ure Boundary Charge Test Fire /20% in 2 Years I sition BC Remote Position Indication, Altemate/2 Years 7 !
bire Boundary Class 1 Relief Valve Tests /5 Years and 20% in 2 Years !
kition BC Remote Position Indication, Altemate/2 Years 7 ,
me Boundary Class 1 Relief Valve Tests /5 Years and 20% in 2 Years )
kition BC Class 2/3 Relief Valve Tests /10 Year (and 20%lin 4jh BC Class 2/3 Relief Valve Tests /10 Year {and.20%lin 43 Year)s i
'B Remote Position Indication, Exercise /2 Years 3 pre Boundary Exercise Full Stroke (See Notes) sition Operability Test /10 Years B Remote Position Indication, Exercise /2 Years 3 ure Boundary Exercise Full Stroke (See Notes) kition Operability Test /10 Years B Remote Position Indication, Exercise /2 Years 3 pre Boundary Exercise Full Stroke (See Notes) kition Operabdity Test /10 Years B Remote Position Indication, Exercise /2 Years 3 kne Boundary Exercise Ibil Stroke (See Notes) p
' ition Operability Test /10 Years B Remote Position Indication, Exererse/2 Years 3 ure Boundary Exercise Full Stroke (See Notes) mition Operabdity Test /10 Years Revision: 11 9702269304qrO nr=rt.1997 3.9 145
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- 3. Design of Structures, CImponents, Equipment, and Systems Table 3.9-16 (Shee VALVE INSERVICE TEST Valve Tag Safety-Related Number Description (I) Valve Type Missions Safet RCS-PbV013B ' Third Stage Automatic Depassurization System Isolation Remote Mamtam Open Active Maintain Oose RCS Pres Transfer Open Remote i RCS-PbV014A Fourth Stage Automatic Depressurization Syrtem Isolation Remote Maintain Open Remote i RCS-PLV014B Fourth Stage Automatic Depressurization System Isolation Remote Maintain Open Remote i RCS-PbV014C Fourth Stage Automatic Depressurization System Isolation Remote Maintain Open Remote F RCS-PbV014D Fourth Stage Automatic Depressurization System Isolation Remote Maintain Open Remote F RCS-PLV150A Reactor Vessel Head Vent Remote Maintain Open Activeh Maintain Oose RCS Pres Transfer Open Remote F RCS-PbV150B Reactor Vessel IIcad Vent Remote Maintain Open Active {
Maintain Oose RCS Pres Transfer Open Remote F RCS-PLVISOC Reactor Vessel Head Vent Rernote Maintain Open Active {
Maintain Oose RCS Pres Transfer Open Remote F RCS-PbVIS0D Reactor Vessel Head Vent Remote Maintain Open Maintain Oose Active RCS Pres Transfer Open Remote G RCS-K03 Safety Valve Discharge Chamber Rupture Disk Relief Transfer Open Active !
RCS-K04 Safety Valve Discharge Chamber Rupture Disk Relief Transfer Open Active ,
RNS-PLV001A RNS Hot Leg Suction Isolation - Inner Remote Mmntain Oose Active l Transfer Oose RCS Pres Remote G RNS-PLV001B RNS Hot Leg Suction Isolation - Inner Remote Maintain Oose Active Transfer Close RCS Pres Remote 8 RNS-PLV002A RNS Hot Ieg Suction and Containment Isolation - Outer Remote Maintain Oose Active Transfer Oose RCS Pres Containm Remote $
RNS-PLV002B RNS Hot 12g Suction and Containment Isolation - Outer Remote Maintain Oose Active Transfer Oose RCS Pres Contairun Remote 6 W Westinghouse l
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"l** * **'8 ASME iST ANSTEC w on e category insen4ce T.ung Type and rrequency istnou. APERTURE B Remote Position Indication, Exercise # Years e Boundary Exercise Full Stroke (See Notes) 3 h F
CARD kion '
Operability Test /10 Years kion B Remote Position Indication, Exercisen Years h Av@sb% on Aperture Octd Wion B Remote Position Indication, Exercise # Years kion B Remote Position Indication, Exercise /2 Years tion B Remote Position Indication, Exercise /2 Years g
e Boundary B Remote Position Indication, Exercise /2 Years Exercise Full Stroke / Cold Shutdown 4 ,7 7/
tion Operability Test /10 Years h
e Boundary B , Remote Position Indication, Exercise /2 Years 4 II Exercise Full Stroke / Cold Shutdown tion Operability Test /10 Years h
