NSD-NRC-97-5153, Forwards Formal Transmittal of Correspondence Previously Sent Informally Over Period of 970423-0512
| ML20148D566 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 05/27/1997 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD-NRC-97-5153, NUDOCS 9706020003 | |
| Download: ML20148D566 (60) | |
Text
. . _ _ . . -
O Westinghouse Energy Systems Bm 355 Pittsburgh Pennsylvania 15230 0355 Electric Corporation NSD-NRC-97-5153 DCP/NRC0888 Docket No.: SIN-52-003 May 27,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T. R. QUAY
SUBJECT:
INFORMAL CORRESPONDENCE
Dear Mr. Quay:
Please find enclosed a formal transmittal of correspondence we have previously sent to you informally.
This informal correspondence was sent over the period April 23,1997 through May 12,1997.
Attachment I provides the index of the attached material as you have requested.
Lp-ihian A. McIntyr , Manager Advanced Plant Safety and Licensing
.imi , .p Attachment O' Enclosure i cc: N. J. Liparuto, Westinghouse (w/o A michment, Enclosure) )
l M. M. Slosson, NRC (w/o Enclosure; \
l 200127
"~"
lEllllEllllllllll!Illlllllll
- rgmuuajk A
l L
a i
o 0
Attachment I to Westinghouse Letter DCP/NRC0888 DATE ADDRESSEE DESCRIPTION 5/12/97 Huffman Markups of SSAR chapter 18 that reflects NRC comments.
5/7/97 Scaletti Information related to open item 4122. Was originally submitted to NRC in our letter of 2/21/97. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed" ,
5/7/97 Scaletti information related to open item 3481. Was originally submitted to NRC in our letter of 3/5/97 and~ revision 1I of the l SSAR. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed".
5/7/97 Scaletti information related to open items 363,358 and 1142. Was l originally submitted to NRC in our letter of 4/25/97. Request NRC review material and provide definitive action or provide !
direction to change NRC status to " Action N" or " closed" I 5/6/97 Huffman Information related to RAI 440.26. l l
5/8/97 Huffman Information request for ROSA test results.
l 5/6/97 Quay Resend of 5/1 fax on status reporting. l 5/6/97 Huffman Tech specs ending in Mode 4.
5/8/97 Huffman Advance draft of revision to response to RAI 410.17.
5/8/97 Huffman OITS report for the HICB items that are not statused as resolved.
5/7/97 Huffman Advance draft of revision to response to RAI 410.17.
4/23/97 Huffman Looking for time for phone call on advanced draft of response to the two ERG /ASI letters.
4/29/97 Huffman Schedule for tech spec telephone calls on 4/30.
4/30/97 Huffman Notes from 4/30 tech spec telephone call.
5/6/97 Huffman Resend of 5/2 open item status and comments on open items 224 and 233.
4/25/97 Huffman Schedule for tech spec telephone calls on 5/8.
4/28/97 Huffman Notice of 5/6 telephone call on tech spec 5.0.
5/5/97 Huffman Expectations for 5/6 RTNSS meeting.
l 12084 W PF
- 5/1/97 Scaletti information related to open itcm 1025. Was originally submitted to NRC in revision 6 of the SSAR,3/29/97. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed".
5/5/97 Scaletti Information related to open item 562 and 4116. Was originally submitted to NRC in revision 1I of the SSAR,2/28/97 and letters of 10/14/96 and 2/21/97. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed" 5L/97 Scaletti information related to open item 1797. Was originally submitted to NRC in revision 12 of the SSAR,4/30/97.
Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or l
" closed". !
5/6/97 Scaletti information related to open item 564. Was originally submitted to NRC in revision 12 of the SSAR,4/30/97 and letter 2/19/97. Request NRC review material and provide definitive action or provide direction to change NRC status to
" Action N" or " closed".
5/6/97 Scaletti information related to open item 783. Was originally submitted to NRC in revision 11 of the SSAR,2/28/97 and
- etter 2/19/97. Request NRC review material and prerHe definitive action or provide direction to change NRC status to ,
" Action N" or " closed". I 5/7/97 Sebrosky information related to telecon on MCR habitability testing.
Request for date for call.
5/2/97 Scaletti Information related to open item 368. Was originally submitted to NRC in revision 11 of the SSAR,2/28/97.
Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or
" closed" 5/6/97 Scaletti information related to open item 563. Was originally submitted to NRC in revision 11 of the SSAR,2/28/97 and letter 10/14/96. Request NRC review material and provide definitive action or provide direction to change NRC status to
" Action N" or " closed" l
l 4/28/97 Huffman Information to support long term cooling T/H uncertainty
! telecon. Request for date for call.
5/7/97 Sebrosky Comments on 4/15/97 PRA meeting summary.
5/6/97 Huffman Draft markups for 5.2.2 and 5.3 for telecon on 5/7.
5/5/97 Huffman Roadmap and proposed SSAR revisions on grid stability incorporating NRC comments from 5/1 phone call.
I l
1228 A WPF
i
,/
7b : 6i// Nuffonn Er Jim Orn9arcq y 1 from : Ob!n bf3 hY $VC kI'fCb I
/dere Ls on advanco ceyj of tks 52hf Ch J8 / nark p ad ik fraasm&a/
/eMer. Jk wcef revisims uil/ be 3ipd A csyicd tomorrw so prabably ike sed y% ad J1['u eds. ?!ZY 1
mornwa .
I beneve % />adup CCfffC fCfNCb ffSd/Uh df C6nrt4k btAf ,p/ dX d5lC N/m 7b Chkfm bebYf r %d FM out.
Thads AC wa s/a p
mm w ume-4 1
Westinghouse Energy Systems Box 355 I
Electric Corporation Pinsbuqpi Pennsylvania 15230-0355 j
NSD NRC 97-5119 DCP/NRC0856 Docket No.: STN-52-003 s l May 9,1997 I Document Control Desk l U.S. Nuclear Regulatory Commission l Washington, DC 20555 ATTENTION: T. R. QUAY l
SUBJECT:
RESOLUTION OF EDITORIAL HFE COMMENTS RECEIVED 04/24/97
Reference:
- 1. Letter from NRC to Westinghouse (Huffman to Liparulo), " Editorial Comments on the AP600 Human Factors Engineering Documentation," dated 04/24/97.
Dear Mr. Quay:
. 1 Attached are the following documents, issued to resolve the comments received in the referenced l
l letter:
WCAP-14651, Revision 2 -
WCAP 14701, Revision i l WCAP-14401, Revision 3 1 l '
. SSAR Chapter 18 markups as summarized in Table 1 of this letter With this submittal, the Westinghouse status for DSER open item 5247 is changed to " Confirm-W" I with an action for Westinghouse to ensure the attached SSAR markup is incorporated into the AP600 SSAR Revision 13, scheduled for May 30,1997. This markup does not reflect changes being incorporated into SSAR Revision 13 to resolve NRC comments on the HFE ITAAC. Please contact i Robin K. Nydes at (412) 374-4125 if you have any questions regarding this transmittal.
Brian A. McIntyre, Manager !
Advanced Plant Safety and Licensing jml ,
l Enclosure
, cc: Jim Bongara, NRC (IL, IE) i Bill Huffman, NRC (ll, SE)
(-- John O'Hara, BNL (1L, iE) _
i Jim Higgins, BNL (IL, IE) l N. J. Liparuto, Westinghouse (w/o Attachment) l w
f0+$0 .
l
- l TABLE 1 !
SSAR PAGE REVISIONS MADE TO RESOLVE NRC COMMENTS FROM REFERENCE 1 OF 1 NSD-NRC-97-5119 l
1
? l Comment Revised SSAR Page(s) Notes i
1 18.1 2 1
, 2 18.1-4 j 3 18.2-1, 18.12-1 l l 4 18.2-12, 18.2-5 l
l 5 18.2-4, 18.2-19 i 6 18.4-2 I 7 18.8-1, 18.8 16, 18.8 20 8&9 18.8 17, 18.8-18 The change to 18.8-18 also addresses the consistency issue {
of using " remote shutdown facility" vs. " remote shutdown workstation" Consistency with SSAR Chapter 7 and the ITAACs was confirmed.
10, 11, & 12 18.11-1 1
1 12 18.11-5 (Figure 18.11-1) 13 18.1-4, 18.8-25, 18.12-8 Reference list changed for l
WCAP-14651 changes made.
1 l
l 12, 14, & 15 18.8-26, 18.11-2 Reference list changed for WCAP 14701 changes made.
15 18.8-29 16 18.d-30 17 18.1-4, 18.8-25, 18.11-2 References list changed for WCAP-14401 changes made.
Consistency with the HFE ITAAC was confirmed.
n/a 18.12-10 A typographical error was corrected.
mer
r . J 18, Hn=== Fcetors Engineering
\ '
l l The layout and environmental design of the main control and the remote shutdown rooms, and 1 I the supplementary support areas, such as the technical support center, are sites of application 1 of the traditional disciplines of human factors engineering.
I I Input from the designers is provided to the Combined License applicant that includes I decisions made in the design of the AP600 that affect those interfaces in the Combined l License applicant's scope. This includes input on the operating staff training program and on I
I the development of the plant operating procedures.
l i Because of the rapid changes that are taking place in the digital computer and graphic display 1 l technology employed in a modern human system interface, design certification of the AP600 l 1 focuses upon the process used to design and implement human system interfaces for the l l AP600, rather than on the details of the implementation. As a result, this chapter describes (
l the processes used to provide human factors engineering in the design of the AP600.
I 1 1
I This chapter describes the application of the human factors engineering disciplines to the design of the AP600. The basis for the human factors engineering program is the human
)
I factors engineering process specified in Reference 2. Figure 18.1-1 illustrates the 10 elements l l of the human factors engineering program. These elements conform to the elements specified l l in Reference 2. The organization of this chapter parallels these elements. In addition to the I i 10 elements of the program review model, this chapter includes a description of the minimurn l l inventory of controls, displays, and alarms present in the main ontrol roog\The following ,.
l l I provides an annotated outline of the chapten ---___ ! ,
I Ardaf 1. I I 18.2 Human Factors Engineering Program Management g gg!
I Section 18.2 presents the AP600 human factors engineering program plan that is I used to develop, execute, oversee, and document the human factors engineering [
I program. This program plan includes the composition of the human factors 'l l engineering design team.
I i 18.3 Operating Experience Review
-l I Section 18.3 and Reference 3 present the results of a review of applicable operating I experience. This operating experience review identifies, analyzes, and addresses i human factors engineering-related problems encountered in previous designs.
l 1 1 18.4 Functional Requirements Analysis and Allocation l
I l i Section 18.4 and Reference 4 present the results of the functional requirements I analysis and function allocation process applied to the AP600. Thg functional I requirements analysis defines the plant's safety functions, decomposes each safety I function, compares the safety functions and prneauaa with currently operating I Westinghouse pressunzed water reactors, and provides the technical basis for those I pracana< that have been modified. The function allocation documents the I
Revision: 9 August 9,1996 18.1-2 3 W8S!!ngh0058 1
t
- 18. Human Factors Engineering 18.12 k.
Inventory Section 18.12 presents the minimum inventory of 9 enhxed80ntrols, displays, and alarms The design basis and the selection criteria used to identify the minimum inventory bare presented.
18.1.1 References ffUenfIr5 Ot ima.ni Onf7tl rs oatt Mh a l~ OL reintbfe sitshk cens un rkSfs.fdor ,
1.
Reason. J. T., " Human Error," Cambridge, U.K., Cambridge University Press,1990.
I 2.
NUREG-0711. " Human Factors Engineering Program Review Model," July 1994, U.S. NRC.
- 3. WCAP-14645, " Human Factors Engineering Operating Experience Review Report for the l AP600 Nuclear Power Plant," Revision 2, December 1996.
I 4. WCAP-14644, "AP600 Functional Requirements Analysis and Function Allocation "
l Revision 0. September 1996.
- 5. WCAP-14694, " Designer's Input To Determination of the AP600 Main Control Room l Staffing 1.evel," Revision 0, July 1996.
- 6. WCAP-14651, " Integration of Human Reliability Analysis with Human Factors l Engineering Design Implementation Plan," Revisiorpf, September 1996' (
1 7.
z M6y 7 WCAP-14690, " Designer's Input To Procedure Development for the AP600," Revision 0, I June 1996.
- 8. WCAP-14655, " Designer's Input to "Ihe Training of 'Ibe Human Factors Engineering i Verifkation and Validation Personnel," Revision 1, August 1996.
- 9. WCAP-14401, " Programmatic level Description of the AP600 Human Factors l Verification and Validation Plan," Revision 2[Januar'y 1997.
N. ifs l
/
Revision: 11 February 28,1997 18.1-4 3 %biii@0088
~
- 18. Human Fcetors Ergineering Nau r Utst de(f.(ch
! 18.2 Human Factors Engineering Program Management I i l
The purpose of this section is to describe the goals of the AP600 human fagors engineering I program, the technical p;ogram to accomplish these goals, the he= frm @ ring -
I design team, and management and organizational stmeture that support the implementation i of the technical program.
I I Human factors engineering is the system angineering of human system interfaces. The l 1 program management tools and procedures that govem the design of AP600 systems apply I to the human factors engineering activity. This approach is expected to integrate the design I of human system interfaces with other plant systems.
I l 18.2.1 Human Factors Engineering hogram Goals, Scope, Assumptions, and Constraints i
i 18.2.1.1 Human Factors Engineering Program Goals I
i The goal of the human factors engineering program is to provide the users of the plant I operation and control centers effective means for acquiring and understanding plant data and I executing actions to control the plant's processes and equipment.
I I The objective is to enable personnel tasks to be accomplished within time and performance I criteria.
. I i 18.2.1.2 Assumptions and Constraints l
I There are a number of inputs to the human factors engineering design process that specify I assumptions or constraints on the human factors engineering program and the human system I interfaces design.
I I Major design inputs include regulatory guidelines, guidance from utilities and utility I representative groups, utility requirements documents, and AP600 plant systems design I specifications. The requirements resulting from these design inputs are captured in human I system interfaces specification documents and functional requirements documents.
l While assumptions and constraints specified by design inputs are provisionally treated as I design requirements, the appropriateness of these requirements is evaluated as part of the I human factors engineering design process. Results of human factors engineering activities I such as operating experience review, task analyses, rapid prototyping and concept testing, and
, I verification and validation activities are used to provide feedback on the adequacy of initial I human system interfaces design assumptions and constraints. If results of human factors I engineering analyses or evaluations indicate that initial human system interfaces design i assumptions or constraints are inadequate, then the human system inteifaces design I requirements are revised utilizing the standard AP600 design configuration change control I process.
Revision: 9 18.2 1 August 9,1996 3 Westingh0Us4
Qt f& k },as44 agW IS i tSf o'h bl ', bi C?f Qt x l, h6 E. C Fy .'ac lui,4 f / e /MA h /0d#182 "umbactorsknheebs i R!f!fM + M a6 6,w Jwe n me- n ew . -
~
l l }I )
I the implementation technology to be assumed for the human system interfaces is derived from
! I assessment of existing technology and anticipated advancements. An emphasis is placed on l l utilization of proven, reliable technology. The decision on the specific technology to be l I employed is made on a case-by-case basis after available technology altematives are evaluated.
l I 18.2.1.3 Applicable Facilities l
l Facilities included in the scope of the AP600 human factors engineering program are the main l I control room (MCR), the technical support center (TSC), the remote shutdown facility, the l l operational support center, the emergency operations facility (EOF), and local control stations.
! I _The cumiscovy vysiouuus fodOy dwisu, unjudag specification of a icerier, is o >
l l - respen:!b!!!ty of the Combined Lke~ App!km Communication wnh the emergency I operations facility is the responsibility of the Combined Licente applicant. Section 13.3 1 discusses the responsibility for emergency planning.
I i 18.2.1.4 Applicable Human System Interfaces l
l 1 The scope of the human system interfaces encompasses the instrumentation and control I systems which perform the monitoring, control, and protection functions associated with all 1 modes of plant normal operation a.s we.Il as off. normal, emergency, and accident conditions.
l I Both the physical and the cognitive caaracteristics of those humans involved in the use, I control, maintenance, test, inspection, u.d surveillance of plant systems are accommodated.
I i 18.2.1.5 Applicable Plant Personnel I
l The AP600 human factors engineering program and the design of the human system interfaces l I includes the selection, synthesis, and distributioa of process data to plant operations personnel l as well as other plant personnel. nese addit'.onal users include management, engineering, I maintenance, health physics and chemistry personnel.
I i 18.2.1.6 Technical Basis I
l The human factors engineering; program is performed in accordance with accepted industry l
I standards, guidelines, and practices. The references listed at the end of each Chapter 18 l section and within any supporting documentation and reports are used to guide the human I factors engineering program. He human factors engineering process specified in Reference 1 I is used.
I i 18.2.2 Human System Interfaces Design Team and Organintion I
- I ne human system interfaces design team is part of the AP600 systems engineering function I and has similar responsibility, authority, and accountability as the rest of the design I disciplines. Figure 18.2-1 depicts the process used by the human system interfaces design i team members. Figure 18.2-2 shows the orgamprion of the human system interfaces design I team and its relationship to the AP600 design organization.
Revision: 9 August 9,1996 18.2-4 3 7,biim$iollS8 l
1 . - - -
. 18. Humm Fcet:rs Engi eering i 18.2.2.1 Responsibility I
I The mission of the human system interfaces design team is to develop the main control room I and ancillary control facilities (such as remote shutdown workstation) that support plant I personnel in the safe operation and maintenance of the plant. He human system interfaces I design team is responsible for coordinating the human factors aspects associated with I designing the structures, systems, and components that make up the main control room and I ancillary control facilities.
I I ne human system interfaces design team is responsible for-I l . Development of human system interfaces plans and guidelines I
- Oversight and review of human system interfaces design, development, test, and I evaluation activities
! . Initiation, recommendation, and provision of solutions for problems identified in the I implementation of the human system interfaces activities I . Assurance that human system interfaces activities comply with the human system I interfaces plans and guidelines I 18.2.2.2 Organizational Placement and Authority I
l The organization of the human system interfaces design team and its relation to the AP600 l design organization is depicted in Figure 18.2-2. The structure of the organization may I change, but the functional nature of the human system interfaces design team is retained I through the change. De human system interfaces design team consists of an instmmentation I and control system manager, advisors / reviewers team, core human system interfaces design i team, and human system interfaces technical lead, ne technical disciplines described in i subsections 18.2.2.3 and 18.2.2.4 are organized by function within the core human system I interfaces design team. He core hum system interfaces design team and the I advisors / reviewers team report to the mstmmentation and control system manager. He i human system interfaces technical lead works within the human system interface design I function and reports to thegstmmentation and control system manager through the I manager of the human system interface design function. Thejilan? Instrumentation and I control system manager is responsible for the design of the AP600 instrumentation and control I systems which include the human system interfaces. Dgmstrumentation and control I system manager reports to the AP600 project manager.
I 1
1 The manager of the human system interface design function, who performs the function of l l technical project management for the human factors engineering design process, is responsible l for the overall human system interfaces design and for integration of the human system l
( l interfaces design with the ovenil plant design. De advisors / reviewers team is responsible i
I for overseeing the general progress of the human system interfaces design, providing guidance I within the core human system interfaces design team, reviewing and providing comments on i documents, specifications, and drawings pertaining to the human system interfaces design, and I providing supplemental expertise in particular areas of design. De responsibility of the core I human system interfaces design team is to produce the detailed design of the human system Revision: 9 W Westinghouse 18.2-5 August 9,1996
4
'. 18. Human Fact:rs Engineering i 18.2.3.1 General Pr,ocess and Procedures l
l The nstrumentation and control system function is responsible for development I of,the AP600 instmmentation and control (I&C), including human system interfaces, and I coordinating and integrating AP600 instmmentation and control and human system interfaces I with other AP600 plant design activities. The overall operation of the project instrumentation I and control systems function is defined. He function includes human system interfaces I design of control rooms and control boards, instrumentation and control design, and control I toom/ equipment design. The function includes definition of an engineering plan, review of I inputs, production of system documentation, verification of work, procurement and i manufactming follow-up, and acceptance testing. An iterative feature is built into the process.
1 I Documents produced as part of the instmmentation and control and human system interfaces I design process include:
l I
- Operating experience review documents 1
- Task analysis documents I
- Functional requirements documents
!
- Human system interfaces design guidelines documents l
- Design specification documents I
- Instmmentation and control architecture diagrams l
- Block diagrams l
- Room layout diagrams I e Instrumentation lists I
- System specification documents l
l De procedures goveming instrumentation and control engineering work specify methods for I verification of work. The types of verification include:
I I
- Design verification by design reviews I
- Design verification by independent review / alternative calculations l
- Design verification by testing I
I System Specification Documents l
I System specification documents identify specific system design requirements and show how I the design satisfies the requirements. They provide a vehicle for documenting the design and I they address information interfaces among the various design groups.
I I System specification documents follow established format and content requirements. The ,
I content of a system specification document includes:
l l
- Purpose of the system I
- Functional requirements and design criteria for the system 1
- System design description including system arrangement and performance parameters I
- Layout Revision: 9 August 9,1996 18.2 12 T Westinghouse
I .
luman Factors Erginaring l
iterative stages of the human factors engineering process. Potential points of iteration are indicated in Figure 18.2-3. Further details on the activities, inputs, and output documents associated with the various elements of the human factors engineering program are provided in the sections corresponding to each human factors engineering element.
l Figure 18.2-3 provides a program milestone schedule of human factors engineering tasks showing relationships between human factors engineering elements and activities, products, and reviews. Internal design reviews are performed at various points throughout the design
! process.
N
. eFi 11 t !!
18.2.6 Combined License Information
'lle Combined License applicant referencing the AP600 certified design is responsible fofthe emergency ppe ations facility)desi *In'cluding specification of the location lit aco rdt,t C f i U'O S ObOb hW,un AdDO cr kitet*1 r/R rern; l 18.2.7 References /
- 1. NUREG-0711, " Human Factors Engineering Program Review Model," U.S. NRC.
- 2. WCAP-14645, " Human Factors Engineering Operating Experience Review Report For The AP600 Nuclear Power Plant."
- 3. WCAP-14694, " Designers Input to Determiriation of the AP600 Main Control Room Staffing Level."
- 4. WCAP-14644, "AP600 Functional Requirements Analysis and Allocation."
- 5. Reason, J.T., " Human Error," Cambridge, U.K., Cambridge University Press,1990.
l 6. WCAP-14822, "AP600 Quality Assurance Procedures Supporting NRC Review of AP600 l SSAR Sections 18.2 and 18.8," Revision 0, February 1997.
f f}tk. Dm b s*tt'h hist!1 Y II M #NNCarf t't ef't 0N d'Y IS V C(I'1!
oH(nNe i \
wv)
N UW r uf tr\e hc for 4he <x ecuhCh n
hM11dd en bwh jro raigmu+e0 by ' s a h e'r, IL 2.
k d5 i
l Revision: 11 W Westinghouse 18.2 19 February 28, L997
- 18. Human Fcetors Engineering l 1 . et Mechanisms available for reconsidering, and if necessary, changing AP600 function allocations in response to operating experience, and the outcomes of ongoing analyses -
and trade studies 18.4.1 Combined License Information 4 j l
This section has no requirement for information to be provided in support of the Combined License application.
l
! 18.4.2 References
- 1. NUREG-0711, " Human Factors Engineering Program Review Model" 1994.
I 2. WCAP-14644, "AP600 Functional Requirements Analysis and Function Allocation."
l Revision 0, September 1996.
- 3. NUREG/CR 3331, "A Methodology for Allocation of Nuclear Power Plant Control Functions to Human and Automated Control," 1983.
l l
l l
I I
Revision: 11 February 28,1997 18.4-2 T Westirighouse l
l
' 18. Human Fceters Engineering 18.8 Human System Interface Design l l
This section provides an implementation plan for the design of the human system interface (HSD and information on the human factors design for the non HSI portion of the plant. He human system interface includes the design of the operation and control centers (OCS) and i each of the human system interface resources. Execution and documentation of this I implementation plan is the responsibility of the Combined License applicant.
I l
l The operation and control centers includes the main control room, the technical suppon center, the remote shutdown facility, operational support center local control stations and associated workstations for each of these centers. The AP600 hu i system interface resources include: '
. Wall panel information system
. Alarm system .
Plant information system Computerized procedure system 8 *"
9/ y"^W #M,, 4y
/
Soft controls / dedicated controls Qualified data processing system ne wall panel information station presents information about the plant for use by the operators. No control capabilities.are include 1. The wall panel information station provides dynamic display of plant parameters and alarm information so that a high level understanding of current plant status can be readily ascenained. It is located at one end of the main control area at a height such that both operators and the shift supervisor can view it while sitting at their respective workstations. This panel provides information important to maintaining the situation awareness of the crew and for supporting crew coordination. The wall panel information station provides a dynamic plant display of the plant. It also serves as the alarm system overview panel display. He display of plant disturbances (alarms) and plant process data are integrated on this wall panel information station display. The wall panel information station is a nonsafety related system. It is designed to have a high level of reliability.
The mission of the AP600 alarm system, together with the other human system interface resources, is to provide the operations and control centers operating staff with the means for acquiring and understanding the plant's behavior. He alarm system improves the performance of the operating crew members, when acting both as individuals and as a team, by improving the presentation of the plant's p'rocess alarms. The alarm system supports the control room crew members in the following steps or activities of Rasmussen's operator decision-making model (Referese 25):
. The "alen* activity, which alens the operator to off-normal conditions
- The ' observe what is abnormal" activity, which aids the user in focusing on the important issue (s)
Revision: 10 W W85tingh0058 18.8-1 Lecember 20,1996
i
- 18. Human Fcctors Engineerirg i
The technical basis for software specifications are verified with plant data (for example, heat-l up and cool down limits, steam generator setpoints and high- and low-level alarm setpoints).
The AP600 human system interface is designed so that the plant data is a separate data file l independent of the software specifications.
i l 18.8.2.6 Minimum Information ne AP600 human system interface resources used to address the Safety Parameter Display System requirements are the alarm system, plant information system, and the computerized procedure system. The AP600 human system interface displays sufficient information to determine plant safety status with respect to the Safety Parameter Display System safety functions. De safety functions and respective parameters presented in Table 2 of I Reference 32 is used as a starting point for the AP600. He human system interface design implementation plan is desenbed in subsection 18.8.1 and includes the integration of Safety Parameter Display System requirements into the human system interface. He Safety Parameter Display System design issue of " minimum information" is tracked by the human factors engineering issues tracking system.
18.8.2.7 Procedures and Training As stated in Sections 13.2 and 13.5, the development of training programs and plant procedures are the responsibility of the Combined License applicant. Reference 30 describes l how training insights are passed from the designer to the Combined License applicant.
Reference 31 provides input to the Combined License applicant for the development of plant operating procedures.
18.8.3 Operation and Control Centers NW He human system interface includes th< design of the operation and control cente (operation and control centers). The design of each of these control centers is conductej using the human system interface implementation plan presented in subsection 18.8.1. Kptission for _
each of the operation and control centers in the AP600 is provided in the nextyght subsec-tions. Coupled with each mission staternent is a brief description of the major tasks and design features that are supported by that center.
18.8.3.1 Main Control Room Mission and Major Tasks 1
The mission of the main control room is to provide a seismically qualified habitable and comfortable location for housing the resources for a limited number of humans to monitor and control the plant processes.
De major tasks performed in the main control room include monitoring, supervising, manag-ing, and controlling those aspects of the plant processes related to the thermodynamic and energy conversion processes under normal, abnormal, and emergency conditions. Operating staff can monitor, supervise, manage, and control processes that have a real-time requirement for protecting the health and safety of operating personnel. 'Ihe main control room supports I
Revision: 11 February 28,1997 18.8-16 3 Westinghouse J
l
09 3% N O ec.Ts6 Ibo Sr & Hf low 1 N #d h ' d,I f h^T
/r O 6 '
.. . e e, l fesg]seMf! ,'j'"$?'E Mt# ^y f'*1V Hu nhato
/ffl % l'&*f'*t]'h E e j
- l#ha % ah e n pft,..$ .
i Q'] "*' ' *1firl'rer /
3 Jy, o ya.m}
/
l alarms. displays, controls, and procedures. Dese resources are located in a control area outside of the main control room.
18.8.3.8 Local Control Stations Mission and Major Tasks l De mission of the local control stations is to provide areas, outside of the main control room.
l the remote shutdown room, and the radwaste control area. for operations personnel with the appropriate resources to perform monitoring and control activities. Activities that are implemented on local control stations are reviewed to verify that their removal from the main control room is consistent with the operator staffing and performance considerations. Human system interface locations are provided for single task operations such as the operation of a efitikVaIVE Ib'Ede5 G' w 'y t.*l y Qs n r,: k s Q c. ,1; 18.8.4 Human Facters Design for the i o Human System Interface Portion of the Plant 18.8.4.1 General Plant Layout and Design The AP600 design process incorporates a human engineering approach to operations and maintenance. Maintainability design guidelines and human factors and as-low-as-reasonably-achievable (ALARA) checklists are used to meet the requirements of a human engineered environment. The design objectives include reducing worker exposure and eliminating unnecessary inspection and maintenance tasks.
18.8.4.1.1 Maintainability Design features such as component selection, layout and standardization increase the probability that targeted repair times are achieved. These features coupled with a preventative maintenance program help the AP600 meet its objectives for operation and maintenance.
Design requirements from the utility industry and industry design practices establish criteria for layout, changeout, and replacement for parts and components; access for major pieces of equipment; and vehicle passage.
, Critical path outage models are prepared for the AP600. A typical refueling and maintenance outage schedule is used by design engineers. The model indicates maintenance windows for major outage events. Maintenance and testing of equipment and necessary plant operations (for example, refueling, heatup, and cooldown) are scheduled within the outage window.
18.8.4.1.2 Accessibility and Equipment Laydown Provisions AP600 maintainability design guidelines assist designers in identifying top-level layout requirements for equipment accessibility. Component engineers specify space requirements for routine maintenance, inservice inspection, testing and component replacement.
Frequency of inspection and maintenance dictates whether permanent platforms, ladders, and scaffolding are provided.
Revision: 10 December 20,1996 18.8-20 3 Westirighouse
l
- 18. Humrn Factors Engin:ering . i~
l the operator's decision making process, and promotes the interaction with other plant i personnel, while preventing distractions by non-operating personnel. The main control room provides the interfacing resources between the operation of the plant and the maintenance of l the plant. Its areas include the main control area the switching and tagging area, the shift t i
supervisor's office, the shift supervisor's clerk's office, and the operations staff's area (see Figure 1.2-8). Habitability systems are described in Sections 6.4 and 9.4.
18.8.3.2 Main Control Area Mission and Major Tasks l
l Mr i
De mission of the main control area is to provide the support facilitie necessary for the l operators to monitor and control the AP600 efficiently and reliably. Fi 6.4-1 provides a view of the main control area. The main control area includes reactor operator workstations and the elpan /"/ciaiaim..; A and BDiIe' supervisor's workstation, the associated with the wall panel informa. ion system. The layout, size and ,
ergonomics of the operator workstations and the wall panel information system depicted in l this figure does not reflect the results of the human system interface design implementation l plan desenbed in subsection 18.8.1. The actual size, shape, ergonomics and layout of the l operator workstations and the wall panel information system is an output of the i implementation plan.
l The major task of the main control area is to provide the human system interface resources
\
l that determine the plant state and implement the desired changes to the plant state during both l normal and emergency plant operations. The main control area provides alarms to alert the operator to the need for funher investigation. Plant process data displays permit the operator to observe abnormal conditions and identify the plant state. He controls enable the operator l to execute actions. The process data displays and the alarms provide feedback to enable the ,
operator to observe the effects of the control actions. 1 EmL q +4te- deactor operator y/orkstations A a.a Mtain the displays .ed controls to st maneuver the plant, and shut down the plant. Reference 44 presents input from the designer to the Combined License applicant for the determination of the staffing level of the operating ;
j crew in the main control room. Each workstation is designed to be manned by one operator. 1 There is sufficient space and operator interface devices for two operators. The physical j makeup of,"/ci=ica; A r.d " is identical. The human system interface resources available at each workstation are:
%e %cy(c h i equahltGofV M $3 1
\
Plant information system displays Control displays (soft controls)
Alarm system support displays I
=
Computerized procedure displays Screen and component selector controls _-
l
\the rewbr acch
- UN. .
The supervisor workstation is identical to "/em;5dA 65%xcept that its controls are locked out. The supervisor workstation contains both internal plant and extemal plant communications systems.
Revision: 10 l W Westfrighouse 18.8-17 December 20,1996 l
= ?
, 18. Human Fcctars Engineering l
re n % o p e k uTu % .hin Upon failure of either "M=is A e. '//cas w 0"The failed workstation is locked out, I and the supervisor workstation controls are unlocked. This modified workstation configuration maintains independent, redundant workstations.
l A dedicated safety panel is located in the main control area. He qualified data processmg system visual display units and the dedicated safety system controls are provided in this panel.
These visual display units are the only monitoring display devices in the main control room that are seismically qualified and provide the post accident monitoring capabilities in accordance with Regulatory Guide 1.97. Dedicated system-level safety system control l
switches are located on the dedicated safety panel to provide the operators with single-step l safety system actuation capabilities. A minimum inventory of these dedicated displays and controls are presented in Section 18.12.
1 There is storage space for supplies, protective clothing and some spare parts. Cabinets are !
provided for necessary documents, and a drawing laydown area is provided for the operators' l use. Restroom and kitchen facilities are provided for the main control room operations crew.
18.8.3.3 Switching and Tagging Area Mission and Major Tasks The mission of the switching and tagging area is to provide an interface between plant maintenance and plant operations personnel. Figures 1.2-8 and 6.4-1 provide the layout of the switching and tagging area. The operations ' staff monitors and approves the state of systems, major components, and equipment. The maintenance staff is informed of maintenance required by the operations staff. De means for initiating, tracking, and logging ,
maintenance work orders is provided. I The major task of the switching and tagging area is to ease the management and implementation of the switching and tagging operations. The switching and tagging area I generates notifications that equipment is not available due to testing, maintenance, or i equipment failure. These notifications alert plant operating personnel to the unavailability of l equipment. Notifications are provided to plant maintenance personnel, alerting them that ,
operating personnel are aware of the equipment status. He switching and tagging area l facilitates a systematic and organized approach to removing equipment from service as well as retuming it to service.
18.8.3.4 Remote Shutdown 6 Mission and Major Tasks d
The mission of the remote shutdown.wormanen is to provide the' resources to bring the plant to a safe shutdown condition after an evacuation of the main control room. De remote shutdown workstation resources are based on an assumed evacuation of the main control room without an opportunity to accomplish tasks involved in the shutdown except reactor trip.
Subsection 7.4.3 discunes safe shutdown using the remote shutdown workstation, including i design bases informatior>.
Revision: 10 December 20,1996 18.8-18 T Westinghouse l
- 18. Human Factors Engineering i,
- 18. Electric Power Research Institute, " Advanced Light Water Reactor Utility Requirements Document, Vol. III. ALWR Passive Plant, Chapter 10: Man-Machine Interface Systems,"
1 Revision 6, December 1993.
- 19. International Electrotechnical Commission, " Design for Control Rooms of Nuclear Power Plants," IEC Standard 964,1989.
- 20. Intemational Electrotechnical Commission, " Operating Conditions for Industrial-Process Measurement and Control Equipment," IEC Standard 654-1,1979.
- 21. Proctor, D. H. and Hughes, J. P., " Chemical Hazards of the Workplace," 1978.
- 22. 29CFR1910. " Occupational Safety and Health Standards," 1975.
- 23. WCAP-14651, " Integration of Human Reliability Analysis With Human Factors l Engineering Design Implementation Plan," Revision,K Sj.ptemTr 1996'.
z- my 7
- 24. WCAP-14401, " Programmatic Level Description of the AP600 Human Factors i Verification and Validation Plan," Revision % Januarfl997.
b 'A]&
- 25. WCAP-14695, " Description of the Westinghouse Operator Decision Making Model and l Function Based Task Analysis Methodology," Revision 0, July 1996.
l .
- 27. NUREG-0737, Supplement li " Requirements for Emergency Response Capability."
- 28. NUREG-0696, " Functional Criteria For Emergency Response Facilities."
- 29. NUREG-0711, " Human Factors Engineering Program Review Model," July 1994.
- 30. WCAP-14655, " Designer's Input for the Training of the Human Factors Engineering I Verification and Valids. tion Personnel," Revision 1. August 1996.
l 31. WCAP-14690, " Designer's Input to Procedure Development for the AP600," Revision 0, 1 June 1996.
l 32. NUREG-1342, "A Status Report Regarding Industry Implementation of Safety Parameter Display Systems."
l 33. Rasmussen, J.,1986, "Information Processing and Human-Machine Interaction, An j Approach to Cognitive Engineering," (New York, North-Holland).
i
- 34. O'Hara, J. M. and Wachtel, J.,1991, " Advanced Control Room Evaluation: General Approach and Rationale" in "Proceedmgs of the Human Factors 35th Annual Meeting,"
pp.1243-1247, (Santa Monica, CA Human Factors Society).
I l
l Revision: 11
! W Westingh00S8 18.8-25' February 28,1997
==
- - - - =- . - . - . _- - - --
- 18. Human Factors Engineering
- 35. Woods, D. D. and Roth, E. M.,1988, " Cognitive Systems Engineering," Helander, M.
(l!
(ed.), " Handbook of Human-Computer Interaction," pp.3-43, (New York, NY, Elsevier Science Publishing Co., Inc.).
- 36. Woods, D. D., Wise, J. A., and Hanes, L. F.,1982, " Evaluation of Safety Parameter 4
Display Concepts," NP-2239, (Palo Alto, CA, Electric Power Research Institute).
l
- 37. Woods, D. D. and Roth, E. M.,1986, "Ihe Role of Cognitive Modeling in Nuclear Power Plant Personnel Activities," NUREG-CR-4532, Volume 1, (Washington, D.C.,
U.S. Nuclear Regulatory Commission).
i
' 38. Woods, D. D., Roth, E. M., Stubler, W. F., and Mumaw, R. J.,1990, " Navigating Through Large Display Networks in Dynamic Coraol Applications" in " Proceedings of the Human Factors Society 34th Annual Meeting," pp. 396-399 (Santa Moaica, CA, Human Factors Society).
4
- 39. Reason, J. T.,1990, " Human Error," (Cambridge, UK, Cambridge University Press).
- 40. Stubler, W. F., Roth, E. M., and Mumaw, R. J.,1991, " Evaluation Issues for Computer-Based Control Rooms" in " Proceedings of the Human Factors Society 35th Annual Meeting," pp. 383-387, (Santa Monica, CA, Human Factors Society).
1
(
- 41. Woods, D. D.,1982, " Application of Safety. Parameter Display Evaluation Project to Design of Westinghouse Safety Pammeter Display System," Appendix E to " Emergency Response Facilities Design and V & V Process," WCAP-10170, submitted to the U.S.
Nuclear Regulatory Commission in support of their review of the Westinghouse Generic Safety Parameter Display System Non-Proprietary, (Pittsburgh, PA, Westinghouse Electric Corp.).
i
- 42. U.S. Department of Defense,1989, " Military Standard 1472D; Human Engineering Design Criteria for Military Systems, Equipment and Facilities," (Washington, D.C., U.S.
Department of Defense).
- 43. American National Standanis Institute,1988, " ANSI /HF 1001988, American National Standard for Human Factors Engineering of Visual Display Terminal Workstations,"
(Santa Monica, CA, Human Factors Society, American National Standards Institute).
- 44. WCAP-14694, " Designer's Input To Determination of the AP600 Mam Control Room Staffing Level."
- 45. WCAP-14701, " Methodology and Results of Defining Evaluation Issues for.the AP600 Human System Interface Design Test Program."j guipin 1 j jfh y /g i
I 46. Reid, G. B. and Nygren, T. E.,1988, "Ihe Subjective Workload Assessment Technique:
1 A Scaling Procedure for Measuring Mental Workload," also in Hancock, P. A., and 1 Meshkati, N., (eds.), " Human Mental Workload," (Amsterdam, North Holland).
2 Revision: 11 February 28,1997 18.8-26 3 Westinghouse
. . . . . - . . .- . . ~ - - - . . - .. .- ..
- 18. Huixta Fcetors Engineering
?- -
i i
l l HFE Vertfication and Validation l l 1 HSI Timsk HFE HSI l Steport I HSI implenwntation Desgn Funcuonal Design & Integration A Wrifkation Veritication l (Hardware & Software) g
^ l 1r 1r l
- Resolve design issues
- Design concepts g" **'NN -4
- Establish adequacy of j high-fidelity, " Wtem g
- Functional design uncept and requirements functional requrements l #'9 A* l 1 P l 3 r i I I CowTests I ution l I
Mamin-the4aop test of concrete example of funcdonal desgn: l l 1 I
- Rapid prototypes l l
- Part-taskarmAations l I Final Plant HFE HigMidety senuator I l for smilar plant I
- Factory l aooeptance test I I
.m I acceptance test I I I Figure 18.8-2 l (cn(f/ y-AP600 Man In the Loo [ Testing and Verification and Validation Activities
{ Revision: 9 W W95tingh00S8 18.8-29 August 9,1996
- 18. Human Factors Engineering
! I i
l Detection and Monitoring / Situation Awareness l
Alert Observe Identify State i
Marm System WPt3 wptg I
WPIS Plantinformacon System Narm System QDPS Plant intormation System CDPS c e enx. u )
l l
Interpretation / Planning i l Implications Goal Selection Plan Success Selectl of State computerued PWur Path Formulate Plant Informaeon System I " 8Y'#"' Computanand Procedums Actions ,
l , ~~
, Plant Worms 6an Syshm i
! Control l
l Execute Actions sort conrom
- co,ema l
l Feedback i
Monitor Goal Monitor State Vertfy Action Achievement wPis sect w Marm system um co, g Plant intormsson System
- 8 QDPS Ptst Wormsbon System cors ,
Figure 18.8 3 i
Mapping Human System hierface Resources to Operator Model A h het$$tbn- VlisklA Revision: 9 August 9,1996 18.8-30 $ WestinghouS8 l -
- 18. Human F ctors E giceering i
(3
\V) i 18.11 Human System Interface Design Test Program l Dis section describes de p!r fc AP600 human system / interface design test program.
l 1 His test program consists of two distinct parts:
I I .
Concept tests to be performed as part of the human system interface design process (as I
described in subsection 18.8.1).
l l l l .
Verification and validation (V&V) tests to be performed at the completion of the AP600
, I design process.
! l desipi i V ne goal of the human system interface test program is to systematically evaluate human I factors concems that affect plant performance and incorporne the results into the design of I the human system interface.
I l
py@'s sfuja<e 8 Plan (fagilities and plant staff activities are addressed in the humgm M -
I f f L i f y W in::6:e Facilities included in the scope of 6 A^^ test program are i the main control room, technical support center, the remote shutdown facility, and local I control stations. Staff activities included are those activities required to operate under normal, abnormal, and emergency conditions. g g l l Re "- ,v f:: S: AP600 human system interfaceY ocuses f on the following human i t /" I system interface resources:
L l l l Plant information system (including functional and physical displays of plant processes) l l . Alarm system l l . Computerized procedure system 1 . Dedicated and soft (computer based) controls l . Wall panel information system I . Qualified data processing system l }ww q im , nepesf l As illustrated in Figure 18.11-1, a two phase process is used to derme thetest program. Phase l , 1 is called issue definition. Its purpose is to integrate major operator activities with the I human system interface resources in order to establish a set of human performance evaluation I issues. Phase 2 addresses test development. He purpose of this phase is to develop testing I plans for each of the evaluation issues identified in Phase 1. Reference 1 presents a 1 description of the methodology, analysis, and the results of executing these two phases. The I results include the identification of 17 human performance evaluation issues and a description l l of the testing approach and requirements for addressing each of the evaluation issues. He i 17 human performance evaluation issues are listed in Table 18.11-1.
l l l The 17 human performance issues are organized under five headings:
1 l . Evaluations for detection and monitoring I . Evaluations for interpretation and planning O
V Revision: 9 Y W85!!ngh00S8 18.11-1 August 9,1996 1
18 Human Factors Engineering Evaluations for controlling plant state
(
Evaluations of conformance to human factors engineering design guidelines Evaluations for validation of the integrated human system interface The first 15 issues are grouped into the first three headings above.
As described in subsection 18.8.1, man-in-the-loop concept tests are performed as part of the human system interface design process. Rese concept tests are organized around the first 15 i human performance issues. Reference 2 provides a description of the AP600 man-in-the-loop test plan which includes the concept tests.
Evaluation issues 16 and 17 describe evaluations that are performed as part of the AP600 human factors verification and validation and fall under the last two headings above. A programmatic level description of the AP600 verification and validation program is provided l by Reference 3. Figure 18.8-2 shows the man-in-the-loop concept testing and the verification I and validation activities conducted as part of AP600 human factors engineering program. !
Using the programmatic level description, it is the responsibility of the Combined License applicant to develop an implementation plan for the AP600 human factors engineering verification and validation. The Combined License applicant is responsible for the execution !
and documentation of the plan. I 18.11.1 Combined License Information ,- !
l 1 '
Combined License applicants referencing the AP600 certified design will address the development, execution and documentation of an implementation plan for the verification and validation of the AP600 human factors engineering program. He programmatic level description of the AP600 verification and validation program, presented and referenced by Section 18.11, will be used by the Combined License applicant to develop the implementation plan.
l l
18.11.2 References l
- 1. WCAP-14701, " Methodology and Results Of Defining Evaluation Issues For the AP600 I l
El 199f.
Human System Interface Design.L Testr Program,"
7 Revision 1 2. WCAP-14396, " Man-In-The-Loop Test Plan Description," Revision 2, January 1997.
- 3. WCAP-14401, " Programmatic Level Descriptionjf the AP600 Human Factors l Verification and Validation Plan," Revision / January 1997.
4 3 fff l Revision: 11 February 28,1997 18.11-2 3 W6ai'gdiOUS8 u
s i I
_ _ . l
. 18. Human Factsrs EngineeriEg l
Phase 1. Issue Definition Define HSI Map HSI Define Major 4 Evaluation Evaluation 4
4 Resources Resources i to Operator issues as issues i l
j ActNities (Jnks Between i (Model HSt Resources
} Employ of Support) and Operator d
Human Permance Performance 3
Model Identify l Major Classes l of Operator Actnnbos 3
Phase 2. Test Development
=
Develop Evaluation Define Evaluation Denne Evaluation Document
- issue into Testabb + Approach for + Requiremente for + Evaluation 4
Hypothesis and ConceptTesting and Concept Testing Descriptione Performance Portormance Testing: and Performance Requrements Venfication Testme:
. Valdation .Vertflestion
. Validation
( k "I -
" ' " " ~ ~ ~ "
r M":es = ' *# ;; ' " - -4 Sy5 .n u n: Sue $es.cd Te >+ fro ds m
) ) Revision: 9 August 9,1996 3 Westhghouse 18.11 3
l '
18.' Hum:s Fact:rs Engineering l l l 18,12 Inventory 18.12.1 Inventory of Displays, Alarms, and Controls l An inventory of instruments, alarms, and controls for the AP600 systems is provided in the respective system piping and instrumentation diagrams, ne AP600 system design engineers determine the specific sensors, instrumentation, controls, and alarms that are needed to operate the various plant systems. He instmments, alarms, and controls for each system are documented in the piping and instrumentation diagram. An instmment, alarm, and control is specified by the system design engineer if it is needed to control, verify, or monitor the operation of the system and its components. System functions and their respective functional requirements are considered by the system designer when determining the need for a specific instrument, alarm, or control.
He role of the Human Factors Engineering (HFE) design team in the determination of the .
total inventory list is one of verification. As described in Section 18.5, the Huma Fnw Engin::-ing design team has functionally decomposed the plant. He top four levels of this model for the AP600, are shown in Figure 18.5-1. Each Level 4 function has a function-based task analysis (FBTA) performed as described in the Task Analysis Implementation Plan.
I Considering the plant operating modes and emergency operations, the function-based task analysis:
- Identifies the functions goals Identifies the processes used to achieve each goal Documents the performance of a cognitive task analysis of each process The cognitive task analysis of each process answers the monitoring / feedback, planning, and controlling questions. The answers to these questions identify the data for each functional process (instrumentation, indications, alarms, and controls) needed by the operator t'o make decisions. The results of the cognitive task analysis phase of each function-based task analysis are used to verify the inventory list of instruments, controls, and alarms developed by the AP600 system designers and documented in the respective design documents.
18.12.2 Minimum Inventory of Main Control Room Fixed Displays, Alarms, and Controls l Background l ne man machine interface systein design includes the appropriate plant displays, alarms, and controls needed to support a broad range of expected power generation, shutdown, and accident mitigation operations. Soft control displays and plant information displays are generated by a computer and can be changed to perform different functions, allow
- control of
' different devices, or display different information. Dese displays appear on display devices such as cathode ray tubes, flat panel screens, or visual display units. Alarms are used to direct operator attention. Soft controls are provided through devices such as a keyboard, touch screen, mouse, or other equivalent input devices. He majority of the operations for both the Revision: 10 18.12-1 December 20,1996 3 'AbiinniOUS8
l l
- 18. Haman Facton Engineering 18.12.5 References
- 1. WCAP-14651, " Integration of Human Reliability Analysis With Human Factors 1 Engineering Design Implementation Plan," Revision [ September 1996.
2 rthy 7
- 2. WCAP-13793, "The AP600 System / Event Matrix," June 1994.
l l
l
)
c
.# 1
, e i
4
\.
i 1
4 2
1
(
Revision: 11 February 28,1997 18.12-8 3 Westinghouse
- 18. H man Fcetors Ergineerirg Table 18.12.21 (Sheet 2 of 2) l MINIMUM INVENTORY l Parameter Control Display Alarm
- Manual safeguards actuation x Manual CMT actuation x l Manual main control room emergency x I habitability system actuation ( ,
~
I 5=;. ;=y h&!p"5 gc- : :c:= den
'l Manual ADS actuation (13 and 4) x Manual PRHR actuation x Manual containment cooling actuation x j Manual IRWST injection actuation x l Manual containment recirculation actuation x Manual containment iso?ation x Manual main steamline isolation x Manual feedwater isolation x )
I l Manual containment hydrogen igniter x
I (nonsafety related) e I l
.Q;'
l l Notes:
I 1. Although this parameter does not satisfy any of the selection criteria of subsection 18.12.2, its importance to l l manual actuation of ADS justifies its placement on this list. j l 2. These parameters are used to generate visual alerts (safety-related displays for the main control room; nonsafety- l l related displays for the remote shutdown workstation) that identify challentes to the critical safety functions.
I 3. These instuments are not required after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (Subsection 7.5.4 includes more information on the class IE I valve position indic tion signals, specified as part of the post accident monitoring instrumentation.) 4 1 4. This manual actuation capability is not needed at the remote shutdown workstation. , f 1
i Revision: 11 February 28,1997 18.12-10 1/ Wes21ghollS8
e FAX to DINO SCALETTI i May 7,1997 1 l
CC: Sharon or Dino, please make copies for: Diane Jackson l Ted Quay l Don Lindgren l Bob Vijuk Brian McIntyre OPEN ITEM #4122 (RAI #260.89) l To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
l Documentation" by 5/97, we believe that NRC must acknowledge receipt of all '
Westinghouse submittals by May 30,1997. This is just 23 calendar days away (18 business days). The relevant documentation related to Open Item #4122 (RAI #260.89) was provided in Westinghouse letter NSD-NRC-97-4993 dated February 21,1997. Pertinent pages of this letter are attached. It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed." {
l Jim Winters 412-374-5290 i
1
7 AP600 Opes hem Trackirg Sy: tim Dst:btse: Exec tive Summary DIte: 5/7/97 l
Selection: litem no] between 4122 And 4122 Sorted by Type a ,
item DSLR Section . Title / Description Resp NRC (W)
- O'
. WIN. _._ D_Pe. yptatus - - , - . - _ , _ .
WW - ,_ -_ Jyus _- 13,5___,, Igr No f _ __ ,,_Dac_ , j 4122 NRR/IlQMH 3.2.1 RAl-OI Lindgren Closed Action W NSD-NRC-97-4993
' ~ ~ '
- RAl# 260 89 SSAR Section 3112. "A'pplication of ClassifEation," PMi2d, state [Epar("SExture[sistems, and components clEiIEd' ~
' equipment class A, B, or C or seismic Category I are basic components as defined in 10 CFR 21.* Please clanfy how a " Basic Component" as
[ defined in 10 CFR Part 21 can also be classified as Equipment Class D, as defined in SSAR Section 3116.
~~
~ ~ ~ ~ ~- ' ~~ '~
[ Closed - See letect NSD.NRC-97-499fdated 2I2'th7. - - --
- - . - . . - - - . . . -.-. - -. .-. - - - . - . . - - - - ]. . l t
t I
i L i
M >
I i
E t
4 t
I i
s 1
i i
I i
h l
I Page: 1 Total Records: I r
~
k
-I
i
- h- l l
- f t
Westinghouse Energy Systems sa 333 Electric Corporation omswin pennsrania isuc cass :
i
- NSD-NRC-97-4993 DPC/NRC0747 Docket No.
- STN-52-003 j February 21, :997 i i
Document Control Desk U. S. Nuclear Regulatory Commission :
Washington, DC 20555 TO: T.R. QUAY l
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL I
, INFORMATION ON THE AP600.
]
l
Dear Mr. Quay:
Enclosed are three copies of the Westinghouse responses to open items on AP600 topics. Responses
- to nine RAls are included in this transmittal. RAI 410.261 provides information on Section 9 of the I
! SSAR. Responses to RAI 440.571, Revision 1, discusses the OSU Test Analysis Report. Responses i to RAls 260.83,84,85,86,87,88, and 89 address questions on Section 3 of the SSAR. l l
The NRC technical staff should review these responses as a part of their review of the AP600 design.
l These responses close, from a Westinghouse perspective, the addressed questions. The NRC should i
inform Westinghouse of the status to be designated in the "NRC Status
- column of the OITS.
Please contact Brian A. McIntyre on (412) 374-4334 if you have any questions concerning this transmittal.
fSb Brian A. McIntyre, Manager Advanced Plant Safety and Licensing l jml Enclosures
( ;
i cc: T. Kenyon, NRC - (w/o enclosures)
W. Huffman, NRC - (w/ enclosures) l N. Liparulo, Westinghouse - (w/o enclosures) l l- ._ :
3 f l 00
~
s< >
Dl$ r( g u vi
t 6
NRC REQUEST FOR ADDITIONAL INFORMATION l
.=a i El !
Question 260.89 l Re: 4122 SSAR Section 3.2.2.2, " Application of Classification." Page 3.2-5. states, in part. " Structures, systems, and components classified equipment class A. B, or C or seismic Category I are basic components as defined in 10 CFR 21." Please clarify how a " Basic Component" as defined in 10 CFR Part 21 can also be classified as Equipment Class D, as defined in SSAR Se-tion 3.2.2.6.
1 1
Response
SSAR Section 3.2.2.2, Application of Classification, applies the " Basic Component" designation to safety related l Class A. B. and C components only. SSAR Section 3.2.2.6 defines Class D as nonsafety related structures, systems I and components containing radioactivity where a conservative analysis must show that the potential for failure, due j to a design basis event, does not result in exceeding the normal offsite doses. Consistent with the requirements of '
10 CFR 21, the Class D classification is not applied to designated " Basic Compnents" quality requirements.
SSAR Revision: NONE i
1 l
h 260.89
. . _ _._. . ~ ._ ._. _ _ _ . _ __
i FAX to DINO SCALETTI l
May 7,1997 CC: Sharon or Dino, please make copies for: Diane Jackson l Ted Quay Don Lindgren l
Bob Vijuk j Brian McIntyre l
OPEN ITEM #3481 (RAI #410.295) l To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
i Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 23 calendar days away (18 business days). Open Item #3481 (RAI #410.295) was answered by Westinghouse letter NSD-NRC 5012 dated March 5,1997 (attached herewith). In addition to the material provided in the letter, l the pertinent material is included throughout Table 3.2-3 and in Sub-Section 3.2.4 of the SSAR
! These changes were included in Revision 11 of the SSAR. It is requested NRC review this material and provide defimitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
c#*
Jim Winters 412-374-5290 l
l l
l l
l / 9
AP600 Opea It:m Tracking Syst:m D t.ab se: Exec tiveSumm ry D;.te: Sn/97 +
i Selection: [ item no] between 3481 And 3481 Sorted by Type i l
Item DSLR Section Title /Ikscription Resp (W) NRC No Branch Question Type Iktail Status Engineer Status Status . g citer.N_o /. ,
__ __ _ _ _ _ _ _ . __ _ . _ __ _ D_ ate _ .
3481 NRR/SPl.B 3.2 RAl4)I Lindgren Closed . Action W NSD-NRC-97-5012
~~ ~
l lRAls 410.295 - NRC letter 8/15/1996, SSAR Tab'le 3 2-3, AP'600 riassifliation of Mechanical and Huld Systems [ Components, and Equipme'n't?
!a. Westinghouse needs to revise Table 3.2-3 to provide the classification of the following fluid systems and their associated generalized equipment:
- 1. Radiologically Controlled Area Ventilation System (VAS)
- 2. Contaimnent Recirculation Cooling System (VCS)
- 3. IIcalth Physics and flot Machine Shop IIVAC system (VilS)
- 4. Radioactive Waste Building IIVAC System (VRS)
- 5. Turbine Building Ventilation System (VTS)
- 6. Annen/ Auxiliary Nonradioactive Ventilation System (VXS)
- 7. Liquid Waste Management System 8 Gaseous Waste Management System
- 9. Radiation Monitoring System
- 10. Main Steam System 7 II. Condensate Storage System !
b 12. Reactor Coolant Pressure Boundary (RPCB) trakage Detection and Monitoring System ;
- b. In the previous version, there was a
- location
- column in the table, which is useful to the reviewer. It was removed from the table in Revision 8.
(Bring the location information back to the table.
hi~Re'sp5 s~e pdvided liy~NSD-NRU-97-50liof55/97I--~ ~ ~ ~ -]
1 I
i L
s t
i i
i Page: 1 Total Records: I
x\
k' Westinghouse Energy Systems sex 355 l Pittst:urgh Pennsylvania 15230-0355
! Electric Corporation l
l l NSD-NRC-97-5012 DCP/NRC076 Docket No.: STN-52-003 March 5,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T.R. QUAY
SUBJECT:
RESPONSE TO RAI 410.295 CLASSIFICATION OF MECHANICAL AND FLUID SYSTEMS l
Dear Mr. Quay:
Attached are responses to NRC RAI 410.29 and a related open item (OITS #289). Also attached is a l copy of SSAR subsection 3.2.4 that shows the changes made in Revision 11 that implements the conunitment in the response. Table 3.2-2 has been revised in SSAR Revision 11 to include all fluid and mechanical systems, identify the classification for all components, and provide classification of fire dampers. Table 3.2-3 has also been revised in SSAR Revi/Jn 11 to include the changes due to l other responses and design changes identified to date.
This response will permit Plant Systems Branch to close these items and finalize its input to FSER section 3.2. l If you have any questions please contact D. A. Lindgren at (412) 374-4856.
Brian A.
/p-cIntyre, Manager Advanced Plant Safety and Licensing jml Attachment cc: Diane Jackson, NRC (w/ Attachment)
W. Huffman, NRC (w/ Attachment) i i
)lOta f 1
/
, mo se ope.* e-
+
[ .
l Attachment to NSD-NRC-97-5012 l
RAI# 410.295 (3481) - NRC Letter 8/15/1996, SSAR Table 3.2-3, AP600 Classification of Mechani-cal and Fluid Systems, Components, and Equipment:
! a. Westinghouse needs to revise Table 3.2-3 to provide the classification of the following fluid systems and their associated generalized equipment:
! 1. Radiologically Controlled Area Ventilation System (VAS)
- 2. Containment Recirculation Cooling System (VCS)
- 3. Health Physics and Hot Machine Shop HVAC system (VHS) l
- 4. Radioactive Waste Building HVAC System (VRS)
- 5. Turbine Building Ventilation System (VTS) '
- 6. Annex / Auxiliary Nonradioactive Ventilation System (VXS)
- 7. Liquid Waste Management System l
- 8. Gaseous Waste Management Systerr l 9. Radiation Monitoring System
- 10. Main Steam System II. Condensate Storage System
- 12. Reactor Coolant Pressure Boundary (RPCB) LeakaFe Detection and Monitoring System L
- b. In the previous version, there was a " Location" column in the table, which is useful to the reviewer. It was removed from the table in Revision 8. Bring the location information back to the table.
Westinyhouse Resoonse
- a. Table 3.2-3 focuses on the classification of safety-related (Class A, B, or C) components and equipment in mechanical and fluid systems. Items that are AP600 equipment Class D or equivalent are identified in a general basis. The table is being revised to include the fluid and mechanical systems in AP600. Items that are not Class A, B, C, or D are not individually identified. Systems that are electrical or instrumentation systems are not included in this table.
This is consisteet with the guidance in Regulatory Guide 1.70. The components in the incore instrumentation f.ystem that have a pressure boundary function are included in the table.
Responses for the specific systems follows. The text in subsection 3.2.4 referencing the contents of the table has also been revised.
- 1. The radiologically controlled area ventilation system (VAS) is included in the table. The Class D toom coolers and valves that provide a Class D function are identified. The balance of the equipment in the VAS is Class E.
- 2. The equipment in the containment recirculation cooling system (VCS) is Class E or Class L.
l
- 3. The equipment in the health physics and hot machine shop HVAC system (VHS) is Class E.
- 4. The equipment in the radioactive waste building HVAC System (VRS) is Class E, Class L, 7
- Class F, or Class R.
f7 ma. I
O D
~
Attachment to NSD-NRC-97-5012
- 5. The equipment in the turbine building ventilation system (VTS) is Class E. Class L, or Class F.
- 6. The annex / auxiliary nonradioactive ventilation system (VXS) is included in the table. The air handling units and dampers that provide a Class D function are identified. The balance of the equipment in the VAS is Class E .
- 7. The liquid radwaste system (WLS) is included in Table 3.2-3.
- 8. The gaseous radwaste system (WGS) is included in Table 3.2-3.
- 9. The radiation monitoring system (RMS) is not a fluid system and is not included in Table 3.2-3. The system is discussed in Section 11.5. Most radiation detectors are included as a part of the system they monitor. Table 11.5-1 identifies the safety-related monitors. i Subsection 7.1.4 provides information on qualification and other requirements for safety- I related monitors.
l
- 10. The main steam system (MSS) is the nonsafety-related portion of the main steam line and l associated piping. Equipment in the MSS is Class E. The safety-related portion of the l
main steam line is included in the steam generator system (SGS). l
- 11. The condensate system (CDS) is not safety-related. Equipment in the CDS is Class E.
- 12. The reactor coolant leak detection function is provided by components and subsystems in a ;
number of systems. It does not rely on a dedicated leak detection system. Subsection 5.2.5 addresses the leak detection approach and identifies the systems which include the equipment required for leak detection.
- b. The location infonnation in the previous revision of the table merely identified in which j building a component was located. Safety-related equipment is either in the containment or auxiliary building. The location for all br most of the equipment will be identified on the line with the system name. Systems such as the plant gas system that have components in scleral of the buildings are identified as being located in "Various" buildings. More specific informa-tion on the locations of safety related equipment is provided in Table 3.11-1 and the associated information in Table 3D.5-1. Location by fire area in provided in Appendix 9A.
OITS 281 and NRC letter dated October 17,1996, item 7. e. (2) (OITS 289) identified the request to include fire dampers in. Table 3.2-3.
l l
l Westinghouse Resnonse Fire dampers have been added to the table in SSAR Revision 11. The approach used is similar to the approach used for Class D equipment.
S l 2 nou l
m
e.
Attachment to NSD-NRC-97-5012 l
1 3.2.4 Application of AP600 Safety-Related Equipment and Seismic Classification System The application of the AP600 equipment and seismic classification system to AP600 systems and components is shown in Table 3.2-3. Table 3.2-3 lists safety-related and seismic category I mechanical and fluid system component and associated equipment class and seismic category as well as other related information. The table also provides information on the systems that I contain Class D components. Additional information on the Class D functions of the various I systems 'can be found in the description in the SSAR for the systems. Mechanical and fluid I systems that contain no safety-relued or Class D systems are included in the table and general I information provided on the system. Supports for piping and components have the same classification as the component or piping supported. Supports for AP600 e'quipment Class A, B, l
and C mechanical components and piping are constructed to ASME Code,Section III, Subsection NF requirements. The principle construction code for supports for nonsafety-related components and piping is the same as that for the supported component or piping.
I Following the name of each system is the building location of the system components. Some of I the systems supply all or most of the buildings. This is indicated by identifying the locaticn as I various. Where a system includes piping or ducts that only passed through a building without I including any components that building is generally not included in the list.
l l The following list includes the systems in Table 3.2-3. The three letters in the beginning of each line is the acronym for the system. The systems included in Table 3.2-3 are listed alphabetically I by three letter acronym. Th: :y tc= p!-ted !- it.:!!c: have no compen:n : in C!= : A, B, C, l er D and h=: nc !!:: Of ecmponent: :n !uded 5 ie ::b! . Those systems. marked with an I asterisk
- are electrical or instrumentation systems and are not included in Table 3.2-3. The I components in the incore instrumentation system that have a pressure boundary function are I included in the table. See Section 3.11 for identification of safety-related electrical and instrumentation equipment.
NSSS/ Steam Generator Controls And Auxiliaries l BDS Steam Generator Blowdown System
! CNS Containment System l CVS Chemical and Volume Control System l PCS Passive Containment Cooling System PXS Passive Core Cooling System RCS Reactor Coolant System RNS Normal Residual Heat Removal System RXS Reactor System -
SGS Steam Generator System Nuclear Control and Monitoring
- *DAS Diverse Actuation System i 2 11 S Incore Instrumentation System
- OCS Operation and Control Centers
- PMS Protection and Safety Monitoring System PSS Primary Sampling System
- RMS Radiation Monitoring System f[f me. 3 i _ x
i, s
- i l
l . . , ;
l Attachment to NSD-NRC-97-5012 -
- SJS Seismic Monitoring System
- SMS Special Monitoring System Main Power Cycle and Auxdiades CDS Condensate System CFS Turbine Island Chemical Feed System ,
CPS Condensate Polishing System DTS Demineralized Water Treatinent System DWS Demineralized Water Transfer and Storage System FWS Main and Startup Feedwater System
. GSS Gland Seal System HDS Heater Drain System MSS Main Steam System MTS Main Turbine System RWS Raw Water System l Turbine Island Vents, Drains and Relief System TDS Class 1E and Emergency Power Systems
CES - Condenser Tube Cleaning System Circulating Water System l CWS SFS Spent Fuel Pit Cooling System l SWS Service Water System ;
TCS Turbine Building Closed Cooling Water System Auxdiary Steam ASS Auxiliary Steam Supply System Generation and Tran=nierian j
- ZAS Main Generation System
- ZBS Transmission Switchyard and Offsite Power System
- ZVS Excitation and Voltage Regulation System Radweste i WGS Gaseous Radwaste System i WLS Liquid Radwaste System WRS Radioactive Waste Drain System WSS Solid Radwaste System .
l HVAC VAS Radiologically Controlled Area Ventilation System VBS Nuctsar Island Nonradioactive Ventilation System VCS Containment Recirculation Cooling System VES Main Control Room Emergency Habitability System VFS Containment Air Filtration System 7p
- m. 4
~w O
w-ll.i
.h Attachment to NSD-NRC-97-5012 VHS Health Physics and Hot Machine Shop HVAC System VLS Containment Hydrogen Control System VRS Radwaste Building HVAC System VTS Turbine Building Ventilation System VUS Containment Leak Rate Test System VWS Central Chilled Water System VXS Annex / Auxiliary Nonradioactive Ventilation System VYS Hot Water Heating System' VZS Diesel Generator Building Ventilation System Turbine-Generator Controls and Auxiliary CMS Condenser Air Removal System HCS Generator Hydrogen and CO2 Systems HSS Hydrogen Seal Oil System LOS Main Turbine and Generator Lube Oil System
- TOS Main Turbine Control and Diagnostics System l l Material Handling l FHS Fuel Handling and Refueling System l MHS Mechanical Handling System 1
Piping Services CAS Compressed and Instrument Air Systems
! DOS Standby Diesel and Auxiliary Boiler Fuel Oil System FPS Fire Protection System PGS Plant Gas Systems PWS Potable Water System Non-Class IE Power Systena
- ECS Main AC Power System -
- EFS Communication Systems
- EGS Grounding and Lightning Protection System l
! *EHS Special Process Heat Tracing System
- ELS Plant Lighting System
- EQS Cathodic Protection System Non-Nuclear Controls and Monitoring
- DDS Data Display and Processing System I *MES Meteorological and Environmental Monitoring System
- Plant Control System
- PLS
. *SES Plant Security System SSS Secondary Sampling System
- TVS Closed Circuit TV System l- 5 mu
.. . . - . .- . .-. . . - . .- - . _ _ . . .= .-.. .
J
- t. I l.
1 1
Attachment to NSD-NRC-97-5012 )
l Non-Radioactive Drains DRS Storm Drain System l
RDS Gravity and Rt.< ? Drain Collection System SDS Sanitary Drainage System WWS Waste Water System l ":: rj:::.= ph::d l :!!:: hav: n: ::=p:=:n': ;- C!::::: .1, 3, C, c.- D .-d hav: n:
!!:: :l::=p:n:n:: int!ud:d en 'h: Tab!: 3.2 3. Those systems marked with an asterisk l (*) are electrical or instrumentation systems and are not included in Table 3.2-3.
l 1
l l
l J
l l
I l
uos. 6
. o
1 L.
1 To . BM WoWrmn 1
(t. 20 Gar \in bh ~Hs is &c Sumynec Sun . Awocked are l n 69 Rev o res rmse in RAI %0,% C ATWS3 !
and i l
A a cevised resyse b %d RAI. (I dad WM if Ms nos been fwma Ly p rovcdod) .
I vocerstand Sommer Wavks on ATtcs reprc.
Can vot -trgo & cepub anc. sim I ;
-% info in on RAr_ respcme' p y issue T-Fnch, We plan +o crea4r a reput very mach like he crtachec RAI respme revisicrn (Ifed i pi; %is saf601 If Summec wods i
sor nehng ese, pieoa han him ogi bai
+o nut . bo you Now dwb en cesemse rev e Arws rex + mid 6 ' cue' ?
@aAs,$ bit \hG35 S/G J
p\ & ab y 1
. 15. Accide:t Analyses 15.8 Anticipated Transients Without Scram 15.8.1 General Background An anticipated transient without scram (ATWS) is an anticipated operational occurrence during which an automatic reactor scram is required but fails to occur due to a common mode fault in the reactor protection system. Under certain circumstances, failure to execute a required scram during an an:icipated operational occurrence could transform a relatively minor i transient into a more severe accident. ADVS events are not considered to be in the design basis for Westinghouse plants.
I 15.8.2 Anticipated Transients Without Scram in the AP600 For Westinghouse plants, the ATWS rule (10 CFR 50.62) requires the installation of ATWS mitigation systems actuation circuitry (AMSAC), which consists of circuitry that is separate from the reactor protection system, to trip the turbine and initiate decay heat removal.
The basis for the ADVS mle requirements, as outlined in SECY-83-293 (Reference 1), is to reduce the risk of core damage due to ADVS to less than 10-5per reactor-year.
I He AP600 includes a diverse a'ctuation system which provides the AMSAC protection features mandated for Westinghouse plants by 10 CFR 50.62 plus a diverse reactor scram (see Section 7.7). Rus, the ATWS mie is met.
15.8.3 Conclusion i The AP600 is equipped with a diverse actuation system, which provides the functions of AMSAC. He ATWS core damage frequency for the AP600 is well below the SECY-83-293 goal of 10-5 erp reactor year. De AP600 ADVS core damage frequency is discussed in I Chapter 33 of the Probabilistic Risk Assessment (PRA). He AP600 design meets the ABVS rule (10 CFR 50.62), and its ADVS core damage frequency safety goal basis.
I 15.8.4 Combined License Information I
l Dis section has no requirement for additional,information to be provided in support of the i Combined License application.
I 15.8.5 References l 1. Dircks, WJ., " Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram (ADVS) Events," SECY-83-293, USNRC, July 19,1983.
1 Revision: 9 15.8 1 August 9,1996 Y W96fingh00S4
l C \-T A -9 3 - 12 6 LEV-0 g NRC REQUEST FOR ADDITIONAL INFORMATION l , , . _
=
Question 440.26 Section 15.8 of the SSAR states that AP600 plant design includes a diverse actuation system, which provides for all of the AMSAC protection features mandated for Westingbouse plants plus a diverse reactor scram. and thus meets the ATWS rule. However. it does not provide an ATWS analysis to demonstrate that AP600 ATWS response is consistent with that considered by toe staff in its formulation of the 10 CFR 50.62 design requirements for current plants. Provide such an analysis.
Response
An anticipated transient without scram (ATWS)is a beyond-design basis event initiated by a postulated anticipated operational occurrence (i.e., Condition 11 event, or a fault of moderate frequency) during which an automatic reactor scram is required, but fails to occur because of a common mode fault in the reactor protection system. To mitigate the consequences of postulated ATWS events, the AP600 l is equipped with a diverse actuation system (DASI. The DAS provides a diverse scram actuation and l the equivalent functions of AMSAC required by 10 CFR 50.62. The AP600 ATWS core damage frequency is discussed in Section F.2.22 of the PRA study. The ATWS core damage frequency for the AP600 is well below the SEC7-83-293 goal of 10-' per reactor-year. The AP600 design meets the ATWS rule (10 CFR 50.62) and its ATWS core damage frequency safety goal basis.
A deterministic complete loss of normal feedwater ATWS analysis will be performed. The purpose of this analysis will be to demonstrate that the AP600 ATWS response characteristics are comparable to the responses of other Westinghouse plants and are therefore consistent with the bases considered I by the staff in the formulation of 10 CFR 50.62 design requirements. The analysis methodology used I l will be based on the analyses presented in previous Westinghouse submittals (References 440.26-1 l and 440.26 2). This analysis will be submitted to the staff in August 1993.
l l
Reference 440.26-1 WCAP 8330, " Westinghouse Anticipated Transients Without Trip Analysis," August l 1974.
440.26-2 Anderson, T.M.,
- ATWS Submittal," Westinghouse Letter No. NS TMA 2182 to S.H.
l Hanauer of the NRC, December,1979.
SSAR Revision: None l
I l
440.26 1 W-Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION CN-TA-93-1n - .~> d 0173 l Response Revision 1 1
, Cuestion 440 26 Secuen iM of me SS AR uates that AP600 plant design includes a diserse xtuauon $3 stem. which proudes for
.tll ot' be A.\tS AC protecuen features mandated for Westingbouse plants plus a diserse reactor wtam. and thus ,
meets the ATWS rule. Howeser, it does not proude an ATWS analysts to demonstrate dat AP600 ATWS I response is consistent with (bat considered by the staff in its formulauon of the 10 CFR 50.62 design requtrements for current plants. Proude such an analysis.
Response
General Background
i request. & deterministic complete loss of normal feedwater ATWS analysis has been performed. The purpose of this analysis is to demonstrate that the AP600 ATWS response charactenstics are comparable to the responses of other Westinghouse plants and are, therefore consistent with the bases considered by the staff in formutation of 10 CFR 50.62 design requirements. The analysis methodology used is based on analyses presented in previous Westinghouse submittals (References 440 26 4 & 440 26-5).
For Westinghouse plants. the ATWS rule (10 CFR 50.62) requires the installation of ATWS mitigation systems actuaton circuitry (AMSAC), which is separate from the reactor protectsn system, to inp the turbine and initiate resadual heat removal. The AP600 design includes a Diverse Actuation System 1 (OAS) that provides the AMSAC protection features mandated for Westinghouse plants by 10 CFR 50 62 plus, among other functions, a drverse reactor scram. The basis for the ATWS rule requirements.
as outlined in SECY 83 203 (Reference 440.26-3), is to reduce the nsk of ATWS related core damage to less than 10 per reactor year.
Identification of Causes and Accident Description The most limiting ATWS events for Westinghouse plants were found in a previous study (Reference 440 26-4) to be the heatte transients caused by a reduction in the heat removal capability
, of the secondary (steam) side of the pat. Because of the strong negative moderator temperature feedback in a PWR. heatup accidents resulting from a loss of heat sink cause the nuclear heat generation rate to decrease until the reactor power matches the heat extraction from the passive residual heat removal system. These events proceed relatively slowty in the AP600 due to the large water inventory in both the pnmary and secondary sides.
Results for previous Westinghouse designs presented in References 440 26-4 and 440.26 5 indicate that the reactor reaches a steady state condition with no impairment of reactor coolant system integrrty or significant fuel damage. Consistent with the very low probability of an ATWS event, these analyses employ several best estimate assumptions. Turbine tnp and the passive residual heat removal system are assumed to be available to mitigate the event.
440.26(RI)-1 W Westif)ghouse
1 C T~ - T e v- o s - y--J 0174 NRC REQUEST FOR ADDITIONAL INFORMATION Round:0 N
Question Sof: 09/23/92 Response Revision 1 1
The AP600 transient calculations assume actuaton of the passive residual heat removal system and a turcine top as the pnmary actons needed to mitigate the effects of an ATWS. Both of these functions l are assumed to actuate on a signal generated by the OAS when the decreasing steam generator water level reaches the wide range low level setpoint. This same OAS signal also generates a diverse l reactor tnp that is not modeled in the base case presented in this report. Such a diverse reactor tnp function is not needed and therefore not implemented in the standard Westinghouse AMSAC system its inclusion in the OAS provides an increased level of protection for the AP600. This function represents an independent means of inrtiating RCCA in.eertion, in the unlikely event that the reactor protecten system fails to generate a required reactor tnp signal.
There are two distinct failure categones identrfied in the Subsection F 2.22 of the AP600 Probabilistic I Aisk Assessment (PRA) (Reference 440.26-6) which could prevent the control rods from inserting after receipt of a reactor tnp signal dunng an antcipated transient. The two categones are a failure in erther the mechanical or electncal porten of the reactor tnp system. For the case of a mechanical failure;
} turbine tnp startup feedwater, and the passive resdual heat rernoval system are assumed to be available via the control and protecten systems. For the case of an electncal failure; the failure in the reactor tnp system is conservativeh assumed to be common to the entire reactor protecton sys%m.
Therefore', the DAS must provide turbine tnp and passive resdual heat removal system actuaton on
- low steam generator wde range level. Since the OAS setpoints are outsde the range of normal reactor protecten system setpoints, turbine tnp and passive residual heat removal system actuation are delayed for the case of an electncal failure and the resulting transient is more severe. Only the timrting case of an electncal failure is presented in this report.
Limitino Cnteria
~
Consistent with prevous assessments and the crttens used to define a successful event outcome for PRA purposes, it is conservatively assumed that if any one of the following requirements are not met, unacceptable core damage can occur dunng an ATWS transient (Reference 440 2610):
- 1. The peak reactor coolant system pressure must not exceed the pressure limit corresponding to the service lim #t stress of the ASME Boiler and Pressure vessel Code for Level C (' emergency condition *) events (Reference 440.26-7). The pressure limrt assumed in Reference 440 26-10 is 3200 psig.
- 2. Reactor coolant system heat removal must be adequate, both before and after the core is brought subentical. For the AP600, long term heat removal is provided by the passwo residual heat removal system. Note that for the ATWS event, this long term cooling requirement must be met without RCCA inserton.
- 3. Actions must be initiated to achieve subentcalfty within an acceptable time penod. For AP600, operator actions are available to manually initiate boration independent of the reactor protection system.
440.26(RI) 2 W Westnghouse
I l-NRC REQUEST FOR ADDITIONAL INFORMATION EN " # -9 3 - :1 r . ms 1
Res - s. ev,sen , g With respect to achieving subenticality. even though the OAS provides a diverse reactor trip signal. it is conservatively assumed for analysis purposes that the shutdown condrtion must be achieved without RCCA insertion. In the AP600. boron injection from the CMTs produces the required shutdown condition. The operator can manually actuate the CMTs via the OAS. An attemative means to achieve subenticality is for the operator to initiate boration using the makeup pumps in the Chemical and l Volume Control System (CVS).
Method of Analysis l
The LOFTRAN code (Reference 440 26 2), including the modtfications for the AP600 passrve safety systems as desenbed in Appendix 158 of the SSAR. is used to compute the reactor transient response I to the ATWS event for AP600. The event analyzed is a complete loss of normal feedwater (LONF) with i reactor tnp signals being generated, but no RCCA inserton actually taking place. Previous studies (References 440 26 4 and 440 26 5) have shown that this event typcally produces the maximum reactor coolant system pressure for Westinghouse PWRs.
I Major assumptions made in this analysis are: j
- 1. The transient is initialized from nominal full power condrtions. I
- 2. An AP600 specific Doppler feedback model for system conditions indicathe of an ATWS is input as a function of power and core inlet mass flow.
3 The moderator temperature coefficient (MTC) used in the analysis is 7.3 pcm/*F. This j value represents a coefficient, amved at iteratkely, that gives a peak reactor coolant system pressure of approximately 3200 psig danng the ATWS. The selection of MTC in this manner is consistent with the analysis methodology used in previous Westinghouse submrttals (References 440 26 9 and 440 2610). The magnitude of the MTC used in this analysis is comparable to that used in the previous submrttals.and is within the expected range for AP600. j
- 4. The ANS 5.1 1979 decay heat model (+ 2 sigma) is used. (Reference 440.26 8)
- 5. Both pressunzer safety vanes are assumed available. The relief model assumes 3 percent and 10 percent pressure accumulation for steam and water relief, respectrvely. l The AP600 does not include power operated relief vanes for the pressurizer. l
- 6. Main feedwater suppfy to both steam generators falls to zero in 4 seconds, with no l
main feedwater afterwards.
l l
I 1
440.26(RI) 3 W WestifighetJse
1 CN-To-93-123 NRC REQUEST FOR ADolTIONAL INFORMATION 0175 Round: 0 l M Question Set: 09/23/92 Response Revision 1 7 The OAS is assumed to actuate on the wide range steam generator low level DAS signal The analysis setpoint is considered to be conservatively low. and will therefore l delay OAS actuaten. Given a fixed MTC. a higher level setpoint and an earlier DAS actuaton would produce a lower peak pressure for a gNen ATWS event. AlternatNely.
if a case is targeted to produce the peak pressure of 3200 psig, a higher level setpoint j would allow meeting the limit pressure with a less negatrve moderator temperature coefficient.
- 8. CAS setpoints are typicalh set outside the range of normal reactor protection system setpoints or, as an alternative, the OAS functions include delays that allow time for the reactor protection system function to actuate. Since a conservatively low OAS wide range steam generator level analysis setpoint is assumed, no additonal delay on this signal is assumed.
l
- 9. The turbine is assumed to trip 4 0 seconds after the wide range steam generator low fevel DAS setpoint is reached.
- 10. The passive residual heat exchanger vatves are assumed to be fully opened 10 seconds after the low wide range steam generator level setpoint is reached.
! 11. The design value of 40 percent steam dump to the condenser is modeled for I
consematism.
l 12. Steamline isolaten. which would actuate on the low steamline pressure signal, is conseNatNety not modeled in itse limiting ATWS case for AP600. If assumed in the l analysis, steamline isolation would occur pnor to the predicted turbine tnp (per item e9. above). Steamline isolation, dunng this event. has the same beneficial effect on the reactor coolant system pressure transient as a turbine tnp. That is, the steamline isolation would " bottle up' the steam generator, thereby reducing the secondary system heat removal rate. and produce a more rapid reactor coolant system l heatup. The result is increased reactivity feedback that would produce a more rapid reducten in the core power.
- 13. Startup feedwater is conservatively assumed to be unavailable.
14 Following turbine trip. steam relief through the spnng loaded steamline safety vanes is assumed if the steamline pressure exceeds the safety valve setpoint (1100 psta) with a 3 percent allowance for accumulaten.
l 440.26(RI) 4 gg I
NRC REQUEST FOR ADDITIONAL INFORMATION C 1V - T A - 9 3 - 12 d Response Revision 1 15 As an additional conservatism. CMT actuation (safety injecten) and the associated '
l reactor coolant pump tnp is not modeled for the duration of the transient analyzed.
l Ounng the analyzed event. the low steamline pressure safety iniection signal is l
( generated by the pnmary protecten system. but the resulting CMT actuation and )
reactor coolant pump inp signals are ignored. Analyzing the event with the pumps l
l operating is consistent wrth ATWS analyses for standard Westinghouse PWRs that
( have snown the ATWS cases with loss of off site power and reactor coolant pump coastdown to be less limiting than those that maintain forced pnmary system flow. If modeled a reactor coolant pump tnp produces addrtenal heatup and the resulting feedback generates additional negatNe reactrVity. As part of this-AP600 ATWS I analysis effort, a case with CMT actuation and RCP tnp modeled was performed to confirm this assertion.
I Results The ATWS case analyzed for AP600 is a complete loss of normal feedwater flow to both steam generators. The sequence of events for this analysis is presented in Table 1. The first reactor tnp setpoint reached is low steam generator water level (narrow range) at 45.5 seconds. Although a reactor ;
tnp signal is generated, the ATWS scenano dictates that control rod inserton is assumed to fail.
Similarly, the associated turbine tnp due to the reactor tnp signal generated at this time in the event is ignored due to an assumed failure in the electncal system.
Though the injecten of borated water is conservatively not modeled in the analyzed case, the low steamline pressure 'S' that could actuate the CMTs. is generated at 62.2 seconds. Following this signal, there is a nominal demy time of about 12 seconos before the CMT valves open and injection flow begins. A similar delay applies to the reactor coolant pump coastdown that is automatically initiated by the reactor protection system in parallel wrth CMT actuation. If credit for CMT actuation had been taken, boration from the CMT and RCP coastdown would have helped to minimize the pressunzaten.
The low steam generator water level (wide range) DAS setpoint is reached at 73.4 seconds. Turbine tnp and passive residual heat removal system actuation are initiated upon receipt of this signal after delay times of 4 and 10 seconds, respectrvefy. These two functions lead to a successful termination of the heatup transient. The peak reactor coolant system pressure is reached at 119.0 seconds. The predicted peak reactor coolant system pressure is 3198 psia.
The nuclear power transient is presented in Figure 1 1. As the reactor coolant system heatup proceeds, the negatrve moder1stor density ocefficient produces a decrease in reactor power. The core power falls to a value below the rate of passive residual heat removal as shown in Figure 12.
Eventually, core power reache,2n equilibnum dictated by the combined heat removal capability of the passive residual heat terreval system and the steam generators.
440.26(RI) 5 3 Westingh00se
CN-TA-93-in - n .i ms i l
NRC REQUEST FOR ADDITIONAL INFORMATION Round: 0 M
t Question Set: 09/23/92 Response Revision 1 l Maximum reactor coolant system pressure and Tm (the average of the inlet and outlet temperatures for a loop) as functions of time are presented in Figures 13 and 1-4. respectuely These two figures show that the reactor coolant system heatup and pressunzation is terminated in conjunction with the reduction in core power Pressunzer pressure, water volume, and relief rate throughout the transient are shown in Figures 15.
- 16. and 17. respectuely. When the pressunzer becomes water solid (97.5 seconds), the entire reactor coolant system begins to pressunze quickly. After the heatup is terminated. the pressunzer pressure is reduced below 2500 psia and the pressunzer safety valves reseat at 195.0 seconds. As l the reactor coolant system continues to cooldown, the pressunzer eventually regams steam space (288 seconds).
! The total reactor coolant system mass flow as a function of time is shown in Figure 18. For l conservatism, the reactor coolant pumps are assumed to operate throughout the transient. The total l mass flow is maintained above 80. of nominal throughout the transient.
In summary, the AP600 response to a postulated loss of normal feedwater ATWS event is similar to previous Westinghouse PWR designs. The MTC used in the analysis ( 7.3 pcm/'F), which gives a peak reactor coolant system pressure of approximately 3200 psig,is cons 4 stent wrth achieving an ATWS core damage frequency well below 10 per reactor year.
Altemate Case I
l The OAS also provides a diverse reactor scram, which would still be available despite an electncal failure in the normal reactor protection system. To demonstrate the expected plant response to a less l of normal feedwater event coincident with a failure in the electncal part of the normal reactor tnp
! system, an attemate case was analyzed assuming the presence of the diverse reactor scram initiated by the OAS. The same moderator temperature coefficient as the base case was maintained for the alternatNe Case.
l The results of an attemative case which assumes the presence of the diverse reactor scram are presented in Figures 21 through 2 8. The time sequence of events is presented in Table 2. The sequence of events is identical to the base case up until the time of diverse reactor scram I (75.4 seconds), after which the transient is essentialty terminated. The pressunzer never becomes l
water solid, thus, there is no sharp increase in pressure. The peak reactor coolant system pressure for this case is 2571 psia, which is a 627 psia benefit compared to the base case wrthout the dNerse scram. The peak pressure occurs at 64.5 seconds,9 seconds before the OAS signal is generated.
This shows that the diverse reactor scram terminates the pressure transient independent of turbine Inp l or passive resedual heat removal system actuation and eliminates any challenge to the pressure limit.
l 440.26(R1) 6 W Westirighouse l
NRC REQUEST FOR ADDITIONAL INFORMATION CN-TA-93-129 0 l'? 9 Response Revision 1 Conclusions The analysis results show that the AP600 produces acceptable responses to the limrting pressure ATWS event. The results of the base case are similar to typical Westinghouse PWRS thus. the AP600 ATWS response is consistent with that considered by the staff in its formulation of the 10 CFR 50 62 design requirements for current plants. The AP600 OAS provides the AMSAC protection features mandated for Westinghouse plants by 10 CFR 50 62 plus, arnong other functions. a diverse reactor scram. The results of the attemative case. assuming the presence of the diverse reactor scram.
demonstrate the added capabiltty for the OAS diverse reactor scram to mitigate the consequences of an ATWS event. Given the presence of the OAS, the results of the PRA. as discussed in Subsection F 2 22 of Reference 440.26-6. show that the ATWS core damage frequency contributen for the AP600 is well below the goal of 10 The AP600, therefore, meets the ATWS rule (10 CFR 50.62) and its ATWS core damage frequency safety goal basis.
SSAR Revision: None 440.26(RI) 7 W Westifighouse
C N- T A 12^ _
NRC REQUEST FOR ADDITIONAL INFORMATION Round:0 1
N Question Set: 09/23/92 Response Revision 1 A ef erences.
440 26 1 Simplified Passive Advanced Light Water Reactor Plant Program. AP600 Standard Safety Analysts Report.' OE AC03-90SF18495. June 26.1992.
440 26 2. Burnett, T. W T . 'LOFTRAN Code Desenption,* WCAP//907 P A (Proprietary) and WCAP 7907 A (Nonpropnetary), Apnl 1984.
440 26 3. Circks. W. J., " Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram (ATWS) Events,' SECY 83 293, USNRC, July 19,1983.
440 26 4. WCAP 8330. ' Westinghouse Anticipated Transients Without Trip Anahsis.*
August.1974.
440 26 5 Anderson, T. M.. *ATWS Submrttal.' Westinghouse Letter No. NS TMA 2182 to S. H.
Hanauer of the NRC, December,1979.
l 440 26 6. 'AP600 Probabiliste Risk Assessment,' Con'.;act No. DE ACO3 90SF18495, June 26, 1992.
440 26 7. *ASME Boiler & Pressure Vessel Code, and American National Stadard.' ACl Standard 359 83, Secton ill, Divison 1, Subsection NB 3224, Juh,1980.
440 26 8. ANSl/ANS 5.1 1979. August 1979, 'American National Standard for Decay Heat Power in Light Water Reactors.'
440 26-9. WCAP 11992, ' Joint Westinghouse Owners GroupM/estinghouse Program: ATWS Rule Administraton Process,' Oecember,1988.
440 26 10 WCAP 11993, " Joint Westinghouse Owners GroupWestinghouse Program:
Assessment of Compliance With ATWS Rule Basis for Westinghouse PWRs,'
Oecember,1988.