e Boundary B Remote Position Indication, Excretse/2 Years Exercise Full Stroke / Cold Shutdown 4 83 tion Operability Test /10 Years h
e Boundary B Remote Position Indication, Exercise /2 Years 4 81 Exercise Full Stroke / Cold Shutdown tion Operability Test /10 Years BC Inspect and Replace /5 Years BC Inspect and Replace /5 Years B Remote Position Indication, Exercise /2 Years 15 e Boundary Exercise Full Stroke / Cold Shutdown tion B Remote Position Indication, Exercise /2 Years 15 e Boundary Exercise Full Stroke / Cold Shutdown tion B Remote Position Indication, Exercise /2 Years 15,16 e Boundary Exercise Full Stroke / Cold Shutdown Isolation tion B Remote Position Indication, Exercise // Years 15,16 e Boundary Exercise Full Stroke / Cold Shutdown Isolation tion Revision: 11 or rt.1997 9 7 0 2 x o 0 3 0 4 _O, 3.9-147
- 3. Design of Structures, Components, Equipment, and Systems Table 3.9-16 (Sheet VALVE INSERVICE TEST Valve Tag Safety Related Number Description (l) Valve Type Missions Safet RNS-PbV003A RCS Pressure Boundary Valve 'Ibemial Relief Check Maintain Oose Active Transfer Open RCS Pres Transfer Oose RNS-PL-V003B RCS Pressure Boundary Valve Thermal Relief Check Maintain Oose Active Transfer Open RCS Pr:s Transfer Oose RNS-PbV011 RNS Discharge Containment Isolation Valve - ORC Remote Maintain Oose Active Transfer Oose Cnn'ainm Safety Se Remote 11 RNS-PbV013 RNS Discharge Containment Isolation -IRC Check Maintain Oose Active Transfer Open Containm Transfer Oose Safety Se RNS-PI V015A RNS Discharge RCS Pressure Boundary Check Maintain Oose Active Transfer Oose RCS Pres RNS-PIrV015B RNS Discharge RCS Pressure Boundary Check Maintain Oose Active Transfer Oose RCS Prea RNS-PI V017A RNS Discharge RCS Pressure Boundary Check Maintain Oose Active Transfer Open RCS Pres Transfer Oose j RNS-PI V017B RNS Discharge RCS Pressure Boundary Check Maintain Oose Active Transfer Open RCS Pre Transfer Oose l
RNS-PI V021 RNS Hot leg Suction Pressure Relief Relief Maintain Co;e Active j Transfer Open Cnntainm Transfer Oose Safety Se RNS-PIrV022 RNS Suction Header Containment Isolation - ORC Remote Maintain Oose Active Transfer Oose Containm Safety Remote -
RNS-PI-V023 RNS Suction from IRWST - Containment Isolation Remote Maintain Oose Active Transfer Oose Containm Safety W Remote RNS-PI-V045 RNS Pump Discharge Relief Relief Maintain Oose Active i I
Transfer Open Transfer Oose RNS-PI V046 RNS Heat Exchanger A Oiannel Head Drain Isolation Manual Maintain Open Active Transfer Open 2
i t -- t i 1
5 of 16)
REQUIREMENTS ASME IST I%nctions(2) Category Inservice Testing Type and Frequency IST Notes BC Check Exercise / Refueling Shutdown 23
- B = ndary ANSTEC i
~
BC Check Exercise / Refueling Shutdown 23 APERTURE tre Boundary h CARD i A Remote Position Indication, Exercise /2 Years 27 Also Aval!able on a Isolation Containment Isolation Ieak Test (See Notes) Aperture Card Irakage Exercise Full Stroke / Quarterly pition i
AC Containment Isolation Leak Test (See Notes) 27 pt Isolation Ched: Exercise /Quartedy Ieakage BC Check Exercise / Refueling Shutdown 24 se Boundary BC Check Exercise / Refueling Shutdown - 24 h Boundary BC Cbeck Exercise / Refueling Shutdown 24 l Boundary ge BC Check Exercise / Refueling Shutdown 24 se Boundary AC Containment Isolation Leak Test /2 Years 17,27 E Isolation Class 2S Relief Valve Tests /10 Yearsjand 20%in 4 Years Irakage A Remote Position Indication, Exercise /2 Years 27
' Isolation t Containment Isolation leak Test (See Notes)
!Ieakage Exercise Full Stroke / Quarterly btion A Remote Position Indication, Exertise/2 Years 17,27 t Isolation Containment Isolation Ieak Test (See Notes)
!Izakage Exercise Full Stroke / Quarterly ition BC Class 24 Relief Valve Tests /10 Y and 20%in1Ye" l 0175 !