I
\
440.26(RD 8 3 Westirigh00se
C N - T A - 9 3 - 19'e.d .s ,
NRC REQUEST FOR A00lTIONAL INFORMATION 018i Response Revision 1 Table i Time Sequence of Events for the Loss of Normal Feedwater Anticipated Transient Without Scram Event Base Case j l
Event Time (s) l Ntain feedwater supply to all steam generators is terminated 0-4 Low steam generator water level (narrow range) reactor trip setpoint reached 45.5 ifailure oi RCCA insertion assumed) )
l Pressunzer safety valves open 60.5 l Low steamline pressure "S" setpoint reached 61.7 (signal ignored in analysis)
Low steam generator water level (wide range) DAS setpoint reached 73.4 l CMT actuation on low steamline pressure "S" signal conservatively not modeled 73.7 Steam generator tube uncovery 74.5 Turbine trip assumed to occur on DAS generated signal 77.4 Passive residual heat exchanger valves opened 83.4 Pressunzer fills with water 97.5 Peak RCS pressure is reached (3198 psia) I19.0 Steam generator dryout 148.0 Pressurizer safety valves rescat 195.0 Pressurizer regains steam space 238.0 440.26(R1) 9
C N -T ^ -9 ') - 19. ' us2 l NRC REQUEST FOR ADDITIONAL INFORMAflON I Round: 0 l l
Question Set: 09/23/92 i t
Response Revision 1 l l
i Table 2 l Time Sequence of Events for the Loss of Normal Feedwater I l
Anticipated Transient Without Scram Event l
l - Alternate Case: Diverse Reactor Scram Assumed i l l Event Time ts)
, 1 Stain feedwater supply to all steam generators is terminated 04 l Low steam generator water lesel (narrow range) reactor trip setpoint reached 45.5 I j (failure of RCCA inserdon assumed) j Pressurizer safety valves open 60.5 I l \
l Low steamline pressure "S" setpoint reached 61.7 l (signal ignored in analysis)
Peak RCS pressure is reached (2571 psia) 64 5 Low steam generator water level (wide range) DAS setpoint reached 73.4 l
CN1T actuauon on low steamline pressure "S" signal conservadsely not modeled 73.7 l Diserse reactor scram assumed to occur on DAS signal 75.4 l
Turbine trip assumed to occur on DAS generated signal 77.4 l Pressunzer safety valves rescat 79.5 l
l Passise residual heat heat exchanger valves opened 33.4 l l
1 440.26(RI) 10 W Westloghetjse
l CN-T A 126 NRC REQUEST FOR ADDITIONAL INFORMATION l
Response Revision 1 l
Figure 1 1 AP600 LONF ATWS: Base Case Nuclear Power 12 l
- 1 l l 1
I 1
- 08 Z
1 3 )
b l 3 l 5 0.6 t
i i 4 l
S I
04 -
1 02 -
l 9
I o 50 100 150 200 250 300 Time (sec) l l
i i
440.26(RI) 11 W-We@gh0058
CN-TA-93-123
. 0134 ,
NRC REQUEST FOR ADolTIONAL INFORMATION Round:0 Question Set: 09/23/92 Response Revision 1 l
Figure 12 l AP600 LONF ATWS: Base Case !
Core Heat Flux and PRHR Heat Removal 0 25 02 -
Core Heat Flux
[ 0.15 t
4 3 0.1 --
l PRHR Heat Removal 0 05 -
0 O 50 YO 150 200 250 300 Time (soc) l 440.26(RI) 12 W WE910gh00se l
l
NRC REQUEST FOR ADDITIONAL INFORMATION CX-TA-93-123 O..
t.
mn:amsmas Response Revision 1 Figure 13 AP600 LONF ATWS: Base Case Maximum RCS Pressure 3.400
. 0 -
Peak Pressure Time (11111 LIEl 3198 1190 S
j 2.800 -
5 .
I 2.600 -
1 5
3 2.400 2.200 -
2,000 O 50 100 150 200 250 300 Time (sec) 440.26(R1)-13
CN-TA-93-123 013i NRC REQUEST FOR ADDITIONAL INFORMATION Round: 0 l Question Set: 09/23/92 i Respona Revision 1 l
Figure 14 AP600 LONF ATWS: Base Case Tavg 660
%0 -
620 -
, 600 -
6.
en
~ 580 -
/
560 -
540 -
I I
520 0 50 100 150 200 250 300 Time (soc) 440.26(RI)-14 W
Westinghouse 1
l I
l
1 l
C I- T A - 9 3 ~ o ^.
0187 NRC REQUEST FOR ADDITIONAL INFORMATION .
1
.- ? l Response Revision 1 l
l I
l 4
Figure 15 AP600 LONF ATWS: Base Case Pressurizer Pressure 3.400 3.200 -
l l
g 3.000 -
b -
3 2.800 -
f
! 2.600 -
5 l
E 2.400 -
2.200 -
2.000 0 50 100 150 200 250 300 Time (sec) 440.26(RI)-15 W Westloghouse
I CN-T A 123 g g, l NRC REQUEST FOR ADDITIONAL INFORMATION Round:0 i M Question Set: 09/23/92 Response Revision 1 l
l 1
1 Figure 1-6 AP603 LONF ATWS: Base Case Dressurizer Water Volume 1 A00 5
1200 -
a b /-
- 1.000 T
3 e
E 800 --
600 o 50 100 150 200 250 300 Time (sec) i I
i l
I 440.26(RI)-16 3 Westloghouse l
l
. C N - T .A - 9 3 '_21 l NRC REQUEST FOR AD0lTIONAL INFORMATION Response Revision 1 ,
Figure 17 AP600 LONF ATWS: Base Case Pressurizer Steam and Water Relief N20 --
3 li -
-5
-3 o
15 3
~
E.
h10 3
3 w
3 j5 -
1 0
O SO 100 150 200 250 300 Time (sec) l l
440.26(RI)-17 9ase l
I
1 C X - T A - 9 3 ~ ~. 3. ' ouo NRC Rl! QUEST FOR ADDITIONAL INFORMATION Round: 0 W
Question Set: 09/23/92 Response Revision 1 Figure 18 AP600 LONF ATWS: Base Case Core Coolant Mass Flow i
i l
l 1 l
l E
O -
e 08 -
E r
2 1 A
)
2
\
l j 06 '
l 2 '
o l
j on -
l 1
i 5 l u f
1 02 -
I i
0 0 50 100 150 200 250 300 Time (sec) l l
1 1
440.26(RI) 18 W Westinghouse
l .
l li." b", , ,
NRC REQUEST FOR ADDITIONAL INFORMATION 0191 i Response Revision 1 1
l l
Figure 21 l
A0600 LONF A1WS: Diverse Scram Assumed Nuclear Power 12 1
N l l g 0.8 -
)
, o
- . 1
, 5 1 5 0.6 --
L 4
2 l
0.4 -
1 0.2 -
i 0 50 100 150 200 250 300 Time (sec) i l
i 440.26(RI) 19 w_ Westinghouse
CN-TA-93-:"" 0192 NRC REQUEST FOR AD0lTIONAL INFORMAilON Round: 0 M Question Set: 09/23/92 Response Revision 1 i
)
Figure 2 2 AP600 LONF ATWS: Diverse Scram Assumed Core Heat Flux and PRHR Heat Removal 0 25 0.2 --
Core Heat Flux 5o'S t,
5 I
1 l 3
0.1 PRHR Heat Removal 0.05 - -
1 0
O S0 100 150 200 250 300 Time (sec) l l
l i
l l
440.26(RI)-20 W.
Westloghetse l
1 4
l NRC REQUEST FOR ADDITIONAL F A 10 0* ,9 0133 l
Response Revision 1 i
l i
l Figum 2 3 AP600 LONF A1WS: Olverse Scram Assumed 3.400 3 200 -
} 3.000 -
S 5
Q 2.800 -
j Peak Pressure Time l l
- n
~
M UltCl l I
2571 64.5 N .600 2 -
9 c
I j 2,400 -
2.200 -
2,000 O 50 100 150 200 250 300 Time (sec) i uo.26(RI)-21 l T walopouw
l lo CN-T A 123 0134 NRC REQUEST FOR ADDITIONAL INFORMATION Round:0 M
(
Question Set: 09/23/92 Response Revision 1 1
I l
l Figure 2 4 l AP600 LONF ATWS: Diverse Scram Assumed I 660 640 -
l 620 -
1 6% -
s
?
- 580 -
560 -
l 540 -
i
~
l i
1 520 i 0 50 100 150 200 250 300 Time (sec) l l
l l
l i
(
l 440.26(RI) 22 N==WGSilfighDUS8 l
l l
l U N - T P. 3 ~.
,. i 1
NRC REQUEST FOR ADDITIONAL INFORMATION 0195 l l
Response Revision 1 i
Figure 2 5 AP600 LONF ATWS: Diverse Scram Assumed Pressurizer Pressure 3,400 3.200 -
g 3.000 -
b e
3 2,800 -
i
$ 2,600 -
5 3
I 2,400 -
l 2.200 -
2,000 0 50 100 150 200 250 300 Time (sec) t
uo.26(RI)-23 W
Wedoppose l
0196 j NRC REQUEST FOR ADDITIONAL INFORMATION l Round: 0 Question Set: 09/23/92 Response Revision I
{
l l
l l
Figure 2-6 AP600 LONF ATWS: Olverse Scram Assumed Pressurizer Water Volume 1.400 - )
1 l
5
- I l
{ 1.200 8
> i a
l /
1,000 --
E I
800 --
600
- O 50 100 150 200 250 300 Time (sec) i l
l 440.26(RI)-24 W M @ouw l
' j om 0 am 0 O ,, '
~-
.. \
NRC REQUEST FOR ADDITIONAL INFORMATION I
OI37 '
Response Revision 1 1
1 l
l I
Figure 2 7 l AP600 LONF ATWS: Diverse Scram Assumed l Pressurizer Steam and Water Relief Mt 20 eo e
- is k
l10 --
M 3
M 3
c
$5 -
l t 0 '
0 50 100 150 200 250 300 ;
rim. (.=> l l
440.26(RI)-25
! W.
wisinghouse
l
- C N - T J - 9 3 - 12:1 NRC REQUEST FOR ADDITIO AL INFORMATION om l l
Round:0 M Question Set: 09/23/92 Response Revision 1 l
I Figure 2-4 AP600 LONF ATWS: Diverse Scram Assumed Core Coolant Mass Flow _
l Z
O .
- t. 0.8 - I 1
3 2
x !
j 0.6 --
5 b,0.4 -
5 0
02 -
1 I
0 50 100 150 200 250 300
, Time (see)
! 440.26(RI)-26 ggp i
l l
l
I p AP600 Open It na Tracki:g Systra Database: Execttiva S:nimary D:t;: 5/8/97 Selectic::: [ item no] between 5247 And 5247 Sorted by item #
Item DSER Section Title / Description Resp (W) NRC No. Branch Type Detail Status Engineer Status Status Leues No. / Date Question 5247 NRR/IlliFB 18 RAI-OI MMIS/Kerch Action W Action W I
{ Respond to NRC " Editorial Comments on the AP600 Iluman Factors Engineering Documentation
- received by letier dased april 2
~
I fe'lecon with Kerch/BongarraEid $d SAR and WCAP changes, agreed upon. Markups were transmitted inthally for management review.
(Comments rec'd and final changes being incoporated. Will provide a letter to NRC transmitting martu I
finto SSAR Rev 13,end-May. rkn 5/8/97 Page: I Total Records: I
t i
d FAX to DINO SCALETTI May 7,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don Lindgren ;
Bob Vijuk Brian McIntyre OPEN ITEM #363 (M10.2-6)
- 358 (M10.2-6)
L #1142 (DSER 10.2.10-1) 1 l To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
l Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 23 calendar days away (18 business days). The relevant documentation related to Open Items #363 (M10.2-uj, #358 (M10.2-6), and
- 1142 (DSER 10.2.10-1) is in Westinghouse letter NSD-NRC-97-5089, dated April 25,1997.
Appropriate pages of this letter are included. This letter provides the information to close Open Item #358 (M10.2-6). Open Items #363 (M10.2-6) and #1142 (DSER 10.2.10-1) have remained
" Action W" pending the closure of Open Item #358 (M10.2-6). Therefore, it is requested NRC review this material and provide definitiv: action for Westinghouse or provide direction to change the status of these items. We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 l
I s
l lv 9 i
AP600 Opca l'em Tracki;g Sy; tem Datb:.se: Exec tiv2 S;mm:ry Date: 5/7/97 Selectioa: litem no] between 363 And 363 Sorted by Type item DSI R Section 1 PJe/ Description Resp NRC (W)
No Branch Question T)pc Detail Status Engineer Status Status
--- - _ . . - - -- - Let_ter N.o._. /. Da_te 363 NRR/SPLil 10 2 MTG OI Winters,3. Action W Action W
~
~
~
M io 2-6 (TURIllNE GENERAlOR DLSIGN llASIS) De responses to Q410.139, Q4 to 143, and Q410.144'were received after'the DSER was l prepared, and are umkr staff review. Open items and questions may be de}cloped as a result of the review of those responses;
_ _ l
~ ~ ~ ~
, Closed (Q410.l39i- The staff found resMse to RAI 41d. I39, TGIRCS'coinpaiability, acceptable per' I2/13N4 meetinglNo'SSAR revision
' required ) 'l Closed (Q4 t o 144)- The s' aft found response to RAI 410144, compliance with URD, acceptable per 12/13/94 meetmg (No SSAR res-ision required)
, Closed (Qt10.143)- The sttif found response to RAI 410143, compliance with Standard Review Plan, partially acceptable per 12/I3S4 meeting l Remaining com se de tailed in meeting items M10.2-1 through M10.2-4 (07TS 358,359,360,361)
NRC - Action W "m zt tele, Ch.10 and Ch 4 are inconsistent on MWth.
l
[Clased - NSSS power, as :hown in SS AR Chapter 10, includes core heat + RCP heat added - RCS heat loss SSAR Chapter 4 cos ers core design Irq general, the core design should only discuss the 1933 Mwt rating SSAR Chapter 10 discusses steam and power conversion and it is appropriate to i juse the NSSS rating of 1940 Mwt. De use of 1933 in Chapter 4 and 1940 in Chapter 10 is consistent, Correcting the table in Chapter 4 would be
! incorrect. His is the basis for the Non-ILOCA analyses for Chapter 15 and for preparing Chapter 4 and in the core design.
N Action W - Pending resolution of OITS# 358.
b s
O Page. I Total Records: I
AP600 Opes It:mi Tracking Systen Datbase: Execrtive Szene:ry Date: 5/7/97 Selection: [ item not between 358 And 358 Sorted by Type item D5ER Section Titic/ Description Resp (W) NRC No. Branth Question Type Detail Status Engineer Status Status
. - - . . -- - - . . -.- - - ...------. ---. -- - htter No. / . Date 358 NRR/SPLB 10 2 mig 4)I Winters. L Action N Action W NSIM4RC-97-5089
^
lM10l2-1 (TURBINE OVERSPEED' TRIP')' The AP600' turbine generator AEs~mA*.ve a Manical'ovcNgud trip desice as desNibed in SRF~
Section 10.2. Paragraph III 2 c.1he applicant should provide the bases for not I r,mg a mechanical overspeed trip device. Specifically, the concern of diversity and common rnoic failure needs to be addressed.
= : = = = :: - . = = == = . 7-:=========..- : = 2. .
z=--------'==-.:
, Closed - SSAR revision (10.2.2.53) provides requested justification for cIcctronic trip devices per 12/13/94 meeting agreement.
i Related item. Closed - Concerns related to turbmc missiles (OITS 2030) were discussed at Westinghouse /NRC Senior Management Meetings.
NRC - Action W - Will send qualitative discussion and demonstrate quantitatively that diversity is equal or
- better when calculation is complete l Action N - Response provided by NSD-NRC-97-5089 of 4/25/97. jaw
?
P l
ce !
L J
l l
\
l r
Page: 1 Total Records: I
%_s .
AP600 Open Item Trackirg System D.2t . base: Execztive Szmaxr7 Dat : 5/7/97 .
Selecties: [ item noj between 1142 And 1142 Sorted by Type '
licm DSER Section Titic/ Description Resp NRC (W) l Na Branch Questum {ype W Status Engineer Status
_ _ {tatus , ,
lxtter gf_ _ _ _ Date iI42 NRR/SPLl! 10 2.10-1 DSER-OI Winters, L Closed Action W f
'(TURBINE OVERSPEED TRIP EXTRACTION NONRETURN C1IOSING TIME, val.VE TESTING INTERVAL, STAIFQUESTIONS)De [
fstaff has not yet determined the acceptability of the design of the turbine-generator. g
~ ~
I ision to address tlw staff condand to refled$'ckin vaiA
~
Ac't ion-W (MIO 2-2)'Westin' g house to provide addiiioddiAussiAin SSAR're f closing time per 12/I3/94 meeting agreement. .
Action-W(M10.2-3)- Westinghouse to provide inforrnaten to address the staff concern and may change the valve inspection interval per 12/I3/94
^
meeting agreement. ,
Action-W(M10.2-4)- Westinghouse to specify a valve test interval in the SSAR and provide justification per 12/I3/94 meeting agreement Closed (Q410. I39) - The staff found response to RAI 410.139. TG / RCS compatability, acceptable per 12/13/94 meeting. (No SSAR revision required )
!' Closed (Q410.144) 'the staff found response to RAI 410.144, E- ,' - e with URD, acceptable per 12/13/94 meeting. (No SSAR revision required.) (
Action-W(Q410.143)-The staff found response to RAI 410.143, c4- with Standard Review Plan, partially acceptable per 12/13/94 meeting. [
Rememing concerns are detailed in meeting items M10.2-1 through M10.2-4 (OITS 358,359,360,361). WCAP 13054 to be reviewed for l consistency. F Closed - See items 359,360,361, and 363 i Actiion W - Pending resolution of OITS # 358 - 36I . . . . . - I
- . . . ~ . .. - - - , - . . - ._.
Y
~
G i I
i i
t i
L Page: 1 Total Records: I
- - . . . - - _ _ . . .-. - ---:-- -_ . - . _ _ , . . _ . -.n --a.a. - - - . . . - -
- - - _ . - - - - - - - _ _ _ - - . - _ . . _ ~ _ _ . - _ . _ _ . - - _ _ - _ - . _ - - - . _ _ _ _ - . - _ _ - . . - . . - . - - _ - _ _ - - _ . - . _ _ _ . . -
!r
. ,O yo ,
Westinghouse . Energy Systems sa 335 l
Electric Corporation Arswin Pennsrea ismess 1
- DCP/NRC0831 a l
Docket No.: SIN 52-003 ;
. April 25,1997
! Document Control Desk '
i U.S. Nuclear Regulatory Commission l Washington, DC 20555 -
ATTENTION: T. R. QUAY
SUBJECT:
TURBINE OVERSPEED TRIP - KEY ISSUE 14 l
Dear Mr. Quay:
The AP600 design utilizes an electronic turbine overspeed trip in lieu of a mechanical turbine i overspeed trip. This is inconsistent with the recommendation found in Criteria til.2.c under Review !
Procedures in Standard Review Plan Section 10.2. This has been identified as Key issue 14 in your l
i letter of December 6,1997.
l
! ' Westinghouse has performed an assessment of the probability of a destructive failure of the turbine 1 l with either the electronic overspeed trip or the mechanical trip. The results of this assessment show that the two designs are equivalent during operation. This assessment did not consider errors during testing. The use of an electronic overspeed trip system in lieu of a mechanical overspeed trip device is in compliance with the requirements in General Design Criteria 4 for protection from missiles. This design is also in conformance with the Acceptance Criteria in Standard Review Plan 10.2.
l Attached is a summary of this issue and a description of the AP600 design. This information should permit the NRC staff to complete the review of the turbine overspeed protection, close Key issue 14, and close the associated open item (OITS #358). The Westinghouse status for OITS #358 will be changed to Ac' tion N.
E Please' contact Donald A. Lindgren at (412) 374-4856 with any questions.
- f. f '
Brian A. McIntyre, Manager Advanced Plant Safety and Licensing
- jmi i
i __ cc: D. Jackson, NRC (w/ Attachment) ._.
hh 1
SN! & __ ._.. _
l s
l* Attachment to NSD-NRC-97 5089 KEY ISSUE 14: TURBINE OVERSPEED TRIP (01-358)
NRC Statements:
"M10.2-1 (TURBINE OVERSPEED TRIP) The AP600 turbine generator does not have a mechanical overspeed trip device as described in SRP Section 10.2, Paragraph III.2.c. The applicant should provide the bases for not having a mechanical overspeed trip device. Specifically, the concern of divers ty and common mode failure needs to be addressed."
Per NRC Letter dated entitled Westinghouse AP600 Turbine Overspeed Protection - Key issue #14 dated April 11,1997, "... Westinghouse has not provided any quantitative engineering analysis or operating data to support the assertions that the AP600 turbine overspeed trip reliability and diversity is equal to or better than the turbine overspeed protection recommended in SRP Section 10.2.
Additionally, Westinghouse has not addressed the concern of common mode failure for twc, electrical trips as stated in the DSER."
Issue:
AP600 design utilizes an electronic overspeed trip system in lieu of the mechanical overspeed trip device, ne SRP recommends a mechanical trip device.
AP600 Position:
, The AP600 design maintains high reliability in overspeed protection during normal operations and l reduces the chance for overspeed events during testing by replacing the mechanical overspeed trip !
device with a new electronic overspeed trip system. The new electronic overspeed trip system is diverse from the overspeed protection control portion of the existing electrohydraulic control system (DEH).
Discussion:
The following discussions provide a technical description of the electrohydraulic control system and the new electronic overspeed trip system. Included in the discussions are the redundancy, diversity and testing requirements of the electronic overspeed trip system design and the results of a Westinghouse probabilistic risk assessment.
Electrohydraulle Control System:
Turbine speed control is provided by the governing action of the electrohydraulic control system. He DEH has two modes of operation to protect the turbine against overspeed. He first mode is the speed control that functions to maintain the desired speed. He second mode is the overspeed protection control (OPC) which operates if the normal speed control should fail or upon a load rejection. There are two solenoid valves in the OPC and they are arranged in parallel and are deenergized closed under 4 normal operating conditions. In the closed position, the solenoid valves block a path to drain of the OPC trip header fluid, and pressure can be established under the interceptor valve and governor valves m 1
(, l t o
e
(
Attachment to NSD-NRC-97-5089 servo-actuators. In the event of an OPC action, such as occurs if the unit reaches 103% of rated speed or, if the generator breaker is opened when internal pressures in the turbine indicate that it is carrying above 30% load, the solenoid valves open releasing the OPC trip header fluid to drain. This causes the immediate closing of the interceptor valves and governor valves. Check valves between the auto stop emergency trip (AST) fluid circuit and the OPC Ouid circuit retain the pressure in the AST line and the throttle valve and reheat stop valve remain open. With a reduction in speed, to rated speed; the solenoid valves close, the interceptor and governor valves reopen, and the governor salves take over control of the turbine and keep the unit at rated speed.
Three speed sensors located at the governor pedestal of the high pressure (HP) turbine monitor turbine speed. The turbine speed is calculated via the Westinghouse Digital Processing Family (WDPF) computer.
This DEH portion of the AP600 speed control and overspeed control design is not unique to AP600 and has been utilized in numerous nuclear plants.
Electronic Overspeed Trip System:
The electronic overspeed trip system replaces the mechanical overspeed protective trip device typically used in nuclear plants.
There are six trip solenoid valves contained in the trip control block mounted on the trip system skid located on the governor pedestal. Four of the solenoid valves are energized from the electronic overspeed trip system also called the emergency trip system (ETS). The remaining two are controlled by the OPC ponion of the DEH controller as described above.
The ETS uses the four trip control block solenoid valves called auto-stop emergency trip (AST).
These solenoid valves (PY49321, PY49322, PY49323 and PY49324) are energized closed during normal steam turbine operation. When the solenoid valves are closed, they block a path to drain of the auto-stop emergency trip header fluid, and pressure can be established under the steam inlet valve actuators. When the solenoid valves are deenergized and open, the header Huid goes to drain and causes the throttle valve, reheat stop valve, governor valve and interceptor valve to close. Through the DEH, the decrease in pressure results in the opening of the turbine drain valves, and t!'e closing of the turbine non return valves.
The four AST solenoid valves are arranged into a series / parallel con 6guration. To provide redundant protection, they are arranged into two redundant circuits with two solenoid-activated valves per circuit.
PY49321 and 23 correspond to circuit I and PY49322 and 24 correspond to circuit 2. It can be seen from Figure I that both circuits must trip before the auto-stop emergency trip header pressure collapses to close the steam turbine inlet valves. Tripping occurs if at least one solenoid valve in each circuit is deenergized. His fail-safe design provides the system with both reliability and testability.
l Reliability is enhanced in that the failure of any one of the solenoid valves in either circuit will not cause a turbine trip or prevent a valid trip. Testability is achieved by the fact that one circuit can be tripped without actually tripping the turbine. Furthermore, the solenoids in each ciauit can be tested individually by locking-out one at a time from the circuit test. Testing each solenoid valve mu 2 3 ) (O
l i
Attachment to NSD-NRC-97-5089 L
individually confirms the tripping function without requiring an actual trip.
i The ETS monitors turbine speed via two speed sensors located at the turbine turning ge calculated via redundant AIRPAX microprocessors. Both the speed sensor location and which calculates speed are therefore diverse from the DEH OPC speed sensor location a The AST header pressure between circuits 1 and 2 is monitored by two pressure switches (PS4 and PS49330B). The DEH senses the change in fluid pressure to confirm that the circuit being tested has indeed tripped, and to prevent testing one circuit when the other circuit is being tested.
The two solenoid OPC valves are not controlled from the ETS described in this section. The controlled directly from the DEH OPC as previously described in this document.
Speed Sensors:
It should be noted that although the speed sensors used in the ETS are redundant, tl'ey are not d from the DEH OPC speed sensors, identical speed sensors were selected based on their in actual service. As noted below, the probability of failure of the overspeed protection system is controlled by the failure rate of the turbine inlet valves and by blockage in the emergency trip fluid line, not failure of the speed sensors.
i 1
1 t
I me 3 iv to
4 Attachment to NSD.NRC.97-5089 a,A 4 'J .A 4 n.:y.'sidvit. T
+o 4 cg a m . . sc sa i 30 . .
e <:s , .a:: 3 cia." w v s av s
-h yase _
u eq:e,
-.- .r t, g . ja. ,
.(v ") ..
m T T
_' D a n-V
,T.. .
E - N-
+
- " f amu ,
! D
.pe '4 3 .*c(4 & r l eda 'AP :(#C[
i ""' ! ~_
5 l
ww '+o 'sn k o '~ _ _ ~_.
, , ...., 4 ,;g, ~F
" o.w 1 i -
,,m a s . . . . .
' p ,, ,$YE O
, ~*O nW, O "
M:Wl I s _ g -
.}-_c ,
Gr '3s0
, , ci : .
i
)
i
- Wac,io
'gp , ,g }
1 Tn.mITI*:X8 l
, i '
l
! , gn i i i +
- u
~ -
1 ! .x co m T Tl i --g .m.
,S H ,n=gy.y. l .=* ~ ~& nnes , e ry.
i 1
_ . H- . , ,
o -." . .-
.[ . . . _ .;_- _._..f... . . . . . 1 , ,
so :=
. . :ar aq q x .pa., -
AN
- C.M..' .5 FIGURE 1. TRIP SYSTEM HYDRAULIC INTERFACE l
! 1874A 4 c;a [b
_ .m . . __ __ __. . . _ _ _ _ . _ . _ . _ _ _ _ _ _ _ _ . _ _ _ __._. ... _ _
Attachment to NSD-NRC 97-5089 Advantages of New Electronic Overspeed Trip System: i Each tripping function can be individually tested on-line from the operator / test panel without
, tripping the turbine by separately testing each circuit of the appropriate trip function. The solenoid j valves may be individually tested.
- e - Protection is provided against spurious trips during testing.
j =
Testing of the emergency trip system trip functions is donewithout disabling the trip protection in L the circuit not in test or the DEH overspeed protection control thereby always maintaining .
overspeed protection for the turbine-generator. .
l
. . Unlike testing of the mechanical overspeed device, testing is performedwithout the need for an l
- operator to manually operate the manual overspeed trip lever located at_the turbine front standard. l This " front standard test" is a cause for concern because not resetting the trips after the test or at the proper time can allow the turbine to trip.
)
Probabilistic Risk Assessment (PRA) Results: .l A recent Westinghouse probability assessment of the new electronic overspeed trip system for the l AP600 shows the conditional probability of destructive overspeed to be unchanged by replacing the l - mechanical trip device with the new electronic overspeed trip system. In both cases the calculated i 4
value is 1.2x10 . Additionally, the mechanical overspeed trip device and the new electronic overspeed I trip system were shown to have failure probabilities on the same order of magnitude (10$. It should l
' be noted that the dominant element in the total probability of destructive overspeed is the failure of the i turbine inlet valves to close on demand and by blockage in the emergency trip fluid line. If the mechanical overspeed trip device or the electronic overspeed trip system were 100% reliable, the conditional probability of destructive overspeed would decrease minimally.
l
Conclusion:
The speed control and overspeed protection function of the DEH combined with the electronic overspeed trip system redundant speed switches, trip circuits, and power supplies provide a level of j redundancy and diversity at least equivalent to the recommendations for turbine overspeed protection found in 111.2 of Standard Revi ew Plan (NUREG-0800) Secti on 10.2. Additionally, the issues and problems with overspeed protection systems identified in NUREG-1275 (Reference 3) have been .
addressed to minimize turbine overspeed events during turbine overspeed protection system testing. 1 As noted in NUREG-1275, testing of overspeed protection systems in operating plan'.s has resulted in j excessive overspeed in several cases. Replacing the mechanical overspeed device with the electronic overspeed trip system reduces the potential of accidental plant trips due to operator error.
mu S
( 0 -) g
f Westinghouse FAX COVER SHEET e
RECIPIENT INFORMATION SENDER INFORMATION DATE: MN [ /hh7 NAME: h_ p,%
TO: p I LOCATION: ENERGY CENTER -
7/LL )1 dbFMN l
I EAST 3 p, g j
PHONE: FACSIMILE: 9,/ - JD l-ilf-2002,. PHONE: Office: (tl12.B79-53;o I
COMPANY: Facsimile: win: 284 4887 l outside: (412)374 4887 LOCATION: W dgg E ,f- %
l l
l I l Cover + Pages 1+ l The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:
l WIN: 284 5125 (Janice) or Outside: (412)374 5125.
COMMENTS:
i chb ~ e /2aa u .
4 l
... . = _ .. - - . .. . . - . - . - - . . - --
1 l
l' l The following questions need resolution before we can complete analysis of the ROSA test transmitted 05/08/97.
- 1. Is the indicated flow the total flow through all 45 tubes?
- 2. Is the indicated pressure absolute?
l 3. I'm assuming that the indicated lengths are tube lengths rather than elevations.
i
- 4. What is the elevation of the water level?
- 5. I'm assuming that the pool temperatures remain constant throughout the test. (?)
- 6. Are tests AP-BO-01 and AP-CL-04 the same test or the same conditions?
- 7. Is the water level essentially constant?
- 8. Where are positions "F", "E", and "C" defined?
l l i
l l
l l
l l
i I
l 1
l l
l I
t i
i
6 k
5/1/97 Ted, The phone calls today went a long way to helping us understand where we are in the review of chapter 9. Chapter 9 was a good choice to start with as there were issues that still need to be addressed by both organizations and it therefore represented a reasonable test of the process. There is nc desbt as to who has the action and what it is for the sections we covered. Your presence and particiption were greatly appreciated!
Attached is a copy of a table that we (on this end) believe can capture the essence of where we (together) are in the review as a management tool. It allows the progress to be tracked on an FSER section basis, while keeping track of who has how many actions. I tried to keep it simple so we do not get lost in the numbers.
! What is on the attachment reflects what happened during today's phone call along with what we (on this end) think we know about chapters 7 and 18. I realize there are open items for chapter 7, but they are related to Westinghouse submitting the ITAAC and not to specific items. As you and I discussed, Chapter 18 is done, pending Westinghouse formally submitting, as an SSAR revision, the SSAR markups. This will be in revision 13 at the end of May.
The purpose of sending this to you is to get any suggestions you, the PMs and Marylee may have.
Collectively, we need some metric to measure status and progress that we can report to Sam and Howard on a regular basis. It will be work to keep whatever tool we decide on up, but I believe it will definitely be worth it at the end of the project.
Again, thanks for your help today, y Talk it over on your end and let me know what you think!! <
e cc:
W? Bob y/ f, Jim
~
Cindy
% P {y Robin Sue l
0613BAM.wPF/May 1.1997 1
AP600 REVIEW STATUS May 1,1997 FSEP Last Total Action W Action N FSER STATUS Section Update OITS items Start Draft DRPM DRPM with w/o 01 01 mummmmmmme mummmmmmummummmmmmmmmammmmmemummmmmuummmmmmmm 7 5/1/97 29 0 11 X X X 9.1 5/1/97 27 0 19 X 9.2.1 5/1/97 8 2 1 X X 9.2.2 5/1/97 6 2 0 X X 9.2.3 5/1/97 1 0 0 9.2.8 5/1/97 2 0 0 X X 9.3.1 5/1/97 15 2 6 9.3.3 5/1/97 4 0 0 l 9.3.6 5/1/97 3 0 0 l
l I
9.5.3 5/1/97 3 0 2 18 5/1/97 109 1 33 X X X Total 5/1/97 207 7 72 6 5 2 0613B AM WPF/May 1,1997 2
i l
l
^o', sit Huf b an kre are sine oF % rev'vd Spec.s +o shoa j how we San da resolve those which apph in mobe 4 onu "end" in 0106E 4. N & all 4 hose 5>ecs are in here huF % prow'cb.s a resolutim s+ca egy.for NRC cacorrenu .
% compleb +kas is peraly % bcs
( t.co 3.o.3 h % mobs s') . Enjoy, biinMyde5 s/G p o-F . D3 1
ESFAS Instrumentation 3.3.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME z 54 573 B. (continued) B.B.1 Be in MODE 3. JKI hours B AND E 60 B.t.2 Be in MODE 4. 9(Ihours 7 s. 2.fREsTaME MAsteLINelhTM 8 AMR$
paus = oracew.z sim j C. All Engineered Safety C.1 Restore 1 Actuation 7l hours c_
Features Actuation Subsystem in the Cabinets (ESFACs) inoperable division .
battery backed logic to OPERABLE status. ,
groups in one division inoperable. @
IL C.2.1 Be in MODE 3. lHThours AND
/8 C.2.2 Be in MODE 4. j# hours
' dP9
- c. 2.1 # sistE I Ac M % I M 55 h 4*L 4MS
'W75SdD27 W4?# " 375 D. One required 0.1 Verify the interlocks I hour 4-interlock inoperable. are in the required i state for existing plant conditions.
E 7 D.2.1 Place any functions J' hours associated with inoperable interlocks in bypass.
-ANG-
-0.2.2 Re:ter: Mt:rk d ts- ICO heurs -
'#C"a"LC stetui.
0R (continued)
APG00 3.3-18 08/96 Amendment 0
ESFAS Instrumentation 3.3.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME 7 s7S D. (continued) D.3.1 Be in MODE 3. 11tf hours L AND l 13 ,
D.3.2 Be in MODE 4. lylhours l
l b SM E. One channel E.1 Place channel in ghours 'O inoperable'. -
trip.
y ,
I (= STS F. One required channa F.1 Place channel in E' hours .E inoperable. bypass. l l
- ANO -
, ~ . .
F.2.1 "e.t c,c; ;t,;7,- i % ^ "" "a r noro n o, e g ",} ' "'
g
- g. it.1 BE N *adCJ, gN OR
~
ANO 12--
F.2./.1 Be in MODE 3. JJ1fhours f.2.Z #EiN4180EM t#Fa6 i And2. AND
- t. 2.3 ttml cMt. fM jg 7 y/g F.2./.2 g in MODE 4. 1,d hours .
W~
F 2 .1, f.2 kU
- f. STS G. All battery backed G.1 Restore 1 Functional I hours C logic groups in one Logic Subsystem in Protection Logic the inoperable Cabinet (PLC) cabinet to OPERABLE inoperable status.
1
~OR
~
/ 2.
G.2.1 Be in MODE 3.
l l AND Sd hours
/8 l G.2.2 Be in MODE 4. pshours 1 Ano l
\ c,2.3 f2,Wsw n (continued) l l h AP600 3.3-19 08/96 Amendment 0 l mn,- oanoxoo.ame
ESFAS instrumentation 3.3.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME
(. srs R. One channel R.1 Place channel in g hours inoperable. A/we.
bypass.
W 5.A.i Red i.ve e slianncl t^- 300 hGUr3 -
OI MBLE etetus.-
OR R . 2./ Initiate action to be 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> in MODE 5 with RCS open and visible level in pressurizer.
S. One required channel S .1 %, Place channel in g ours STS inoperable. bypass. NWE ANO, S.2.1 Dastere ch::::1 ts 400 heucs - -
OPEMBLE stetu;.
e OR-
/ Z_
S.2.X.1 Be in MODE 3 R1i hours AND Jo S.2.(.2 Be in MODE 4 with the 192 hours0.00222 days <br />0.0533 hours <br />3.174603e-4 weeks <br />7.3056e-5 months <br /> RCS cooling provided y the RNS.
i
! 7 S.2.3 ( E s n e r t,s. @ i4 K # N W#
c.uovude 3 *#84854s. (continued)
S 771
- S .
l h AP600 3.3-28 08/96 Amendment 0 wimi-amim l
ESFAS Instrumentation 3.3.2 ACTIONS (continued) l i
l CONDITION REQUIRED ACTION COMPLETION TIME ;
CR&fS T. One channel T.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> STS 3,3, y inoperable.
OPERABLE status. 4g l
_OR I7 T.2.1 Verify atlernate 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> radiation monitors are OPERABLE. -
AND .
T.2.2 Verify control room 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- l isolation and air supply initiation !
' manual controls are operable. .
OR 78 T.3.1 Be in MODE 3 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> l
] y ggSrett CdMM AND p oP5t@i STAM. 89 T.3.2 Be in MODE 4 g hours ,
MA \
T. t . 3 g /08 /AU l U. One manual initiation U.1 Restore manual ghours' M !
device inoperable. initiation device to q O
- OPERABLE status.
0_R L
c.2 "e. ; i., . :.,;;' J: hour.s ,
4aiti: tic: d^"f 0 ^;
e wiv:'ent to the eaat-^1 r:ce devie : m ia the Ree te ,
shutdnwa ugyketst4nn _
c.; ^M"A"J& -do .-
-dedicated : perm,w
. whn 4e 4- -*=+4- ane
??N :" ',#Jf ' ^^ F f k i
^-! Y. : lv";l '.^v^ '. (v%^ T *, *
'fild.y--du -l. Y. ,1 -
~Amule_ae __
(continued) h AP600 3.3-29 08/96 Amendment 0 mi .uun -im
ESFAS Instrumentation 3.3.2 CONDITION REQUIRED ACTION COMPLETION TIME
- U. (continued) OR l St
! U.3.1 Be in MODE 3 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> AND 7L >
ghours l U.3.2 Be in MODE 4 with the l RCS cooling provided l
y y the RNS.
7 M.3.3 RAsTeta M 4 w +4. 9 4 Nond.1 ou m en w ow w.c 7D C ff4d8/,8 @ $.
e I
i 6
0 l
l t
I
, f,P6 { , _ ,,, 3.3-30 08/96 Amendment 0
RCS Leakage Detection Instrumentation 3.4.10 ACTIONS (continued) ;
CONDITION REQUIRED ACTION COMPLETION TIME B. Required containment -------------NOTE-------------
atmosphere LCO 3.0.4 is not applicable. 575 3,%/5 radioactivity monitor -----------------------------
inoperable. 8L !
l 4
B.1.1 Analyze grab samples of Once per containment atmosphere. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR -
B.I.2 Perform SR 3.4.8.1. Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> !
AND 3.0 M V3
, 8.2 Restore containment 159 See :
atmosphere !
radioactivity monitor j to OPERABLE status. ;
l h
1 C. Required Action and C.1 Be in MODE 3. fhours associated Completion Time not met. AND
' 12-C.2 Be in MODE 4. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
.A n - ,
- n n n
~,
AWO
\
C.3 RSs7ME wontedts as Hods Lf4thst niETrermas
/As374a**tFM11M TC OPEGdLE 574ms.
AP600 3.4-16 08/96 Amendment 0 www.oac e -
Containment 3.6.1
. )
3.6 CONTAINMENT SYSTEMS 3.6.1 Containment LC0 3.6.1 Containment shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQijlRED ACTION - COMPLETION TIME A. Containment A.1 Restore containment to I hour inoperable. 0.PERABLE status. J
(*
- 8. Required Action and B.1 Be in MODE 3. ghours associated Completion Time not met. AND
/- 2 B.2 Be in MODE 4. / hours Y
6* 3 R85"4 L C* ***"
t o ofMs08LE $147*45. Y l
l i
l l
AP600' 3.6-1 08/96 Amendment 0 mn.. o.co.oi ao.aon.
~
Containment Air Locks 3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3 ----------NOTE---------
Air lock doors in high radiation areas may be verified locked closed by administrative means.
l l Verify an OPERABLE door Once per is locked closed in each 31 days affected air lock.
C. One or more C.1 Initiate action to Immediately containment air locks evaluate overall inoperable for containment leakage rate reasons other than per LCO 3.6.1 Condition A or B.
AND C.2 Verify a door is closed I hour in the affected air lock.
AND C.3 Restore air lock to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.
D. Required Action and D.1 Be in MODE 3. / hours associated Completion Time not met. AND 2
3 D.2 Be in MODE 4. /phours
$UA Q$ RssTda 4 4 A.oc.x.8)
Tu orsgnelt s,m s . .2fo Wes i
@ AP600 3.6-5 08/96 Amendment 0 m m. u.o2o.or,o os
. 1 Containment Isolation Valves 3.6.3 1 ACTIONS !
CONDITION REQUIRED ACTION COMPLETION TIME l
C. (continued) AND One or more C.2 --------NOTE---------
penetration flow Isolation devices in paths with one high radiation areas may containment isolation be verified by use of valve inoperable. administrative means.
- Verify the affected . Once per
! penetration flow path is 31 days isolated.
D. Requ. ired Act. ion and 0.1 Be in MODE 3. hours i
associated Completion .
Time not met in MODES AND 1, 2, 3, and 4. /M l- D.2 Be in MODE 4. / hours l- dS!2 p.3 RMM Wh'M *M isetAnd JMLUG6) A 1 36 ;)ost.g ofEuttii. Sten'$ _
l l
h AP600 3.6-9 08/96 Amendment 0 uoi s.. si.co.m -n..
l Containment Pressure 3.6.4 l
l 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure l
l LC0 3.6.4 Containment pressure shall be [> -0.2 psig and) 5, +1.0 psig.
APPLICABILITY: MODES 1, 2, 3, and 4.
l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
/
, A. Containment pressure A.1 Restore containment 74 hourg not within limits. sure to within g gg A.I B. Required Action and B.1 Be in MODE 3. / hours associated Completion Time not met. AND '
lL l B.2 Be in MODE 4. 34' hours I deLR_
3.3 RG5WE NWM M * **
ntsssute 7D caowd
>omors. -
SURVEILLANCE REQUIREMENTS l
SURVEILLANCE FREQUENCY !
/ 2. 575 ;
SR 3.6.4.1 Verify containment pressure is within JA' hours 5'4.3,6,Y,/
limits.
Reviewer Note: The low pressure limit is not needed for plant locations for ,
which the lowest possible ambient temperature is approximately 20 *F.
i I
h AP600 3.6-12 08/96 Amendment 0 mni. a.uo.o. ,o m.
Containment Air Temperature 3.6.5 l, 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature !
. LCO 3.6.5 Containment average air temperature shall be i 120*F.
i I
i APPLICABILITY: MODES 1, 2, 3, and 4. 1 l
ACTIONS 1
CONDITION REQUIRED ACTION .
COMPL'ETION TIME 8
A. Containment average A.1 Restore containment. )(hours. I air tempe'rature not average air. temperature ^
i
' within limit. to within limit. .
575 3.6.5 I A.I B. Required Action and B.1 Be in MODE 3. hours !
associated Completion l Time not met. AND l
/ 2.
l 8.2 Be in MODE 4. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> I
a'3 essw a =*s**
ggg Ast 75hlP6RA%M % W R$
l TO WTWsd LIMIT *.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I
l SR 3.6.5.1 Verify containment average air 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> temperature is within limit.
l i.
b AP600 3.6-13 08/96 Amendment 0 m n. u.ox.o.,e.e.ca.
l
-- - _ . . _ _ - - ._ .. _ ~ .
PCS - Operating 3.6.6 ACTIONS (continued)
CONDITION REQVIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. ghours associated Completion l Time of Conditions A, AND 8, or C not met. go l D.2 Be in MODE 4. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR yo srs 3.<,4. c.
p p W E Pc5 W N MU b LC0 not met for ggyg 7, reasons other than A, NM MM
- B, or C.
MW 5) l
- SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify the water storage tank temperature ----NOTE----
a 40 *F and s 120*F. Only required when the ambient temperature is s 32*F or 2 100'F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
. sis s R. 3.5 4. /
7 04r5 i
SR 3.6.6.2 Verify the water stor' age tank volume 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-a 400,000 gallons, g l s A.3.5. e t 3t Om SR 3.6.6.3 Verify each passive containment cooling 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> system, power operated, and automatic valve in each flow path that is not locked, sealed, or otherwise secured in h 3. 5.2.1 position, is in the correct position.
(continued)
AP600 3.6-15 08/96 Amendment 0 w o.amo. m.
pH Adjustment 3.6.9 l 3.6 CONTAINMENT SYS7EstS l 3.6.9 pH Adjustment i
LC0 3.6.9 The pH adjustment shall be OPERABLE.
l l APPLICABILITY: MODES 1, 2, 3, and 4.
1 ACTIONS -
1 CONDITION REQUIRED ACTION COMPLETION TIME i
A. Thh volume of A.1 Restore volume of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i i trisodium phosphate trisodium phosphate to ;
I not within limit. within limit. .
G.
B. Required Action and B.1 Be in MODE 3. Afhours associated Completion Time of Condition A AND not met.
OR B.2 Be in MODE 4. /(2-p hours LCO not met for 4"O l reasons other than A. S. 3 R53784 Erd 4#A5" " " *#8 l rs oPsted22 rrem3-l i
l f
(h)AP600 3.6-22 08/96 Amendment 0 se011:sseesess10030000 t00400300
I j -
Main Steam Safety Valves (MSSVs) '
1 l 3.7.1 l-l i
3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (HSSVs) )
l l
l LC0 3.7.1 The MSSVs shall be OPERABLE as specified in Table 3.7.1-1 and Table 3.7.1-2.
l t
1 1
APPLICABILITY: MODES 1, 2, 3, MODE 4 with RCS not cooled by RNS.
ACTIONS
\
........................................N0TE-------------------------.-...---..
Separate Condition entry is allowed for each MSSV. .
I
............................................................................... l l . CONDITION REQUIRED ACTION COMPLETION TIME j 44 !
A. One or more required A.1 Reduce power to less than )fhours i MSSVs inoperable. or equal to the applicable
% RTP listed in r5 3.7.1 Table 3.7.1-1. A/
1 4, i
- 8. Required Action and B.1 Be in MODE 3. E hours associated Completion Time not met. AND
/2 OR &(hours B.2 $f'in MODE 4.
One or more steam fib /j@
l- generaters with less .3(;, ycwe.$
i t
i .
t
()AP600 3.7 1 08/96 Amendment 0
-ow.- n .ow m , ...