B Exercise Full Stroh5tedysRd:Eg Sht:dp ?
Revision: 11 9702260304 1
or fi. 1997 i 3.9-149 l
T^
.1 Design of Structures, Cn , r ^
. Equipssent, and Systems
- 1 Table 3.9-16 (Shed i
VALVE INSERVICE TESij Valve Tag Safety Related Neanber Description (I) Valve Type Missions Safd SGS-LV255B ' Startup Feedwater Control Remote Mainamin Oose Active-to Transfer Ocee Remote (
)
VES-LV002A Pressure Regulatmg Valve A Press. Reg. Throttle Flow Active ll VES-LV002B Pressuse Regulating Valve B Press. Reg. Throttle Flow Active VES-LV005A Air Delivery Isolation Valve A Remote Maintain Open Active-tb Transfer Open j VES-LV005B Air Delivery Isolation Valve B Remote Maintain Open Active.Qd Transfer Open VES-PL-V008A Refill Check Valve A Check Maintain Oose Active Transfer Open Transfer Oose VES-LV0088 Refill Qaedt Valve B Chedc Mainamin Oose Active Transfer Open Transfer Oose VES-LV022A Pressume ReliefIsolation Valve A Remote Maintain Open Active-0s:
Transfer Open VES-LV022B Pressure ReliefIsolation Valve B Remote Mamtam Open Active-to.
Transfer Open VES-LV040A Air Tank Safety Relief Valve A Relief Maintain Oose Active Transfer Open VES-LV040B Air Tank Safety Relief Valve B Relief Maintain Oose Active Transfer Open VES-LV041A Air Tank Safety Relief Valve A Relief Maintain Oose Active Transfer Open VES-LV041B Air Tank Safety Relief Valve B Relief Maintain Oose Active Transfer Open VES-LV042 RefillIJne Vent Isolation Valve Manual Maintain Oose Active l Transfer Oose i W Westingh0use
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l i
i l4 cf 16) l l
REQUIREMENTS ASME IST Functions (2) Category Inservice Testing Type and Frequency IST Notes Pailed B Remote Position Indication. Exercise /2 Years ANSTEC kition . Exercise R11 Stroke / Quarterly Operability Test /10 Years g
B Exercise Full Stroke / Quarterly p CAFID Operability Test /10 Years B Exercise Full Stroke / Quarterly Also Available on Operability Test /10 Years Aperture Card Failed B Remote Position Indication, Exercise /2 Years Exercise Full Stroke / Quarterly Operability Test /10 Years B Remote Position Indication, Exercise /2 Years failed Exercise Full Stroke / Quarterly ;
Operability Test /10 Years BC Check Exercise / Refueling Shutdown 21 i
BC Check Exercise / Refueling Shutdown 21 l
Failed B Remote Position Indication, Exeretse/2 Years ;
Exercise Full Stroke / Quarterly Operability Test /10 Years Pailed B Remote Position Indication, Exercise /2 Years Exercise Full Stroke / Quarterly Operability Test /10 Years BC Class 2/3 Relief Valve Tests /10 Years 0%1in 4 Y BC Class 2/3 Relief Valve Tests /10 Ye d 20%;in'.4, Years BC Class 2/3 Relief Valve Tests /10 Y 0%'.in 4 BC Class 2/3 Relief Valve Tests /10 Yearsh 20Rin 4 years B Exercise Full Stroke / Quarterly I
Revision: 11 9 7 02 2 6 0 3 0 4 _o( "'"i/in
- ~
- 3. De.,lg: of Structrres, C:mponents, Equipment, and Systems
- 2. Valves listed as having an active or an active-to-failed safety-related function provid the safety-related valve transfer capabilities identified in the safety-related mission column. Valves having an active-todailed function will transfer to the position identified in the safety-related mission column on loss of motive power.
- 3. ADS stage 1/2/3 valves (RCS-V001 A/B, V002A/B, V003 A/B, V011 A/B, V012A/B, V013A/B) will be full stroke tested every 6 months on a staggered basis. His is a relief request from the ASME code. Exercise testing of these valves represents a risk of loss of reactor coolant and depressurization of the RCS if the proper test sequence is not followed. For this reason, the frequency of this valve exercise testing should be mimmized. Conversely, the PRA assumes that valve reliability for these valves is a function of test freqv.:ncy. He recommended frequency has been incorporated into the AP600 PRA. De PRA results show that the AP600 ricets its safety goals and that the results are not overly dependent on the ADS stage 1/2/3 valves.