Mein steam isolation values (MSIVs) 3.7.2 l ACTIONS (continued) ,.-
CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Be in MODE 3. ' hours
! associated Completion l Time of Condition C AND or 0 not met.
E.2 Be in MODE 4 with RCS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> cooling provided by the RNS.
Ado y,3 A S$7be t MEWS TD 3 L McAS 6PERMLE. C 7?97us .
l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 l.
.-----NOTE------------------
, Only required to be performed prior to entry into MODE 2.
)
l Verify MSIV closure time 1 5 seconds on In accordance an actual or simulated actuation signal, with the Inservice Testing Program l
l t .
t l
(!)AP600 3.7 6 08/96 Amendment 0
.,ommmm
l' Main Feedwater Isolation and Control Valves (MFlV and MFCV) l 3.7.3 i*
j ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME i
D. Required Action and 0.1 h
l Be in MODE 3. / hours
! associated Completion Time not met. AND l 0.2 Be in MODE 4 with RCS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
! cooling provided by RNS.
DEA.
l d. 3 9 Ear 4RE mFW@ AMO sc yag l mycu%) Tu 09546f42 ,
5747m1.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 -----------------NOTE------------------
Only required to be performed prior to entry into MODE 2.
l Verify the closure time of each MFIV and In accordance l MFCV is 1 5 seconds on an actual or with the simulated actuation signal. Inservice
,. Testing Program l
I l
l AP600 3.7-8 08/96 Amendment 0 asoluesresent e030703 r04400300 l
c_-____-___--_-----___-___ _ -
Main Cont.ol Room Habitability System (VES) 3.7.6 l
- j 3.7 PLANT SYSTEMS 3.7.6 Main Control Room Habitability System (VES)
LCO 3.7.6 Two Main Control Room (MCR) Habitability System trains shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
During movement of irradiated fuel assemblies.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One VES train A.1 Restore VES train to 7 days inoperable. OPERABLE status.
B. MCR air temperat'ure B.1 Restore MCR air 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
. not within limit. temperature to within limit.
C. Loss of integrity of C.1 Restore MCR pressure 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> MCR pressure boundary to OPERABLE boundary. status G
- 0. Required Action and D.1 Be in MODE'3. $ hours associated Completion Time of Conditions A, AND B, or C not met in l 2, MODE 1, 2, 3, or 4. D.2 Be in MODE 4. J4 hours N REsrMM UE5 To
- o. - u , . .u. no E. Required Action and E.1 Suspend CORE Immediately associated Completion ALTERATIONS.
Time of Conditions A, B, or C not met AND during movement of irradiated fuel. E.2 Suspend movement of Immediately irradiated fuel assemblies.
1 1
(continued) j i
(b)AP600 3.7-11 08/96 Amendment 0 o n. a .omo. ,o.mu..
Main Control Room Habitability System (VES) 3.7.6 l ', .
ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME G
F. Two VES trains F.1 Be in MODE 3. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, 2, 3, or 4. AND IL F. Be in MODE 4. M hours T3 Risue oe)LVEs~t.M) 3 (o desut5 r= o FM4 A A # 7d*5 - -
G. Two VES trains G.1 Suspend CORE Immediately inoperable during ALTERATIONS.
i movement of -
l irradiated fuel. AND ,
G.2 S,uspend movement of Immediately irradiated fuel assemblies. .
- l l
i l
l I
l t
h AP600 o . nom ,o.<.e..
3.7-12 08/96 Amendment 0 oio
~.. -. . - . _ _ _ _ _ _
l Startup Feedwater Isolation and Control Valves 3.7.7 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME 6 l l C. Required Action C.1 Be in MODE 3. ghours
( and associated Completion Time AND not met.
C.2 Be in MODE 4 with RCS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1 cooling provided by the RNS. l C., 3 R.EsrM& VhW H5)Tu 2 L >pagS ,
o PG44&E svfras. !
( .
SURVEILLANCE REQUIREMENTS l l l SURVEILLANCE- FREQUENCY
.SR 3.7.7.1 Verify both startup feedwater isolation In accordance 4 with the and control valves are OPERABLE.
Inservice l Testing Program l
t l l
l l
l 1
i l.
h AP600 3.7-16 08/96 Amendment 0 apot weswt 4030707 r06400300
.. . .. . ~.
DC Sources - Operating 3.8.1 l*. 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 DC Sources - Operating LCO 3.8.1 The Division A, B, C, and D Class 1E DC power subsystems l shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1
i '2.
, A. One DC electrical A.1 Restore DC electrical power subsystem -M- hours power subsystem to inoperable. OPERABLE status. 575 4c 0.5.81 D.t *D.2.,
& AG /.9 3 h..RequiredActionand .1 Be in MODE 3. jfhours
\ associated Completion Time not me'.. AND
\ c. 12-g.2 Be in MODE 4. 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br /> 6 rwo Oc. 6LGereces LI RGEMt2 Dc St.EadoN. 2. Ho"tb { },"*
Posk< SatsysTSAs FW sueSYsTem ro A.;
/No!ER$dLE. offAA4LE s%7us .
.ANO C. 3 RES* Tota. PL MW Sb * $ l i= W e/ sg as W s,) 70
~
estEx46LS t*r& m 1- a i
!I h AP600 3.8-1 08/96 Amendment 0 mi-i i --
inverters - Operating 3.8.3 l
l-ACTIONS (continued) .
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and 8.1 Be in MODE 3.
. associated Completion hours Time not met. AND IL B.2 Be in MODE 4. M hours AJA
=
l g3 Ras 7wte_ /a/a4rE4. re 34'g M tg o ?EQdLE T7hus .
SURVEILLANCE REQUIREMENTS SURVEILLANCE .
FREQUENCY 3.8.3.1 . 7
. SR Verify correct inverter voltage, 41- days frequency, and alignment to required AC instrument and control buses. .
575 sRS.S.7.)
r_=----- - -
I i
h AP600 3.8-8 08/96 Amendment 0 mi-noman-m 6
T.
! v a .i iwu c iv o a p 6c.o - vper4 ting l 3.8.5 l
ACTIONS (continued)
CONDITION .r REQUIRED ACTION COMPLETION TIME !.
i
- f. Required Action and .1 Be in MODE 3.
associated Completion I hours Time not met. AND i 12-f Be in MODE 4.
M hours a.3 CE574E ossMoams % goazs ;
p f sussnrom
%T&S . m oveeme \
jf. Two Divisions with 9.1 Enter LC0 3.0.3.
inoperable distribu- Immediately 1 tion subsystems that !
result in a loss of i safety function. i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.5.1 Y 7 Verify corr et breaker and switch -3+ days alignme and voltage to required DC an ass 1 , C instrument and control bus lectrfcal power distribution STS SR. 3.8.9. I subsystems.
I l
r 1
h AP600 3.8-12 08/96 Amendment 0 mi-i mees-=
inverters - Operating 3.8.3 1 l
ACTIOTONS (comtinued) .-
CON (DITION REQUIRED ACTION COMPLETION TIME B. Requiretd Action and B.1 Be in MODE 3. hours associated Completion Time nait met. AND IL j B.2 Be in MODE 4. M hours
. _As!O RssmitL tdMTEL is I
g3 cae4dLE Tinna. 3C Huts 1 SURVtu1LLANCE REQUIREMENTS SURVEILLANCE ,
FREQUENCY 7 I SR 3.8.3.1 Verify correct inverter voltage, -M- days frequency, and alignment to required AC ihstrument and control buses. s75 sRJ.s.7.1
= --
l l
h A AP600 3.8-8 08/96 Amendment 0 mr a - m.
6
. . . . . . _ - . . _ . - - _ - . _ _ ~ . - . . . - _ . - - - . . - . . - . -- - . . - . _ .
- l. viate iuution sy>tems - uperating ,
3.8.5 l ACTIONS (continued) f
, CONDITION REQUIRED ACTION COMPLETION TIME i
. Required Action and .1 Be in MODE 3. I hours associated Completion Time not met, AND f Be in MODE 4. hours
- 2. '5 CEsME. 0#57tti&orW 34 Ars l y y suasnross sr m s. Ta erwenne 5 Two Divisions with 9.1 Enter LC0 3.0.3. Immediately inoperable distribu-tion subsystems that ,
l result in a loss of safety function.
l 1
l 1
l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY V
7 SR 3.8.5.1 Verify corr et breaker and switch -M days alignme and voltage to required DC an ass 1,. C instrument and control bus lecrrfcal power distribution S T 3 S R. 3. 8.9. I subsystems. ' ,
I i
l l
i L
i h AP600 3.8-12 08/96 Amendment 0 Ape 1wtie830005 /064eefte i
'. )
NRC REQUEST FOR ADDITIONAL INFORMATION t
M.. ...
Question 410.17 Revision i Position C.9 of RG 1.45 states that the technical speciGcations should address the availability of various types ofinstruments for RCPB leakage to ensure adequate coverage at all times. Describe how the AP600 design will meet this regulatory position (Section 5.2.5)
Response
SSAR Chapter 16, Technical Speci6 cation 3AQ,3.4.10 de6nes the operability requirements for RCS leakage detection instrumentation. ! 2dd!+!^- This instrumentation, used to identify reactor coolant pressure boundary leakage, is designed so that its operability may be determined at all times. Should a detector fail (signal outside its calibrated range or self>-^ : ered detect trouble de+e:ted), the plant ;
instrumentation system will alarm in the main control room that the specific leak detectiormaaw display l mkwa is questionable. The alarm prompts the operators to observe other sensors providing leak detection information. Technical Speci6 cation 3.4,9,3.4.10 requires instruments ofdiverse monitoring methods to i be operable to provide a high degree ofcosfidence that smallleaks are detected in time to allow actions 1 to place the plant in a safe condition when RCS leakage indicates possible reactor coolant pressure ,
boundarydegradation 2!!c re the ! erb;;e te b : cer;;ed e cer 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />; +'enf^m eper2t ~ Erre suf9cient l t!: e +^ de+e- := lr n2!! '+ re r-e - th e rerter cce!rt g *e rd to t2Ee ce~e +:ve r+!^~ : r
^ de. '; =- e .
The primary methods ofdetecting leaks during stable plant operating conditions are sump level changes l i
andreactor coolant inventory (pressuri:er level) changes. The primary methodsfor detecting leaks during varying system conditions such as makeup and letdown operations andpower transients are sump level changes and N13/F18 detection.
Table 410.17-1 summari:es the validity of the diverse methods of detecting RCPB leakage.
1 SSAR Revision: Revisions to SSAR Section 5.2.5 and Chapter 16 are attached.
410.17-1 3 Westinghouse Rev.1 i
l 1
NRC REOUEST FOR 4 $_TIONAL INFORMATION A
Table 410.17-1 Validity ofRCPB Leak Detection Afethods ali of B Leak & tecti <m .MetM
% RATED A VERAGE TilERhfAL RCS Af0 DES TITLE RE4CTIVITY POlVERl") TEhfrERATURE Sump Level RCS Imentory CONDITION (K,) (*F) (J),(e) NI3/Fl3 (g) 1 Power Operation > 0.99 >5 NA Valid Valid (f) Valid 2 Startup > 0.99 <5 NA Valid Invalid Valid 3 flot Standby < 0.99 NA > 420 Valid Imalid Yo!id 4 < 0.99 NA 420 > T, > 200 Vahd invalid Valid Safe Shutdown (b) 5 Cold Shutdown (b) < o_99 NA < 200 NA NA NA NA NA NA NA NA 6 RefuehngIC) NA (a) Exct uding decay heat.
(b) All reactor vessel head closure boltsfully tensioned (c) One or more reactor vessel head closure bolts less thanfully tensioned ,
i l
410.1F W Westinghouse Rev.1i =
i l
___ ._._..u-..-.i i.:
NRC REQUEST FOR ADDITIONAL INFORMATION (J) Sump level change detection ofleaks not valid Juring extremely cold outside conditions uhenfrostforms on interior ofcontainment vessel.
(<> Lear detection method not valid Juring containment atmosphere purge operatiora or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ofend ofpurge.
(f) Power > 20%
(g) RCS at steady state; i e. no temp. change, no pressure change, no makeup or ferdown W Westinghouse 410.17-3 Rey,1
)
NRC REQUEST FOR ADDITIONAL INFORMATION i
= %.?
{
l l
5.2.5 Detection of Inkage Through Reactor Coolant Pressure Boundary j 1
The reactor coolant pressure boundary (RCPB) leakage detection monitoring provides a means of I detecting and to the extent practical, identifying the source and quantifying the reactor coolant leakage, The detection monitors perform the detection and monitoring function in conformance with the requirements of General Design Criteria 2 and 30 and the recommendations of Regulatory Guide 1.45.
Leakage detection monitoring is also maintained in support of the use of leak-before-break criteria for high-energy pipe in containment. See subsection 3.6.3 for the application of leak-before-break criteria.
Leakage detection monitoring is accomplished using instrumentation and other components of several systems. Diverse measurement methods including level, How, and radioactivity measurements are used for leak detection. The equipment classification for each of the systems and components used for leak detection is genera ly determined by the requirements and functions of the system in which it is located.
There is no requirement that leak detection and monitoring components be safety-related See Figure 5.2-I for the leak detection approach. The descriptions of the instrumentation and components used for leak detection and monitoring include information on the system.
To satisfy position 1 of Regulatory Guide 1.45, reactor coolant pressure boundary leakage is classified as either identificd or unidentified leakage. As depned in SSAR Chapter 16.1, Technical Specipcation Section 1.1, identified leakage includes:
- Leakagec re- '~ed :;:'em: such as pu-'; ;-'-' e c'~ ' ~:e! seal or valve packing seal leaks that are captured and conducted to collection systems or a sump or collecting tank
- Le4 2;r :-'^ 9!! y :, ' : --d : c&y :;:'e : P'- :; :*- ' 47e) < : '" " * ;e :: e*
C^"dd^ ed 'e M p"' e f 'b e ' O gp M ' ! de"'; Urd '"4?ge - 'be hr ^r ^# 'b? '^C b "'C2!
rp "atie- ? ' ? *: ?dd!';^ 2! ' ' ge mur' M cc :!de ed :- 'be n!u?- ^r ,L7.--,~
ree!rt:- -*ey M'r ^)
- RCS leakage through a steam generator to the Secondary System or through the PRHR heat exchanger to the IRWST. -
- Leakage mto the containment atmospherefrom sources that are specipcally located and known not to interfere with the operation ofleakage detection systems or pressure boundary leakage. (Up to 10 gpm ofiden:iped leakage is considered allowable by Technical Specsfication 3.4.8 because leakage isfrom known sources that do not interfere with detection of unidentiped leakage and is well within the capability of the RCS Makeup System.)
Other leakage is unidentified leakage.
i 410.17-4 Rev.1 3 Westinghouse I
. . _ . _~ _. . - . . . - . - _ - _ . -. . - - , --. . .
- l 1
l l
NRC REQUEST FOR ADDITIONAL INFORMATION ME 4E!
= g i
5.2.5.1 Collection and Monitoring of Identified Ler kage i
identified leakage other than intersystem leakage is collected in the reactor coolant drain tank. The I reactor coolant drain tank is a closed tank located in the reactor cavity in the containment. The tank vent j is piped to the gaseous radwaste system to prevent release of radioactive gaa to the containment ;
atmosphere. For positions I and 7 of Regulatory Guide 1.45, the liquid level in the reactor coolant drain j tank and total flow pumped out of the reactor coolant drain tank are used to calculate the identified :
leakage rate. These parameters are available in the main control room. The reactor roolant d/ain tank, I pumps, and sensors are part of the liquid radwaste system. The following sections omH;.c the various sources of identified leakage other than intersystem leakage.
1 5.2.5.1.1 Valve Stem Leskoff Collection Valve stem leakoff connections are not provided in the AP600.
5.2.5.1.2 Reactor Head Seal The reactor vessel flange and head flange are sealed by two concentric seals. Seal leakage is detected by two leak-off connections: one between the inner and outer seal, and one outside the outer seal. These lines are combined in a header before being routed to the reactor coolant drain tank. An isolation valve is installed in the common line. During normal plant operation, the leak-off valves are aligned so that leakage across the inner seal drains to the reactor coolant drain tank.
A surface-mounted resistance temperature detector installed on the bottom of the common reactor vessel 1 seal leak pipe provides an indication and high temperature alarm signal in the main control room I indicating the possibility of a reactor pressure vessel head seal leak. The temperature detector and drain l line downstream of the isolation valve are part of the liquid radwaste system.
The reactor coolant pump closure flange is sealed with a welded canopy seal and does not require leak-off collection provisions.
Leakage from other flanges is discussed in subsection 5.2.5.3, Collection and Monitoring of Unidentified Leakage.
5.2.5.1.3 Pressurizer Safety Relief and Automatic Depressurization Valves Temperature is sensed downstream of each pressurizer safety relief valve and each automatic l depressurization valve mounted on the pressurizer by a resistance temperature detector on the discharge l pipingjust downstream of each valve. High temperature indications (alarms in the main control room) l l identify a reduction of coolant inventory as a result of seat leakage through one of the valves. These I detectors are part of the reactor coolant system. This leakage is drained to the reactor coolant drain tank l l
\ l a
410.17-5 3 Westinghouse Rev.1 l
i i
l i
t .
NRC REQUEST FOR ADDITIONAL INFORMATION
=E 1
l during normal plant operation and vented to containment atmosphere or the in-containment refueling water storage tank during accident conditions. This identined leakage is measured by the change in level of the reactor coolant drain tank.
5.2.5.I.4 Reactor Coolant Pump Drain Leakage from the reactor coolant pump drain is directed to the reactor coolant drain tank. This identified leakage is measured by the change in level in the reactor coolant drain tank.
5.2.5.1.5 Other Leakage Sources In the course of plant operation, various minor leaks of the reactor coolant pressure boundary may be detected by operating personnel. If these leaks can be subsequently observed, quantified, and routed to the containment sump, this leakage will be considered identined leakage.
5.2.5.2 Intersystem Leakage Detection Substantial intersystem leakage from the reactor coolant pressure boundary to other systems is not expected. However, possible leakage points across passive barriers or valves and their detection methods are considered. In accordance with position 4 of Regulatory Guide 1.45, auxiliary systems connected to the reactor coolant pressure boundary incorpo ate design and administrative provisions that limit leakage.
Leakage is detected by increasing auxiliary system level, temperature, Cow, or pressure, by lifting the relief valves or increasing the values of monitored radiation in the auxiliary system.
The normal residual heat removal system and the chemical and volume control system, which are connected to the reactor coolant system, have potential for leakage past closed valves. For additional information on the control of reactor coolant leakage into these systems, see subsections 5.4.7 and 9.3.6 and the intersystem LOCA discussion in subsection 1.9.5.1.
5.2.5.2.1 Steam Generator Tubes An important potential identified leakage path for reactor coolant is through the steam generator tubes into the secondary side of the steam generator. Identined leakage from the steam generator primary side is detected by one, or a corsoination, of the following:
High condenser air removal discharge radioactivity, as monitored and alarmed by the turbine island vent discharge radiation monitor Steam generator secondary side radioactivity, as monitored and alarmed by the steam generator blowdown radiation monitor Secondary side radioactivity, as monitored and alarmed by the main steam line radiation monitors 410.17-6 Rev.1 T Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
!? 9 Radioactivity, boric acid, or conductivity in condensate as indicated by laboratory analysis Details on the radiation monitors are provided in Section 11.5, Radiation Monitoring.
5.2.5.2.2 Component Cooling Water System Leakage from the reactor coolant system to the component cooling water system is detected by the component cooling water system radiation monitor, by increasing surge tank level, by high flow downstream of selected components, or by some combination of the preceding. Refer to Section 11.5, Radiation Monitoring, and subsection 9.2.2, Component Cooling Water System.
5.2.5.3 Collection and Monitoring of Unidentified Leakage To detect unidentified leakage inside containment, in accordance with position 3 of Regulatory Guide 1.45, the following diverse methods may be utilized to quantify and assist in locating the leakage:
Containment Sump Level i
Reactor Coolant System Inventory Balance l Containment Atmosphere Radiation Other methods that can be employed to supplement the above methods include:
Containment Atmosphere Pressure, Temperature, and Humidity ,
Visual Inspection The reactor coolant system is an all-welded system, except for the connections on the pressurizer safety l valves, reactor vessel head, pressurizer and steam generator manways, and reactor vessel head vent, which l are flanged. During normal operation, variations in airborne radioactivity, containment pressure, temperature, or specific humidity above the normal level signify a possible increase in unidentified leakage rates and alert the plant operators that corrective action may be required. Similarly, increases in containment sump level signify an increase in unidentified leakage. The following sections outline the methods used to collect and monitor unidentified leakage.
5.2.5.3.1 Containment Sump Level Monitor Leakage from the reactor coolant pressure boundary and other components not otherwise identified inside the containment will condense and flow by gravity via the floor drains and other drains to the containment sump.
A leak in the primary system would result in reactor coolant flowing into the containment sump. Leakage is indicated by an increase in the sump level. The containment sump level is monitored by two seismic Category I level sensors in accordance with position 6 of Regulatory Guide 1.45. The level sensors are l
410.17-7, T Westinghouse gey,1 r
(
NRC REQUEST FOR ADDITIONAL INFORMATION powered from a safety-related Class 1E electrical source. Rese sensors remain functional when subjected to a safe shutdown earthquake in conformance with the guidance in Regulatory Guide 1.45. The containment sump level and sump total Dow sensors located on the discharge of the sump pump are part of the liquid radwaste system.
Failure of one of the level sensors will still allow the calculation of a 0.5 gpm in-leakage rate within I hour. The data display and proccssing system (DDS) computes the leakage rate and the plant control system (PLS) provides an alarm in the main control room if the average change in leak rate for any given i
measurement period exceeds 0.5 gpm for unidentified leakage. The minimum detectable leak is 0.03 gpm. l Unidentified leakage is the total leakage minus the identified leakage. The leakage rate algorithm subtracts the identified leakage directed to the sump.
To satisy positions 2 and 5 of Regulatory Guide 1.45, the measurement interval must be long enough to permit the measurement loop to adequately detect the increase in level that would correspond to 0.5 gpm ,
leak rate, and yet short enough to ensure that such a leak rate is detected within an hour. The '
measurement interval is less than or equal to I hour.
When the sump level increases to the high level setpoint, one of the sump pumps automatically starts to pump the accumulated liquid to the waste holdup tanks in the liquid radwaste system. The sump discharge flow is integrated and available for display in the control room in accordance with position 7 of Regulatory Guide 1.45.
Procedures to identify the leakage source upon a change in the unidentified leakage rate into the sump include the following:
Check for changes in containment atmosphere radiation monitor indications, Check for changes in containment humidity, pressure, and temperature, Check makeup rate to the reactor coolant system for abnormal increases, Check for changes in water levels and other parameters in systems which could leak water into the ,
containment, and '
1
- Review records for maintenance operations which may have discharged water into the containment.
l l
l 410.17-8 3 Westinghouse Rev.1_1h) l l
l l
t
l NRC REQUEST FOR ADDITIONAL INFORMATION i l
. = \
F d
5.2.5.3.2 Reactor Coolant System Inventory Balance Reactor coolant system inventory monitoring provides an indication of system leakage. Net level change in the pressurizer is indicative of system leakage. Monitoring net makeup from the chemical and volume control system and net collected leakage provides an important method of obtaining information to establish a water inventory balance. An abnormal increase in makeup water requirements or a significant change in the water inventory balance can indicate increased system leakage.
The reactor coolant system inventory balance is a quantitative inventory or mass balance calculation. This approach allows determination of both the type and magnitude of leakage. Steady-state operation is required to perform a proper inventory balance calculation. Steady-state is defined as stable reactor coolant system pressure, temperature, power level, pressurizer level, and reactor coolant drain tank and in-containment refueling water storage tank levels. The reactor coolant inventory balance is done on a periodic basis and when other indication and detection methods indicate a change in the leak rate.
The mass balance involves isolating the reactor coolant system to the extent possible and observing the change in inventory which occurs over a known time period. This involves isolating the systems connected to the reactor coolant system. System inventory is determined by observing the level in the pressurizer. Compensation is provided for changes in plant conditions which affect water density. The change in the inventory determines the total reactor coolant system leak rate. Identified leakages are monitored (using the reactor coolant drain tank) to calculate a leakage rate and by monitoring the intersystem leakage. The unidentified leakage rate is then calculated by subtracting the identified leakage rate from the total reactor coolant system leakage rate. The minimum detectable leak is 0.1 gpm.
Since the pressurizer inventory is controlled during normal plant operation through the level control system, the level in the pressurizer will be reasonably constant even ifleakage exists. The mass contained in the pressurizer may fluctuate sufficiently, however, to have a significant effect on the calculated leak rate. The pressurizer mass calculation includes both the steam and water mass contributions.
Changes in the reactor coolant system mass inventory are a result of changes in liquid density. Liquid density is a strong function of temperature and a lesser function of pressure. A range of temperatures exats throughout the reactor coolant systenr all of which may vary over time. A simplified, but acaptably accurate, model for determining mass changes is to assume all of the reactor coolant system is at % .
t The inver. tory balance calculation is done by the data display and processing system with additional input l
from unsors in the protection and safety monitoring system, chemical and volume control system, and
! liquid radwaste system. The use of components and sensors in systems required for plant operation provides conformance with the regulatory guidance of position 6 in Regulatory Guide 1.45 that leak detection should be provided following seismic events that do not require plant shutdown. 1 1
410.17-9 T Westingflouse gey,1 l
l l
l l
NRC REQUEST FOR ADDITIONAL INFORMATION s =7 ,
l 1
I 5.2.5.3.3 Containment Atmosphere Radioactivity Monitor I Leakage from the reactor coolant pressure boundary will result in an increase in the radioactivity levels l inside containment. The containment atmosphere is continuously monitored for airborne gaseous l radioactivity. Air flow through the monitor is provided by the suction created by a vacuum pump.
l Gaseous and Nn/Fu concentration monitors indicate radiation concentrations in the containment atmosphere.
% gr ^'- e' r- - epend r2pid!y !? ~'e ee!rt prenure beund "y 'e"?g- No and F,, are is a j
neutron activation products which 46 are proportional to power levels. ^ dd!!:^- "", N u S r 2 re!2 e e'y l 92+h2!fnee 73 eg-eeque_,iy :n .mg equ :Hur a r2pid!y. An increase in activity inside containment l would therefore indicate a leakage from the reactor coolant pressure boundary. Based on the concentration of Nn/Fu and the power level, reactor coolant pressure boundary leakage can be estimated.
h.Nu&re "e n; g :'-~ hr 2 h!;h r '" "- e- e erter : epr't!n; 2! 2 pe" e- nge h!;'m-
'hr 20 p-~-* The N13 and F18 monitor is seismic Category I. Conformance with the position 6 guidancekf Regulatory Guide 1.45 that leak detection should be provided following seismic events that do not require plant shutdown is provided by the seismic Category I classification. Safety-related Class j IE power is not required since loss of power to the radiation monitor is not consistent with continuing '
operation following an earthquake. Ocm 20 p ee-! pe" e 'e' ^' e e he" ,2 !e 'er - 0.5 ;p-er be d-'ed Op-'+: ; exp : y hr !-d!~-3 t'e ner2;e ^ng i te ' 4?g-(r gy _p!!n; t c" " - +ed 'e d err: 33 un:3_,m d e ! 4'ge +^ 6e e^atamme^,1 c.g_ ge .m7-,m. egg!rt ( ,m_ rge e'" r- 0 ' 'nd 02 gpr "' e "a e :-trat!^- ' " ~e-~- by 2t 'es-t 25 p r-t set e r e":' n; 0.i ;p ' 4 ;e Erkgreed rd 2!-^e
. mn ',np e-*re r - :n;02 ;p ' "?;e 9e : e r:^: re
!g;., .ne :m :,:. . :mg r ,n e ug., g.,:,g. ,. ,-, u n :,w Radioactivity concentration indication and alarms for loss of sample flow, high radiation, and loss of indication are provided. Sample collection connections permit sample collection for laboratory analysis.
The radiation moniter can be calibrated during power operation.
5.2.5.3.4 Containment Pressure, Temperature and Humidity Monitors Reactor coolant pressare boundary leakage increases containment pressure, temperature, and humidity, values available to the orcrator through the plant control system. Atfullpower the minimum detectable leak is 0.1 gpm when tne radionuclide concentration in containment reaches equilibrium. The Nn and F,, monitor can detect a 0.5 gpm leak when the plant is above 20% power and the concentration of radiogas in containment is at equilibrium.
An increase in containment pressure is an indication of increased leakage or a high energy line break.
Containment pressure is monitored by redundant Class IE pressure transmitters. For additional discussion see subsection 6.2.2, Passive Containment Cooling System.
l l
410.17-10 Rev.1 W Westinghouse i
l
NRC REQUEST FOR ADDITIONAL INFORMATION The containment average temperature is monitored using temperature instrumentation at the inlet to the containment fan cooler as an indication of increased leakage or a high energy line break. This instrumentation as well as temperature instruments within specific areas including steam generator areas, pressurizer area, and containment compartments are part of the containment recirculation cooling system.
An increase in the containment average temperature combined with an increase in containment pressure indicate increased leakage or a high energy line break. The individual compartment area temperatures can assist in identifying the location of the leak.
Contairu..cra hunidity is monitored using temperature-compensated humidity detectors which determine the water-vapor content of the containment atmosphere. An increase in the containment atmosphere humidity indicates release of water vapor within the containment. The containment humidity monitors are part of the containment leak rate test system.
The humidity monitors supplement the containment sump level monitors and are most sensitive under conditions when there is no condensation. A rapid increase of humidity over the ambient value by more than 10 percent is indication of a probable leak.
Containment pressure, temperature and humidity can assist in identifying and locating a leak. They are not relied on to quantify a leak.
l l
l 3 W6stinghouse 410.17-11 Rev.1 ),
l l
l l
l NRC REQUEST FOR ADDITIONAL INFORMATION
_ = = -n l
B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.8 RCS Operational LEAKAGE BASES
= = = = = - = = = . = - = = _ = = _ = = - = .=== _
BACKGROUND Components that contain or transport the coolant to or from the reactor core comprise the RCS.
Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.
During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specines the types and amounts of LEAKAGE.
10 CFR 50, Appendix A, GDC 30 (Ref.1), requires means for detecting and, to the extend practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.
The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.
A limited amount of LEAKAGE inside containment is expected from auxiliary systems that cannot be made 100% leaktight. LEAKAGE from these systems should be detected, located, and isolated )
from the containment atmosphere, if possible, to not interfere with RCS LEAKAGE detection.
This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).
410.17 12
, Rev.1 W Westinghouse i
i t
l NRC REQUEST FOR ADDITIONAL INFORMATION l
APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses SAFETY ANALYSES do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA. The amount of LEAKAGE can affect the probability of such an event.
The safety analysis for an event resulting in steam discharge to the atmosphere assumes a 1000 gpd primary to secondary LEAKAGE as the initial condition.
Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leak contaminates the secondary fluid.
The SSAR (Ref. 3) analyses for the accidents involving secondary side releases assume 500 gpd primary to secondary LEAKAGE in each generator as an initial condition. The dose consequences I resulting from the accidents are reported in Reference 3.
l The RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement.
LCO RCS operation LEAKAGE shall be limited to:
- a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.
LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets are not pressure boundary LEAKAGE.
- b. Unidentified LEAKAGE 0.5 gpm of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air N13/Fl8 radioactivity monitoring and containment sump level monitoring equipment, can detect within a reasonable time period. This leak rate supports leak before break (LBB) criteria. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.
1 1
410.17-13 3 Westinghouse g,y, q l
1 .
j l
l NRC REOUEST FOR ADDITIONAL INFORMATION l
li= -- i
& Fq l
BASES l
_ _ _ . _ _ _ . _ . _ _ . _ _ _ l l
l LCO c. Identified LEAKAGE 1 (continued) l Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and I is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE. Violation of this LCO could result in continued degradation of a component or system.
- d. Primary to Secondary LEAKAGE through Both Steam Generators (SGs) )
Total primary to secondary LEAKAGE through both SGs amounting to 1000 gpd produces acceptable offsite doses in the Steam Line Break (SLB) accident analysis.
Violation of this LCO could exceed the offsite dose limits for this accident. Primary to )
secondary LEAKAGE must be included in the total allowable limit for identified 1 LEAKAGE.
- e. Primary-to-Secondary LEAKAGE through One SG The 500 gpd limit from one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture. I
- f. Primary to IRWST LEAKAGE through the PRHR Heat Exchanger HX The 500 gpd limit from the PRHR HX is based on the assumption that a single crack leaving this amount would not lead to a P?J!R tube rupture under the stress condition of an RCS pressure increase evect. If leaked through many cracks, the cracks are very small, and the above assumpt on is conservative. This is conservative because the thickness of the PRHR HX tubs is approximately 60% greater than the thickness of the SG tubes. Furthermore, a PRHR HX tube rupture would result in an isolable leak and would not lead to a direct release of radioactivity to the atmosphere.
l 410.17 14 Rev.1 3 Westilighouse l
NRC REQUEST FOR ADDITIONAL INFORMATION g g 1 .eo ;
BASES (continued)
APPLICABILITY In MODES 1,2,3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.
In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant l pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.
1 l
l ACTIONS A.1
\ \
Unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE in excess of the LCO limits must be reduced to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This Completion Time is based on risk considerations and allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.
B.1 and B.2 If any pressure boundary LEAKAGE exists, or if unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE cannot be reduced to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the reactor l must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that leakage past seals and gaskets is not oressure boundary LEAKAGE. The reactor must be brought to MODE 3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and to MODE 5 wiriin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors which tend to e' ,rade the pressure boundary.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without ACTIONS challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.
410.17-15 3 Westingtiouse Rev.1 l
l l-
1 NRC REQUEST FOR ADDITIONAL INFORMATION i l
+
SURVEILLANCE SR 3.4.8.1 I REQUIREMENTS l Verifying RCS LEAKAGE within the LCO limits ensures the integrity of the RCPB is maintained.
Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.
l Unidentified LEAKAGE and identified LEAKAGE are determined by performance of a RCS water j inventory balance, Primary to secondary LEAKAGE is also measured by performance of an RCS !
water inventory balance in conjunction with effluent monitoring within the secondary steam and I feedwater systems.
The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure. Therefore, this SR is not required to be performed in MODES 3 and 4 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation near operating pressure have been established.
l Steady state operation is required to perform a proper inventory balance; calculations during l maneuvering are not useful and a Nott requires the Surveillance to be met when steady state is ,
established. For RCS operational LEAKAGE determination by inventory balance, steady state is l
defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, and l
with no makeup and or letdown.
An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere N13/F18 radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These LEAKAGE detection systems are specified in LCO 3.4.10 "RCS l LEAKAGE Detection Instrumentation." The containment atmosphere Nf3/Fl8 radioactivity l leakage measurement is valid onlyfor plant power > 20% in Af0DE 1. l l
The containment sump level change method ofdetecting leaks during Af0 DES I, 2, 3 and 4 is not ;
valid while containment purge occurs or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end of containment purge.
The containment atmosphere Nf 3/F18 radioactivity leakage measurement during Af0DE 1 is not l
valid while containment purge occurs or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end of containment purge.
The containment sump level change method ofdetecting leaks during Af0 DES 1, 2, 3 and .I is not valid during extremely cold outside ambient conditions whenfrost isforming on the interior ofthe l containment vessel.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance i of early leakage detection in the prevention of accidents. A Note under the Frequency column l states that this SR is required to be performed during steady state operation.
410.17-16 Rev.1 W Westinghouse
, N s
i
l*
, NRC REQUEST FOR ADDITIONAL INFORMATION l
gr =g
= =1 l
^
\
l This SR provides the means necessary to determine SG OPERABILITY in an operational MODE. I The requirement to demonstrate SG tube integrity in accordance with the Steam Generator Tube Surveillance Program emphasizes the importance of SG tube integrity, even though this ]
, Surveillance cannot be performed at normal operating conditions. )
l i
l
)
1 l
i l
l l
l t
j 410.17-17 W
Westinghouse Rev.1 l
i l
i l
l NRC REQUEST FOR ADDITIONAL INFORMATION y "hi j 1 , . ..
l l
B 3.4 REACTOR COOLANT SYSTEM (RCS) i B 3.4.10 RCS Leakage Detection Instrumentation BASES
=.== - .
. _ = _ , = = = _ _ = =.,=. .
BACKGROUND GDC 30 of Appendix A to 10CFR50 (Ref.1) requires means for detecting, and, to the extent practical, identifying the source of RCS LEAKAGE.
Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting LEAKAGE detection systems.
LEAKAGE detection systems must have the capability to detect significant reactor !
coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure. Thus, an early indication or warning signal is necessary to permit proper evaluation of all unidentified l LEAKAGE. '
l Industry practice has shown that water flow changes of 0.5 gpm can be readily l detected in contained volumes by monitoring changes in water level, in flow rate, or I in the operating frequency of a pump. The containment sump used to collect I unidentified LEAKAGE, is instrumented to alarm for increases of 0.5 gpm in the normal flow rates. This sensitivity is acceptable for detecting increases in unidentified LEAKAGE. i l
The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation. Reactor coolant radioactivity F/8 has a longer half hfe and is the dominant source usedfor leak detection. 'e' e!: ;"
be !ce dring : ti !2! er+c : temp --der e re" ": $e e+- "-+" r+" "ed cc--^::^ p edr+: here '-- fe ed 1-d 9:en p c A epper r e r~ ' e' - - +
,i,ag:q ,m.
, :-g 2. g,aa:_; ger -m The production of N13 and F/8 is proportional to the reactor power level. N13 has a short half life and comes to equilibrium quickly. F18 has a longer hfe and is the dominant source usedfor leak detection. Instrummt sensitivities for gaseous monitoring are practical for these i EAKAGE detection systems. The Radiation Monitoring System includes monitoring N13/F18 gaseous activities to provide leak detection. l 410.17-18 Rev.1 Westinghouse
i NRC REQUEST FOR ADDITIONAL INFORMATION l l .._ ...-
, APPLICABLE The need to evaluate the severity of an alarm or an SAFETY ANALYSES l indication is important to the operators, and the ability to compare and verify ,
I with indications from other systems is necessary. The system response times and l i
sensitivities are described in the SSAR (Ref. 2).
The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leak occur.
RCS LEAKAGE detection instrumentation satisfies Criterion 1 of the NRC Policy l Statement.
l LCO One method of protecting against large RCS LEAKAGE derives from the ability of instruments to rapidly detect extremely small leaks. This LCO requires instruments (
l of diverse monitoring principles to be OPERABLE to provide a high degree of I confidence that small leaks are detected in time to allow actions to place the plant in i a safe condition, when RCS LEAKAGE indicates possible RCPB degradation. !
The LCO is satisfied when monitors of diverse measurement means are available. l
! Thus, the containment sump level monitor, in combination with an N13/F18 gaseous activity monitor provides an acceptable minimum. Containment sump level monitoring j is performed by two redundant. seismically qualified level instruments.
l l
APPLICABILITY Bec4use of elevated RCS temperature and pressure in MODES 1,2,3, and 4, RCS LEAKAGE detection instrumentation is required to be OPERABLE.
l In MODE 5 or 6, the temperature is < 200 F and pressure is maintained low or at atmospheric pressure. Since the temperatures and pressures are lower than those for MODES 1,2,3, and 4, the likelihood of leakage and crack propagation are much l smaller. Therefore, the requirements of this LCO are not applicable in MODES S l
and 6.
i 410.17-19 3 3 Westinghouse Rey,1 ,
I e
l NRC REQUEST FOR ADDITIONAL INFCRMATION l I
== u= 1
- E 1 Sump level monitoring is a valid methodfor Af0 DES 1, 2, 3 and 4. Containment l l
atmosphere Nf 3/F18 monitoring is valid only in Af0DE 1 with reactor power > 20%
Reactor coolant system inventory monitoring via the pressuri:er level changes is valid in Af0 DES 1, 2, 3 and 4 only when the RCS conditions are stable; le. temperature constant, pressure constant, no makeup and no letdown.
The containment sump level change method ofdetecting leaks during A10 DES 1, 2, 3 and 4 is not valid while containment purge occurs or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end of containment purge.
The containment atmosphere Nf3/Fl8 radioactivity leakage measurement during l Af0DE I is not valid while containment purge occurs or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end l of containment purge. j The contamment sump level change method ofdetecting leaks during Af0 DES 1, 2, 3 and 4 is not valid during extremely cold outside ambient conditions when frost is ;
forming on the interior of the containment vessel.
l l
ACTIONS A.1 and A.2 1
i With the required containment sump level channel inoperable, no other form of sampling can provide the equivalent information; however, the containment atmosphere Nf3/F18 radioactivity monitor will provide indications of changes in LEAKAGE.
Together with the atmosphere monitor, the periodic surveillance for RCS inventory balance, SR 3.4.8.I, must be performed at an increased frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to provide information that is adequate to detect LEAKAGE.
Restoration of the sump channel to OPERABLE status is required to regain the function in a Completion Time of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> after the monitor's failure. This time is acceptable, considering the frequency and adequacy of the RCS inventory balance required by Action A.I.
Required Action A.1 is modified by a Note that indicates that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the containment sump channel is inoperable. This allowance is provided because other instrumentation is available to monitor RCS Leakage.
B.1.1, B. I.2, and B.2 With one gaseous N13/F18 containment atmosphere radioactivity-monitoring instrumentation channel inoperable, alternative action is required. Either grab samples of the containment atmosphere must be taken and analyzed or ECS inventory balanced, 410.17-20 Rev.1 W Westinghouse l
l
t j NRC REQUEST FOR ADDITIONAL INFORMATION l 2 =
in accordance with SR 3.4.8.1, must be performed to provide alternate periodic information.
l With a sample obtained and analyzed or an RCS inventory balance performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor may be an RCS operated for up to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> to allow restoration of the radioactivity monitor.
The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval for grab samples or RCS inventory balance provides periodic information that is adequate to detect LEAKAGE. The l'38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> Completion Time recognizes at least one other form of leak detection is available.
ACTIONS B.I.1, B.I.2, and B.2 (continued)
Required Action B.1 and Required Action B.2 are modified by a Note that indicates that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the gaseous N13/F18 containment atmosphere radioactivity monitor channel is inoperable. This allowance is provided because other instrumentation is available to monitor for RCS LEAKAGE.
C.I and C.2 If a Required Action of Condition A or B cannot be met within the required Completion Time, the reactor must be brought to MODE 4 where the probability and }
consequences of an event are minimized. To achieve this status, the plant must be I brought to at least MODE 3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to MODE 4 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner without challenging plant systems.
SURVEILLANCE SR 3.4.10.1 REQUIREMENTS l SR 3.4.10.1 requires the performance of a CHANNEL CHECK of the containment atmosphere N13/F/8 radioactivity monitor. The check gives reasonable confidence that the channel is operating properly. The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on instrument reliability and risk and is reasonable for detecting otT normal conditions.
SR 3.4.10.2 SR 3.4.10.2 requires the performance of a CHANNEL OPERATIONAL TEST (COT) on the atmosphere N/3/F18 radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. The test verifies the alarm setpoint 410.17-21 3 Westinghouse gey,1 1 l
I i
t i
l NRC REQUEST FOR ADDITIONAL INFORMATION g;= =g l
l and relative accuracy of the inst:ument string. The Frequency of 92 days considers risks and instrument reliability, and operating experience has shown that it is proper for detecting degradation.
SURVEILLANCE SR 3.4.10.3 and SR 3.4.10.4 REQUIREMENTS (continued) These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS Leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 24 months is a typical refueling cycle and considers channel reliability.
Again, operating experience has proven that this Frequency is acceptable.
REFERENCES 1. 10 CFR 50, Appendix A, Section IV, GDC 30,
- 2. Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary LEAKAGE Detection Systems," U.S. Nuclear Regulatory Commission.
- 3. AP600 SSAR Chapter 15 " Accident Analysis."
i l
l 410.17-22 l Rev.1 3 Westlaghouse l
AP600 Opca It m Trackirg Syst m D:t: base: Execrtive Srmm ry Dat:: 5/7/97 Selectier: [nrc st code]<>' Resolved
- And [NRC 13 ranch) like 'NRR/li!Cir Sorted by DSER Section item DSER Section Titic/ Description Resp NRC (W)
No Dranch Question 13pc Detail Status Engineer Status Status letter No / Date 2024 NRR/IllCB 16 DSER4150 Deutsch Ikopped Dropped
~ '~ ~ ~
j28I Design of the Diverse Actuation System l The DAS has t<en identified by Westinghouse to be a RTNSS-important system for ATWS considerations. The statTneeds additionel
- infortnatim regardjng the design and reliability (Sce DSER Open item 7.7.2-1)
, Closed - SSAR Chapter 7 revised to address.
Per i1/21 telecon. NRC has action to discuss lack of DAS/ Tech Spec relationship within the statT rLn 12/2.
- Forwarded report copy of this item to NRC for confirmation that NRC status should be Action-N to determine if this item is
[ Dropped - Dropped per telephone directive from Diane Jackson on 4/23/97.
2434 NRR/IllCB 16 I MTG-OI TECilSPEC Closed Action W
- ~
[ Provide Westinghouse position on admin controls section requirements for a softw are failure root cause analysis program _
j
- ~~ ' ~~ ~~
1Clo'ed With' issuance of the 1cch Specs in SSAR Rev.9 ~-~ ~
s
_ __]
2436 NRR/IllCB 16 I MTG-OI TECilSPEC/Birsa Closed Action W
[Esaluate the actior s for one' channel inoperable versus si gle channels ofinu{iott functions inoperable due to reactor trip group failure
~ ~
~ ~ - ~ ~ " ~ ~ ~ ' - ' ~ ~ ~ ~
~ [ ~ 'l
[ Closed 515iwunce of the TectiSMin 55'AR Re' v
~~ 5. 'j 2437 NRR/IllCB 16 i MTG-OI TECilSPEC/Birsa Closed Adion W npuddeWures
~
_that((cornplete tjmes conside{ut e onktunes to repa{ _ ._ __ _]
~~)
- ~ ~ ~~ ~~~ ~
[losedditlUssuance of the Tech Spain'55IR Rev. 9.
2438 NRR/IllCB 16 I MTGot TECilSPEC/Birsa Closed Action W Consider providing (in Bases) rationale for automatic system actions versus operator actions ('which may reverse the system automatic actims)(e g.,
hLe when a channel trips in the auto function only to be bypass as inop by the operator based upon action steps)
~ ~
Closed - With issuance of the Tech Specs in SSAR Rev. 9 2439 NRR/IllCB 16 i M TG-OI TEC11 SPEC /Birsa Closed Action W
~ ~ - -
Definition of channel is from the sensor through the output ot the reactor trip subgroup Need to include where applicable (1.0 definitions,3 3 hasesj etc ) -, ,.
j
~ '
Closed yWith issuance of the Tech Specs in 5SAR Rev 9. l 2440 NRR/IllCB 16 i MTG-OI TECllSPEC/Birsa Closed Action W
~ ulde in t eh A 6P 00 specs.