01
- 4. %is note applies to the reactor vessel head vent solenoid valves (RCS-V150A/B/C/D). Exercise testing of these valves at &
power represents a risk of loss of cactor coohnt and depressurization of the RCS if the proper test sequence is not followed.
gpucn tesung may also result m incidves dWeloping through seat leaks Exercise testing of these valves will be performed at cola snutdown. -
- 5. Bis note applies to squib valves in the RCS and the PXS. The squib valve charge is removed and test fired outside of valve.
Squib valves are not exercised for inservice testing. Their position indication sensors will be tested by local inspection.
- 6. His note applies to the CVS reactor coolant pressure boundary isolarba valves (CVS-V001, V002, V081, V082). Closing these valves at power will result in an undesirable temperature transient on the RCS due to the interruption of purification flow. Derefore, quarterly exercise testing will not be performed. Exercise testing will be performed at cold shutdown.
- 7. His note applies to the pressurizer safety valves (RCS-V005A/B) and to the main steam safety valves (SGS-V030A/B, V031 A/B, V032A/B). Since these valves are not exercised for inservice testing, their position indication sensors are tested by local inspection without valve exercise.
- 8. %is note applies to CVS valve (CVS-V081). %e safety functions are satisfied by the check valve function of the valve.
- 9. His note applies to the PXS accumulator check valves (PXS-V028A/B, V029A/B). To exercise these valves, flow must be provided,through these valves f== We :== :=10 the RCS. These valves are not exercised during power operations 6 because the accumulators cannot provide flow to the RCS since they are at a lower pressure. In addition, providing flow t.o-the R during. power anmtinn wnnkt cause undesirable thermal transients _on the RCSJ9uring cold rEntdowns, a full flow '
stroke test is impractical because of the potential of adding sigruficant water to the rcd. and lifting the RNS relief valve.
%ere is also a risk of injecting nitrogen into'the RCS. A partial stroke test is practical during longer cold shutdowns (248 j hours in Mode 5). In this test, flow is provided from test connections, through the check valves and into the RCS. Sufficient flow in not available to provide a detectable obtuator movement. Th=0 valv= = not enemised during+old-sistdowns-Prior teperforming =ch : :est4heeeeumulater p=== m=t be-sJgnificantly =duced, centrol-pcreer m=: be restored :e 1e c!cced
{
aceumulater-meter cpem:ed vdve =d $c valve m=: be Opened. F "c-i9g ie =: 1e =c=ula:ct p=== ==: be inc==ed, the-aceumulater-water-kvel inc= red, =d de me:ct operated valve c!cced =d i:: pcvc= =meved. In addition, thie4es: in=== he-chanoe-of-signific=:!y pe:tuinting We p es =i= bve! cr lifting 4he PSS relief-valv+if4hc tes: is act condue:ed proper'y. D= to Se c=p!exity of 1 := , exercising-4W- ""-- A" ring ce!d chutdeve= b net ce=id=d ,
=ctien!. Full stroke EexercispAestmg of these valves is conducted durinWutdowns. 'eutage vehen !: in practicair l
- 10. his note applies to the PXS CMT check valves (PXS-V016A/B, V017A/B). Dese check valves are biased open valves and are fully onen during normal operation. Rese valves will be verified to be open quarterly. In order to exercise these check l valves, sipificant reverse flow must be provided from the DVI line to the CMT. These valves are not tested during power W
Westingh0use ,
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1 we.n , .m,o a
operations because the test would cause undesirable thermal transients on the portion of the line at ambient temperatures and change the CMT boron concentration These valves are not exercised during cold shutdowns because of changes that would result in the CMT boron concentration. Because this parameten is controlled by Technical Specifications, this testing is impractical. These valves are exercised during refueling when the RCS boron concentration is nearly equal to the CMT concentration and the plant is in a mode where the CMTs are not required to be available by the Technical Spedfications.
o ers
- 11. This note applies to the PXS containment rectreulation check valves (PXS-Vil9A/B). [ Squib valves b line with the checjk S*7 (valves prevent the use of IRWST water to test tbe valves. To exe ' e these check %1ves an operator must enter tne coui.Jmam, muuvs a cavgr trom the recirculatio Mjnsert test evice into the rectreulation pipe to push n the check alve..fibe test. device is made to interface with the valve without causing valve damage.;'Ibe test Cincorporates loads measuring sensors to mersure the initial opening and full open force.pbese valves are not exe power is bypassed.operations These valves are notbecause ofcoldthe exercised during need shutdown to enter operations highly for the same reasonsradioictive (Tbmuit v:!v : :u- igy g
gNQ areasbana De (Ee-wi$ i ched v !ve Muen uc2: 1e Ec= pau :cr rc:er 6:: reu!d 5: used in $e te::.pse valves are exercised g during retuelmg conditions when the recirculation lines are not requued to be available by Technical Specifications LCOs 3.5.7 and 3.5.8 and the radiation levels are reduced.
C 4
A Mlaby
- 12. ' Ibis note applies to the PXS IRWST injecdon check valves (PXS-V122A/B, V124A/B). To exercise these check valves a IIUtw eg,j test cart must be moved into containment and temporary connecuons made to these check valves. In addition, the IRWST injection line isolation valves must have pwer restored and b: closed. These valves are not exercised during power operations because closing the IRWST injection valve is not permitted by the Technical Specifications and the need to perform significant work inside containment. Testing is not performed during cold shutdown for the same reasons. These valves are exercised during refueling conditions when the IRWST injection lines are not required to be available by Technical Specifications and the radiation levels are reduced.
- 13. PCS stop check valves (PCS-V005, V023) have inservice testing on their stop function. Their check valve function does not provide active functions and is not subject to inservice testing.
- 14. Component cooling water system containment isolation motor-operated valves CCS-V200, V207, V208 and check valve CCS-V201 are not exercised during power operation. Exercising these valves would stop cooling water flow to the reactor r5 coolant pumps and letdown heat exchanger. Loss of cooling water may result in damage to equipment or reactor trip. These valves are exercised during cold shutdowns when these components do not require cooling water.
7
- 15. Normal residual heat removal system reactor coolant isolation motor-operated valves (RNS-V001A/B, V002A/D) are not exercised during power operation. These valves isolate the high pressure RCS from the low pressure RNS and passive core cooling system (PXS). Opening during normal operation may result in damage to equipment or reactor trip. "Ibese valves are exercised during cold shutdowns when the RNS is aligned to remove the core decay heat.
- 16. Normal residual heat removal system containment isolation motor-operated valves (RNS-V002A/B) are not contamment isolation leak tested. The basis for the exception is:
'Ibe valve is submerged during post-accident operations which prevents the release of the containment atmosphere radiogas or nemsol.
The RNS is a closed, seismically-designed safety class 3 system outside containment The valves are closed when the plant is in modes above hot shutdown
- 17. Normal residem! beat removal system containment penetration relief valve (RNS-V021) and containment isolation motor-operated valve (RNS-V023) are subjected to containment leak testing by pressurizing the lines in the reverse direction Revision: 11 9 7 0 2 n 0 3 0 4 -o g ~g; l
c~ -
. mn-I to the flow which accompanies a containment leak in this path.
- 18. ' Ibis note applies to the CAS instrument air containment isolation valves (CAS-V014, V015). It is not practical to exercise these valves during power operation or cold shutdowns. Exercising the valves during these conditions may result in some l air-operated valves inadvertently openmg or closing, resulting in plant or system transients. These valves are exercised during l refueling conditions when system and plant transients would not occur. l l
- 19. Primary sampling system containment isolation check valve (PSS-V024) is located inside containment and consulerable effort l
is required to install test equipment and cap the discharge line. Exercise testing is not performed during cold shutdown operations for the same reasons. 'lhese valves are exercised during refueling conditions when the radiation levels are reduced. l
- 20. This note applies to the main steam isolation valves and main feedwater isolation valves (SGS-V040A/B, V057A/B). h valves are not full stroke tested quarterly at power since full valve stroking will result in a plant transient during normal power operation. Therefore, these valves will be partially stroked on a quarterly basis arxi will be full stroke tested on a cold ;
shutdown frequency basis. The full stroke testing will be a full " slow" closure operation. The large size and fast strokmg '
nature of the valve makes it advantageous to limit the number of fast closure operations which the valve experiences. 'Ibe timed slow closure verifies the valves operabdity status and that the valve is not mechanically bound.
- 21. Post-72 hou( ==.%. valves that require temporary connections for inservice-testing are exercised every refueling outage. These valves requue transport and installation of temporary test equipment and pressure / fluid supplies. Since the valves are normally used very infrequently, constructed of stamless steel, maintained in controlled environments, and of a I simple design, there is little benefit in testing them more frequently. For enmple, valve PCS-V014A is a simple valve that is opened to provide the addition of water to the PCS post-72 hour from a temporary water supply. To exercise the valve, a temporary pump and water supply is ennnected using temporary pipe and fittings, and the flow rate is observed using a temporary flow measuring device to confirm valve operation.