~
- Verify that the NUREG definitions are appropratc and inc ]
[ Closed With issuance of the Tech Specs in SSAR Rev. 9_. _
}
244l NRR/IllCB 16 i M TGot TECilSPEC/Birsa Closed Action W ine term " automatic trip logic" is utilised within the 1S without definition Determine the appropriate term for use in the TS. If automatic trip logic l
!'8 "5Cd, then define the term _
l
. Closed - With issuance of the Tech Specs in SSAR Rev. 9. f Page: 1 Total Records: 55
AP600 Opes It;m Trackirg Sy;tm Dmb2e: Execntive Summ ry Date: Sn/97 Selection: lnrc st codejo' Resolved' And (NRC Branch] hLe 'NRR/lilCir Sorted by DSER Section item DSER Section Titic/ Description Resp (W) NRC No. Branch Question Type Detail Status Engineer Status Staius
- ._ -- .. - - - - i etter_No. /
~ D. _a_te_
2442 NRR/IllCD 16.1 M fG-OI TEC11 SPEC /Birsa Closed Action W
~
Es aluate the potential for failures that could defeat the capability for placing functions into bypsss. This needs'io be considered in the deselopmen' t l
'of the actions (operator could be required to take an action to put the channel in bypass, and be unable to perform the action) At very minimum. thE bases should clearly explain what is meant by placing the channct in bypass. Is taking the action (switch operation) without the system succeeding
%L? This action is there to go from I/3 logic to 2/3 logic (which affords operating fault tolerances). Staying in the 1/3 condition is not unacceptable.
[This concern is valid for both the RPT and ESF.
^~
~ ~ ' ~
~~
~~ ~~j
[ Closed - With issuance of the Tech'Sp'ecs in SSAR' Rev79[
2443 NRR/1HCB 16 i MTG-OI TECllSPEC/Birsa Closed Action W
~ ~
[ Communication of failed bypassed conditions is a concern within the instrumentation sectkui
. 2 _ - - - . _ = - . _: : . - _z-
. _ ' ll Closed _With issuance of the Tech Specs in SSAR Rev. 9.
2444 NRR/illCB 16 I MTGOI TECllSPEC/Birsa Closed Action W
~
ussing actuat echep Qan EKAC ( A1 or [ A2) fails, need to venfy logic cabin'te is operable _ _
~~ ~
]
$EE wiME* ace?'he Tech SMs in3SAR Rc}v}{[ _ _ (( ]
2445 NRR/IllCB 16 I MTG-OI TECllSPEC/Birsa Closed Action W
~
Re-consider development ofI SF actuation table and consideration of failures. Con id 3 eration@ds to include failures that defeat multiple functionsj
~ ~ ~ ~ ~~ ' ~~ ~ ~ ~ ~
[ Closed - With iEuanc~c of the Tech Specs'in SSAR Rev3
~}- ' ~ ~~ ]
2446 NRR/IllCD 16 i MTG-OI TECilSPEC/Birsa Closed Action W
~ ~
{ Agecment reached that ESF channci definition is from sensor through the output ot ESF subgroups (ESH and ESil)
~ ~ ~~ ~
]
Closed - With issuance of the Tech Specs in SSAR Rev[9' ~ _
'}
2447 NRR/IllCB 16 I M iG-OI TECilSPEC/Birsa Closed Action W
- Assignment of ESF functions to ESF subgrups and logic processors should be consistent uith'disersity amoung functions. SPecifically, diverse l functions should be placed so that fluid system diversity is mair.tained through the PMS where appropriate. Fuild system designer etTorts to devekp
, cts should help identify these functions. f
~
~ ~~ ~~
~
Closed - With issu[ance of the Tech Specs in SSAR Rev. 9 _ _
{
2448 NRR/IllCB 16 1 mig-COM TECllSPEC/Birsa Dropped Dropped
~
I ine for logic may be down the middle of the logic cabinet block (not including the output actuation signals)~ j No action required (11 Li identified for deletion, Westinghouse concurred)
, Issue dropped to " Top 50" list-l Closed - With issuance of the Tech Specs in SSAR Rev. 9~ ~ ~ ~
2449 NRR/111CB 16 I mig-OI TECllSPI C/Birsa Closed Actkin W
~
- Manual actuations of ESF should address the dedicated contmis'(s) stem level manual actuations). l
~
[losed - With issuance of the Tech Specs in SSAR Revi9.~ ~ ~ _ .
)
IMge. 2 Total Records: 55
AP600 Open It:m Trackirg Syst:m Dat;b;se: Execntive Simm:ry D:tn Sn/97 Selectio2: [nre st code]<>'Resobed* And [NRC 13ranchlliEc 'NRR/IllCB' Sorted by DSER Section item DSER Sect on Titic/ Description Resp NRC (W)
No. Branch Question Typc Detail Status Engineer Status Status i.etter No. / Date i 2450 NRR/IllCB 16 I MTGot TECIISPEC/Birsa Closed Action W
~ ~ ~
~
Consider card failures that'afIcct multiple manual actuationi(IK) cari faijiire) versus~ fai!ures that affect' individual manua'l actuaiions (input device '
l failure). 3
[ Closed - With issuance of the Tech Specs in SSAR Rev. 9. _
] ;
2451 NRR/llsCB 16 i MTGot TECilSPEC/Birsa Closed Action W 7
- 4th Stage ADS valves will need to be addressed carefully. Evatuate manual actuation precedents for BWRs relative to squib valves.
... _ _ __. _ _._. _ ._ __ _l l Closed - With issuance of the Tech Specs in SSAR Rev. 9. [
]
2452 NRR/IllCD 16 i MTG-OI TECilSPEC/Birsa Closed Action W f
^
II.y signa be co{credgWCA 6
}. u chfteckurefadiauon bgnals_ _ - )
._]
' ~ ~ ~ ~ ~~
~ - . -- t 2453 NRR/IllCB 16.1 MTGot (Cloed!With s issiia~nce'of tfuiTech ' Specs in SSAR'RevT9.
TECIISPEC/Birsa Closed Action W
] ;
~~ ~ ' ~- ~ ~
lVES actuatiorddet.rol r'oo'm' isolation tecluiical sp' ecificaiion'5uId be iricluded with otiter ESF 'functioiiIAs'sImied logic for separatio'n in t
- standard specs is that these fucntions were typically not included in the Westinghouse _ protection systems
- ~
l Closed - With issuance of the Tech Specs in SSAR Rev. 9. ]
2454 NRR/IllCB 16 I MTG-OI TECllSPEC Closed Action N
~ ~
P
~~
__ . _. . _]
[ Closed - With issEce'Ef the YectIS'pecs in SSAR Riv.i ' ]__[_ __.. _ _ ))] -] [~]
2455 NRR/IUCB 16 i MTG-OI TECIISPEC/Birsa Closed Action W
~ ~ ~
Evaluate failure of a display or a QDPS bicol that result in the a un'vailability'of one display. Need to consider safety arguments versus burden of ]
completion time to correct. Spare parts considerations could be signifwant.
l
~ ~~ '
[ Closed - With issuance of the Tecli Specs in SS'AliRev.'9E _ __ ]
2456 NRR/lilCB 16 i M iG-OI TECIISPEC/Birsa Closed Action W
~
With two failures ofdisplays and no safety-related displays available, s the 7 day completiorItime appropriate.
~ ~ ~~ ~
~
_] >
{ Clog- With issuance of the' Tech Specs'in SSAR Rev. 9.'. _ . _ . _ _ _ . _ _ . _ _ _ . _ _ _ _ . _ _ _ . _ . _ _ ~ _
j ;
2457 NRR/IllCB 16 I MTG-OI TECIISPEC/Birsa Closed Action W ;
For PAMS, standard specs were determined using all type A, and category I variables, Consider use of action J Ino LCO 3 0.3) producing a special 4 ireport. Need to consider how this applies to AP600 PAMS and AP600 RG 1.97 categorization.
I 'rL . . . : ' = .^ = = .-- = = .- = = :----r 2: : . . .-- -
[I*'*b II' 55"MOIthe Tc3h Specs in SSAR Rev 9 _
.) i 2458 NRR/IllCB 16 I M TG-Ol TECilSPEC/Birsa Closed Action W
~~
Determine completion times for functions' from Uie~ remote shutsown winistations. Completion tirnes simuld reflect less likelihood of uw. )
[ Closed - With issuance of the Tech Specs in SSAR Rev. 9. l 2459 NRR/IllCB 16 I mig-OI 1ECllSPEClairsa Closed Action W INtermine a defendable position regardmgIE~guiar sEciliand tests n'nd credit for self-diagnostics and proside to NRC l
- S 5' S 9 l Page: 3 Total Records: 55
AP600 Ope:i It m Tracking Sy;t:m Datbase: Exec-tiva S mm:r7 Dat;: Sn/97 Selectio2: [nre st wdejo* Resol ed' And [NRC Branch]like'NRR/lilCB' Sorted by DSER Section item DSI R Section Resp Title / Description (W) NRC No. Branch Question Type Detail Status Engineer Status Status Letter No / Date 2460 NRMUCB 16 i M IG-Of TLCilSPEC/Birsa Closed Action W
~
llow do we test the tester relative to the diodes on the signal corkljtionerUs this a surveillance issue or a maintenance issue [
~
~~
~j
~ ~ ~ ~ ~ ~ ~~
~ - ~[
~
l Closed - With issuance of the Tjch Specs in S$AR Re{9. ~' _ _
]
2461 NRR/lilCD 16 i MTG-OI 1LCllSPI C Closed Action N
- ~
NRC to determme/ consider the credit given to self-diagnostic' feature of the software based system within the development of the DSL}R.
~
~ ~' ~ ~ ' ~ ~ ~ ~ ~ ~
[j 2462 NRR/IIICH 16.1 M1G-OI posep - With issuance of th]e ech Specs in[SSAR Rev. 9 ' ~ ' _ __
TEC11SPtiC/Birsa Closed Actkm W
]
~
Determine appropriateness of continuing to include this statement (allowing for adjustment as necessary of the setpoints are within the required rangd
[and accuracy)in the definition for channel operational test. _
j
~
[ Closed _%}iih issuance of the Tech Specs in SSAR Rev 9. . _ __
]
2463 NRIU1IICH 16.1 MTG-OI TLCIISPIC/Utrsa Closed Action W IConsider NRC recomme'ndation for includmg'one value (allowable valuis'or trip setpoints) Determine the feasability of using safety analysis vaucsi fin brackets for design certification specs j
~~ ' '
mig-COM se th Issuance odTech Skcs in SSAR Rev ~~ _ _ _ ~'l 2464 NRR/IllCB 16 i 1ECllSPEC1McDermot Closed inactiv e
- ~
ilhe TS should include consideration of panel remoial during power operation of the containment arr bafile. It should be noted that the USNRC is l concerned about actions for the purpose of planned maintenarice and may impose restrictions in the future. _
~ ~ ~ ~^ ~
Closcil IWith issuance of die Tech Specs in SSAR'Rew; 9. ~ ~ .
j 1977 NRR/IllCD 20 3-1 DSI R-COL Winters Closed Action N (2031 For Issue 142, the COL applicant should implement an annual program to inspect and test all electronic isolators between Class IE and l
,non-Class iE systems, as well as identify the specific isolation devices used in the design. _
~ ~ ~ ~
- Closed -Issue 142 has becai removed from the Rev. 7 of Section 1.9 4 of the SS'AR. and included in Table I 9-2 Listmg of Unresolved Safety issues {
'and Generic Safety issues, according to what was agreed with the NRC. The item can be considered closed. t
['9'ICC""_with J Kenym, H. Lee K. Deutsch, R. Nydes on I t/21, NRC has the action to review the ITAAC for IE isolators rLn 12/4 .
l 1521 NRR/lilCB 20 3-21 DSER-OI Lindgren/Deutsch Closed Action W jThe staffintroduced significant information coricerning the resolution ofissue'I42 which Westinghouse did not include in it response for the issue.
In terms of the isolation devices and the design of the AP600 instrumentation and control architecture. Westinghouse did not address what happens if I
n communication error occurs, and did not identify the error messages generated and diagnostic tests applied to isolate the cause of the error. The
[ question is would this include errors caused by leakage through an isolatw. The stafT requests that Westinghouse should also address these items in , '
the resolution of tssue 120 for the AP600 design
=: - :=- :=2 -=2
. . . . z. - .
Closed - Revised write-up for issue 142 in Revision 9 includes additional information Per iI!2I telecon mith T. Kenyon and 11. Lee, it is Westinghouse action to determine if this is a COL action item, and if so, to note it as such (or put j in ITAAC) l l' age: 4 1otal Records: 55
AP600 Opea item Trackirg Symm Dat-base: Exec;tive Summary Dr.tz: Sn/97 Selection: [nre si ccdejo' Resolved' And [NRC Branch]like *NRR/illCir Sorted by DSER Section item DSLR Section Titic/Desuiption Resp NRC (W)
No Branch Question lype Detail $tatus Engineer Status Status Date
_-. -_-_-. . _ - .. l etter No _/
1576 NRR/IllCD 20.7-18 DSER-OI 13559 Closed Action W NSD-NRC-96-4818
~
leer Dulletin 90-01, Westinghouse should commn to use Rosemount transmitters manufactured after )uly i1,1989 and adhess the on-line monitoring capability of the AP600 design, because this is an effective method to address the loss of fill-oil in the Rosemount transmitter issue.
~
~ ~ ~ ~ ~
l
- Action W -lhis item will' be addresid in'tlAAugust revision to tlAAP600 SSdR Closed - WCAP-13559 Rev. I issued September i1,1996
^~l
- Per NRC comment received during iI/21 telecon, Westinghouse has the action to determine if this is a COL applicant action and appropnately I
document it as such rin 12/4 i . _ _ _ _ _ _ _.. _ _ . . _ _ _ _ -._ _ . _ . _
4257 NRR/IllCD 7 MTG-COM SSARREV/Deutsch, Ke Confrm-W Confrm-W
~ ~ ~ ~ "~
IAdd WCAP-14080 as reference for SSAR Section 7.1 l t_7 _ .
J
} Westinghouse has confirmed this is an appropriate reference and transmitted the SSAR markups to NRC. This item is opened to ensure the marked l changes are included in the next SSAR rev. rkn 1/15N7.
! Refer to NSD-NRC-97-4947 for changes to be included in SSAR Rev II. rLn I/31/97 2023 NRR/IllCD 7. DSER-0150 ITAAC/Deutsch Dropped Dmpped NSD-NRC-4875
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ - ~ ~~
[No Commitment io Industry Standards fot Digital Systems' ~ ~ ~~
jWhile the SS AR references IEEE standards 279,384,603 and 7% for the design of AD600 I&C systems, the stalTis concemed that there is no lcommunications reference to digital micmprocessor-related standards. Specifically they are concemed about the lack of standards related to multiple protocols, and hardware / software design. The staff wants Weetinghouse to make an explicit commitment to industry hardware and
[ software related standards. No detailed documentation of the process and no phased ITAAC for verification of the design. _ ,
~ ~ ~ ~ ~
1
~
faction WTiteIn 1037' closed!I tat final sentence ofitem. Remaining Ation to ad5ess No dc51eldocumentatioEof the process and no phaNd~
- lTAAC for verification of the design'.
l
- SSAR Ch 7.1 commits to a V&V pmgram, meeting Standards, etc., such that NRC expectations are met. When the ITAAC for PMS is complete, i lthis item will be closed. rLn Sn/96 l
l Closed -ITAAC submitted by NSD-NRC-96-4875 of IlnN6.
l
.Per i1/21 telecon for DSER Ch 7, NRC wants to discuss ITAAC approach with Westinghouse.
' Dropped - Dropped per telephone directive from Diane Jackson on 4/23/97.
" ~ '~ ' ' ' ' ' '~
2025 NRR/IllCD 7. DSER-0150 SSARREV/ Miller Dropped Dropped
~ ~
- 29. Environmental Qualification of DAS Equipment and Sensors l j The DSER indicates that the DAS equipment must be designed and qualified to the environment in uhich it needs to perform lhe Westinghouse' position is that the DAS equipment will be designed to function the environment in which it needs to perform flowever, the DAS equipment will not bc subjected to a fulf-blown 10 {FR $0 49 / IEEE 323 qualification program
~~ ~
Closed - SSIR'Chapte'r 7's'ction e 7.73 I l reviscito adSess[ ~ ~ ' l Per an I I/21 telecon, NRC thinks the DAS sensors and actuated devices (e g , PR11R solenoid valve) should be qualified to a higher (PMS) standard but Westinghouse does not agree.
,By 12/6 fax, W proposed SSAR change to clarify qualification, NRC to review approach. rLn 12/6 Completed in SSAR Rev 10. rin 1/14N7 i
- Whoops! I checked and it didn't get into SSAR Rev 10 it WILL get into Rev II. See NSD-NRC-97-4947. rLn 1/30N7 i Dmppep @mpped per tclephone pyctive frorn Diane Jackson on 4/23N7.
l Page: 5 Total Records: 55
AP600 Ope 2 Item Trcckirg Syst;m Dat: base: Execrtive Simm ry Lut:: 5/7/97 Selecties: [nrc st code]<>' Resolved' And [NRC Branch]like'NRR/ll!CB' Sorted by DSER Section item DSL R Section Resp NRC Title / Description (W)
No Branch Question Type Detail Status Engineer Status Status Ectter No. / Date 1038 NRR/IllCD 7I4-1 DSER-Ol ITAAC/Deutsch, K. Action N Action N NSD-NRC-96-4737
~ ~
iWestinghouse should describe in the SSAR, CDM, and ITAAC the digital shtem des'ign process.
jWestinghouse should provide a detailed descritpiton of the digital system design pgss in the SSAR and CDM with a correspondmg ITAAC.
~
Action W - WCAP-13383, which descritics the digital system design pio~ c eslis Eing updatedI lhe certified design material and I'TAAUs =iIl be ~
modified he SSAR has been modified to reference the design process and to indicate the software design standards the design process conforms
!ao. This information is provided in Revision 3 of the SSAR, Subsection 7.l.2.15. The WCAP and ITAAC revisions must be completed before this l item can be closed out. NRC has requested a presentation when all elements are completed. WCAP-I3383 rev due 5/30/96 ria 5/7/96
- WCAP-13383 in repro 6/14 for 6/l7 release. rkn 6/14/96 Closed - Response provided by NSD-NRC-96-4737.
I Per an iI/21 W/NRC telecon, the NRC thinks the I&C ITAAC is deficient and requested that we 'fix" the ITAAC or justify / explain deviations from ithe SRP 14.3 5 to NRC satisfaction. NRC to provide specific comrnents on the ITAAC. rLn 12/2 1039 NRR/IllCB 7. I 7-1 DSER-OI ITAAC/Deutsch Action N Action N NSD-NRC-96-4737
~ - ~ '
Westinghouse should describe a commercial grade item dedication program for digital systems.
lj Westinghouse has not addressed the commercial grade item dedication program that is necessary to e lrelated and nonsafety-related I&C systems using commercial of-the-shelf equipment. The design, verification, and validation process for COTS ware and hardware shouQbe cleag Mented for desjgn certification _ _ _ - _
~~ ' ~~ ~ ~
! Action W - WCAP-l3383 is tEin[up' dat'esto include a commercial graEitim~ dedication process The SSAR has becrimo'd ified to reference this I
process This information is provided in Revision 3 of the SSAR, Subsection 7.1.2.15. The WCAP revision must be completed before this item can be closed out.
l
.WCAP in repro 6/14 for 6/17 issuance. rLn 6/14 Closed - Response provided by NSD-NRC-96-4737.
}
'Same as item 1038 rLn 12/2 j 104I NRR/IllCB 7.26-1 DSER-Ol ITAAC/Deutsch, K- Action N Action N
~ '
I
'The stafT has not >ct completed its evaluation of the software architecture designI '
...because WCAP 14080 was submitted in July 1994, the staff has not completed its review of the documem and is continuing its evaluation of the software architecture based on both the proposed design and the associated design process. De results from this evalua: ion will be presented in the
. final SER for AP600. -
.: a== - :=====-- . . = . =- _ _ - _ ~ =.
, Closed - Westinghouse has completed necessary submittals to support staff review..
I lPer iI/21 W/NRC telecon, when the NRC agrees with the design process through tncir review of the 11 AACs, this item will be closed. rLn 12/2
' ~
' ~ ~ ~ ~ ~
1043 NRR/IllCD 7.28-1 DSER-OI ITAAC/Deutsch Action N Action N NTD-NRC-95-4464
- Westinghouse should' provide a discuIsion concerning the qualification of digital equipment to the electromagnetic ensironment. f
. Westinghouse has not addressed the issue of electromagnetic environmental qualification and has not committed to the appropriate standards. j Closed - List of standards reviewed by NRC during meeting on May 15-16. Standards incorporated into Resiston 3 of the SSAR, Subsection l 17.1 4 I 6 !
i Per an iI/21 W/NRC telecon, the technical issues are resolved When NRC agrees with design process thru ITAAC review, this item will be closed l l' age: 6 lotal Records: 55
AP600 Ope 2 It:m Trcckirg Sy;t:m D;t .hase: Exec 1tive Simm:ry D;te: Sn/97 Selectica: [nrc si codel<>* Resolved' And [NRC Branch] like 'NRR/IllCB' Sorted by DSER Section item DSER Section Titic/ Description Resp (W) NRC No Branch Questu>n T)pe Detail Status Engineer Status Status Let No. . /_-
.- -. . -_. --.-_ --. . - .- - . .-- -.t.er - - Date . -
1044 NRR/IllCB 728-2 DSER-OI ITAAC/Deutsch, K. Confrm-W Action N NTD-NRC-95-4464
~ ~
~
gWestinghouse should provide information concerning environmental qualification of'PMS components address'ing local' temperature rises aboveAc "
' room ambient experienced by the components during operation.
It is desirable to have additional margin built into the design. De components should, therefore, be qualified by testing to higher temperatures than
!specified in the SSAR for a given room environment. Westinghouse should address this concern in the SSAR. Westinghouse should also provide
! mild environment equipment qualification in the CDM with the corresponding ITAAC.
' - = =:==-===- _ -_: = = == = = == z 2.= 2 ..
- = ; =.= n ,
Closed - Technical information agreeded to by NRC during meeting on May 15-16. Additional technical information regarding the equipment desige margin to loss ofIIVAC has been incorporated into Revision 3 of the SSAR, Subsection 7.1.41.8. rkn 12/2
' Westinghouse needs to decide approach to close this item. rLn 12/6 Action N - NRC still has the action to evalua:e the Westinghouse proposal on procedural fix ofinstrument overheating after 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. (6/2I meeting with W/SPLB/IllCB). Based on 11/21 W/NRC telecon, this approach is reasonabic; see qualification program in SSAR Section 3.11..
, Action W - NRC requested W provide proposed COL item for qualification margin and instrument setpoint data or document in the CDM and
' corresponding ITAAC (W is considering options; did not commit to cither approach). rLn 12/2 Westinghouse does not consider there to be an applicable COL action to identify. Technical information eclared to design margin against a loss of IIVAC was provided in SSAR 7.l .4.16 and is considered technically resolved, as waws previously agreed to by NRC. His item is considered l closed since there is no Westinghouse action required at this time to address this item (since the NRC relates this comm:nt to the PMS ITAAC, the fresponsibic engineer is changed to ITAAC). rkn 1/14/97.
Action W -(from NRC on 1/28/97) Sut,mit revised CMD & ITAAC to include COL action to include additional design margin to accomodate a loss of the normal IIVAC. Provide an alarm if intemal cabinet temperatures reach an excessive value. jww1/28 This mas previously closed and is still considered closed, meaning there is no Westinghouse action identified or required o close this item, all necessary submittals have been made. For background, SSAR Section 7.14.1.6 mas revised in Feb 1996 to address this. Specifically, there is a sentence which reads,'The cabinets containing the digital equipment are provided with temperature sensors which provide an alarm ifinternal l cabinet temperatures reach an excessive value." This is closed. rLa 1/30/97. ;
Per telecon with flulbert Li today, the action is for Westinghouse to include this alarm in the ITAAC. rLn 2/18/97. Rather than include the alarm jspecilically, the EQ info for PMS temperatures is included in the draft ITAAC (pages of which were faxed today). W believes this addresses NRC
- concern. rLa 3/21
] Confirm N - Fax to Kenyon with ITAAC markup wnct on 3/21/97. jww
[ Action N - Status confirmed by Ted Quay in telecon on 5/1/97,1 SER section is in to NRC Projects. jww 1049 NRR/IllCil 7.58-l DSER-O! IIAAC/Lindgren/Deuts Action N Action N
-~
,Westmghouse should describe the d'es'ign feature's of the inco5 ins ~rumentatiori t system. !
In its response to Q492,5 dated My 25,1994, Westinghouse states that information on the employ ment of fixed incore detectors in conjunction with!
an online power distribution monitoring system will be provided to the NRC to support the final SER l
. -. = = 2-- -_ - . .- :z . .
Closed - The technical information was accepted by the I&C Ikanch of NRC during the meeting on May 15-16. This technical information has been,:
[ incorporated into Revision 3 of the SSAR, Subsection 4.4 6.1.
Open for ITAAC based on fax from NRC I/21/97. rLn I I For Chapter 7 this item is resolved (NRC/RSB to communicate any concems with qualification of thermocouples and instrument coolant capabilay l gutspe the scope gfgter Up I2/2 _ ___ - _ __ - __
l I age: 7 Total Records: 55
AP600 Open Item Tr ckizg Sysum D: Abase: Execttive Srmmary D tn Sn/97 Selection: [nre st code]<>' Resolved' And [NRC Branch)like 'NRR/lilCB* Sotted by DSER Section item DSER Section Titic/Descrip' ion Resp (W) NRC i No. Branch Question . Type Detail Statm Engineer Status Status - Letter No. / Date 1052 NRR/IllCB 76.2-1 DSER-OI Schull, T. Closed Action N
~
~-
Westinghouse should provide additional design d'etails of theAumulator isolation vrive iEterlocks' impostant io safety to confirm th'aithe design'~ '
meets the relciant requirements of the SRP, includmg IEEE 279.
~ ~ ~ ~
[Clohd ddit'ional t ic h ion hNbeen incorpos iedVo Rbsh3 of tiw SSIll Subsectionb2.Ibgub.2 was also modifihd to'
- include additional technical detail.
, Action NRC - Per II/21 telecon, NRC to review teclu.ical information already provided since this operator is nonsafety, not impostant to safety, has t
, separate power, positive 3 position indications, and power removed at-power (consistent with Tech Specs) and lunit switch alarms. rLn 12/2 lTechnicalinformation provided. NRC to sJvise ta resolution status rkn 1/14/97 ;
Per fax, NRC considers this open for interlocks concern (FSER open item 7 o.2-l) rkn I/21/97 '
1053 NRR/IllCB 763-1 DSER-Oi Schulz, T. Closed Action N
~ ~ ~ '
{ Westinghouse should provide addhional dcsi F n detaiis of tElliW'ST disch7ge v'alve interlocks important' to safet'y to cddirm that th'e design meets l the relevant requirements of the SRP, including IEEE 279. L Closed - Additional technical informatien has been incorporated im Revision 3 of the SSAR, Subsection 7.6.2.2. Figure 7.2-1 was also modified t-i !
include additional technical detail Action NRC - See 1052. rin 12/2 Technical information has been provided. NRC to advise regarding resolution status. rkn 1/I4!97. [
,Per fax, NRC considers this open for interlocks concern (FSER open item 7.6.2-1). rin 1/2I!97 1055 NRR/lilCB 7.7.2-1 DSER-OI ITAAC/Delose, Frank Action M Action N NTD-NRC-95-4464
~ ~ ~ ~ ~~ '" ~ ^~
[WestMse'should priv~ide ' additional infEmation concerniEgddesignif the DASE ~l
' Closed - Technical information accepted by NRC during meeting on May I 5-16. This additonal tech sical detail has been incorporated into Revisiosj 3 of the SSAR, Subsection 7.7.1.11.
l
[NRC action to n: view ITAAC. Per iI/21 telecon, this item is now subject to DAS ITAAC comment resolution / completion. sin 12/2 5095 NRR/IllCB II AAC RAI-OI Deutsch Action W Action W
~~
l l(
SUBJECT:
FOLLOW UP QUESTION REGARDING IllE' P600 INSPECTIONS,'TESli ANALYSS [
, ITAAC) i lnstrumentation and Controls Branch Camments
- 640 371he certified design material (CDM) and the inspection, tests, analyses and acceptance criteria (ITAAC) for the AP600 I&C systems should I ! provide information on the design process and implementation, with appropriate tests, inspection and acceptance criteria, based on supporting i Iinforrnation in SSAR Chapter 7 and Section 14.3. The material should include information on the design controls, desclopment, and qualification i
[ processes for IAC hardware, software, and other design features. __ l l _ _ _ _ _ _ . _ _ _ _ . __ _
5096 NRR/lilCB ITAAC RAl-OI Deutsch Action W Action W
^ '
i SUBJECT 5 FOLLOW UP QUEST 1ON REGIRDl5G rile P600lNSPECIl6NS, TESTS, ANALYSES, AND ACCLI'TANCE CRITI RfA l '(ITAAC)
- Instrumentation and Controls Branch Comments
- 640.38 The CDM should address the hardware and software development process to be used in the design, testing, and installation of R&C equipmeni [
!and should also include the description of the design process to be followed for hardware and software development, design commitments, the i i
! inspections, tests, and analysis to be performed to verify that the design is consistent with the commitments, and acceptance criteria against which
!the design will bejudged.
I l The commitment in the ITAAC should reflect the elements, activities, and documentation required of the various phases of the life cycle as
[shown in Figure I of SRP Section 14.3.5._, __ _ _ _ _ _
Page: 8 Total Records: 55 l ,
l . . . _ . . - _ , - - . - - . - - - ._ ___
AP600 Opea Itim Tracki;g Syst:m Dat: base: Exec tive Szmm ry D:te: Sn/97 Selection: [nre st code l<>' Resolved' And [NRC Branch]like'NRR/lilCir Sorted by DSER Section item DSI R Section Titic/ Description Resp NRC (W)
No Branch Question Type Detail Status Engineer Status Status ! ctter No. /
_. _Date 5097 NRR/IllCB ITAAC RAI4)I Deutsch Action W Action W
~
{
SUBJECT:
FOLLOW UP QUESTION REGARDING Tile AP600 INSPECTIONS,11ESTS, ANALYSES,- ND ACCEPTANCE CRITERIA' ~
'(ITAAC) instrumentation and Controls Branch Comments
[64039 Provide criteria in the CDM and SSAR to guide the design process throughout the digital l&C systems life cycle stages. The ITAAC should
' provide the acceptance criteria for verifying the design through the stages while the S'iAR adds the set of guidehnes and standards that will proside
'more detailed criteria for the development of the design.1he ITAAC for software and hardware for the I&C sy stems should verify the design stages within the overall design process as specifkd in the WCAP-13383, Resision I:
(a) Design requirement phase (b) Definition phase (c) Development phase (d) Test phase (integration, serificatism, and validation)
In addition to the four phases listed above, the stalT believes that two more phases should be added.
(c) Installation phase (f) Operation and maintenance phase.
The ITAAC for soRware development should include, but not be limited to the following elements:
- software quality assurance plan (SQA) software management plan (SMP) software configuration management plan (CMP) software development plan (SDP)
- verification & validation plan (V&VP)
- software safety plan (SSP) software operation and maintenance plan (SOMP) 5098 NRR/litCB ITAAC RAl4)I Deutsch Action W Actioa W i
SUBJECT:
FOLI.OW UP QUESTION REGARDING Tile AP600 INSPI CTIONS. TESTS, ANALYSES, AND ACCI PTANCE CRITERIA i l(ITAAC) jinstrumentation and Controls Branch Comments
.640 401hc CDM should address the development and quahfication process for R&C equipment 1he discussion should include: ;
(a) design processes and acceptance criteria to be used for safety-related systems using programmable microprocessor-based control equipmenp (b) a program to assess and mitigate the eft'ects of electromagnetic interference on I&C equipment, (c) a program to establish setpoints for safety-related instrument channels, (d) a program to qualify safety-related !&C equipment for in-service environmental conditions, including mild environmental conditions with the potential for local hot spots due to abnormal conditions.
(c) a program to verify the conformance of the safety-related R&C systems in accordance uid guidance prosided in IEEE standards 279 and l03.
,6 (1) a program to verify the independence between redundant dnisions. In addition to sepaintMn requiremerts, the isolation a3pects should be la lso addressed 5099 NRR/IllCB ITAAC RAl4)I Deutsch Action W Action W
~
i SUBJLIT: I'OLLOW UP QUESTION REGARDING TIIE AP600 INSPECTIONS, TESTS, ANAL YSlis, AND ACGPTANCE CRllLRIA l
'(ITAAC) j Instrumentation and Controls Branch Commerits
,640 411he CDM should iriclude an Instrumentation and Control Systems Architecture I lock Diagram simil. :a !igure 7 l-1 in the SSAR. ;
Page: 9 Total Records: 55
i AP600 Opea it:m Trackirg Syst:m Dr.t: base: Exec:tive Summnry D;tn Sn/97 !
Selection: [nre st codej<>' Resolved' And [NRC Branch) like 'NRR/IIICB' Sorted by DSER Section item DSER Section Resp Title / Description (W) NRC .
No. Usanch Question Typc Detail Status Engineer Status Status '
letter No / Date 5100- NRR/IllCB ITAAC RAICI Deutsch Action W Action W
~ ~ ~
P600 lNSPECTIO5S, TESTS, ANAL'YSES,'AND NCCEPT5NCE'CIOTERIA l(
SUBJECT:
ITAAC) FOLLOW UP QUESTIOA REG RDINCITilE
- Instrumentation and Controls Branch Comments ;
440 42 in addition to the PMS and DAS, the R&C CDM and ITAAC should include the following R&C systems:
PLS - Plant Control System DDS - Data Display and Processing System OCS - Operations and Control Centers System ;
115 -Incore Instrumentation System SMS - Special Monitoring System i i
5101 NRR/IIICB ITAAC RAI-OI liayes Action W Action W
~
~ ~
f(SUBJECTi' ITAAC) FOLLOW UP QUESTION REG RDING Tile'AP600 INSPECTIONS TESTS, ANALYSES!h j instrumentation and Controls Branch Comments 640.43 The CDM and ITAAC should include the communication system that verifies the communication between the main control room and the flocal control stations, and the remote shutdown station and the local control stations 5102 NRR/lilCB ITAAC RAl-OI Deutsch Action W Action W i
~ - ~
SUBIECU: FOLLOW'UP QUESTION REGAliD15G Tile APEXUNSPECTIONS, TESTSTANAl YSES, ND ACCEPT NCECRIIIRIA ~
(ITAAC)
Instrumentation and Controls Branch Comments 640.44 in the CDM for PMS, the description of the logic and control should have more detail when addressing automatic decision-making and trip g
logic functions, and manual initiation functions associated with the safety actions of the safety-related systems. i 5103 NRR/IllCB ITAAC RAI4)I Deutsch Action W Action W l
l(
SUBJECT:
, ITAAC) FOLLOW UP QUESTION REGARDING TIIE AP600 INSPECTIONS, TESTS, ANALYSE Instrumentation and Controls Branch Comments
'640.45 The CDM and ITAAC for the DAS should follow the commercial grade item dedication program as defined in the WCAP-13383 Revision I.
[The DAS CDM should address defense-in-depth considerations for protection against common mode failures in the PMS.
L 6
Page: 10 Total Records: 55
)
l.
l Lovato, Janet M.
l From: N , " vuni K.
Sent: ednesday, Ma 07, 997 4:57 PM To:
Cc: t;N, ".; gov'
'wch@nrc.
aWr, Roberty B.; Lovato, Janet M.
Subject:
C..a w i o o.4.10 s
38 41017r1 wpf
<WP Attachment Enclosed >
Bill- Here is the most current version of the still draft RAI response it provides some explanation for Gene Hsii and Ron Young for TS 3.4.10.
Janet - Would you please print this and put it in the informal correspondence file? Thanks, Robin l
l l
I l
l e
4 n
Page1
O I
1*
l l NRC REQUEST FOR ADDITIONAL INFORMATION
! Question 410.17 Revision 1 l Position C.9 of RG 1.45 states that the technical specifications should address the availability of various types ofinstruments for RCPB leakage to ensure adequate coverage at all times. Describe how the AP600 design will meet this regulatory position (Section 5.2.5).
Response
SSAR Chapter 16, Technical Specification 3,4,9, 3.4.10 defines the operability requirements for RCS leakage detection instrumentation. ' 2dd!*!c", This instrumentation, used to identify reactor coolant pressure boundary leakage, is designed so that its operability may be determined at all times. Should a detector fail (signal outside its calibrated range or self r^ :+^ red detect trouble de+eeted), the plant instrumentation system will alarm in the main control room that the specific leak detectior- ^e:!c display
& is questionable. The alarm prompts the operators to observe other sensors providing leak detection information. Technical Specification 3,4,9,3.4.10 requires instruments ofdiverse monitoring methods to be operable to provide a high degree ofconfidence that smallleaks are detected in time to allow actions i to place the plant in a safe condition when RCS leakage indicates possible reactor coolant pressure boundarydegradation 2!!e":: the !-Enge te be n erged e cer 2 4 heur ; +'erefere, epenter h2 ce ruft!ent
+ re te dete : e r rnu!! SE: 2e frem +he -er+~ cee!2n+ 7 r+e- nd te t2ke ce~ec+: ce r4er : 2n
^det - 1n e-The primary methods of detecting leaks during stable plant operating conditions are sump level changes and reactor coolant inventory (pressuri:er level) changes. The primary methodsfor detecting leaks during varying system conditions such as makeup and letdown operations and power transients are sump level changes and Nf 3/F18 detection.
Table 410.17-1 summari:es the validity of the diverse methods ofdetecting RCPB leakage.
SSAR Revision: Revisions to SSAR Section 5.2.5 and Chapter 16 are attached.
l l
a 410.17-1 3 Westinghouse Rey,1 t
I I
NRC REQUEST FOR ADDITIONAL INFORMATION A
Table 410.I7-I l'alidity ofRCPB Leak Detection AfethoJs alia f B Leak &tection AletM
% RATED A VERAGE TilERAfAL RCS A10 DES TITLE REACTil'ITY PolVER(a) TEAfPERA TURE Sump Level ! RCS Inventory CONDITION (K,) (*F) (J),(e) Nf3/F18 (g) 1 Pouer Operation > 099 >5 NA Valid Valid (f) l'alid 2 Startup > 0.99 <5 NA Valid invalid Valid 3 flot Standby < 0.99 NA > 420 Valid Invalid l'alid 4 Safe Shutdown (b) < 0.99 NA 420 > T, > 200 Valid invalid Valid 5 Cold Shutdown (b) < 0.99 NA < 200 NA NA NA 6 Refueling (C) NA NA NA NA NA NA (a) Exca uding decay heat.
(b) All reactor vessel head closure bolts fully tensioned (c) One or more reactor vessel head closure bolts less thanfully tensioned 410.1F2 W Westinghouse Rev.1 -
. -~ __ _ - - ..
NRC REQUEST FOR ADDITIONAL INFORMATION (d) Sump level change detection ofleaks not valid during extremely cold outside conditions whenfrostforms on interior of containment vessel.
(e) Leak Jetection method not valid during containment atmosphere purge operations or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ofend ofpurge.
(f) Power > 20%
(g) RCS at steady state; i e. no temp. change, no pressure change. no makeup or letdown 410.17-3 W Westinghouse Rev.1
1 1
l NRC REQUEST FOR ADDITIONAL INFORMATION g_.
5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary The reactor coolant pressure boundary (RCPB) leakage detection monitoring provides a means of detecting and to the extent practical, identifying the source and quantifying the reactor coolant leakage.
The detection monitors perform the detection and monitoring function in conformance with the requirements of General Design Criteria 2 and 30 and the recommendations of Regulatory Guide 1.45.
Leakage detection monitoring is also maintained in support of the use of leak-before-break criteria for high-energy pipe in containment. See subsection 3.6.3 for the application of leak-before-break criteria.
Leakage detection monitoring is accomplished using instrumentation and other components of several i systems. Diverse measurement methods including level, flow, and radioactivity measurements are used i for leak detection. The equipment classification for each of the systems and components used for leak l detection is generally determined by the requirements and functions of the system in which it is located.
There is no requirement that leak detection and monitoring components be safety-related. See Figure 5.2-I for the leak detection approach. The descriptions of the instrumentation and components used for leak ,
detection and monitoring include information on the system. !
l To satisfy position I of Regulatory Guide 1.45, reactor coolant pressure boundary leakage is classified as either identitied or unidentified leakage. As defined in SSAR Chapter 16.1, Technical SpeciRcation l Section 1.1, identified leakage includes:
. Leakager e c'~ed : :' .: :vch as pu p ;W' c- er'e- re! seal or valve packing seal leaks that are captured and conducted to collection systems or a sump or collecting tank
- Le";e :~^ 2u9t!r; :::' : rd :r^ndrj :r : "~e ry'e- 'e+2ge' (m: 'ed:;e :: e'
- ^ :!$e ed te 'e p2-' ef e 'O gp ': * !de~m ed 'e 2;e ' ' e Srr e rn g ,eenm:g!
- pr m at:^- ? M ".!: 23f t:en2! 'e 2;e mu:' Se cent!de ed := e en!u2 tie e r e er'e-ee!1-' -' -~e ; S2!rce '
- RCS leakage through a steam generator to the Secondary System or through the PRHR heat exchanger to the IRWST.
- Leakage into the containment atmospherefrom sources that are specifically located and known not to interfere with the operation ofleakage detection systems or pressure boundary leakage. (Up to 10 gpm ofidentified leakage is considered allowable by Technical Specification M 8 because leakage isfrom known sources that do not interfere with detection of unidentiped leakage and is well within the capability of the RCS Makeup System.)
Other leakage is unidentified leakage.
i 410.17 4 W Westinghouse Rev.1 i -
l
=
l l
1 i
l i NRC REQUEST FOR ADDITIONAL INFORMATION im sg l
5.2.5.1 Collection and Monitoring of Identified Leakage
! Identified leakage other than intersystem leakage is collected in the reactor coolant drain tank. The reactor coolant drain tank is a closed tank located in the reactor cavity in the containment. The tank vent is piped to the gaseous radwaste system to prevent release of radioactive gas to the containment l atmosphere. For positions I and 7 of Regulatory Guide 1.45, the liquid level in the reactor coolant drain l tank and total flow pumped out of the reactor coolant drain tank are used to calculate the identified !
leakage rate. These parameters are available in the main control room. The reactor coolant drain tank, pumps, and sensors are part of the liquid radwaste system. The following sections outline the various sources of identified leakage other than intersystem leakage. I 5.2.5.1.1 Valve Stem Leakoff Collection Valve stem leakoff connections are not provided in the AP600.
5.2.5.1.2 Reactor IIcad Seal The reactor vessel flange and head flange are sealed by two concentric seals. Seal leakage is detected by two leak-off connections: one between the inner and outer seal, and one outside the outer seal. These lines are combined in a header before being routed to the reactor coolant drain tank. An isolation valve is installed in the common line. During normal plant operation, the leak-off valves are aligned so that leakage across the inner seal drains to the reactor coolant drain tank.
A surface-mounted resistance temperature detector installed on the bottom of the common reactor vessel seal leak pipe provides an indication and high temperature alarm signal in the main control room indicating the possibility of a reactor pressure vessel head seal leak. The temperature detector and drain line downstream of the isolation valve are part of the liquid radwaste system. l The reactor coolant pump closure flange is sealed with a welded canopy seal and does not require leak-off collection provisions. l Leakage from other flanges is discussed in subsection 5.2.5.3, Collection and Monitoring of Unidentified Leakage.
5.2.5.1.3 Pressurizer Safety Relief and Automatic Depressurization Valves l
Temperature is sensed downstream of each pressurizer safety relief valve and each automatic depressurization valve mounted on the pressurizer by a resistance temperature detector on the discharge piping just downstream of each valve. High temperature indications (alarms in the main control room) identify a reduction of coolant inventory as a result of seat leakage through one of the valves. These detectors are part of the reactor coolant system. This leakage is drained to the reactor coolant drain tank l
410.17-5 T Westirighouse Rev.1 l
l NRC REQUEST FOR ADDITIONAL INFORMATION 7 ...
during normal plant operation and vented to containment atmosphere or the in-containment refueling water storage tank during accident conditions. This identified leakage is measured by the change in level of the reactor coolant drain tank.
5.2.5.1.4 Reactor Coolant Pump Drain j Leakage from the reactor coolant pump drain is directed to the reactor coolant drain tank. This identified leakage is measured by the change in level in the reactor coolant drain tank.
l 5.2.5.1.5 Other Leakage Sources !
In the course of plant operation, various minor leaks of the reactor coolant pressure boundary may be detected by operating personnel. If these leaks can be subsequently observed, quantified, and routed to the containment sump, this leakage will be considered identified leakage.
5.2.5.2 Intersystem Leakage Detection Substantial intersystem leakage from the reactor coolant pressure boundary to other systems is not i expected. However, possible leakage points across passive barriers or valves and their detection methods are considered. In accordance with position 4 of Regulatory Guide 1.45, auxiliary systems connected to the reactor coolant pressure boundary incorporate design and administrative provisions that limit teakage.
Leakage is detected by increasing auxiliary system level, temperature, flow, or pressure, by lifting the y relief valves or increasing the values of monitored radiation in the auxiliary system.
The normal residual heat removal system and the chemical and volume control system, which are connected to the reactor coolant system, have potential for leakage past closed valves. For additional information on the control of reactor coolant leakage into these systems, see subsections 5.4.7 and 9.3.6 and the intersystem LOCA discussion in subsection 1.9.5.1.
5.2.5.2.1 Steam Generator Tubes An important potential identified leakage path for reactor coolant is through the steam generator tubes into the secondary side of the steam generator. Identified leakage from the steam generator primary side is detected by one, or a combination, of the following:
+ High condenser air removal discharge radioactivity, as monitored and alarmed by the turbine island vent discharge radiation monitor
- Steam generator secondary side radioactivity, as monitored and alarmed by the steam generator blowdown radiation monitor
+ Secondary side radioactivity, as monitored and alarmed by the main steam line radiation monitors 410.17-6 W Westinghouse Rev.1
. _ _. . _ -___ _ _ . .. ~ . - _ _ _ __ _
i l
! 1 l NRC REQUEST FOR ADDITIONAL INFORMATION
=.m um 1
J Radioactivity, boric acid, or conductivity in condensate as indicated by laboratory analysis Details on the radiation monitors are provided in Section 11.5, Radiation Monitormg.