- 22. Exercise testing of the auxihary spray isolation valve (CVS-V084, V085) will result in an undesirable temperature transient on the pressurizer due to the actuation of auxthary spray flow. Therefore, quarterly exercise testing will not be performed.
Exercise testing will be performed during cold shutdowns.
- 23. Thennal relief check valves in the normal residual heat removal suction line (RNS-V003A/B) and the Chemical and Volume i Control System makeup line (CVS-V100) are located inside containment. To exercise test these valves, entry to the containment is required and temporary connections made to gas supplies. Because of the radiation exposure and effort required, this test is not conducted during power operation or during cold shutdowns. Exercise testing is performed during refueling shutdowns.
- 24. Normal residual beat removal system reactor coolant isolation check valves (RNS-V015A/B, V017A/B) are not exercise tested quarterly. During normal power operation these valves isolate the high pressure RCS from the Icw pressure RNS. Opening during normal operation would require a pressure greater than the RCS normal pressure, which is not available. It would also subject the RCS connection to undesirable transients. These valves will be exercised during cold shutdowns.
- 25. This note applies to the main feedwater control valve (SGS-V250A/B). The valves are not quartedy stroke tested since full stroke testing would result in a plant transient during power operation. Normal feedwater control operation provides a partial s stroke confirmation of valve operability. The valves will be full stroke tested during cold shutdowns.
- 26. This note applies to containment compartment drain line check valves (WLS-V071 A/B/C, V072A/B/C). These check valves are located inside containment and require temporary connections for exercise testing. Because of the radiation exposure and effort required, these valves are not exercised during power operation or during cold shutdowns. The valves will be exercised W
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- 3. Design of Structures, Components, Equipment, and Systems during refuelings.
- 27. Containment isolation valves leakage test frequency will be conducted in accordance with the Pnmary Containment I.cakage j Rate Test Program" in accordance with 10 CFR 50 Appendix J. Refer to SSAR subsection 6.2.5. :
1
- 28. This note applies to the chilled water system containment isolation valves (VWS-V058,V062, V082 and V086). Closing any of these valves stops the water flow to the containment fan coolers. This water flow may be necessary to maintain the contamment air temperature within Technical Specification limits. As a result, quarterly exerose testing will be deferred when plant operating conditions and site climatic conditions would cause the containment air temperature to exceed this limit dming testing.
ANSTEC APERTURE g CARD Also Availabb on Aperture Card Revision: 11 Draft,1997 9702260304-OZ 3.9-165
i . .
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- 6. Engineered Safety Features 1
1 In the incontainment refueling water storage tank injection lines, the squib valves are in series l with normally closed check valves. In the containment recirculation lines, the squib valves are in series with normally closed check valves in two lines and with normally closed motor operated valves in the other two lines. As a result, inadvertent opening of these squib valves will not result in loss of reactor coolant or in draining of the incontainment refueling water storage tank.
The type of squib valve used in these applications provides zero leakage in both directions.
It also allows flow in both directions. A valve open position sensor is provided for these valves.
Squib valves are also used to isolate the fourth stage automatic depressurization system hnes.
These squib valves are in series with normally open motor operated gate valves. Redundant-series controllers are provided to prevent spuriously opening of these squib valves. The type ;
of squib valve used in this application provides zero leakage of reactor coolant out of the '
reactor coolant system. The reactor coolant pressure acts to open the valve. A valve open position sensor is provided for these valves.
6.3.2.3 Applicable Codes and Classifications
)
i Sections 5.2 and 3.2 list the equipment ASME Code and seismic classification for the passive core cooling system. Most of the piping and components of the passive core cooling system !
within containment are AP600 Equipment Class A, B, or C and are designed to meet seismic a r i Category I requirements. Components and piping that provides an emergency core coolirig 5 2, z 4 fursetion that is _ Equipment' Class C has augmented weld inspection requirements,1see subsection 3.2.2.5. Some system piping and components that do not perform safety-related functions are nonsafety-related.
The requirements for the control, actuation, and Class IE devices are presented in Chapters 7 and 8.
6.3.2.4 Material Specifications and Compatibility Materials used for engineered safety feature components are given in Section 6.1. Materials for passive core cooling system components are selected to the meet the applicable material requirements of the codes in Section 5.2, as well as the following additional requirements:
=
Parts of components in contact with borated water are fabricated of, or clad with, austenitic stainless steel or an equivalent corrosion-resistant material.
Intemal parts of components in contact with containment emergency sump solution during recirculation are fabricated of austenitic stainless steel or an equivalent corrosion resistant material.
=
Valve seating surfaces are hard-faced to prevent failure and to reduce wear.
Revision: 11 W
Westinghouse Draft,1997 6.3-23 ^=niIN3aJl1:1M21297
- 6. Engineered Safety FItures F
Table 6.2.3-1 i Containment Mechanical Pen.
Explanation of Heading and Acronyms for Table 6.2.31 System: Fluid system penetrating contamment Containment Penetration- These fields refer to the penetration itself Line: Fluid system line Flow: Direction of flow in or out of contamment Closed Sys IRC: Closed system inside containment as defined in SSAR Section 6.2.3.1.1 Isolation Device: These fields refer to the isolation devices for a given penetration Valve / Hatch ID: Identification number on P&ID or system figure Subsection Containing Figure: Safety analysis report containmg the system P&ID or figure Position N-S-A: Device position for N (normal operation)
S (shutdown)
A (post-accident)
Signal: Device closure signal MS: Main steamline isolation LSL: Low steamline pressure MF: Main feedwater isolation LTC: Low Teold PRHR: Passive residual heat removal actuation T: Contamment isolation S: Safety injection signal HR: High contamment radiation DAS: Diverse actuation system signal PL2: High 2 pressurizer level signal S+PLI: Safety injection signal plus high 1 pressurizer level SGL: High steam generator level Notes:
- 1. Contamment leak rate tests are designated Type A, B, or C according to 10CFR50, Appendix J.
- 2. The secondary side of the steam generator, including main steam, feedwater, startup feedwater, blowdown and containment. These systems are not part of the reactor coolant pressure boundary and do not open directly to t generators is vented to the atmosphere outside containment to ensure that full test differential pressure is applie
- 3. The central chilled water system remains water-filled and operational during the Type A test in order to mainta
- 4. The containment isolation valves for this penetration are open during the Type A test to facilitate testing. Bei
- 5. The inboard valve flange is tested in the reverse direction.
- 6. he valves are not subject to a Type CUpstream t]e side of RNS hot leg suction isolation valves is not vel post accIdem vyu. Lea.
- 7. He inboard globe valve is tested in the reverse direction. The test is conservative since the test pressure tends l
[ W85tingh0Use
n =_!!
1 l
1 lheet 4 of 4) l trations and Isolation Valves Closure Time:
Required valve closure stroke time std: Industry standard for valve type N/A: Not Applicable ANSTEC ,
Test: 'Ibese fields refer to the penetration testing requirements APERTUsiE Type: Required test type CARD .
A: Integrated Leak Rate Test -
Also Avenable on B: Local leak Rate Test -- penetration Aperture Card ,
C: Local leak Rate Test -- fluid systems ,
Note: See notes below Medium: Test fluid on valve seat Direction: Pressurization direction Forward: High pressure on contamment side ,
Reverse: High pressure on outboard side l
ampling piping from the steam generators to the contamment penetration, is considered an extension of the e containment atmosphere during post-accident conditions. During Type A tests, the secondary side of the steam l to this boundary.
n stable containment atmospheric conditions.
leak rates are measured separately.
01r5 ,
Bo G~
ted during local leak rate test LLR-T-to retain double isolation of RCS at elevated pressure. Valve is flooded during !
to unseat the valve disc, whereas containment pressure would tend to seat the disc.
.s-niroso2ams,umien Revhion: 11 Draft,1997 9702260304 -07
- 6. Engineered Sally Features I
Table 6.23-1 Containment Mechanica I
- a Containment Penetratloa l
System Une Flow Closed Sys IRC Valve / Hatch Identification SSAR Subsectioi CAS Service air in In No CAS-LV204 9.3.1 CAS-LV205 l Instronent air in In No CAS-LV014 9.3.1 CAS-EV015 CCS IRC loads in In No CCS-LV200 9.2.2 l
CCS-PLV201 1 IRC loads out Out No CCS-PLV208 9.2.2 l CCS-PLV207 CVS Spent resia flush out Out No CVS-LV041 93.6 CVS-LV040 CVS-LVO42 Iendown Out No CVS-LV047 9.3.6 CVS-LVO45 Clwging In No CVS-LV090 9.3.6 CVS-LV091 CVS-lQ100 #0&9)
H2injection to RCS 6 [ No CVS-LV092 9.3.6 CVS-LV094 DWS Demin. water supply In No DWS-LV244 9.2.4 DWS-LV245 MIS Fuel transfer N/A No MIS h b1 m ) 6.2.5 FPS Fire protection standpipe sys. In No FPS-LV050 9.5.1 H5-LV052 PSS RCS/PSX/CVS samples out Out No PSS-LV011 9.3.3 PSS.LV010A,B Cont. air samples out Out No PSS-LV046 9.3.3 PSS-LV008 RCS/ Cont, air sample return In No PSS-LV023 9.3.3 PSS-LV024 l
l l
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i tt 1 of 4) t tions end Isolation Valves Isolation Device Test .,
Pos h a Closure A N4 A SIS *"I 1 IIPe & Note Mediaan Deres,$les M Y h {Arh[
CaC N wA C., n Fo,w.,
nr RTUn5
- c " *^ h CARD OGC T std. C,5 Air Forward F
OGC None N/A d %ygffgg OOC S std. C,5 Ak Forward OOC None N/A OOC S std. C,5 Ak Forward OOC S std.