5.2.5.2.2 Component Cooling Water System 1 Leakage from the reactor coolant systern to the component cooling water system is detected by the component cooling water system radiation monitor, by increasing surge tank level, by high flow ;
j downstream of selected components, or by some combination of the preceding. Refer to Section 11.5, -
l Radiation Monitoring, and subsection 9.2.2, Component Cooling Water System.
l 5.2.5.3 Collection and Monitoring of Unidectified Leakage I To detect unidentified leakage inside containment, in accordance with position 3 of Regulatory Guide '
l 1.45, the following diverse methods may be utilized to quantify and assist in locating the leakage:
l
- Containment Sump Level
+ Reactor Coolant System Inventory Balance
! . Containment Atmosphere Radiation Other methods that can be employed to supplement the above methods include:
. Containment Atmosphere Pressure, Temperature, and Humidity j
= Visual Inspection The reactor coolant system is an all-welded system, except for the connections on the pressurizer safety valves, reactor vessel head, pressurizer and steam generator manways, and reactor vessel head vent, which are flanged. During normal operation, variations in airbome radioactivity, containment pressure, !
- temperature, or specific humidity above the normal level signify a possible increase in unidentified i
leakage rates and alert the plant operators that corrective action may be required. Similarly, increases in containment sump level signify an increase in unidentified leakage. The following sections outline the methods used to collect and monitor unidentified leakage.
5.2.5.3.1 Containment Sump Level Monitor Leakage from the reactor coolant pressure boundary and other components not otherwise identified inside the containment will condense and flow by gravity via the floor drains and other drains to the containment sump.
A leak in the primary system would result in reactor coolant flowing into the containment sump. Leakage is indicated by an increase in the sump level. He containment sump level is monitored by two seismic l Category I level sensors in accordance with position 6 of Regulatory Guide 1.45. The level sensors are 1
410.17-7 l T Westinghouse gey,1 i
l l
l 4
. l I
I NRC REQUEST FOR ADDITIONAL INFORMATION l s;
IE l
powered from a safety-related C! ass IE electrical source. These sensors remain functional w hen subjected '
to a safe shutdown earthquake in conformance with the guidance in Regulatory Guide I AS. The containment sump level and sump total flow sensors located on the discharge of the sump pump are part of the liquid radwaste system.
1 Failure of one of the level sensors will still allow the calculation of a 0.5 gpm in leakage rate within 1 '
hour. The data display and processing system (DDS) computes the leakage rate and the plant control system (PLS) provides an alarm in the main control room if the average change in leak rate for any given measurement period exceeds 0.5 gpm for unidentified leakage. The minimum detectable leak is 0.03 gpm.
Unidentified leakage is the total leakage minus the identified leakage. The leakage rate algorithm subtracts the identified leakage directed to the sump.
l To satisy positions 2 and 5 of Regulatory Guide I AS, the measurement interval must be long enough to
, permit the measurement loop to adequately detect the increase in level that would correspond to 0.5 gpm leak rate, and yet short enough to ensure that such a leak rate is detected within an hour. The measurement interval is less than or equal to I hour.
When the sump level increases to the high level setpoint, one of the sump pumps automatically starts to j pump the accumulated liquid to the waste holdup tanks in the liquid radwaste system. The sump l discharge flow is integrated and available for display in the control room in accordance with position 7 of Regulatory Guide 1.45.
Procedures to identify the leakage source upon a change in the unidentified leakage rate into the sump include the following:
. Check for changes in containment atmosphere radiation monitor indications,
- Check for changes in containment humidity, pressure, and temperature,
- Check makeup rate to the reactor coolant system for abnormal increases,
+ Check for changes in water levels and other parameters in systems which could leak water into the containment, and
+ Review records for maintenance operations which may have discharged water into the containment.
W Westinghouse Rev.1 -
l l
i .
I NRC REQUEST FOR ADDITIONAL INFORMATION 5.2.5.3.2 Reactor Coolant System Inventory Balance 1
Reactor coolant system inventory monitoring provides an indication of system leakage. Net level change l in the pressurizer is indicative of system leakage. Monitoring net makeup from the chemical and volume control system and net collected leakage provides an important rnethod of obtaining information to establish a water inventory balance. An abnormal increase in makeup water requirements or a significant change in the water inventory balance can indicate increased system leakage.
I The reactor coolant system inventory balance is a quantitative inventory or mass balance calculation. His approach allows determination of both the type and magnitude of leakage. Steady-state operation is required to perform a proper inventory balance calculation. Steady-state is defined as stable reactor coolant system pressure, temperature, power level, pressurizer level, and reactor coolant drain tank and l in-containment refueling water storage tank levels. The reactor coolant inventory balance is done on a l periodic basis and when other indication and detection methods indicate a change in the leak rate.
l The mass balance involves isolating the reactor coolant system to the extent possible and observing the change in inventory which occurs over a known time period. This involves isolating the systems connected to the reactor coolant system. System inventory is determined by observing the level in the pressurizer. Compensation is provided for changes in plant conditions which affect water density. The change in the inventory determines the total reactor coolant system leak rate. Identified leakages are monitored (using the reactor coolant drain tank) to calculate a leakage rate and by monitoring the intersystem leakage. The unidentified leakage rate is then calculated by subtracting the identified leakage rate from the total reactor coolant system leakage rate. The minimum detectable leak is 0.1 gpm.
Since the pressurizer inventory is controlled during normal plant operation through the level control system, the level in the pressurizer will be reasonably constant even ifleakage exists. The mass contained in the pressurizer may fluctuate sufficiently, however, to have a significant effect on the calculated leak rate. The pressurizer mass calculation includes both the steam and water mass contributions.
l Changes in the reactor coolant system mass inventory are a result of changes in liquid density. Liquid density is a strong function of temperature and a lesser function of pressure. A range of temperatures exists throughout the reactor coolant system all of which may vary over time. A simplified, but j acceptably accurate, model for determining mass changes is to assume all of the reactor coolant system is at To...
l The inventory balance calculation is done by the data display and processing system with additional input I from sensors in the protection and safety monitoring system, chemical and volume control system, and liquid radwaste system. De use of components and sensors in systems required for plant operation provides conformance with the regulatory guidance of position 6 in Regulatory Guide 1.45 that leak detection should be provided following seismic events that do not require plant shutdown.
I
, l 410.17-9 W Westini; Muse Rey,1 i
l
- . 1 i
i I
NRC REQUEST FOR ADDITIONAL INFORMATION
! l H@
l 5.2.5.3.3 Containment Atmosphere Radioactivity Monitor l Leakage from the reactor coolant pressure boundary will result in an increase in the radioactivity levels inside containment. He containment atmosphere is continuously monitored for airborne gaseous radioactiv ity. Air flow through the monitor is provided by the suction created by a vacuum pump.
Gaseous and Nu/fu concentration monitors indicate radiation concentrations in the containment atmosphere, n e p: er ' ~" erpend rapid!; te rer+c cee!rt pr- ' e 'cund' r Mi ;e. No and F,, are is a neutron activation products which is are proportional to power levels. ^ dd!! en2!!y % hr _ m!2t! e'y th e- h2!f"rer 4 c equr yd " " er' equ!! b-!"-' r2pidly. An increase in activity inside containment would therefore indicate a leakage from the reactor coolant pressure boundary. Based on the concentration of Nn/Fa and the power level, reactor coolant pressure boundary leakage un be estimated.
It+NjEgre 'e-:n; :: :'e- hr 2 h!;h e :: :'y " '-- e ~'e : ep-'t g 2: 2 pe" e rge '!;he-
'b r 20 p ce-' The N13 and Fl8 monitor is seismic Category 1. Conforr.iane with the position 6
^
guidance of Regulatory Guide 1.45 that leak detection should be provided follo vi ng seismic events that do not require plant shutdown is provided by the seismic Category I classification. Safety-related Class IE power is not required since loss of power to the radiation monitor is not consistent with continuing operation following an earthquake. ibeu 20 p ~-t pe"r- 4'e' : e- ' cur, 2 %d 'r: 'hr 0 5 gp--
en b de'^c'ei Opr2 ting e7 e hr !-d!ated tN 2:2;e 6"; te '-'t'ge (rren cr'p!:n;!crer, ce"ec'ed ' t gerL ,-3 ug!3em,me g qg2;e te eoncmm32!-_e-g amm on e _ ,g. g!3g, gym _ryge l p,m y n,,3n7gp_ n u e cg.,~ -, 3,ge- m :n :. :. ,y n i y ,t m ,,, 75 p ce.,t geu ,., g 7,:.,g j 0 ' gp ' 'mge Mdground rd e'-^r' 'O p r-* er ' r::"n; ^ ? gp- 992;;e 9e "c re: 2re m
e n . ,n:- ne g.,g,n :,m mr .n g ejgg,,g.a.g. ,mg:n.:m, 1
1 Radioactivity concentration indication and alarms for loss of sample flow, high radiation, and loss of j indication are provided. Sample collection connections permit sample collection for laboratory analysis.
l The radiation monitor can be calibrated during power operation. l 5.2.5.3.4 Containment Pressure, Temperature and Humidity Monitors Reactor coolant pressure boundary leakage increases containment pres ure, temperature, and humidity, l values available to the operator through the plant control system. AtfuJ power the minimum detectable leak is 0.1 gym when the radionuclide concentration in containment reaches equilibrium. The Nn and Fu monitor can detect a 0.5 gpm leak when the plant is above 20% power and the concentration of radiogas in containment is at equilibrium.
An increase in containment pressure is an indication of increased leakage or a high energy line break.
Containment pressure is monitored by redundant Class IE pressure transmitters. For additional discussion see subsection 6.2.2, Passive Containment Cooling System.
410.17-10 Westinghouse Rev.1
l l
NRC REQUEST FOR ADDITIONAL INFORMATION 1
!== q E
The containment average temperature is monitored using temperature instrumentation at the inlet to the containment fan cooler as an indication of increased leakage or a high energy line break. This instrumentation as well as temperature instruments within specific areas including steam generator areas, pressurizer area, and containment compartments are part of the containment recirculation cooling system.
An increase in the containment average temperature combined with an increase in containment pressure indicate increased leakage or a high energy line break. The individual compartment area temperatures l can assist in identifying the location of the leak. l l
l Containment humidity is monitored using temperature-compensated humidity detectors which determi:w I the water-vapor content of the containment atmosphere. An increase in the containment atmosphere l humidity indicates relene of water vapor within the containment. The containment hum'.dity monitors i are part of the containment leak rate test system.
The humidity monitors supplement the containment sump level monitors and are most sensitive under ;
I conditions when there is no condensation. A rapid increase of humidity over the ambient value by more than 10 percent is indication of a probable leak.
Containment pressure, temperature and humidity can assist in identifying and locating a leak. They are not relied on to quantify a leak. I 1
l l
l l
l 410.17-11 W Westinghouse Rey,1 1
l l
l i
l NRC REQUEST FOR ADDITIONAL INFORMATION gg M
B 3.4 REACTOR COOLANT SYSTEM (RCS) )
i B 3.4.8 RCS Operational LEAKAGE i BASES
= -
)
l BACKGROUND I l
Components that contain or transport the coolant to or from the reactor core comprise the RCS. l Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate l
connecting systems from the RCS.
l During plant life, the jo.nt and valve interfaces can produce varying amounts of reactor coolant )
LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.
10 CFR 50, Appendix A, GDC 30 (Ref.1), requires means for detecting and, to the extend practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.
The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS LEAKAGE into the containment area is l necessary, Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is I necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is , detrimental to the safety of the facility and the public.
A limite<1 amount of LEAKAGE inside containment is expected from auxiliary systems that cannot be made 100% leaktight. LEAKAGE from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS LEAKAGE detection.
This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to pre renting :he accident analyses radiation release assumptions from being exceeded. The consequeices of vioiaing this LCO include the possibility of a loss of coolant accident (LOCA).
410.17-12 Rev.1 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION s: m=
APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses SAFETY ANALYSES do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA. The amount of LEAKAGE can affect the probability of such an event.
The safety analysis for an event resulting in steam discharge to the atmosphere assumes a 1000 gpd primary to secondary LEAKAGE as the initial condition.
Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leak contaminates the secondary fluid.
The SSAR (Ref. 3) analyses for the accidents involving secondary side releases assume 500 gpd primary to secondary LEAKAGE in each generator as an initial condition. The dose consequences resulting from the accidents are reported in Reference 3.
The RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement.
LCO RCS operation LEAKAGE shall be limited to:
- a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.
LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals end gaskets are not pressure boundary LEAKAGE.
- b. Unidentified LEAKAGE 0.5 gpm ei unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air N/3/F18 radioactivity monitoring and containment sump level monitoring equipment, can detect within a reasonable time period. This leak rate supports leak before break (LBB) criteria. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.
410.17-13 3 Westinghouse Rev.1 i
i l
1 i
I m
1 l
NRC REQUEST FOR ADDITIONAL INFORMATION
. =jg l Ei l BASES l l
I LCO c. Identified LEAKAGE ;
(continued) l Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE. Violation of this LCO could result in continued degradation of a component or system.
- d. Primary to Secondary LEAKAGE throue.h Both Steam Generate s ;SGs)
Total primary to secondary LEAKAGE through both SGs amounting to 1000 gpd produces acceptable offsite doses in the Steam Line Break (SLB) accident analysis. ;
Violation of this LCO could exceed the offsite dose limits for this accident. Primary to l secondary LEAKAGE must be included in the total allowable limit for identified i LEAKAGE.
- e. Primarv-to-Secondary LEAKAGE through One SG The 500 gpd limit from one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main .
steam line rupture.
l l
- f. Primary to IRWST LEAKAGE through the PRHR Heat Exchanger HX The 500 gpd limit from the PRHR HX is based on the assumption that a single crack leaving this amount would not lead to a PRHR tube rupture under the stress condition of an RCS pressure increase event. If leaked through many cracks, the cracks are very small, and the above assumption is conservative. This is conservative because the thickness of the PRHR HX tubes is approximately 60% greater than the thickness of the SG tubes. Furthermore, a PRHR HX tube rupture would result in an isolable leak and would not lead to a direct release of radioactivity to the atmosphere.
- 10. m 4 W Westinghouse Rev.1 -
NRC REQUEST FOR ADDITIONAL INFORMATION i=" ii BASES (continued)
APPLICABILITY In MODES 1,2,3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.
In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.
ACTIONS _AJ Unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE in excess of the LCO limits must be reduced to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This Completion Time is based on risk considerations and allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.
B.I and B.2 If any pressure boundary LEAKAGE exists, or if unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE cannot be reduced to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its i potential consequences. It should be noted that leakage past seals and gaskets is not pressure boundary LEAKAGE. The reactor must be brought to MODE 3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and to MODE 5 within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors which tend to degrade the pressure boundary.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without ACTIONS l challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.
l 1
410.17-15 W Westilighouse Rey,1
NRC REQUEST FOR ADDITIONAL INFORMATION 1
ia- == l I
1 SURVEILLANCE SR 3.4.8.1 l REQUIREMENTS Verifying RCS LEAKAGE within the LCO limits ensures the integrity of the RCPB is maintained.
Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be l positively identified by inspection.
l l
Unidentified LEAKAGE and identified LEAKAGE are detennined by performance of a RCS water inventory balance. Primary to secondary LEAKAGE is also measured by performance of an RCS water inventory balance in conjunction with effluent monitoring within the secondary steam and ,
feedwater systems. l The RCS water inventory balance must be met with the reactor at steady state operating conditions )
and near operating pressure. Therefore, this SR is not required to be performed in MODES 3 and 4 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation near operating pressure have been established.
l l
Steady state operation is required to perfonn a proper inventory balance; calculations during l maneuvering are not useful and a Note requires the Surveillance to be met when steady state is i established. For RCS operational LEAKAGE determination by inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, and with no makeup and or letdown.
An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere N/3/F/8 radioactivity and the I
containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These LEAKAGE detection systems are specified in LCO 3 4.10, "RCS LEAKAGE Detection Instrumentation." The containment atmosphere Nf 3/F18 radioactivity leakage measurement is valid onlyfor plant power > 20% in Af0DE I.
The containment sump level change method ofdetecting leaks during Af0 DES I, 2, 3 and 4 is not valid while containment purge occurs *or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end ofcontainment purge. l l
The containment atmosphere Nf 3/F18 radioactivity leakage measurement during Af0DE 1 is not valid while containment purge occurs or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end of containment purge.
l The containment sump level change method ofdetecting leaks during Af0 DES 1, 2, 3 and 4 is not valid during extremely cold outside ambient conditions whenfrost isforming on the interior ofthe containment vessel.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. A Note under the Frequency column states that this SR is required to be performed during steady state operation.
W Westinghouse Rev.1 -
I l
NRC REQUEST FOR ADDITIONAL INFORMATION l s2
=
'iEi m
l SR_ 3.4.8.2 This SR provides the means necessary to determine SG OPERABILITY in an operational MODE.
The requirement to demonstrate SG tube integrity in accordance with the Steam Generator Tube l Surveillance Pregram emphasizes the importance of SG tube integrity, even though this l Surveillance cannot be performed at normal operating conditions.
l l
l l
l 410.17-17 T Westinghouse Rev.1
l NRC REQUEST FOR ADDITIONAL INFORMATION
+
~2
=0 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 RCS Leakage Detection Instrumentation BASES BACKGROUND GDC 30 of Appendix A to 10CFR50 (Ref.1) requires means for detecting, and, to the extent practical, identifying the source of RCS LEAKAGE.
i Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting LEAKAGE detection systems.
LEAKAGE detection systems must have the capability to detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure. Hus, an early indication or warning signal is necessary to permit proper evaluation of all unidentified LEAKAGE.
Industry practice has shown that water flow changes of 0.5 gpm can be readily detected in contained volumes by monitoring changes in water level, in flow rate, or in the operating frequency of a pump. The containment sump used to collect unidentified LEAKAGE, is instrumented to alarm for increases of 0.5 gpm in the i
normal flow rates. This sensitivity is acceptable for detecting increases in unidentified LEAKAGE.
The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation. Reactor coolant radioactivity FIB has a longer halfhfe and is the dominant source usedfor leak detection. 'e' e'c ' ;"
'e 'c" dring : t!c! ere- m2p rd re 2 #e" " re d e , r " r N2ted ee- e t^ predrz hre be- fe ed rd 9: ^ pr~t : 2ppe2 r-em fue' e'--*
e'edding cent 2m:nst!^ e e'-d f ng of-+r The production of N13 and F18 is proportional to the reactor power level. N13 has a short half life and comes to equilibrium quickly. Fl8 has a longer hfe and is the dominant source usedfor leak detection. Instrument sensitivities for gaseous monnoring are practical for these LEAKAGE detection systems. The Radiation Monitoring System includes monitoring N13/F18 gaseous activities to provide leak detection.
l 410.1F18 W Westinghouse Rev.1 -
l 1
l
l i*
l NRC REQUEST FOR ADDITIONAL INFORMATION E+ igg
.y l
APPLICABLE The need to evaluate the severity of an alarm or an SAFETY ANALYSES indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary. The system response times and sens;tivities are described in the SSAR (Ref. 2).
The safety significance of RCS LEAKAGE varies widely depending on its suerce, rate, and duration. Therefore, detecting and monitoring RCS- LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE frc the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leak occur.
RCS LEAKAGE detection instrumentation satisfies Criterion I of the NRC Policy Statement.
LCO One method of protecting against large RCS LEAKAGE derives from the ability of instruments to rapidly detect extremely small leaks. This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that small leaks are detected in time to allow actions to place the plant in a safe condition, when RCS LEAKAGE indicates possible RCPB degradation.
The LCO is satisfied when monitors of diverse measurement means are available.
Thus, the containment sump level monitor, in combination with an N13/F18 gaseous activity monitor provides an acceptable minimum. Containment sump level monitoring is performed by two redundant, seismically quahfied level instruments.
l APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1,2,3, and 4, RCS LEAKAGE detection instrumentation is required to be OPERABLE.
In MODE 5 or 6, the temperature is < 200*F and pressure is maintained low or at atmospheric pressure. Since the temperatures and pressures are lower than those for MODES 1, 2, 3, and 4, the likelihood of leakage and crack propagation are much smaller. Therefore, the requirements of this LCO are r.ot applicable in MODES 5 and 6.
l 410.17-19 3 Westinghouse Rev.1 t
i l
1 I
i NRC REQUEST FOR ADDITIONAL INFORMATION
?
Sump level monitoring is a valid methodfor A10 DES 1, 2, 3 and 4. Containment atmosphere Nf 3/F18 monitoring is valid only in A10DE I with reactor power > 20%.
Reactor coolant system inventory monitoring via thepressuri:er level changes is valid in Af0 DES 1, 2, 3 and 4 only when the RCS conditions are stable; i.e. temperature constant. pressure constant, no makeup and no letdown.
The containment sump level change method ofdetecting leaks during Af0 DES 1, 2, 3 and 4 is not valid while containment purge occurs or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end of containment purge.
The containment atmosphere Nf3/F18 radioactivity leakage measurement during A10DE I is not valid while containment purge occurs or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end of containment purge.
l The contamment sump level change method ofdetecting leaks during Af0 DES 1, 2, 3 and 4 is not valid during extremely cold outside ambient conditions when frost is l
forming on the interior of the containment vessel.
l ACTIONS A.I and A.2 With the required containment sump level channel inoperable, no other form of !
sampling can provide the equivalent information; however, the containment atmosphere N/3/Fl8 radioactivity monitor will provide indications of changes in LEAKAGE. ]
Together with the atmosphere monitor, the periodic surveillance for RCS inventory !
balance, SR 3.4.8.1, must be performed at an increased frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to l provide information that is adequate to detect LEAKAGE. I Restoration of the sump channel to OPERABLE status is required to regain the function in a Completion Time of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> after the monitor's failure. This time is acceptable, considering the frequency and adequacy of the RCS inventory balance required by Action A.I.
l Required Action A.I is modified by a Note that indicates that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the containment sump channel is inoperable. This allowance is provided because other ;
instrumentation is available to monitor RCS Leakage.
B. I .1, B. I .2, and B.2 With one gaseous N13/F18 containment atmosphere radioactivity-monitoring instrumentation channel inoperable, alternative action is required. Either grab samples of the containment atmosphere must be taken and analyzed or RCS inventory balanced.
l l
i 410.17-20 W Westinghouse Rev.1 -
l8 ,
! l I
NRC REQUEST FOR ADDITIONAL INFORMATION r= g i
- E l
in accordance with SR 3.4.8.1, must be performed to provide alternate periodic information.
I 1
With a sample obtained and analyzed or an RCS inventory balance performed every l 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor may be an RCS operated for up to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> to allow restoration !
of the radioactivity monitor. j Re 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval for grab samples or RCS inventory balance provides periodic information that is adequate to detect LEAKAGE. The 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> Completion Time recognizes at least one other form of leak detection is available.
ACTIONS B.I.1, B.1.2, and B.2 (continued)
Required Action B.! and Required Action B.2 are modified by a Note that indicates )
that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the gaseous N13/F18 containment atmosphere radioactivity monitor channel is inoperable. This allowance is provided because other instrumentation is l available to monitor for RCS LEAKAGE. I I C.1 and C.2 If a Required Action of Condition A or B cannot be met within the required Completion Time, the reactor must be brought to MODE 4 where the probability and l consequences of an event are minimized. To achieve this status, the plant must be brought to at least MODE 3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to MODE 4 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner without challenging plant systems.
SURVEILLANCE SR 3.4.10.1 REQUIREMENTS ;
! SR 3.4.10.1 requires the performance of a CHANNEL CHECK of the containment !
l atmosphere N/3/Fl8 radioactivity monitor. The check gives reasonable confidence J that the channel is operating properly. The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on ,
instrument reliability and risk and is reasonable for detecting off normal conditions. I l- SR 3.4.10.2
(
SR 3.4.10.2 requires the performance of a CHANNEL OPERATIONAL TEST (COT) on the atmosphere N13/F/8 radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. The test verifies the alarm setpoint 410.17 21 T Westinghouse gey,1
s !
l NRC REQUEST FOR ADDITIONAL INFORMATION
- 4,.
i l and relative accuracy of the instrument string. The Frequency of 92 days considers risks and instrument reliability, and operating experience has shown that it is proper for detecting degradation.
l SURVEILLANCE SR 3.4.10.3 and SR 3.4.10.4 )
REQUIREMENTS l (continued) These SRs require the performance of a CHANNEL CALIBRATION f, .un of the RCS Leakage detection instrumentation channels. The calibration verities the accuracy of the instrument string, including the instruments located inside containment. The I Frequency of 24 months is a typical refueling cycle and considers channel reliability.
Again, operating experience has proven that this Frequency is acceptable. !
l REFERENCES 1. 10 CFR 50, Appendix A, Section IV, GDC 30. l
- 2. Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary LEAKAGE i Detection Systems," U.S. Nuclear Regulatory Commission.
- 3. AP600 SSAR Chapter 15, " Accident Analysis."
i l
410.17-22 W Westinghouse Rev.1 -
i I
. l l
Nydes, Robin K. _ _ - - . _ _ _ _ - - - - ----
l From: Corletti, Michael M. l Sent: Friday, April 25,1997 9:22 AM l To: N es, Robin K. .
Subject:
R : Telecon for ASI/ ERG l 1
l i
l d
l l
l l
1 From:
I Sent: NA... Robin K.
J.7, " - 23,1997 4:33 PM !
To: 'wdh@nre gov' '
Cc: Corletti. M .
Subjec'c or ASI/ ERG l
Bill, I just faned you an advance copy of the Westinghouse response to the two ERG /ASI letters. And can we l have that telecon with Jim Higgins on Letdown isolation during S/D operations Thursday (4/24) aftemoon i sometime? Mike and I are open so just let me kr.ow what time and you can call us on x4871 (room 3258).
Thanks, Robin 1
l l
l I
i i
i a
Page1 1
l' l Nydes, Robin K. . _ _ _ _
From: es,Ro K.
Sent: T esday, ril 29,1 7 2:31 PM To: 'wch@nrc. gov' Cc: Suggs,Cha .; Nydes, Robin K.; Carlson, William R.; Carlin, Edward L.; Corletti, Michael M.;
rry L.; McIntyre. Brian A; Winters, James W.; Deutsch, Kenneth L.
Subject:
Reacto'r Systems Branch Tech Spec Telecon i
Bill-I got your fax listin which T/Ss SRXB is ready to discuss. I suggest the following format for our telecon tomorrow (April 30) but ave not heard back from everyone:
10:00 Robin Nydes and Chip Suggs LCOs 3.0.3, Bases p 3.0-10, and Spec 5.6.5 (COLR) 10:15 Bill Carlson LCOs in 3.1 and 3.2 (Bill has not confirmed his availability yet).
, 10:30 Ed Carlin LCOs 3.4.1 and 3.4.2 (Ed has not confirmed his availability yet).
l 10:45 Mike Corletti and Ed Carlin LCOs 3.4.3 through 3.4.9 11:05 Mike Corletti LCO 3.9.4 (and can we cover 3.9.2 also?)
11:15 Terry Schulz LCO 3.4.12-14 and all LCOs from 3.5 You also listed LCO 3.3.2 (ESFAS) and 3.4.10 (RCS Leak Detection). These are I&C Branch and Plant Systems Branch responsibilities, and I think we should cover them in the calls with Hulbert (do you have a date for this?)
and with SPLB on May 8. At the end of our telecon tomorrow, Chip and I can talk with you and Gene about his comments related to these T/S. If we can't discuss them to NRC satisfaction, I suggest we continue it in the I&C and SPLB telecons.
You can call us at 10.00 on April 30 on 412-374-3860 (that's room 322 here). Robin l
l l
[
Page1
l Nydes, Robin K. - . _ _ . _
From: No- " % K.
Sent: nesday, April 30,1997 4:16 PM To: 'wch@nrc. gov'
, Cc: McIntyre, Br' , inters, James W.; Suggs, Charles W.; Schulz, Terry L.; Nydes, Rob!n K.
l
Subject:
my 4i50 h econ notes I
l l A telecon with the Reactc w nums Branch (SRXB) was held today to discuss Westinghouse responses to SRXB Tech Spec comments. * : are dispositioned as follows:
l l The responses to questions 1,2,3,4, 5,8,10,11,18,19,20 are ok (for some, Westinghouse needs to fax the '
spec markup).
Three of these are NRC action:
To close 6 (3.2.1), Summer Sun is to confirm acceptability.
To confirm closure of 14 (3.4.13 and 3.4.14), NRC has action to review the SD Evaluation Report.
To close 24 (3.4.14), NRC will re-evaluate to understand Westinghouse position that STS basis for PlVs was followed.
30 will be discussed during the Plant Systems Branch telecon on May 4.
The others are Westinghouse action, big one first:
To close 7,13,16,17 (and all other TS applicable in Mode 4 but ending in Mode 4), Westinghouse will develop restoration times and replace the new action to " initiate action to restore.. " with an action "to restore.. "
l This is BIG.
l To close 21, Westinghouse needs to act to resolve the squib valve operability issues. This discussion expands I
also to check valves. This is BIG. A small part of 21 that I didn't want to get lost is to recognize the ADS stages are not equivalent and break ouTlhe actions, consistent with our approach developed April 9 & 10.
To close 12 (TS 3.4.10), Westinghouse will provide the TS markup with Rev 1 of the response to RAI 410.17. I'm l not sure this will close so it will take some effort to " finish" this.
l To close 9 (3.4.3), Westinghouse will determine an appropriate pressure and fax a spec markup to NRC for mat'l branch review.
To close 15 (3.5.1), Westinghouse needs to set a nitroger prr.,-are limit and discuss CTr/SFs, including defense of Action D as more conservative than LCO 3.0.3.
1 1 To close 22 (3.4.12), Westinghouse will revise the response to eliminate reference to NRC/ Westinghouse tem.:%n I and to state that these vacuum breakers are not assumed in the safety analysis. Note to self - what are they ir?
To close 23 (3 5.2), Wen,tinghouse is to develop an action plan to address apparent inconsistencies between the TS and SSAR p 3.9-163 foot note 10.
This is just a quick rundown of overall approaches to finish these Specs. Robin i
t Page 1 l
i t
l Nydes, Robin K.
From: Wint ,
Sent: T sday, May 06, 99710:30 AM To: ch obin K.
Cc: chm @nrc. "'i,= gov'*
.: Winters, James W.
Subject:
atus for OITS Items 224 and 233 As indicated in OITS for these items, they were discussed with the reviewer (and Jackson and Quay) on Thursday, May 1, and they are open awaiting the issue of our RTNSS WCAP. We expect no further action, but we must show NRC that these two systems didn't make the RTNSS cut. Although a doubie negative situation here, we (Quay and 1) decided to carry these items for completeness. If they bother you because they are labeled "RTNSS", we can change them back to " Winters" and Chapter 9.
Jim x5290 From: Nydes, Rocen K.
sont: Fnday, May 02,1997 5 31 PM To: 'wch@nre gov' Cc: Winters, James W.; Chase, Wayne L.
subject: Status for OITS Items 224 and 233 Hi Bill. I still have OITS items 224 and 233 statused " Action W' for both Westinghouse and NRC. Quite frankly, I have no idea what is required to close these. They seem to be tied to RTNSS but I'm not sure that's a real tie, i.e., I think we should just work to close them assuming RTNSS is resolved.
I assumed these items were yours since they are RTNSS but they may really be SSAR Chapter 9 issues (Are you the PM7). Anyway, I broke these items down as follows.
Please let me know what NRC needs to consider these resolved (if anything). Thanks, Robin 224a & b - Closed 224c&d - I don't think Westinghouse has an action here. Please confirm we addressed this in SSAR Ch 9.
224e -is this NRC action?
224f&g - We don't plan to do anything else for this.
224h - Closed 2241-!s this NRC action?
233a & b - Closed 233c & d - See 224c & d 233e -is this NRC action?
233f & g - See 224f & g 233h - Closed 2331 -Is this NRC action?
Page1
Nydes, Robin K. - - - _ _ - - . . _
From: Nydes, Robin K.
Sent: Frida r 5,1997 9:08 AM To: s, Char!as ; McDermott, Daniel J.; Sejvar James X.; Grover, James L.; Corletti, Michael M.; Israelsoa, Gv n A.; Wills, Mark E.; Nydes, Robin K.
Cc: 'wch@nrc.go '
Subject:
Tech S on wR5 NRC Plant Systems Branch l
Good morning. Are you available to attend some portion of a telecon on the morning of May 8 with the NRC Plant l Systems branch? We are not certain today which Specs are on the agenda but the Plant Systems branch is
- responsible for the following
3.7.1 & 2 Main Steam / Main Feed Dan McDermott (3.7.3 and 3.7.7 are done!)
3.7.4 Secondary Activity Jim Sejvar/ Jim Grover 3.7.5 Spent Fuel Pool Level Mike Corletti/Gordon Israelson 3.7.6 Control Room Habitability Mark Wills 3.9.2/3.9.4 Refueling Operations Mike Corletti Please let me know your availability on the morning of May 8 (a reply netmail is sufficient) and l*ll work with the NRC to set up an agenda, with times for each Spec discussion.
Thanks, Robin l
l l
Page1
i l Nydes, Robin K. _
From: Nydes, Robin K. C Sent: Monday, April 28,199711:24 AM To: ' david.w. bland @snc.com'; Suggs, Charles .; 'wch@nrc. gov'; ydes, Robin K.
Cc: McIntyre, Brian A
Subject:
Tech Spec 5.0 Telecon Please plan to attend a telecon for 2:00pm on Tuesday, May 6 with NRC's Bill Huffman and Jim Bongarra to discuss Tech Spec 5.0.
If NRC will call us on x4871 (room 325B), we'll tie in David Bland (205)992-6697.
l I
i l
l i
I I !
I l
i l
i 1
Page 1 l
l
~
Brian A. McIntyre,12:53 PM 5/5/9? ,5/6 RTNSS mig attendees j L D' on,05 May 99712:53:33 -0400 l o: wch@nrc.go l
l F "B . McIntyre" <mcintyba@wesmail.com>
l
Subject:
5/6 RTNSS mtg attendees l
i l Cc: TRQ@NRC. GOV, meintyba@wesmail.com. vijuktm@wesmail.com, haagel@wesmail.com, schulztl @ westinghouse.com, jms3 @ nrc. gov Bill Huffman, l The attendees for the RTNSS meeting tomorrow will be myself, Cindy Haag and l Terry Schulz. Please ensure we are preregistered.
My expectation for tomorrow's meeting is that we will cover the three areas we discussed on April 3.
- 1. NRC review of proposed regulatory oversights
- WCAP-13856 (September 1993)
- Post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> actions letter of March 14,1997 l Are these appropriate? !
! i
! Where should they be captured since not in SRP, Reg Guide 1.70?
l l
I
- 2. NRC thoughts on how proposed additional oversight would resolve issues on the table Thermal Hydraulic uncenainty Long term cooling l
l Acceptability of focused PRA Acceptability of baseline PRA
- 3. Westinghouse thoughts on what additional systems should be included Brian A. McIntyre l Bell 412.374.1334 WIN 284.4334 FAX Bell 412.374.4887 FAX WIN 284.4887 1 i
j l!
~
~~ ~~ ~
iPrinted for" Brian A McIntyref<incintyba@wesmad.com>_. _ [~ ~ _
- e FAX to DINO SCALETTI
, May 1,1997 CC: Sharon or Dino, please make copies for: Bill Huffman Ted Quay Don Lindgren -
Bob Vijuk Brian McIntyre OPEN ITEM #1025 (DSER 6.5.3-1)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
- Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 31 calendar days away (23 business days). The relevant documentation related to Open Item #1025 (DSER 6.5.3-1) is in Section 6.5 of the SSAR. The material included in this section of the SSAR provides the satisfaction of this action item; other questions on fission product control may be addressed as part of the review of Section 6.2 and are not pertinent to Section 6.5. The pertinent pages of the SSAR are attached. It is requested NRC review this material included herewith and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend
" Action N" or " Closed."
Jim Winters 412-374-5290
AP600 Opea It;m Tracking System Database: Exec tiv2Snaam ry ' Dat:: 4/2"/97 '
Selecties: [ item nol between 1025 And 1025 Sotted by Type -
Item DSER Section Titic/ Description Resp NRC (W) i No Branch Question Type Detail Status Engineer Status Status Letter No. / Date ~
1025 NRR/SPLB 653-1 DSER-Oi 1.indgren/ Butler Closed . Action W
~ ~ -
~
The staffs acceptance of'the fission pro 5ucIcontrol systems and s'tdurN$ilfbe EEd on Ecmclusions of the reids~addres$d in SectA '
(6312_ 62;M I5j of gis report., _ , _ , _ _ _ , _ _ _ , _ _ _ . _ , _ _ _ _ _ _ _ _ _ _ _ , _ _ , _ _ t
~ , ~~ _ _ _ . _~,__,___
I
~ ~~ ~ '
Close'd7Wstmghouse luis compl53 nEds~ut' u nitsis to supdsta5e~w.
f Action W - Although acticus on 6.5 are complete, Westinghouse has yet to complete all of t6.2 and 3 ative 15.4 product controls.,
to fission L
i N
A=.
<g &
l 6
i I
i Page: 1 Total Records: 1 i
i
- 6. Engineer:d sarety Feztures l
t l 6.5 Fission Product Removal and Control Systems 1
6.5.1 Engineered Safety Feature (ESF) Filter Systems i
This subsection is not applicable to the AP600 6.5.2 Containment Spray System in the event of a core degradation accident there could be a significant quantity of radioactivity released to the containment atmosphere. This activity would consist of noble l'
1 gases, particulates, and a small amount of elemental and organic iodine (as discussed in I
subsection 15.6.5.3. most of the iodine would be in the particulate form), The AP600 does l 1 not include a containment spray system to remove airborne particulates or elemental iodine.
1 Removal of airborne activity is by natural processes that do not depend on sprays (that is, I sedimentation, diffusiophoresis, and thermophoresis). These removal mechanisms are discussed I in subsection 15.6.5.3.2, Much of the non-gaseous airborne activity would eventually be deposited in the containment I
sump solution. Long term retention of iodine in the containment sump requires adjustment of the sump solution pH to 7.0 or above. This pH adjustment is accomplished by the passive
- core cooling system and is discussed in subsection 6.3.2.1.4. '
6.5.3 Fission Product Control Systems l The containment atmosphere is depleted of elemental iodine and particulates as a result of the I
passive removal processes discussed in SS AR subsection 15.6.5.3.2. No active fission product I control systems are required in the AP600 design to meet regulatory requirements. The i passive removal processes and the limited leakage from the containment result in offsite doses I less than the regulatory guideline limits.
6.5.3.1 Primary Containment The containment consists of a freestanding cylindrical steel vessel with ellipsoidal heads. The I containment structural design is presented in subsection 3.8.2.
The containment vessel, penetrations, and isolation valves function to limit the release of radioactive materials following postulated accidents. The resulting offsite doses are less than I regulatory guideline limits. Containment parameters affecting fission product release accident analyses are given in Table 6.5.3-1.
Long-term containment pressure and temperature response to the design basis accident are presented in Section 6.2.
l l
The containment air filtration system may be operated for personnel access to the containment I when the reactor is at power, as presented in subsection 9.4.7 For this reason, the radiological l
Revision: 6 Westingh00S8 3 March 29,1996 6.5-1 e
o g iii
- 6. Engineered Szfety Fe:tures I
assessment of a loss-of-coolant accident assumes that both trains of the air filtration system are in service at the initiation of the event. The isolation valves receive automatic signals to close from diverse parameters. The valves are designed to close automatically as described in subsection 6.2.3.
Containment hydrogen control systems are presented in subsection 6.2.4.
6.5.3.2 Secondary Containment There is no secondary containment provided for the fission product control following design basis accident.
The annulus between containment and shield building from the elevation 100'-0" to the elevation 132'-3" acts as a holdup volume to limit the spread of fission products following severe accident. Most containment penetrations are located within this holdup volume. It is served by the radiologically controlled area ventilation system (VAS) described in subsection 9.4.3. Isolation dampers are provided to reduce the air interchange between the holdup volume and environment. Fission product control via holdup within the annulus is considered in severe accident dose analysis but excluded from consideration for design basis accidents dose evaluations presented in Chapter 15 I 6.5.4 Combined License Information I :
1 This section has no requirement for additional information to be provided in support of the I Combined License applications.
l I
l
(
Revision: 7 3 q April 30,1996 c t
(rf { 6.52 ,
W8Stingt100S8 l j
4
/W-. 4
- 6. Erigineered Safity Features l
Table 6.5.3-1 l
PRIMARY CONTAINMENT OPERATION l
FOLLOWING A DESIGN BASIS ACCIDENT Type of structure . . .. . .. .. ..... .. ............ Freestanding cylindrical steel vessel with ellipsoidal heads 3 . 1.73 x 106 Containment free volume (ft )....... .. ... . . . . . ......
Containment leak rate . . . . . . . . .... .. . . . . . 0.12% containment volume per day,0-24 hours 0.06% containment volume per day,1-30 days l
i I
l l
l l /
Revision: 6
[ W95tifigh0088 /
6.5-3 March 29,1996
)u
O FAX to DINO SCALETTI i i
May 5,1997 i
CC: Sharon or Dino, please make copies for: Diane Jackson ;
Ted Quay Don Lindgren Bob Vijuk Brian McIntyre OPEN ITEM #562 (DSER 3.2.1-1) 4116 (RAI #260.83) i To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
l Documentation" by 5/97, we believe that NRC must acknowledge receipt of all
- Westinghouse submittals by May 30,1997. This is just 25 calendar days away (20 business ;
l days). The relevant documentation related to Open items #562 (DSER 3.2.1-1) and 4116 (RAI
- 260.83) is in the SSAR in Sub-Section 3.2.1.1.2 and in Tables 3.2-1 and 3.2.2 (supplied to you some months ago). The pertinent pages of the SSAR are attached. Additionally, these items ,
were discussed in Westinghouse letters NSD-NRC-96-4841, dated October 14,1996 and NSD- '
I NRC-97-4993 dated February 21,1997. Appropriate pages of these letters are included. It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 l .
l 1
W
AP600 Opea lum Tracking Sysum Datbase: Execstiva Srmm ry Datz: 5/5/97 Selection: - [ item noj between 562 And 562 Sorted by Type item DSER Section litic/ Description Resp NRC (W)_
No. Hranch Question 1_3 pe Detail Status Engineer . Status Status _ _ Le_tte_r No,. / .
_. ._ _ _ . _ _ . _ . . _ _ _ _ _ _ _ . _ D. ate _
562 NRR/l MI'B 3.211 DSER-OI Lindgren Closed Action W NSD-NRC-96-4841
~
l Westinghouse should apply the pertinent quality assurance requirements'of Appendix IItol0 CI'R 50 to all Seismic Caaegory Il SSCs.' A t1 commitment to this effect should be added to Section 3.2.1.1.2 and Table 3.2-1 of the SSAR.-
~ ~~ ~~ ' " ~_ _. .. _ . . . _' ~ ~~ ~ ~~~~
Closed - Statement added to seismic' Category !! requirements for QA }
f Action W - The staff does not agree. The pertinent QA requirements of Appendix H should be applied to all Seismic Category Il structaes, sys
!and components. This commitment should be added to SSAR Section 3 2.1.1.2 and Table 3.2-1.
jResolved - See response in truer NSD-NRC 96-4841, dated October 14,1996. Scismic Category Il QA will be the same as the QA for RINSS. <
Action W - The resolution of this issue is pending the staffs evaluation of responses to RAls 260.83 and 260 87.
l Closed j _ - Last response provided by NSC-NRC-97-4993 of 2/21/97.__
t
)
N 2
L 5
Page: 1 Total Records: I
- _ - - - . . _ _ _ - - _ _ _ _ . - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ - _ . . -- - - - .- . ~ . . -
AP600 Opea item Trackirg System Dr.t:_. base: Execrtive Szmmcry Dat : 5/5/97 Selection: [ item no] between 4II6 And 4116 Sorted by Type item DSL R Section Titic/ Description Resp NRC (W)
No. 11 ranch Questnm Typc Detail Status Engineer Status Status Ixtrer No / Date 4116 NRR/IlQMil 32.1 RAl4)I RlhSS/Khrs Closed Action W NSD-NRC-97-4993 RAl# 260 83 is it Westinghousc's position that RP C.4 of Regulatory Guide (RG) 1.29 is incongruous with the " concept of graded QA"? Also,
'please explain what Westinghouse's " concept of Graded QA" is and whcre that concept is defined in the standard safety analysis report (SSAR)
' ~ ~ _ ~
'~ ~
Closed - See letter NSD-NRC-97-4993 dated 2/21/97. l i _ _ - ___ _ . . _ _ . _ . . _. _ .I N
1,.-
=;
l' age. 1 Total Records: I
l
- 3. Design of Stmetures, Components, Equipment and Systems
- Capability to prevent or mitigate the, consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100.
Seismic Category I structures, systems, and components are designed to withstand the appropriate seismic loads, as discussed in Section 3.7, and other applicable loads without loss of function. Seismic Category I structures are sufficiently isolated from non-Category I
- structures, or non-seismic Category I stmetures are designed to Category II requirements so that the structural integrity of the non-seismic Category I structure is maintained during the postulated safe shutdown earthquake (SSE).
Seismic Category I structures, systems, and components meet the quality assurance requirements of 10 CFR 50, Appendix B. The criteria used for the design of seismic Category I structures, systems, and components are discussed in Section 3.7.
3.2.1.1.2 Seismic Category II (C-il)
Seismic Category II applies to plant structures, systems, and components which perform no safety-related function, and the continued function of which is not required. Seismic Category II applies to structures, systems, and components the structural failure of which during a safe shutdown earthquake or interaction with seismic Category 1 items could degrade the functioning of a safety-related structure, system, or component to an unacceptable level, or could result in incapacitating injury to occupants of the main control room.
Seismic Category II structures, systems, and components are designed so that the safe shutdown earthquake does not cause unacceptable structural failure of or interaction with seismic Category I items. Seismic Category II fluid systems require an appropriate level of pressure boundary integrity if located near sensitive equipment.
The criteria used for the design of seismic Category II structures, systems, and components are discussed in Section 3.7.
10 CFR, 50, Appendix B does not apply to seismic Category II structures, systems, and components. Seismic Category II structures, systems, and components have QA requirements similar to those defined for structures, systems, and components covered by regulatory treatment of nonsafety systems (see Section 17.3). The quality assurance requirements for Seisme Category II structures, systems, and components are sufficient to provide that these components vill rneet the requirement to not cause unacceptable structural failure of or interaction with senmic Category I items.
l 3.2.1.1.3 Non-Seismic l
t Non-seismic (NS) structures, systems, and components are those that are not classified seismic Category I or Category II.