CCC None N/A C Air Forward CCC None N/A C-C-C None N/A C-O-C T std. C Air Forward CoC T sed.
C-O-C HR,PL2, std. C Air Forward CoC S+PL1. SOL C-C-C HR,PL2, std.
S+ plt. SOL None N/A CCC T std. C Air Forward C-C C None N/A C-O C None N/A C,5 Air Forward C-O-C None N/A CoC None N/A B Air Forward C-C.C None N/A C,5 Air Forwani C-C-C None N/A C-C-C T ski. C Air Forward C-C-C T std.
OCC T std. C Ak Forward 0-C-C T std.
OCC T std. C Ak Forward 0-C-C None N/A
.witwesuismim Revision: 11 Draft,1997 9702260304 -lo 6.2 185 w
- 6. Engineered Safety Features Table 6.231 Containment Mechanical Pen Cestainment Penetration Syeessa LJae Flow Closed Sys IRC Valve / Hatch Memeincation SSAR Subsectin PXS N 2to accumulasare In No PXS-LV042 63 PXS-LV043 RNS RCS to RHR pump Out No RNS-LV002NB 5.4.7 RNS-LV023 5A.7 RNS-LV022 5.4.7 RNS-LV021 5.4.7 RNS-LV061 5A.7 PXS-LV208A 63 RHR pump to RCS In No RNS-LV011 5A.7 ,
RNS-LV013 J
SFS IRWST/Ref. cav. SFP pump in No SFS-LV038 9.1.3 discharge SFS-LV037 IRWST/Ref. cav. purif. out Out No SFS-LV035 9.13 SFS-LV034 SOS Main swamline 01 Out Yes SOS LV040A 10 3 SOS-LV027A SOS-LV030A,31 A,32A SOS-LV036A SOS-LV240A Main steamline 02 Out Yes SOS.LV040B 10 3 SOS-LV027B SOS-LV0308,31B 32B SOS-LV036B SOS-LV240B Main feedwaser 01 In Yes SOS-LV057A 10 3 Main feedwater 02 In Yes SOS-LV057B 10 3 SO blowdown 01 Out Yes SOS-LV074A 10 3 SO blowdown 02 Out Yes SOS-LVD74B 10 3 c;
Startup feedwater 01 In Yes SOS-LV067A 10 3 Startup feedwater 02 In Yes SOS-LV067B 10 3 l
- . -s C __._______________ _ ______
=- u IJB heet 2 ef 4) rations and Isolation Valves am Isolation DMce yes Poskion Closure N-S.A 838 ""I h Typel & Ne Mh hw Oec T std. C Air Forward C-C4 None N/A
std.
h Air -- 8W C Reverse CoC HR sed. C4 Forward d I
C-C-C None N/A C %lO wp CoC T Reverse
[] )
sad. C C-C-C None N/A C Forward Forward
/
[
CoC HR std. C,4 Air Forward CoC h N/A C,4 4f8C Ayg7gcb; COC T C-O-C None std. C,5 Air Forward WD Ca N/A CoC T std. C,5 Air Forward CoC T std.
OCC MS 5 sec A,2 N2 Forward OGC IEL std.
C-C C None N/A OCC MS std.
('.C-C MS std.
OCC MS 5 see A,2 N2 F '**'d OOC IEL std.
CCC None N/A OSC MS std.
C-C-C MS std.
OCC MF 5 see A,2 HO2 Forward l OCC MF 5 see A,2 HO2 Forward 0-0-C PRHR std. A,2 HO Forward 2
ooc PRHR sad. A.2 HO 2 Forward j
C-O-C LTC, SOL sid. A.2 HO Forward 2
CoC LTC, SOL std. A,2 HO2 Forward
.wirao2amio-azim Revision: 11 Draft,1997 6.2-187 9702260804-1, -
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