The criteria used for the design of non-seismic structures, components and systems are discussed in Section 3.7.
5 i Revision: 11
! February 28,1997 3.2-2 W Westingt10tlS8 l
i
Table 3.2-1 P
[
COMPARISON OF SAFETY CLASSIFICATION REQUIREMENTS E
- E 6
g AP600 ANS Equip- 10 CFR 50 Inspection
?~
3 9 Code ment Safety RG 1.29 Seismic ASME Code, RG 1.26 NRC Appendix & Testing Required 2 E Letter Class Design Reqmnts See. III Class IEEE Re- Quality Group B Require- Test & h N (1) (2) (3) (4) quirements (5) (6) ments Maint. j A SC-1 I I NA GROUP A YES YES(7) (8) p B
C SC-2 SC-3 I
I 2
3 NA IE GROUP B GROUP C YES YES YES(7)
YES(7)
(8)
(8) f R D
- D NNS(2) NA(9) NA(10) (10) GROUP D NO(10) YES(ll) (II) F
\ OTilER NNS(2) NA(13) NA NA NA NA(12) NA NA E fh li G-L ' NA - Not Applicable OTilER includes Classes E, F. L. P R, and W.
F Notes: g Y 1. A single letter equipment classification identifies the safety class, quality group, and other classifications for AP600. See the subsection 3.2.2 c.
ro G for definition. D
- 2. AP600 safety classification is an adaptation of that defined in ANSI 51.1. The NNS defined in the ANSI 51.1 standard is divided into several #
AP600 equipment classifications namely, Classes D E F, L, P, R, and W. 5
- 3. See subsection 3.2.1 for definition of seismic categories.
- 4. ASME Boiler and Pressure Vessel Code, Section 111 defines various classes of structures, systems, and components for nuclear power plants.
It defines criteria and requirements based on the ci-.:;ification. It is not applicable for nonsafety-related components.
- 5. The guality group classification corresponds to tFose provided in Regulatory Guide 1.26.
- 6. "Yes means quality assurance program is required according to 10 CFR 50 Appendix B.
"No" means quality assurance pmgram is not r . quired according to 10 CFR 50 Appendix B.
- 7. Class A B, and C, structures, systems, and compcnents built to ASME Code,Section III are inspected to ASME Code,Section XI requirements.
See the text for additional specification of requirements.
- 8. Class A, B, and C structures, systems, and comjonents that are required to function to mitigate design base accidents have some testing requirements included in the plant technical spccifications. In addition to the requirements in the technical specifications, testing and maintenance requirements are included in an administratively controlled reliability assurance plan.
- 9. See subsection 3.2.1 for cases when seismic Category Il requirements are applicable for Class D structures, systems, and components.
er
- 10. See the text for a discussion of the industry standards used m the construction of Class D structures, systems and components.
- 11. Class D structures, systems, and components have selected reliability assurance programs and procedures to provide availability when needed.
2 These programs are administratively controlled programs and are not included m the technical specifications.
J$g 12. Normal industrial procedures are followed in procuring, designing, fabricating, and testing these nonsafety-related structures, systems, and components.
p 13. Some Class E, F. L P, R, and W structures, systems, and components may be classified as seismic Category II. See subsection 3.7.3.
Q d
- ?,
4 3::
l-l
- 3. Design of Structures, Components, Equipment and Systems l
1 C'
t l
l Table 3.2-2 l l
SEISMIC CLASSIFICATION OF BUILDING STRUCTURES I l
l Structure Category l l
l Nuclear Island C-I l Basemat l
! Containment Interior Shield Building Auxiliary Building Containment Air Baffle I Containment Vessel C-I Plant vent and stair stmeture C-II Turbine Building NS i Annex Building Columns A - F NS Annex Building Columns F - I C-II Radwaste Building NS ,
Diesel-Generator Building NS Circulating Water Pumphouse and Towers NS C-I- Seismic Category I C-II - Seismic Category II NS- Non-seismic Note: -
- 1. Within the broad definition of seismic Category I and II structures, these buildings contain members and structural subsystems the failure of which would not impair the capability for safe shutdown. Examples of such systems would be elevators, stairwells not required for access in the event of a postulated earthquake, and nonstructural partitions in nonsafety-related areas. These substructures are classified as non-seismic.
l
[o (._
l Revision: 11 February 28,1997 3.2-20 W Westingt100S8 1
i
l l
Westinghouse Energy Systems Bo 355 P*5Sp P"a5YN8"'s n23nass Electric Corporation NSD-NRC-96-4841 DCP/NRC0625 Docket No.; STN 52-003 October 14, 1996 Document Control Desk l U.S. Nuclear Regulatory Commission l Washington /D.C/ 20555 ATTENTION: T.R. QUAY
SUBJECT:
RESPONSES TO NRC MECHANICAL ENGINEERING BRANCH l QUESTIONS l
Dear Mr. Quay:
l Attached are responses to a number of items from the NRC Mechanical Engineering Branch discussed in a telephone call with the NRC staff on October 3,1996. The synopsis of the NRC l
position comes from an NRC letter dated August 20,1996. The questions are related to quality
! assurance, reactor vessel internal vibration, CRDMs, and equipment seismic qualification. The -
questions are identified by the numbers from the Attgust 20,1996 letter, DSER open item or RAI number, and Open items Tracking System item number. This response completes our responses to questions related to subsections 3.2.1, 3.9.2, 3.9.4, 3.9.7, and Section 3.10.
This submittal will permit the completion of staff review for the subsections listed and preparation of the Final Safety Evaluation Report input.
Please contact Donald A. Lindgren on (412) 374-4856 if you have additional questions, ff h AfM Brian A. McIntyre, Manager Advanced Plant Safety and Licensing
/nja Attachments .
cc: D. T. Jackson - NRC N. J. Liparulo - Westinghouse (w/o attachments) 4 3
M55A Jaklo honc9 ?
l l Attachment to NSD-NRC-96-4841 l
- 1. Open Item 3.2.1-1 (562) - Appendix B for all Seismic Cat.11 - Regulatory Guide (RG) 1.29, l Position C.4
- In the Open Item Tracking System Database (OITS), Westinghouse reports that this issue is closed i based on a statement added to Seismic Category Il requirements for Quality Assurance (QA). This l information was added to SSAR Section 3.2.1.1.2 in Revision 7, and states that 10 CFR 50, 1 Appendix B does not apply to Seismic Category !! structures, systems, and components. The staff does not agree. As stated in the DSER for this open item, to satisfy Position C.4 in RG 1.29, the l l pertinent QA requirements of Appendix B should be applied to all Seismic Category 11 structures, I l systems, and components. This commitment should be added to SSAR Section 3.2.1.1.2 and Table l 3.2-1. Therefpre, Open Item 3.2.1-1 remains open.
Response
l ,
When the guidance in Regulatory Guide 1.29 position C.4. was developed, the concept of
( graded QA had not been developed. An Appendix B quality assurance program is not needed to provide that seismic Category II systems, structures, and components do not fail in a manner that would reduce the functioning of a safety-related component. The degree of quality assurance provided for AP600 equipment Class D provides an appropriate level of quality assurance for this function. Westinghouse has defined quality assurance requirements l for the regulatory treatment of nonsafety systems, systems, and components. Those
! requirements also are sufficient to satisfy regulatory requirements for seismic Category II.
Revise the fourth paragraph of 3.2.1.1.2 as follows:
10 CFR, 50, Appendix B does not apply to seismic Category II structures, systems, and
- components. Seismic Category 11 ' structures, systems, and components have QA requirements l similar to those danned for structures, systems; and components covered by regulatory treatment of nonsafety systems (see Section 17.3)RThe quality assurance requirements for
. Seismic Category 11 structures, systems, and components are sufficient to provide that these components will meet the requirement to not cause unacceptable structural failure of or interaction with seismic Category I items.
This item is Resolved pending formal SSAR revision.
- 2. Open Item 3.2.1-2 (563)- Appendix B for new and spent fuel storage racks in the OITS, Westinghouse reports that this issue is closed because SSAR Table 3.2-3 lists the new
, and spent fuel storage racks u Seismic Category I, and as such are required to meet applicable .
j portions of Appendix B. The staff agrees that to meet RG 1.29, Appendix B should be applied to I
these components. However, since the new and spent fuel storage racks are classified as AP600 Class D, it is possible that this commitment might be misinterpreted when one consults SSAR Table 3.2-1.
According to this table, AP600 Class D components do not have to rneet either RG 1.29 seismic i'
design requirements or Appendix B. Table 3.2-1 should be clarified by adding a note to state that although the new and spent fuel storage racks are Class D, they are designed as Seismic Category I, and meet the applicable QA requirements of Appendix B.
N 7c g r .
! i l*
e . , 3e
-c l
Westinghouse Energy Systems gj55g Electric Corporation ,,g ,,3 ,33 NSD-NRC-97-4993 DPC/NRC0747 Docket No. STN-52-003 -
l l February 21,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 TO: T.R. QUAY
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600.
Dear Mr. Quay:
Enclosed are three c'opies of the Westinghouse responses to open items on AP600 topics. Responses to nine RAls are included in this transmittal. RAI 410.261 provides information on Section 9 of the
-SSAR. Responses to RAI 440.571, Revision 1, discusses the OSU Test Arnlysis Report. Responses to RAls 260.83, 84, 85, 86, 87, 88, and 89 address questions on Section 3 of the SSAR.
The NRC technical staff should review these responses as a part of their review of the AP600 design.
l These responses close, from a Westinghouse perspective, the addressed questions. The NRC should inform Westinghouse of the status to be designated in the "NRC Status" column of the OITS.
l Please contact Brian A. McIntyre on (412) 374-4334 if you have any questions concerning this l transmittal.
A.
Brian A. McIntyre, anager L
Advanced Plant Safety and Licensing Jml Enclosures cc: T. Kenyon, NRC - (w/o enclosures)
W. Huffman, NRC - (w/ enclosures)
N. Liparulo, Westinghouse - (w/o enclosures) 5 (7 i 9m
""^
i
i l'
I a
ENCLOSURE TO WESTINGHOUSE i LETTER NSD-NRC-97-4993
- I l FEBRUARY 21,1997 l
! I l
l I
1 l
t 1 i
I i l I l
l I
l 1
l h
m
- - . - .~ . -.
- l l
I 6
NRC REQUEST FOR ADDITIONAL INFORMATION m.
l Question 260.83 i
Re: OITS 4116 is it Westinghouse's position that RP C.4 of Regulatory Guide (RG) 1.29 is incongruous with the " concept of graded QA".? Also, please explain what Westinghouse's " concept of Graded QA" is and where that concept is defined in the standard safety analysis report (SSAR).
Response
The " concept of a graded QA" is the equipment ranking system defined in SSAR Section 3.2, " Classification of Structures Components, and Systems." The seismic classification complies with the criteria established in Regulatory Guide 1.29. Seismic Category I requirements are applied to safety-related equipment as defined in Subsection 3.2.1. A safety classification designation is defined, as well as the use of codes and standard.s. that conforms to the requirements of 10 CFR 50.55a. Safety-related components or systems are rated Class A, B, or C as defined in Subsection 3.2.2.2. Class A, B, and C components are treated as Seismic Category I and conform to l the requirements of 10 CFR 50 Appendix B. Components designated as Class D are nonsafety-related equipment I that will meet requirements established by industrial quality assurance standards like the ASME Boiler and Pressure j Vessel Code, Section Vill.
l l
l SSAR Revision: NONE l
l l
i n
260,83
[ Westingflouse
a FAX to DINO SCALETTI May 5,1997 CC: Sharon or Dino, please make copies for: Diane lackson Ted Quay Don Lindgren Bob Vijuk Brian McIntyre OPEN ITEM #1797 (DSER 3.9.2.4-1)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 25 calendar days away (20 business days). The material regarding the Japanese drop tests of control rods during seismic events has ,
been deleted from the SSAR as excessive detail for the SSAR. The relevant documentation !
related to control rod insertion and seismic events is included in Sub-Section 3.9.4.3 of the SSAR. The material included in this section of the SSAR provides the satisfaction of this action item.. The pertinent pages of the SSAR are attached. It is requested NRC review this material l included herewith and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed." ,
1 1
Jim Wmters l I
412-374-5290 i
i l
l l
t f
W i
- . -.v. ;
AP600 Open Itzm Tracking Syst:m D:tsbase: Exec tive S:mmary D:ts: 5/5/97 !
Selectio2: [ item noj between 1797 And 1797 Sorted by Tyoc r
l'e'n ' DSER Section Titic/ Description Resp NRC (W)
No. Branch Question Typc Detail Status . Engineer . Status Status
~. - - -. . - . - . - - _- _ - . . , . letter No. /
- - - c. D. ate ..
l y
1797 NRR/EMEB 392.4-4 DSER-CN Lindgren - Closed Action W NTD-NRC-95-4464
~
{3 9 2 4-4 Westinghouse should revise S the ' SAR as noted in Section 59.2[4 of this repEt dncorporane SSARicvision frorn'RAlil0[94, ,
l description
- rfor : CRDM
= :: = tests ',: :: ) . : :: . .: .. := :: :.: . : :::_z:: L == ; ...
t iClosed - This information to be deleted from SSAR see open item 785, DSER item 3.92.4-1
- 2: :.:= :: . . . ---].
[
T i
I t
t L. !
r N h i
l
-1 I
I t
i i
i t
l' age: I Total Records: I I
l r l
- r. . .__.t.-
- 3. Design of Structures, Components Equipment, and Systems The coil stack assembly is located outside the pressure housing. The assembly does not come in contact with the reactor coolant and does not have any pressure-retaining function.
The operating temperature of the coils is maintained below 392*F.
The coil stack assembly slides over the pressure housing and remains in place without a permanent mechanical or welded attachment. The assembly clearances permit removal of an assembly even when the control rod drive mechanism is at normal operating temperature.
Thus, a malfunctioning coil assembly could be replaced without a complete cooldown of the plant. He clearances between the coil and coil housing are selected to minimize the gap at normal operating temperature to facilitate coil cooling.
3.9.4.3 Design Loads, Stress Limits, and Allowable Deformations The pressure housing ponion of the control rod drive mechanism is a Class I component required to meet the requirements of AShE Code, Section M. Subsection 3.9.3 defines the loading combinations considered ir the evaluation of AShE Code, Section m, pressure boundary components.
For each loading combination, the appropriate stresses due to pressure, component weight, extemal loads, hydraulic forces, thermal gradients, and seismic dynamic forces are evaluated and demonstrated to be less than the applicable stress limits. He cyclic stresses are combined with constant stresses to evaluate the fatigue usage due to cyclic loads. The transients used in the evaluation of cyclic loads are described in subsection 3.9.1. The effect of seismic events is addressed by considering a seismic event with an amplitude equal to one-third of the safe shutdown canhquake evaluated as a Level B event. He seismic l
contribution to the fatigue evaluation is based on five seismic events with an amplitude of one-third the safe shutdown earthquake and with 63 cycles per event. He results of the stress evaluation are documented in a component stress repon, as required by the ASME Code.
ne control rod drive mechanism is supponed by the attachment of the bottom of the assembly to the reactor vessel head and a connection to the integrated head package at the top of the rod travel housing. The integrated head package also provides the support to the cooling air shrouds and control rod drive mechanism electrical supply cables to prevent excessive loading on the control rod drive mechanisms during seismic events.
Hydrostatic tests according to the requirements of the ASME Code verify the pressure boundary integrity of the pressure housing prior to operation. The latch assembly housing is assembled to the reactor vessel head by the vessel supplier and is hydro tested as part of l the vessel hydro test. He rod travel housing seal weld is performed prior to final assembly l following the assembly of the travel housing to the latch assembly housing. The hydrostatic test of the connection of the rod travel housing to the latch assembly is done as part of the
! system hydrostatic test.
Revision: 12 3 April 30,1997 3.9 80 3 Westingh00Se o
- 3. Design of Structures, Components, Equipment, and Systems To assure functional capability of the control rod dive meet anism following a seismic event !
or a pipe break, the bending moments on the control rod diive mechanisms are limited to l those that produce stress levels in the pressure boundary of the control rod drive mechanism ,
less than AShE Code limits during anticipated transient conditions. This limit provides that I the rod travel housing does not bend to the extent that the drive rod binds during insertion I of the control rods. The analysis evaluates the load combinations that include safe shutdown earthquake and pipe break. The pipe break considered is at least as large as the largest pipe in or connected to the reactor coolant system that is not qualified as leak before break line. See subsection 3.9.7 for information on the control rod drive mechanism defection limit requirements for the integrated head package 3.9.4.4 Control Rod Drive Mechanism Performance Assurance Program
'Ihe capability of the pressure housing components to perform throughout the 60 year design objective is confirmed by the stress analysis report required by the AShE Code,Section III.
To confirm the operational adequacy of the combination of fuel assembly, control rod drive I mechanism, and rod cluster control assembly, functional test programs have been conducted.
These tests verify that the trip time achieved by the control rod drive mechanisms meets the l design requirements.
The units are production tested prior to shipment to confirm the capability of the control rod drive mechanism to meet design specification operation requirements. Each production control rod drive mechanism undergoes a production test as listed in Table 3.9-13. l l
The trip time requirement is confirmed for each control rod drive mechanism prior to initial l reactor operation and at periodic intervals after initial reactor operation, as required by the ;
technical specifications. See Section 14.2 for preoperational and startup testing.
I To demonstrate proper operation of the control rod drive mechanism and to provide acceptable core power distributirsns, rod cluster control assembly partial movement checks are performed as required by the Technical Specifications. In addition, periodic drop tests of the rod cluster control assembly are performed at each refueling shutdown to demonstrate continued capability to meet trip time requirements, consistent with safety analyses in Chapter 15.
l r Revision: 12 3 Westingh00Se 3.9-81 April 30,1997
5 l
1 FAX to DINO SCALETTI l May 6,1997 !
CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don Lindgren )
Bob Vijuk i Brian McIntyre OPEN ITEM #564 (DSER 3.2.2-1) l To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 24 calendar days away (19 business days). The relevant documentation related to Open items #564 DSER 3.2.2-1) is in the SSAR l
in Sub-Sections 3.2.25 and 6.3.2.3. The pertinent pages of the SSAR are attached. Ad- i ditionally, these items were discussed in Westinghouse letter NSD-NRC-97-4989, dated February 19,1997. Appropriate pages of this letter are included. It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this l item. We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 l
l 4
1 Ik7
AP600 Opc2 Item Trackirg Syst2m Dct'. base: Execrtive S:mm:ry D:t : 5/6/97 Selectio2: [ item noj between 564 And 564 Sorted by Type item DSLR S:ctum Resp Title / Description (W) NRC No tiranch Question Typc Detail Status Engineer Status Status letter No. / , _ Date
$64 NRR/EMLII 32.2-1 DSI R-Of Lindgren Closed Action W NSD-NRC-97-4989
~ ~
l Westinghouse should revise Table 3 2-3 and other applicable sections and PalDs of the SSAR to reflect the stafTs position on E{CC
~
Closed - AP600 Class C lines that provide an ECCS functusn mill' require' spot radiograph of the welds. This requirement added to 352.2.5 E SSAR '
! revision 7 l Actior W - In a letter to Westinghouse dated August 20,1996, this open item mas reported by the staiTas being resobed. Ilowever, befora this issue
- is consMered resobed, the staff needs the following information and/or clarifications from Westinghouse
- 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): g in-containment refucimg water storage tank (SSAR Fig 6.3-2)
Accumulator (SSAR Fig. 6.3-1)
Accumulator injection piping to discharge check valve V-028 (SSAR Fig. (6.3-1) b Containment recirculating piping and valves to IRWST injection check =ahe V-122 (SSAR Fig. 6.3-1) k Pipmg from ist,2nd & 3rd stage ADVs to IRWST, including depressurizatkm spargers (SSAR Fig. 5.l-5 & 6.3-2)
N . 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 ECC's
!not listed above are included in the commitment to random radiography for all LCCS.
- b. It rvpcars that SSAR Subsection 3 2.2.5 is the only place in the SSAR that contains the above commitment. Since this commitment is not stated in caher Table 3 2-3 or applicable P&lDs, how can the staff be assured that it will be implemented on all AP600 plants?
- This issue will be discussed during the December 5 & 6,1996 meeting f Action W - In a letter to Westinghouse dated August 20,1996, this cren item mas reported by the stalT as being resobed flowes er, before this issue is considered resolved, the staff needs the following information and/or clarifications in the SSAR:
l a The stafThas identified the components and systems listed below as part of ECCS systems that are classified as AP600 Class C ( ASME Class
'!3): In-containment tefueling water storage tank (SSAR Fig 6.3-2)
Accumulator (SSAR Fig. 6.3-1)
Accumulator injection piping to discharge check vahe V-028 (SSAR Fig. 6.3-l)
Containment recirculating piping and valves to in-containment refueling water storage tank (IRWST) injectkm check valve V-122 (SSAR Fig. j 63-1) l Piping from Ist,2nd & 3rd stage automatic depressurization vahes (ADV) to the IRWST, including depressurization spargers (SSA R Fig. 5 l-5 & 6.3-2)
{ Westinghouse is requested to verify in the SSAR Subsection 3 2.2.5. that all of the above components and sy stems and any other Class 3 ECCS ru>t l listed above arc included in the commitment to random radungraphy for all FCCS.
b It appears that SSAR Subsectum 3.2.2.5 is the only place in the SSAR that contains the above commitment Smcc this commitment is not stated lin either Table 3 2-3 or applicable P&lDs, how can the stafT be assured that it will be implemented on all AP600 plants?
. Closed 7 R esponse proi ided in NSD-NRC-97-4989 of 2/19/97. _ _
j Page: 1 Total Records: I
G i
>p .
[
l J Westinghouse Energy Systems Box 355 i
Electric Corporation Pmsburgh Pennsylvania 15230 0355 l
NSD-NRC-97-4989 DCP/NRC0743 Docket No. STN-52-003 February 19, 1997 j Document Control Desk U. S. Nuclear Regulatory Commission i 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:
l 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
- responses are grouped by the enclosures of the NRC letter. Also attached are markups of SSAR l
revisions that will resolve a number of these items. These changes will be included in Revision i1 of the SSAR.
j The resolution of the items addressed in the attachment will permit the NRC staff to provide input to l the FSER for a number o' the subsections.
l If you have any questions please contact D. A. Lindgren at (412) 374-4856.
l l A Brian A. McIntyre, Manager l Advanced Plant Safety and Licensing l
.jml Attachments i
cc: D. Jackson, NRC (w/atta.hments) m amme 4
l nm .. j ,
3r 07 1 ( u v%a w usef (d-
4 4
Attaclunent I to NSD-NRC-97-4989 NRC Letter Enclosure 2 l 3. Open Item 3.2.2-1 (OITS 564) - Classification of Emergency Core Cooling System (ECCS)
Action W
- in a letter to Westinghouse dated August 20,1996, this open item was reported by the staff as being
- re,olved. However, before this issue is considered resolved, the staff needs the following information 1 and/or clarifications in the SSAR
1
- 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):
2 In-containment refueling water storage tank (SSAR Fig. 6.3-2)
- Accumulator (SSAR Fig. 6.3-l') l Accumulator injection piping to discharge check valve V-028 (SSAR Fig. 6.3-1)
Containment recirculating piping and valves to in-containment refueling water storage tank (IRWST) injection check valve V-122 (SSAR Fig. 6.3-1) ]
1 Piping from ist,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.
- b. It appears that SSAR Subsection 3.2.2.5 is the only place in the SSAR that contains the above commitment. Since this commitment is not stated in either Table 3.2-3 or applicable P& ids, how l can the staff be assured that it will be implemented on all AP600 plants? l 1
Westinghouse Resoonse '
- 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 j subsection 3.2.3.5. 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, m 1
l l i
. j
- 3. Design of Structures, Components, Equipment and Systems t
1 1
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 tirats (for example, new and spent fuel racks, the bridge, and the hoist)
Maintain spent fuel sub-critical l
Monitor radioactive effluent to confirm that release rates or total releases are within I limits established for normal operations and transient operation 1
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 ponions thereof are.not within the scope of the ASME Code,Section III, Class 2. 1 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.
of I
'Ihe(d6mponents~and portions o' f systems;that p'rovide emergency, core lcdoling functi6iis,siid 7.2.2-/
l aniihibited f6 have radi6giapYylof a~ rand 6is' sample'of weldiduringiibnstruction incliide th~e following:
=
AccGihulators l
Injectio;n piping fr~oni the ac'5jmblat6is to the reactor coolant. system isolation check valves.in the direct' vessel injecti'on line Pipihg froriithe:ir@pntainthentieftielliig, water storage ~ tank (IRWST) and'rsBlibblatioq
~
scree 6s'to the'reac^tbTco' olint' systerr[isolatioircheck valves in thei!irect visserisj'ecti6n line Pipingifrom,the.S{agejl',j2land.3iautomatic,degiress~tirization systein;salv6QtRtiie IR.WfTEincludingfl[9?ipsrgirsg; W@WST{i[s3forineajfr6m.pohi6ni'6f{idu6tiiral:r6odules thitlsre elements.6Ethe con _taili,rhi[dfihtsrdal strubtures. The inipection requirements for the welds in these strtictural mod 61F('a@pf6videdliiifSu6section 3.8.3.6.2.
L 3.2.2.6 Equipment Class D l Class D is nonsafety-related with some additional requirements on procurement, inspection l or monitoring.
Revision: 11 wmman.oum 7 Draft 1997 3.2-8 [ Westinghouse
e
- 3. Design of Structures, Components, Equipment and Systems
=
r}
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 Monitor radioactive effluent to confirm that release rates or total releases are within limits established for normal operations and transient operation 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.
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. I I The components and portions of systems that provide emergency core cooling functions and ;
I are required to have radiography of a random sample of welds during construction include the I following: ( }
l l
l -
Injection piping from the accumulators to the reactor coolant system isolation check i valves in the direct vessel injection line l
l
- Piping from the in-containment refueling water storage tank (IRWST) and recirculation I screens to the reactor coolant system isolation check valves in the direct vessel injection I line l
I =
Piping from the Stage 1, 2, and 3 automatic depressurization system valves to the l IRWST including the spargers.
I I The IRWST is fonned from portions of structural modules tht are elements of the I containment intemal structures. The inspection requirements for the welds in these structural modules are provided in Subsection 3.8.3.6.2.
m 3.2.2.6 Equipment Class D Class D is nonsafety-related with some additional requirements on procurement, inspection or monitoring.
Revision: 11 February 28,1997 6J[ 3.2-8 [ W85tingh0US8
- 6. Ergineered S:f;ty Features l
m
)
In the incontainment refueling water storage tank injection lines, the squib valves are in series with normally closed check valves. In the containment recirculation lines, the squib valves l i
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. l It also allows flow in both directions. A valve open position sensor is provided for these l valves. The IRWST injection squib valves are diverse from the containment recirculation I
! squib valves. They are designed to different design pressures. l Squib valves are also used to isolate the fourth stage automatic depressu'ization r system lines.
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 l of squib valve used in this application provides zero leakage of reactor coolant out of the j reactor coolant system. The reactor coolant pressure acts to open the valve. A valve open l position sensor is provide for these valves. !
6.3.2.3 Applicable Codes and Classifications Sections 5.2 and 3.2 list the equipment ASME Code and seismic classification for the passive i core cooling system. Most of the piping and components of the passive core cooling system j within containment are AP600 Equipment Class A, B, or C and are designed to meet seismic Category I requirements. Equipment Class C components and pioing, that provide an emergency core cooling function, have augmented weld inspection requirements (see subsection 3.2.2.5). Some system piping and components that do not perform safety-related functes are nonsafety-related.
~~~ n 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:
. Pans of components in contact with borated water are fabricated of, or clad with, austenitic stainless steel or an equivalent corrosion-resistant material.
- Internal parts of components in contact with containment emergency sump solution during recirculation are fabricated of austenitic stainless steel or an equivalen corrosion l resistant material.
l l
Revision: 12 W
WB5tingh00S8 7 [ 6.3-23 April 30,1997
1 4
FAX to DINO SCALETTI May 6,1997 l CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay
, Don Lindgren Bob Vijuk Brian McIntyre OPEN ITEM #783 (DSER 3.9.2.3-2)' ,
To meet the SECY-97-051 sched. ale of " Applicant Submits Final SS~AR Revisions &
Documentation" by 5/97, we be:ieve that NRC must acknowledge receipt of all ;
Westinghouse submittals by May 30,1997. This is just 24 calendar days away (19 business days). The relevant documentation related to Open items #783 DSER 3.9.2.3-2) is in the SSAR l in Sub-Sections 3.9.2.3,3.9.2.4, and 3.9.9. The pertinent pages of the SSAR are attached. Ad-ditionally, these items were discussed in Westinghouse letter NSD-NRC-97-4989, dated February 19,1997. Appropriate pages of this letter are included. It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
r i c>d^^
Jim Winters 412-374-5290 i
j
AP600 Opc2 It:m Trtcki g System Datzbase: Exec tiva S:.mm:ry p;te: 5/6/97 Selectio:: [ item noj between 783 And 783 Sorted by T)pe Item DS1 R Section litle/ Description Resp (W) NRC No tiranch Questam T)pe Detad Status Engineer Status Status Extter No. / Date 783 NRR/I MI B 3.9.23-2 DSLR4)I Lindgren Confrm-N Action W NSD-NRC-96-4677
~
Westinghouse should provide documentation of vibraton prediction analysis for the AP600 intemais, includ:ng anticipated ' responses at each ]
transducer location, their allowable response level, and the basis for the acceptance criteria. In additum, the document should also be referenced in
.the SSAR.
Closed - A summary table of espected vibratum mas added to the vibration assessment report. COL item added to SSAR R'ev. 4 for calculation of l yransducer responses. M101-GER-001, AP600 Reactor intemals Ilow-Induced Vibratens Assessment Report Transmitted 4/l/95. l l Revision 2 of M101-GER-001 Transmitted on March 29,1996 i
{ Action W - 8/20/96 -The final report was submitted with changes but was not included in the references list in a recent revision of the SSAR Section l l3 9 9 as discussed under DSER Open item 3.9 2.3-1 above. Westinghouse has agreed to implement the above stafT requests Therefore, DSI R Open, lltem 3 9 2.3-2 is technically resolved, pending acceptable completion of Westinghouse acti ms relativ e to the above requests.
} Action W - Reference report in SSAR
- Resolved - See response in i etter NSD-NRC-96-4841, dated October 14,1996 WCAP will be referenced in SSAR.
I l Confirm-N - Reference to WCAP-14761 was added in to subsection 3.9.2.3 in Revision 10.
1 l Action W - In SSAR Revision 10, the first paragraph of Subsectum 3.9.2.3 indicates that the 11ow-induced vibration assessment is documented in lWCAP-14761, which is also included in the reference list in SSAR Section 3 9 9 This is acceptab?e. The WCAP-14761 is a replacement of (N) previous report M101-GER-001, which was submitted by Westinghouse and revicwed by the stafTand found acceptabic.
'llowever, 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 l AP600 reactor internals Information of vibration assessment from reference plants, which include il 11 Robinse Doel 3 and 4, etc. may only be
% used in vibration predictum analysis for the prototype and should not be confused with the prototype. The wording in SSAR Sections 3 9.2.3 and )
- ) 9.2 4 should be revised to avoid confusion between the "prototy pe" and the " reference plants." '
Closed - Response provided in NSD-NRC-97-4989 of 2/19/97. _
i l' age: I Total Records: I
l F
- 3. Design of Structures, Components, Equipment, and Systems l
t
! . Heat exchangers
. Filters !
. Passive valves l l
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 parts to the extent that the component cannot perform its required safety function.
Dynamic Testing l
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 l
= Valve position sensors
- Similar appurtenances for other active valves l Combinations of Analysis with Testing Combinations of analysis, 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
- Other active valves 3.9.2.3 Dynamic Resr,onse 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_a_nalytical means alone. Assessment of vibrational response is
% in; ~ -^Tbi n%n of analysis and tesung. Cmugan: er rmults obtained from reference plant vibration measurement programs have been used to confirm e- ,
scale model tests and other prediction methods as well to confirm the adequacy of reference l plant intemals regarding flow induced vibration. The flow-induced vibration assessment is documented in WCAP-14761 (Reference 18).
Reactor components are excited by flowing coolant, which causes oscillatory pressures on
, the surfaces. The integration of these pressures over the applied area provides the forcing l functions to be used in the dynamic analysis of the stmetures. In view of the complexities of the geometries and the random character of the pressure oscillations, a closed form solution of the vibration problem by the integration of the differential equations is not j always practical and realistic.
Revision: 11 February 28,1997 3)9 3.9-34 W Westingh0US8
- 3. Design of Stnactures, Components, Equipment, and Systems For calculation of pump induced pulsations acting on the AP600 reactor internals, the l 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 l 3.9.2.4 Pre operational Flow-Induced Vibration Testing of Reactor Internals j The pre-operational vibration test program for the reactor internals of the A '600 conducted on the first AP600 is consistent with the guidelines of Regulatory v v. 1.20 for a comprehensive vibration assessment program. Design features that have not meviously been tested in the reference plants or subsequent testing are tested to verify the vibration analysis.
Conformance with Regulatory Guide 1.20 is sumr.narized in Section 1.9.1.
The program is directed toward confirming the long-term, steady-state vibration response of the reactor intemals for operating conditions. The three aspects of this evaluation are the following: a prediction of the vibrations of the reactor intemals, a preoperational vibration test program of the internals of the first plant, and a correlation of the analysis and test results.
l W.th respect to the reactor internals preoperational rest program, the first AP600 plant i ieactor vessel internals are classified as prototype as defined in Regulatog Guide 1.20. The l 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 vestel intemals design features and configuration have successfully completed vibration assessment programs including vibration measurement programs. These units have subsequently demonstrated extended satisfactory inservice operation. _ _ _ _ . . _
l The reference plant for the AP600 is H. B. Robinson that has substantially the same siz and operating conditions as the AP600. Structural differences include modific s resunug, im. S ~ C73Efuel, the removal of the thermPt e change to the inverted top hat upper i.stnals support assembly. These design changes were incorporated into the Doel 3 and Doel 4 reactor intemals 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 intemals) and Doel 4 (lower intemals) plants.
The vibrational responses of the AP600 reactor intemals are characterized by the Doel 3 and 4 vibration measurement programs.
He pre-operational test program of the first AP600 plant includes a limited vibration measurement program and a pre and post-hot functionalinspection program. This program satisfies the guidelines for a Regulatory Guide 1.20 Prototype Category plant. He AP600 reactor intemals design does not require supplernental testing including component vibration tests, flow tests, or scale model tests. AP600 plants subsequent to the first plant will be subject to the pre- and post-hot functional inspection program. The program for plants subsequent to the first plant satisfies the guidelines for a Non-Prototype Category IV plant.
O Revision: 11 February 28,1997 k9 3.9-38 3 Westingh0tlSe
- 3. Design of Structures, Components, Equipment, and Systems
- 8. Takeuchi, K., et al., "Multiflex-A Fortran-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708-P-A, Volumes 1 and 2 (Proprietary) and WCAP-8709-A Volumes 1 and 2. (Nonproprietary), February 1976.
- 9. Cooper, F. W., Jr., "17 x 17 Drive Line Components Tests Phase IB 11, til D-Loop Drop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449 (Nonproprietary), i December 1974.
)
4
- 10. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, Nuclear Regulatory I
, Commission, July 1980. l
- 11. "Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More," ANSI N14.6.
- 12. NRC BULLETIN NO. 88-08: Thermal Stresses in Piping Connected to Reactor
, Coolant Systems,' June 22,1988, including Supplements 1,2, and 3, dated: June 24, 1988; August 4,1988; and April 11,1989.
I 13. Deleted
- 14. NRC BULLETIN NO. 8811: Pressurizer Surge Line Thermal Stratification, l December 20,1988. '
- 15. "AP600 Implementation of the Regulatory Treatment of Nonsafety Related Systems Process," WCAP-13856 September 1993.
I
- 16. NRC IE Bulletin 79-13. " Cracking in Feedwater System Piping," June 25,1979 and 1 Revisions 1 and 2, dated August 30,1979 and November 16,1979. I l
- 17. " Investigation of Feedwater Line Cracking in Pressurized Water Reactor Plants,"
(Proprietary) WCAP-9693 June 1980,
{
l l
- 18. "AP600 Reactor Intemals Flow-Induced Vibration Assessment Program."
WCAP-14761, March,1996. i l
1
, Y 4
[ W6Stingh00Se
(( 3.9 103 Revision: 12 April 30,1997
. - - . - - . - . . - . . - . - . - . . . _ - . . -. ~ - .- .
> 1 N
j W
i 4
i i
Westinghouse Energy Systems Box 355 i
- Electric Corporation Pittsburgh Pennsy!vania 15230-0355 j
i NSD-NRC-97-4989 j DCP/NRC0743 i I
Docket No.: STN-52-003 February 19, 1997 1 '
3 Document Control Desk !
- - U. S. Nuclear Regulatory Commission l Washington, DC 20555 l 1
TO: T.R. QUAY l
l
SUBJECT:
RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION AND OPEN l j ITEMS ASSOCIATED WITH SSAR CHAPTER 3 i-
Dear Mr. Quay:
1 In a letter dated February 7.1997, the NRC provided additional requests for additional information 1 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 1
resolved and do not require a response. The responses to other items will be provided later. The ]
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.
i 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. j If you have any questions please contact D. A. Lindgren at (412) 374-4856.
wf Y
- Brian A. McIntyre, Manager i Advanced Plant Safety and Licensing jml
- Attachments
- cc
- D. Jackson, NRC (w/ attachments)
M
=
m iion w '3
o 1
l Attachment I to NSD-NRC-97-4989 1 1
NRC Letter Enclosure 5
- 1. DSER# 3.9.2.3-2 (783) - Flow-induced vibration prediction analysis l Action W l 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 l SSA.R Section 3.9.9. This is acceptable. The WCAP-14761 is a replacement of previous report I MIDI-GER-001, which was submitted by Westinghouse and reviewed by the staff and found l
acceptable. I 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 I
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.
The wording in SSAR Sections 3.9.2.3 and 3.9.2.4 should be revised to avoid confusion between the i
" prototype # and the " reference plants."
Westinghouse Responic 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.
Westinghouse Resoonse 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
Atlion_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.
_ 7 d9 6
i
- 3. Design of Structures, Components, Equipmint, and Systems
- Heat exchangers
- Filters
+ 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 parts to the extent that the component cannot perform its required safety 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 appurtenances for other active valves Combinations of Analysis with Testing j Combinations of analysis, static testing, and dynamic testing are used for seismic quali6catior, 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
- Other active valves 3.9.2.3 Dynamic Response Analysis of Reactor Internals under Operational Flow Transients and Steady-State Conditions The vibration characieristics 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 reference plant vibration measurement programs have been used to confirm the validity of W scale model tests and other prediction methods as well to confirm the adequacy of reference 7.9.3, p 2.
i plant internals regarding flow induced vibration [!n the fc!!cwing exuzica :he : rm
"=fer:ne: p!=:" is cquwn!=: :0 the :crm prc:ctype a und =d defined in St=da-d Revi r Pl= 3.9.2 =d Regula: cry Guide !.20 for ibm:ica ==zm:n: cf rer. :ct ..;;;rn6. The i flow-induced vibration assessment is documented in WCAP-14761 (Reference 18).
Reactor components are excited by flowing coolant, which causes oscillatory pressures on the surfaces. The 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 . wn omo9. a n.czo797 7 9
Draft,1997 3.9-34 W Westlagt100SB
!t l ..
l
',, 3. Design of Structures, Components, Equipment, and Systems ;
1 I
The reactor coolant canned motor pumps of the AP600, have the same rotational speed and i the same number of impeller blades as in previous plants. Therefore, a significant change !
in vibration is not expected. The forcing function frequencies are similar to previous plants. !
For calculation of pump induced pulsations acting on the AP600 reactor internals, the !
pulsation level at the pumps is taken to be the same as the level previous shaft seal pumps. I 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 l
4 The pre-operational vibration test program for the reactor internals 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 vibratiun 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 reactor vessel internals are classified as prototype as defined in Regulatory Guide 1.20. The 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 extended satisfactory inservice operation. _
g j H e m e:c eferenchlant for the AP600 is H. B. Robinson that has substantially the S'I same si and operatmg conditions as the AP600. Structural differences include modifications resulting from the use of 17x17 fuel, the removal of the thermal shield and the change to the inverted top hat upper intemals support assembly. These design changes were incorporated into the Doel 3 and Doel 4 reactor intemals as well as the AP600.
De effects of these design evolutions from the reference plant were shown by instrumented preoperational testing at the Doel 3 (upper intemals) and Doel 4 (lower intemals) plants.
He vibrational responses of the AP600 reactor internals are characterized by the Doel 3 and 4 vibration measurement programs.
l The pre-operational test program of the first AP600 plant includes a limited vibration
! measurement program and a pre- and post-hot functionalinspection program. His program l
satisfies the guidelines for a Regulatory Guide 1.20 Prototype Category plant, he AP600 l
reactor intemals design does not require supplemental testing including component vibration i
Revision: 11 wmumo%ane2cm f Draft,1997 3.9-38 [ W85tiflgh01.lSe
W FAX COVER SHEET l
i W Westinghouse RECIPIENT INFORMATION SENDER INFORMATION DATE: M4/ 7, / F/ 7 NAME: 6 e w s / l A _ ,c ,9 ,
i TO: LOCATION: ENERGY CENTER -
06 )C6/205/t)/ EAST 37-6 8 PHONE: FACSIMILE: $ -/ - 3 u -V/f- 2ce2 PHONE: Office: (71 - $N- f;i a COMPANY: Facsimile: win:
4[C.
/ - [cayer rs 284 4887 outside: (412)374-4887 l LOCAT10N: (Jdirg Fi.i c , M b Cover + Pages 1+ h .
The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:
WIN: 284 5125 (Janice) or Outside: (412)374 5125.
COMMENTS:
V2, %ff Rt.c.c a mf ir cwS Artz & rrd' edi~h .
-ro t)-r b iiv c' J L -rE7. 2 C cu H<O/ rd 2 57'WPf= 0N 7//E NIC E. mad i 7t)6 it i r'/ re 57 fvt 4-a n 6 7 it t 61 50 f/047.*
i b&tx.-6 40Nb M <=0 on klfC'D Si.sr t bAur Ow.y' T& w d6 s) 00+ 4es w w 70 oas uw/n.soz. oz Nz.r-q/Arsjm s) Ca (Avon SS42 c+urw -ro M%m i >c i- A l
0 ft eQ9 5d= CMu- ME 10 .3 =~T JP 4 YELc~~c tw 75 /2 L&7v12 7~H ' A W 3~lEGL -10 'Di .5 C v = $ TM'4 H .5 W &E
AP600 Approach to First Plant Only Testing The AP6(X) employs unique passive safety systems to mitigate the consequences of design basis accidents, without the need for safety-related components that require ac power such as pumps or fans. To support the design certification of the AP600, Westinghouse embarked on a thorough and extensive testing program of the passive safety systems such as the passive core cooling system and the passive containment cooling system. These tests were used for validation of the Westinghouse computer codes used to predict the performance of the passive safety systems following design basis accidents.
l The various design certification tests were conducted at various scaled conditions, and I provide a sufficient database of information to properly validate the computer codes ;
necessary to predict the thermal-hydraulic phenomenon of the passive safety systems. In a l few instances, unique system performance characteristics of the passive systems are specified I to be observed during startup testing of the AP600. These demonstration tests of unique phenomenological characteristics are used to supplement the testing performed for design certification. These phenomenological characteristics are such that they do not vary from plant to plant because of the standardization of the AP600 design, and as such, these special tests (designated as first plant only tests) are not required on follow plants.
The AP600 approach to first plant only testing is consistent with the approach taken on current PWRs, where phenomenological characteristics are performed on a first plant-type basis for such items as natural circulation cooling, and reactor internals vibration. The NRC has accepted that the unique phenomenological characteristics associated with natural circulation cooling need only be tested on a single plant of each plant-type. Likewise, reactor internals vibration is tested on the first plant that employs a new reactor internals design, and subsequent reactor internals vibration on plants that employ the same reactor l internals is not required. In the AP600, first plant only tests have been specified for the l following phenomenological characteristics:
- IRWST heatup characteristics
- CMT recirculation
- CMT heated draindown
- ADS stage 1-3 blowdown
- Long-term heatup of the main control room, I&C equipment rooms and Class IE de equipment rooms
- Pressurizer surge line and spray line stratification
- Reactor vessels internals vibration
- Natural circulation
- PRHR heat exchanger performance
- Load follow with grey rods l
l l
l
e Draft Revised Response to OITS 1244 D OITS 1244/DSER Open item 14.2.8-6(RAI 260.26):
In the DSER, the staff found that Westinghouse should provide testing of the main control room emergency habitability system on subsequent AP600 plants.
Westinghouse responded in their August 13, 1996, letter that Section 14.2.9.1.6,
" Main Control Room Emergency Habitability System Testing," was revised to include appropriate testing for each plant, but that a long-term demonstration of this system would be conducted only for the first plant.
Staff review of this section determined that sufficient assurance does not exist to conclude that the heat loads in the main control room area are identical for all AP600 plants. Therefore, Section 14.2.9.1.6 should be modified to include app 3cability of this testing to subsequent AP600 plants, or Appendix 1 A in the SSAR should provide appropriate justification for this exception to RG l.68, Appendix A, Item 1.n.(14)(f).
OITS 1244/DSER Open item 14.2.8 6 remains open.
Westinghouse Response:
The AP600 does not provide active, safety-related HVAC for the main control room, I&C equipment rooms, and class IE de equipment rooms. The habitability of the main control rooms is provided by operation of the MCR emergency habitability system, and by the passive heat sinks associated with the main control room structure. Likewise, the environmental conditions that the qualified I&C equipment and class IE equipment will be exposed to are based on the passive heat sinks associated with the building and structures that house this equipment.
In the AP600, a design basis heatup analysis of the main control room, I&C equipment rooms, and class IE de equipment rooms is performed .md the results are discussed in SSAR Section 6.4. This analysis assumes maximum. bounding heat loads for the equipment that could be located in the main control room and equipment rooms. The AP600 Certified Design Material for the Main Control Room Emergency Habitability System (Section 2.2.5, item Nc, and item 8e in Table 2.2.5-4) specifies that an evaluation will be performed using as-built information and heat loads from installed equipment for the i) MCR, ii) I&C equipment rooms, iii) Class IE de equipment rooms. In addition, this evaluation considers the as-built passive heat sinks associated with these rooms, as specified in Certified Design Material Section 3.3, Nuclear Island Building Structures. The acceptance criteria for this
{
heat sink capacity analysis results are that: i) the temperature rise for the MCR is less than or equal to 15"F for the 72-hour period; ii) the maximum temperature for the 72-hour period for ;
the I&C rooms is less than 125"F: iii) the maximum temperature for the 72-hour period for i the Class IE de equipment rooms is less than or equal to 125"F. ]
This evaluation ensures that the as-built information for the pertinent buildings and structures, as well as the as-built MCR equipment and I&C and class IE heat loads are less than that
I l
assumed in the design basis heatup analysis for the MCR and the safety-related equipment rooms noted.
l The first plant only test specified in 14.2.9.1.6 is a test of the long-term heatup characteristics of the main control room, I&C and class IE quipment rooms. It is performed to demonstrate l the heatup characteristics of these rooms when they are subjected to a known heat load. This j test can be used to provide data for comparison to the design basis analyses. However, testing is not required on subsequent plants, since these plants are required to be built to the l l
requirements specified in the Certified Design Material. As the passive heatup of these j rooms is not dependent on the proper operation of a system, but is rather a function of the heat loads and passive heat sinks provided in the design, a verification of these parameters (via the ITAAC process) is sufficient to verify the safety of an AP600 built to the specifications contained in the Certified Design Material.
Moosed AP600 SSAR Chances An exception to R.G.1.68, Appendix A, Item 1.n.(14)(f) has been added to SSAR Appendix l 1 A that states this test needs to be performed for the first plant only provided the design I basis heat loads used as assumptions in the heat sink capacity analysis bound the actual as-built information and heat loads.
{
I
1 Introduction and Cencral Description of Plant !
Criteria Referenced AP600 l Section Criteria Position Clartf1 cation / Summary Description of Exceptions App. A.I.h Conforms The characteristics of the AP600 passive safety systems allow the support systems such as the
" cooling water systems, the heating, ventilating, and air conditioning and the ac power sources to be nonsafety related and simplified. The capability of these systems is established by testing. Cold water interlocks are not a design feature of the AP600.
App. A.I.i Conforms The AP600 has no secondary containment.
Therefore, this guideline applies only to primary containment. The following systerns or functions are not design features of the AP600 and are therefore not tested:
. Containment and containment annulus, vacuum breaker
. Containment supplementary leak collection
. Standby gas treatment
. . Secondary co'ntainment system
- Containment annulus and cleanup
. Bypass leakage tests on pressure suppression
. Ice co'ndenser systems
. Containment penetration cooling App. A.I.J Conforms Recirculation flow control, traversing incore probes, automatic dispatching control systems and hotwell level control are not design features of the AP600.
App. A.I.k Conforms App. A.I.I Conforms Condenser otT gas systems are not a design feature of the AP600 .
App. A.I.m Conforms App. A.I.n (e.rcept A.l.n(14)$) Conforms Seal water, boron recovery, shield cooling, refueling water storage tank heatmg, and equipment for establishing and maintaining subatmospheric pressures are not design features of the AP600.
App. A.I.n(14)W Exception A long term demomtration test of the MCR habitabdity system to maintain the control environment is acceptablefor thefirst plant only provided that the as-budt information and design l basis heat loads used as the assumptioru m the heat sink capacity analysis for the MCR do not
\ changefor subsequent plants.
I Draft Revision W
WSSthgh0080 May 1,1997 2-22
4 FAX to DINO SCALETTI May 2,1997 CC: Sharon or Dino, please make copies for: Diane Jackson
!' Ted Quay Don Hutchings Don Lindgren Bob Vijuk Brian McIntyre OPEN ITEM #368 (M10.4.2-2)
- 370 (M10.4.3-2)
- 1151 (DSER 10.4.2-1)
- 1152 (DSER 10.4.3-1)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30, 1997. This is just 30 calendar days away (21 business days). The relevant documentation related to Open Item #368 (M10.4.2-2) is in the SSAR in Table 3.2-3 (supplied to you in Revision i1 of the SSAR, February 28,1997, over two months ago). The pertinent pages 'of the SSAR are attached. It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of these items. We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 m)7 4
.=
~ AP600 Opca item Trackirg System DMib2se: Execative Srmmnry Date: 5/2/97 Selectier htem nol between 368 And 368 Sorted by Type Item DSER Section Titic/ Description Resp NRC (W)
No tiranch Question Typc Detail Status Engmect Status Status Le ter No / Date 368 NRR/SPEB 1042 M fCWI 13054 Action W Action W NSD-NRC-96-4806
- Mio 4.2-2 (MAIN CONDENSER EVACUATION SYSTEM) iThe response to Q410 257 was received after the DSER was prepared, and is under stalY review Open items and questions may be descioped as a result of the review of those responses.
. .. _ _ . . . . .. .. - . . = _. .
Action-W - Westinghouse to resise WCAP-13054, Compliance with SRP Acceptance Criteria, to reflect that RG 1.26 will not be met, in accordance
- with agreements of 2/22/95 meeting. No SSAR revision required Closed - Response prosided by letter NSD-NRC-%-4806 dated September 5.
This provides additional justification of classification of the main condenser evacuation system as a Class E system NRC has questioned this classification in light of the Standard Resicw Plan and Regulatory Guide 1.26 in this area Regulatory Guide I 26 indicates that Class D standards i
- should be applied to water - or steam <entaining components not classified in a higher class that contain or may contam radioactise material. The
- main condenser evacuation system and the gland steam system in AP600 do not contain radioactive material. They may contam radioactive mascrial,
- to be introduced into the system without appropriate steam system isolation in such an event. the evacuation and gland steam systems may draw in
- radioactive material from the steam spaces in the condenser or turbine.
1 g f AP600 SSAR section 3 2.2.7 contains the conditions for which *may contain radioactise material
- do not require ct:ssatication or component
' Class D.
l
- 1. The system is only potentially radioactive and does not nonnally contain radioactive inatcrial Ihis is true for the esacuation and gland seal :
4 systems in a pressurizcJ water reactor like AP600 where the steam generator provides and effective barrier to the primary coolant l2. The system has shown in plant operations that the operation with the system containing radioactive material meets or can meet unrestricted arcra
'rcicase limits, Studies have shown that once a small steam generator leak is detectable it is still small enough to meet unrestricted arca release
- limits. Automatic and techspec limits arc excccded well before unrestricted arca release limits For very large steam generator leals,othe operatmg parameters provide trip and isolation actions before exceedmg unrestricted area release limits
- 3. An evaluation of the system confirms that the system contains features and components that Lecp the consequences of failure as lomas reasonably j i
.actuesable. This poruon of AP600 is very similar to many operating pressurized mater reactor plants There is a low probabihty ofleakage and therc are ra&ation monitors and radiation limits on the main stcam system and evacuation exhausts. This has resulted in good performance record of no ! !
re! case from this source greater that unrestricted area release limits. 4 i
As a result of meeting these three conditions for not being Class D, the es acuation and gland seal systems are Class E.
{ Action W - Respond to concerns in NRC letter of I t/!1/96 (NRC/lKP/0613). ,
)
r t
i l'aFe 1 lotal Records. I f
l , .
l i 'AP600 Opes item Tr:cking Syztem D:t:brse: Execztive Simm ry Date: 5/2/97 l- . Selection: [ item nol between 370 And 370 Sorted by Type Item DSER Section Title /Descripton . Resp NRC l (W) i No. Branch Question T3pc Detad Status . Engineer Status ^ Leu 5 tm No I
~.Date 370 NRR/SPLB 1043 MIG 4)I 13054 Closed Action W NSD-NRC-96-4806
' ' ~ ~
l Mio 4.3-2 (TURB:NE STEAM SEALING S'YSTEM) )
ilhe responses to Q410.258 and Q410 259 were received after the DSL R vias prepared, and are under stalYrevicw Open items and questions may Id 7
' developed as a result of the retiew of those responses l
Acten-W(410 258)- Westinghouse to resise WCAP-13054, Compliance with SRP Acceptance Criteria, to reflect that RG I.26 mill not be met,in i
. !accordance with agreements of 2/22/95 meeting. No SSAR revision required I l Action-W(410 259) - Westinghouse will re< valuate its position regarding diagram and PalD for turbine steam seahng system in -h with [
- agreements of 2/22N5 meeting. !
jClosed - Response provided by letter NSD-NRC-96-4806 dated September 5,1996 [
, Action W . Pendmg resolution of OITS# 368
___ -l ;
I M f r
I L
I i
i L
L I
t Pge.I 1otal Records. l .
4 i
_ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ -_. -. _ _ . . _ . . _ ~ . . . ._ ,
AP600 Ope:2 Item Trackirg Systzm Dat:bise: Exec tiveSrmmery Dsts: 5/2/97 ;
Selection: [ item noj between II5I And IISI Sorted by Type - I Item DSI R Section Titic/ Description Resp (W) NRC No tiranch f
Questm>n Type Detail Status Engineer . Status Status I.ctter No / l
. Date t151 NRR/SPLB 10 4.2-1 DSER-OI 13054 Closed Action W NSD-NRC-96-4806 l(MAIN CONDENSER EVACUATION SYSTEM, CMS, QUALITY GROUP CLASSIFICATION) The staff has not yet' determined the Iacceptabilny of the design of the main condenser evacuatu>n system.
.- . ==,:=-=.=.- . =;=. r z. .= :=.-.. z.=. ::.=.. l l Action-W - Westinghouse to revise WCAP-13054, Compliance with SRP Acceptance Criteria, to reflect that RG I 26 will not be met, in accordance : '
- with agreements of 2/22/95 meeting No SSAR revision required l
l Closed - Response provided by letter NSD-NRC-96-4806 dated September 5,1996.
t l Action w - Pendmg resolution of OITS# 367 & 368. , _
l l
E 4 .
l e
G s
t i
f f
L f
f t
l' age. I I'otal Rnords I e
t AP600 Opea Item Trackirg Syst m Ditsbrse: Exec tiveSrmarry D tz: $i2/97 Selection: l item nol between i152 And I152 Sorted by Type licm DSI R Section Title / Description Resp - NRC (W) '
No. Branch Question Type Detail Status . En'gineer Status Status t etter No /' Date. [
1852 NRR/SPLH 1043-1 DSER4)I 13054 Closed Action W . NSD-NRC-96-4806 ;
(TURBINE STEAM SEA 1.ING SYSTEM, GSS, QUAUTY GROUli CLASSIFICATION AND DESIGN'!NFO)1he staff has not yct'dctermined~
~
'the acceptability of the turbine steam scaling system.
i .-J 71.
. _ . . . . _ ._ _ _ . . . __ _7_ _
! Action-W(410.258)- Westmghouse to revisc WCAP-13054, Compliance with SRP Acceptance Criteria, to reficct that RG 1.26 mill not bc met,in
'accordance with agreements of 2/22M5 meeting. No SSAR resision required +
f'agreements Action-W(410.259)- of 2/22/95 meeting Westinghouse will re-evaluate its position regarding diagram and P&lD fiw turb; i
[ Closed - Response provided by letter NSD-NRC-96-48% dated September 5.1996 f
- Action W - Pendmg resolution of OITS# 369 & 370. g P
F i
v\ !
3e-i t
i i
i I
t l
i I
t I
l' age I l otal Records. I
4
- 3. Design cf Structures, Ccmponents, Equipm:nt, and Syst:ms l
Table 3.2-3 (Sheet 2 of 64)
AP600 CLASSIFICATION OF hECHANICAL AND FLUID SYSTEMS, COMPONENTS, AND EQUIPMENT Tag Number Description AP600 Seismic Principal Con- Comments Class Category struction Code Comp;nent Cooling Water System (Continued)
CCS PL-V207 CCS Containment Isolation B I ASMEm2 Valve Oatlet Line IRC CCS-PL-V208 CCS Containment Isolation B I ASME m-2 Valve - Outlet Line ORC CCS-PL-V209 Containment Isolation Valve B I ASMEm2 Test Connection - Outlet Line CCS-PL-V257 Containment Isolation Valve B I ASME m-2 Test Connection - Inlet Line CCS-PY-C01 Containment Supply Header B 1 ASME m MC Penetration CCS-PY-C02 Containment Return Header B I ASME m, MC Penetration i Balance of system components are Class E l Condensate System (CDS) location: Turbine Building I System components are Class E I Condenser Tube Cleaning System (CES) Location: Turbine Building I System components are Class E I Turbine Island Chemical Feed System (CFS) Location: Turbine Building l System components are Class E Condenser Air Removal System (CMS) Location: Turbine Building l N/A Condenser Vacuum Breakers E NS ANSI 16.34 Balance of system components are Class D (l
i t.ontainment System (CNS) Location: Containment CNS-MV-01 Containment Vessel B 1 ASME m, MC CNS-MY-Y01 Equipment Hatch B I ASME m MC CNS-MY YO2 Maintenance Hatch B I ASME m, MC CNS-MY YO3 Personnel Hatch - 135' 3" B 1 ASME m. MC CNS MY YO4 Personnel Hatch - 107'-2" B I AShm m, MC l Spare Containment B I ASME m MC I Penetrations I Condensate Polishing System (CPS) location: Turbine Building l System components are Class E Revision: 11 February 28,1997
{)7 3.2-22 3 W95tiligh00S8
_ - - - - - -= . -. . . - . - . . - _ _ _ -
- 3. Design of Structures, Compone ts, Equipmeat, and Systems (bTG3 l
Table 3.2-3 (Sheet 7 of 64)
AP600 CLASSIFICATION OF MECHANICAL AND FLUID SYSTEMS, COMPONENTS, AND EQUIPMENT Tag Number Description AP600 Seismic Principal Con- Comments Class Category struction Code FPS-PL-V052 Fire Water Containment B I ASME III-2 Supply Isolation - Inside FPS-PY-C01 Fire Protection to B I ASME III, MC 4
Containment Penetration l Balance of system components are Class E I Main and Startup Feedwater System (FWS) Imation: Turbine Building n/a Startup Feedwater Pumps D NS Hydraulic Institute
- Standards n/a Valves Providing SFW D NS ANSI 16.34 AP600 Equipment Class D Function 1 Balance of system components are Class E Gland Seal System (GSS) Location
- Turbine Building System components are Class D l Generator Hydrogen and CO 3Systems (HCS) Locatba: Turbine Building I System components are Class E I Hester Drain System (HDS) Location: Turbine Building I System components are Class E I Hydrogen Seal Oil System (HSS) Location: Turbine Building I System components are Class E I Incore Instrumentation System (IIS) Location: Containment I n/a IIS Guide Tubes A 1 ASME III-1 i I n/a l' Thimble assemblies D NS Manufacturer l Std.
l i
1 l
[ W8Stingh00S8 3.2-27 February 28,1997
- . .__ - . =. ..- .- . . . -.
1*
I
\
FAX to DINO SCALETTI May 6,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay l
Don Lindgren Bob Vijuk Brian McIntyre l
OPEN ITEM #563 [DSER 3.2.1-2)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 24 calendar days away (19 business days). The relevant documentation related to Open Items #563 DSER 3.2.1-2) is in the SSAR, in Sub-Section 3.2.2.0 and in Table 3.2-1. The pertinent pages of the SSAR are Revision [ Additionally, these items were discussed in Westinghouse attached.
! dated October 14, 1996. Appropriate pages of this letter are included. It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 l
i 1
9 l
, AP600 Open Itzm Trackirg Sy:tzm Dat . base: Execttive S:mmnry Dat2: 5/6/97 Selection: litem no] between 563 And 563 Sorted by Typc ,
item DSLR Section Title / Description Resp (W) NRC l No, Branch Queston Type Detail Status Engineer Status Status Extter No.- / Date 563 NRR/EMEB 32.1-2 DSLR4)I Lindgren Confrm-N Action W NSD-N RC-96-484 I I
^ ~
{ At a minimum, the new and spent fuel storage racks should meet th e applicable quahty assurance requirements of Appe'ndit B to 10 CfR Part 50,in
[
laddetion to being classified as Seismic Category I. Westinghouse should add a note to Sheet 19 of Table 3.2-3 of the SSAR to reflect this position.
~ - - ~ - . . . . . .
~
Closed -The fuel rack classification'in Table 3.2!3'indicases tlA they are Seismic Category 1. Seismic Category I is rcquired to h'avc Appendix B t
,QA program. A separate not:is not required. '
! Acton W - Since the new and spent fuel storage racks are classified as AP600 Class D, it is possible that this commitment might be misinterpreted jwhen one consults SSAR Tabic 3.2-1. According to this table AP600 Class D components do not hase to meet either RG 1.29 scismic dcign i l requirements or Appendix B Table 3 2-1 should be clarified by adding a note to state that although the new and spent fuel storage racks are Class
{ D, they are designed as Seismic Category I, and meet the applicable QA requirements of Appendix B.
t Resolveded - See response in Ixtter NSD-NRC-96-4841 dated October 14,1996. Add requirement for Appendix B for seismic Category 1 Class D fitems iConfrm-N - Subsection 3.2.2 6 was revised in Revision 10 to specifically state that Appendix B apphes to Class D seismic Category I b^CIIU"
- N.'_*5 3" "N'['S*f .is pendjng g s%_ s eduatQof,yues to RAls 260 88 and 260 89.
\
N n -
V i
v i
!i I
t i
i I
i f
i i
i Page: 1 Total Records: I i
F E
-.v-- e
l a
j 4
3 Design of Structures, Components, Equipment and Systems r i
- I
- 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 ,
l
! doses greater than normal limits (for example, new and spent fuel racks, the bridge, and l the hoist) t
+ Maintain spent fuel sub-critical
- Monitor radioactive effluent to confirm that release rates or total releases are within limits established for normal operations and transient operation '
l
- 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.
- 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.
1 The components and portions of systems that provide emergency core cooling functions and I are required to have radiography of a random sample of welds during construction include the ,
I following: !
l l
l - Injection piping from the accumulators to the reactor coolant system isolation check I valves in the direct vessel injection line I
- Piping from the in-containment refueling water storage tank (IRWST) and recirculation I screens to the reactor coolant system isolation check valves in the direct vessel injection I line 1
l
- Piping from the Stage 1, 2, and 3 automatic depressurization system valves to the l IRWST including the spargers.
I l 'Ihe IRWST is formed from pottions of structural modules that are elements of the I containment intcmal structures. The inspection requirements for the welds in these stmetural I 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 r.anitoring.
Revision: 11 3 ?) *)
February 28,1997 3.2-8 W Westingh0US0
l l l
l ,
- 3. Desig2 of S:ructures, C:mponents, Eq 1pment cad Systems I
l For Class D structures, systems, and cornponents contaming radioactivity, it is demonstrated l
by conservative analysis that the potential for failure due to a design basis event does not result in exceeding the normal offsite doses per 10 CFR 20. His criterion is in conformance I I with the definition of Class D in Regulatory Guide 1.26. l A structure, system or component is classified as Class D when it directly acts to prevent unnecessary actuation of the passive safety syst:ms. Structures, systems and components which support those which directly act to prevent the actuation of passive safety systems are also Class D. The inclusion of these nonsafety-related structures, systems, and components in Class D recognizes that these systems provide an important first level of defense that helps to reduce the calculated probabilistic risk assessment core melt frequency. These r.ructures, systems, and components are nonnally used to support plant cooldown.and depress irization and to maintain shutdown conditions during maintenance and refueling outages.
For Class D structures, systems, and components considered to be risk significant as defined in the reliability assurance plan (see Section 16.2). Provisions are made to check for operability, including appropriate testing and inspection, and to repair out-of-service structures, systems, and components. Rese provisions are documented and administered in the plant reliability assurance plan and operating and maintenance procedures.
Some Class D structures, systems, and components are assumed to function in a severe l containment environment. The design requirements for these components include operation in such an environment. An evaluation is done to confirm that the structure, system, or
) component can be expected to function in such an environment.
Standard industrial quality assurance standards are applied to Class D structures, systems, and components to provide appropriate integrity and function although 10 CFR 50, Appendix B and 10 CFR 21 do not applyf 0 CFR 50, Appendix b ano av trK 2 L do appiv to uan v 2 gructures, svstems. and components that are seismic Category IIIhese mdustnal quahty anuumcc standards are consistent with the guidehnes for NRC Quality Group D. The industry standards used for Class D structures, systems and components are widely used industry standards. Typical industrial standards used for Class D systems and components are provided as follows:
- Pressure vessels - ASME Code,Section VIII
+ Piping - ANSI B 31.1. Power Piping, (Reference 5)
- Pumps - API 610 (Reference 6), or Hydraulic Institute Standards (Reference 7)
Valves - ANSI B16.34 (Reference 8) a f Revision: 12 W W8Stingh00Se 3.2-9 April 30,1997
Table 3.2-1 P COMPARISON OF SAFETY CLASSIFICATION REQUIREMENTS k
@ Code ment Safety RG 1.29 Seismic ash 1E Code, RG 1.26 NRC Appendix & Testing Required 2 E Letter Class Design Reqmnts Sec. III Class IEEE Re- Quality Group B Require- Test & O N (1) (2) (3) (4) quirements (5) (6) ments hiaint. j A SC-1 I I NA GROUP A YES YES(7) (8) O o
B B SC-2 I 2 NA GROUP B YES YES(7) (8)
'8 C SC-3 I 3 IE GROUP C YES YES(7) (8) g D NNS(2) NA(9) NA(10) (10) GROUP D NO(10) YES(1I) (11) F M NA OTHER NNS(2) NA(13) NA NA NA(12) NA NA NA - Not Applicable OTilER includes Classes E. F. L. P. R, and W.
H Notes: g Y 1. A single letter equipment classification identifies the safety class, quality group, and other classifications for AP600. See the subsection 3.2.2 c.
for definition, rn G 7
- 2. AP600 safety classification is an adaptation of that defined in ANSI 51.1. The NNS defined in the ANSI 51.1 standard is divided into several AP600 equipment classifications namely, Classes D E, F. L. P R, and W. [
- 3. See subsection 3.2.1 for definition of seismic categories. E
- 4. ASME Boiler and Pressure Vessel Code, Section 111 defines various classes of structures, systems, and components for nuclear power plants.
It defines criteria and requirements based on the classification. It is not applicable for nonsafety-related components.
- 5. The guality group classification corresponds to those provided in Regulatory Guide 1.26.
- 6. "Yes means quality assurance program is required according to 10 CFR 50 Appendix B.
"No" means quality assurance program is not required according to 10 CFR 50 Appendix B.
- 7. Class A, B, and C, structures, systems, and components built to ash 1E Code, Section 111 are inspected to ASME Code,Section XI requirements.
See the text for additional specification of requirements.
- 8. Class A, B, and C structures, systems, and components that are required to function to mitigate design base accidents have some testing requirements included in the plant technical specifications. In addition to the requirements in the technical specifications, testing and maintenance requirements are meluded in an administratively controlled reliability assurance plan.
- 9. See subsection 3.2.1 for cases when seismic Category 11 requirements are applicable for Class D structures, systems, and components.
?
cr
- 10. See the text for a discussion of the industry standards used in the construction of Class D structures, systems and ccmponents.
- 11. Class D structures, systems, and components have selected reliability assurance programs and procedures to provide availability when r,eedeti.
2 These programs are administratively controlled programs and are not included m the technical specifications.
- 12. Normal industrial procedures are followed in procuring, designing, fabricating, and testing these nonsafety-related structures, systems, and J$ < components. '
M$ 13. Some Class E, F. L. P, R, and W structures, systems, and components may be classified as seismic Category II. See subsection 3.7.3. E hN l ll
l =
Westinghouse Energy Systems Ba 355
"'"50Sh P'""5Y""'a i5230 a355 Electric Corporation
. NSD-NRC-964841 DCP/NRC0625 Docket No.: STN-52-003 October 14, 1996 Document Control Desk U.S. Nuclear Regulatory Commission Washingto'n,'D.C/ 20555 ATTENTION: T.R. QUAY
SUBJECT:
RESPONSES TO NRC MECHANICAL ENGINEERING BRANCH QUESTIONS
Dear Mr. Quay:
Attached are responses to a number of items from the NRC Mechanical Engineering Branch discussed in a telephone call with the NRC staff on October 3,1996. The synopsis of the NRC position comes from an NRC letter dated August 20,1996. The questions are related to quality assurance, reactor vessel internal vibration, CRDMs, and equipment seismic qualification. The questions are identified by the numbers from the Augtist 20,1996 letter, DSER open item or RAI number, and Open items Tracking System item number. This response completes our responses to questions related to subsections 3.2.1, 3.9.2, 3.9.4, 3.9.7, and Section 3.10.
This submittal will permit the completion of staff review for the subsections listed and preparation of the Final Safety Evaluation Report input. l l
Please contact Donald A. Lindgren on (412) 374-4856 if you have additional questions.
// h 4%'/.
Brian A. hic!ntyre, Manager Advanced Plant Safety and Licensing l
/nja Attachments ,
cc: D. T. Jackson - NRC N. J. Liparulo - Westinghouse (w/o attachments) i NS$A h)
4 1
Attachment to NSD NRC-96-4841
- 1. Open item 3.2.1-1 (562) - Appendix B for all Seismic Cat.11 - Regulatory Guide (RG) 1.29, Position C.4 in the Open item Tracking System Database (OITS), Westinghouse reports that this issue is closed l based on a statement added to Seismic Category 11 requirements for Quality Assurance (QA). This information was added to SSAR Section 3.2.1.1.2 in Revision 7, and states that 10 CFR 50, Appendix B does not apply to Seismic Category 11 structures, systems, and components. The staff does not agree. As stated in the DSER for this open item, to satisfy Position C.4 in RG 1.29, the pertinent QA requirements of Appendix B should be applied to all Seismic Category 11 structures, systems, and components. This commitment should be added to SSAR Section 3.2.1.1.2 and Table 3.2-1. Therefpre, Open Item 3.2.1-1 remains open.
Response
When the guidance in Regulatory Guide 1.29 position C.4. was developed, the concept of graded QA had not been developed. An Appendix B quality assurance program is not needed to provide that seismic Category 11 systems, structures, and components do not fail in a manner that would reduce the functioning of a safety-related component. The degree of quality assurance provided for AP600 equipment Class D provides an appropriate level of quality assurance for this function. Westinghouse has dermed quality assurance requirements for the regulatory treatment of nonsafety systems, systems, and components. Those requirements also are sufficient to satisfy regulatory requirements for seismic Category II. J Revise the fourth paragraph of 3.2.1.1.2 as follows:
I i
10 CFR, 50, Appendix B does not apply to seismic Category 11 structures, systems, and components. Seismic Category 11 situctures, systems, and' components' have QA requirements similar to those defined for structures systems', and components' covered by regulatory treatment nonsafety<systemsfsee Section 17.3)sThe quality assurance requirements for Seismic Category 11 structures, systems, and components are sufficient to provide that these components will meet the requirement to not* cause unacceptable structural failure of or l interaction with seismic Category I items.
This item is Resolved pending formal SSAR revision. l Open Item 3.2.1-2 (563)- Appendix B for new and spent fuel storage racks 7
2.
In the OITS, Westinghouse reports that this issue is closed because SSAR Table 3.2-3 lists the new and spent fuel storage racks as Seismic Category 1, and as such are required to meet applicable .
portions of Appendix B. The staff agrees that to meet RG 1.29, Appendix B should be applied to these components. However, since the new and spent fuel storage racks are classified as AP600 Class f D, it is possible that this commitment might be misinterpreted when one consults SSAR Table 3.2-1.
According to this table, AP600 Class D components do not have to meet either RG 1.29 seismic design requirements or Appendix B. Table 3.2-1 should be clarified by adding a note to state that k although the new and spent fuel storage racks are Class D, they are designed as Seismic Category 1, and meet the applicable QA requirements of Appendix B.
\ } f 1 t
i J
d I
I l Attachment to NSD-NRC-96-4841
Response
l 'The tr.it in subsection 3.2.2 will be revised to clarify that Appendix B applies o Class D systems and components that are seismic Category I.
Revist, the seventh paragraph of 3.2.2.6 as follows:
Standard indu: trial quality assurance standards are applied to Class D structures, systems, and components to provide appropriate integrity and function although 10 CFR 50, Appendix B and 10 CFR 21 do not apply. 10 CFR 50, Appendix B and 10 CI'R 21 do apply to Class D structufes, systems; and components that are seismic Category I. . Tl ese industrial quality assurance standards are consistent with the guidelines for NRC Q tality Group D. The industry standards used for Class D structures, systems and compcnents are widely used industry standards. Typical industrial standards used for Class D systems and components are provided as follows:
This item is Resolved pending formal SSAR revision.
21 Open Item 3.9.2.3-2 (783)- Flow-induced vibration prediction analysis Action Westinghouse The staff has received and reviewed Revision 1 (May 1,1995) of the draft report of "AP600 Reactor Internals Flow-Induced Vibration Assessment Program" during and after the May 10, 1995, meeting.
The staff's evaluation of the revised draft report conci'tded that Westinghouse should finalize the report by incorporating the following: (1) add a summary table of vibration prediction analysis results as included in Westinghouse letter dated June 1, 1995, (2) revise the " Introduction" section and other parts of the report for consistency with the SSAR revision, such as including statements designating the reactor internals of the first AP600 plant as a prototype, and (3) to show additional sensors at the guide tubes in Table 8.1 for monitoring their vibrations, which will be consistent with the uvised SSAR Table 3.9-4 of the SSAR. The final report was submitted with changes but was not inc uded in the references list in a recent revision of the SSAR Section 3.9.9 as discussed under DSER Open Item 3.9.2.3-1 above. Westinghouse has agreed to implement the above staff requests. Therefore, DSER Open Item 3.9.2.3-2 is technically resolved, pending acceptable completion of Westinghouse actions relative to the above requests.
Response
The report on vibration assessment will be included in the SSAR. Revise the first paragraph of subsection 3.9.2.3 of the SSAR as follows 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 donc using a combination of analysis and testing. Comparisons of results obtained from reference plant vibration measurement programs have been used to confirm the validity of scale model l
tests and other prediction methods as well to confirm the adequacy of reference plant internals regarding flow induced vibration. In the following discussion th :crm " reference plant" is i
d9 2
S ,.
NRC REQUEST FOR ADDITIONAL INFORMATION Question 260.88 Re: OITS 4121 While the staff may agree that " industrial quality assurance standards are consistent with the guidelines for NRC Quality Group D", it is not clear hov. you concluded that such standards, without NRC endorsement, satisfy the provisions of Appendix B to 10 CFK $0. Please Clarify.
Response
Subsection 3.2.2.6 of the SSAR indicates that Class D systems, structures and components are subject to the requirements of standard industrial QA standards. An exception to this is when the Class D system, structure or component is classified as Seismic Category !. A Category I classification requires that 10 CFR 50 Appendix B be invoked as the appropriate QA requirements. The SSAR does not imply the industrial quality standards meet the provisions of 10 CFR 50 Appendix B.
SSAR Revision: NONE l
h
NRC REQUEST FOR ADDITIONAL INFORMATION
= - t i
Ouestion 260.89 Re: 4122
. SSAR Section 3.2.2.2, " Application of Classincation," Page 3.2-5. states, in part, " Structures, systems, and components classified equipment class A, B, or C or seismic Category I are basic components as denned in 10 CFR 21." Please clarify how a " Basic Component" as defined in 10 CFR Part 21 can also be classified as Equipment Class D, as defined in SSAR Section 3.2.2.6.
Response
SSAR Section 3.2.2.2, Application of Classification, applies the " Basic Component" designation to safety related Class A. B, and C components only. SS AR Section 3.2.2.6 defines Clas D as nonsafety related structures, systems and components containing radioactivity where a conservative analysis must show that the potential for failure, due to a design basis event, does not result in exceeding the normal offsite doses. Consistent with the requirements of 10 CFR 21 the Class D classification is not applied to designated " Basic Components" quality requirements.
SSAR Revision: NONE l
W-Westinghouse I
1 W l W85tingh0USS FAX COVER SHEET D
RECIPIENT INFORMATION SENDER INFORMATION l DATE: April 28,1997 NAME: Barry Sloane TO: Bill Huffman LOCATION: Westinghouse Energy Center )
COMPANY: NRC Monroeville, Pa.
LOCATION: PHONE: (412) 374-4047 / !
WIN 284-4047.
PHONE: 301-415-1141 ,
l FAX: 301 415-2002 / 2300 FAX: (412) 374-5099 Cover + 8 Pages = 9 Pages Total ,
Comments:
1 Bill, i Attached information is for the AP600 Long Term Cooling T/H Uncertainty telecon.
I Plaase call me to set up a time for the call after you've had a chance to look this over.
If at all possible, we would prefer Tuesday AM, Wednesday AM or PM, or Thursday AM.
Thanks.
l N
8:rry Sloane l
- i l
AP600 Long Term Coehne Thermai/Hydrad6o Uncertawity 4
i l OBJECTIVES 1
There are two main objectives of the long-term cooling thermal / hydraulic uncertainty analysis:
o Provide an analytical basis for the PRA success criteria for long term gooling following beyond design-basis events; and o Demonstrate that cons;deration of long term cooling thermal / hydraulic uncertainty in the PRA success criteria analyses i
does not significantly affect the conclusions of the PRA. !
l t
l l
l l
4 808 Apre21,1997 1
. AP400 Long-Term Coating Thermal /Hydrcsdis uncertamey
. GENERA'. APPROACH l
.The approach to be taken for LTC-T/HU was generally defined in letter report NSD-NRC-97-4928 (" Draft Report.-- Resolution of T/H Uncertainty issues for AP600 Passive System Reliability," 1/2/97).
There are some differences between the short-term cooling T/H uncertainty analyses and the long-term cooling T/H uncertainty analyses, as explained in the following discussion.
l l
The long-term cooling analysis will use the expanded Focused PRA event tree approach as developed for the short-term cooling T/H l uncertainty analysis. This will permit risk-informed decisions about which' sequences require analyses demonstrating successful core l cooling such that the PRA results are not significantly affected by LTC-T/HU.
l l
L I I
SOS Apre21.1997 2 I l
l e AP600 Long Term Cooeing Thermal /Hydredic Uncertainty
^
GENERAL APPROACH (Continued)
The steps that are similar to the short-term cooling T/H uncertainty l effort include:
o Grouping focused PRA success sequences with similar equipment availability; .
o identifying',long term recirculation scenarios that are successful.
using only passive syst' ems, and which would have a significant impset on core damage frequency or large release frequency, based or the focused'PRA resul'ts, jf they were considered failure.instead l of success; (These have been termed " risk-significant success sequences.")
l o Performing T/H analyses on the risk-significant long-term recirculation cases, with assumptions and inputs as consistent as possible with those used for the design-basis analysis (DBA) and the short-term T/H Uncertainty analyses; l o Bounding significant sources of uncertainty with assumptions and inputs including use of Appendix K decay heat assumptions, and equipment failure assumptions representative of the limiting
- sequences in each group; o Assessing the PRA significance of the results.
l l
l' sos Apes 21,1ss7 3
l 8
AP;00 Long-Term Cocaing Thermai/Hydridie Uncertamty GENERAL APPROACH (Continued) i o Some differences from the short-term cooling T/H uncertainty approach are necessary because:
l Multiple T/H analyses (e.g., for various break sizes and locations, equipment failures, operator actions, and so forth).without uncertainties have not been made for each LTC PRA success criterion, as was done for short-term co'e cooling; and, consequently,
. 1
" Low-margin" sequences have not been pre-defined relative to i long-term cooling.
I o As a result of the above items, and because these analyses must support the PRA success criteria for LTC, all LTC success sequences from the focused PRA expanded event trees will be considered in I
the sequence groupings.
o Any screening of low-risk-significant sequences will be done by l
! group rather than by individual sequences, such that the
! contribution of screening to core damage or large release uncertainty l is small.
l BDS Apre 21.1997 4 l
_ , , _ . .~. _. . . . . - -- - ..
J APSOO Long-Term Coahng Thermel/Hydrogh Unce, tawny GENERAL APPROACH (Continued) l o WCOBRA/ TRAC will be used, running in window mode, in accordance with' guidelines and practices used for the DBA o Success will be based on maintaining adequate core cooling, generally consistent with DBA requirements per 10CFR50.46 o For risk-significant scenarios with failed containment isolation, it is necessary to determine the effects of the isolation failure on containment pressure, the water inventory loss from containment, and the resulting effect on containment sump water levels. ,
1 Calculations will be made to determine the necessary J WCOBRA/ TRAC inputs, with appropriate conservatisms to allow for I calculational uncertainty.
l O
l l
BOS Apre 25,1997 5
)
AP400 long-Term Coseine Thermai/Hyalradio Uriomrtaewy !
I I
wCoBRA/ TRAC CASES l
The following cases are being analyzed with WCOBRA/ TRAC. These
)
i are cases for which we anticipate successful core cooling with no core j
( uncovery. l l
Containment isolated ;
l o 2" Cold-Leg LOCA: .
l 1 CMT or Accumulator Failed to inject (3 CMT/Accum. injected);
EITHER:
e 2 ADS-4 valves open'AND 1 Barik of ADS Stage 1/2/3 valves open, OR ,
1 e 3 ADS-4 valves open; recirculation through 2 open recirculation paths in 1 of 2 recirculation lines; PCS operates i
This case (counting contributions from all non-DVI LOCAs) represents potential focused PRA impacts, if not successful, of
\
~ 4% CDF, ~ 15% LRF i
DVI line breaks with only 2 ADS-4 valves open have low potential risk significance l
l This case addresses baseline PRA success criteria for ADS following LLOCA (potential impacts of ~17% baseline PRA CDF,
~ 10% baseline PRA LRF) sos Apre 2s.1ss7 e
- e APeoo Lone Term coesine twmyer m. une ri ny WCOBRA/ TRAC CASES (Continued) l l
l Containment isolated (Continued) l 1
o DVI Line Break:
1 CMT or Accumulator failed to inject (3 CMT/Accum, injected),
- 3 ADS-4 valves open, 0 ADS-1/2/3 valves open (failure of all stage 1/2/3 valves),
recirculation through:
- 2 open recirculation paths in 1 recirculation line -
no flow into RCS through broken DVI path after compartment
^
floods PCS operates This case represents relatively small potential focused PRA impacts, but potentially sign!ricant baseline PRA impacts, if not successful ,
This case is being run primarily to support basis for baseline PRA success criteria; T/H conditions are similar to DVI Line Break analyzed for DBA i
i BOS Apre 21,1997 7
6 AP900 Long-Term Cooding Thermat/Hydredee Uncertaanty WCOBRA/ TRAC CASES (Continued)
. Containment Not Isolated j o DVI Line Break:
- 1 CMT and 1 Accumulator injected,
- 4 ADS-4 valves open,
- 0 ADS-1/2/3 valves open, recirculation through: ,
- 2 open recirculation paths in 1 recirculatio,n line,
- no flow into RCS through' broken DVI path after compartment l l
floods -
I PCS operates
- This case represents potential focused PRA impacts, if not successful, of ~ 2% CDF, ~ 27% LRF
- Other failed containment isolation cases are not analyzed because l they either:
- represent small PRA impact, or
- include conditions substantially more favorable (e.g., non-DVI, more CMT/ Accumulator water in sump, more 1
recirculation paths) i l sos Apra 21,1ss7 s
e,
?
r INFORMATION PROVIDED TO JOE SEBROSKY :
i Attached are Westinghouse comments on the NRC's AP6(0 PRA Meeting Summary for the meeting held on April 15,1997. Per the NRC letter, it states a "dr. aft of this meeting summary was provided to Westinghouse to allow them the opportunity to ensure rh.t the representations of their comments i and discussions were correct." The attached are Westinghouse comments on the letter.
A copy of this information was provided to Joe Sebrosky (NRC) during a meeting on May 6,1997. j l
C. Haag (Westinghouse)
SU/97 I
I i
es w es:43 m.ees pei e
nac romu ses u.s. Nuct.eAn nEouteToav commissa oAre rien TELECOPlER TRANSMITTAL TWE a
WARNING: Most facsimile machines produce copies oil thermal paper. The image produced is highly unstable and will deteriorate significantly in a few years. Reproduce copice onto plain paper prior to filing as a recond.
TO NAMI: TELEPHONE 2 G NAME ANO LOCAllOdi OF COMPANY (# MBC)
TELECOPY NUMBEA YEAinCATiON NUMBER FROM 1 NAME TELEPHONE AMLSrOP o'S l TELECOPY DATA NUW8ER OF FAGES PRORTY j NMEDIATE THIS PAGE + PAGES - TOTAL TER (SAsc@)
$8ECIAt e4&TRUCT60NS 4
T
\S 'kh1 j MWMW '
N N kdotA.) od P coma bAce \
sgw 1& d ,
V PROBLEMS DISPOSITION OF BRIGINAL w any protueme occur or W you do not recem M tre paese, o*8. Anet tut ***Py has noen sent, praeees the eetgenal as requesend basew. paese == snecent me espnet me as eseemsJ RETURN TO SENDER TELEPHONE CALL AND SENDEA WILL PCK UP DISCARD PROCESSED BY(w/TA' L3J VEnirlED BY (INmALs; s,sc conw ass itsan Nwerno on mcw wen MAY 5 '97 8: 42 PAGE.001
~
] ~
em m:43 m.ees 002 ;
i
. p\ UNITED STATES
(
!p *
- ! NUCLEAR REGULATORY COMMISSION l wAssunorom, o.c. muss.am k.....
i.
i i APPLICANT: Westinghouse Electric Corporation FACILITY: AP600 k
SUBJECT:
Supt %RY OF APRIL 15, 1997, MEETING WITH WESTINGHOUSE TO DISCUSS ISSUES ASSOCIATED WITH THE AP600 PROBASILISTIC RISK ASSESSMENT ,
j (PRA) l i
4 1 l
) The subject meeting was held on April 15, 1997, in the Rockville, kryland,
- I offices of the Nuclear Regulatory Commission (NRC) between represennatives of ;
i Westinghouse, and the NRC staff. Attachment 1 is a list of meeting attendees, i Attachment 2 and 3 are the handouts provided during the meeting by the staff
- and Westinghouse respectively. ,
2 Prior to the meeting the staff sent Westinghouse a letter dated April 3 1997, which detailed three areas of the PRA in which the staff had concerns. ,The areas were the focused PRA, steam generator tube rupture (SGTR) accident
- sequence modeling, and " safety insights" developed from the PRA.
l 4 Focused PRA The focused PRA, which is a sensitivity study of the baseline PRA, was recognized as being important by both Westinghouse and the staff in resolving
- the issue of Regulatory Treatment of Non-Safety Systems (RTNSS). The staff j was especially concerned with the impact that revisions to risk-important j failure data had on the results of tie focused PRA. In addition, the staff was concerned about the loss of AC power assumption that Westinghouse used in
- the focused PRA.
l The risk-important failure data that were discussed during the meeting i included the data for the following components: In-containment refueling j
3 '.
water storage tank trdection check valves, squib valves, and Reactor Coolant j T. Pump (RCP)* breakers. The staff believed additional justification is needed to support the failure data for the above components. (Westinghouse presented j the data in attachmont 4 during this discussion). The staff agreed to do an a f " g i final review of theifailure data for the check valves, squib valves, and RCPbr i breakars. Westinght se agreed to assemble the pertinent data on the above components from thei PRA into a single package e assist the staff in e reviewing the mater 1. Following the review, th staff consitted to forward 3 their findings to stinghouse. u p, ,
- pnermdy &Smed
] The loss of AC >ower essumption used in the focused PRA was then discussed.
j In the focused >RA, Westinghouse assumed that, after the initiating event l occurred, all non-safety systems (including AC power) failed. However,
. special consideration was given to the Anticipated Transient Without scram j events where main AC power was assumed available to the rod control system.
I f
4 MAY 5 '97 8: 42 PAGE 002 1
1
es e m:43 %N m.ees ge3 i (pV' l(l7 6 8% (fM8 l ~t-l The sta believed that the loss of AC power assumption eliminated the j likelihood some high risk important failures such as the failure of one or
- more reactor colant pumps t . trip. Westinghouse believed that the modeling l of the loss of A power was c nsistent with the agreements reached and ECY papers concerning RTN35.- Westinghouse further stated documented in th(e]the focused PRA was never i j that they believed !
conservatises associated with the loss of certain non-safety systems with the i i conservatises (e.g. adverse systems interactions) associated with other non- l safety systems being available. The staff agreed to evaluate the less of AC l power assumption and how it could be treated in the " Adverse Systems Interac-tions* portion of the RTNS$ review to capture events like the reactor coolant l pumps falling to trip, within the context of the focused scope PRA. We scTR accident seauence modelino '
W 7%<r M #jfboo U n*O} g,,,g a The staff was concerned with lack of the 1-hydraul'ic analyses supporting the assumption that the AP60 design can use on-safety related systems only as the first line of defen to mitigate SG accidents. Westinghouse believed that the AP600 nique features anMhet they were answering similarquestionsaboutSGTReventsfromtheReactor/SystemsBranch. The staff agreed to have the Reactor Systems Branch inv ~1ved in the review of the thermal-hydraulic analyses that Westinghouse will providing in response to concerns regarding the PRA SGTR modeling.
Kafety insichts
- The staff was concerned with a delay in Westinghouse's submission of the l safety insights for external events. The staff was also concen.ed that during a review of the safety insights for the internal events, several important PRA i assumptions may have been overlooked and not included in the submittal. '
Westinghouse responded that the safety insights for the external events had i been transmitted April 11, 1997. Westinghouse whee believapFthat the safety i insights list they developed for internal events was an inclusive list of th ;;t=ti;l b;=t :f t'e efety insighty;i;t
. = :;n. = th Ster; %;i;; 0;tr;l 0;n;t."=ttJan n ;;..=,;;d ;b Westinghouse proposed ,
that the staff 614s:aWEset of insights andgehen g share /AhtW44st with Westinghouse. The%ww44eWwouhb6 hen-h v ,r:d rf "r tied a r eld-p Ris w.a tMMMi.- Unless Westing >use determined that there was technically incorrect information in the sta s list there would be ne new 1
s%g/ 69 L Yktn. SW2W w & covers $ Y M % d'y $2 40 9) he MRC O f M3 7
~m 4 QJf% .
.n.e e MAY 5 '97 8: 43 PAGE.003 1
' j ON 08:43 GO.088 D04 I cl/n{h 1
}$ Y{f O a- Gst~, n -ez.
1 meetings or information tran . The staff agreed to provide feedback to l Westinghouse on its proposa draft of this meeting susinary was provided to Westinghouse to allow them the opportunity to ensure that the representations of their cossments and discussions were correct.
soph M. $ebrosky, ect Manager tandardizatten Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket Mo.52-003 Attachments: As stated cc w/ attachments:
See next page l
I l
(
i MAY 5 '97 8143 PAGE.004 l
t nW i Westinghouse '"5 R50?A V
- s I
- .( s - 5 sa7 asBrian A. Mc Intyre4_"s l
FAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION DATE: May 5,1997 NAME: Tom Hayes TO: Bill Huttman LOCATION: Westinghouse l PHONE: PHONE: 412-374-4420 l COMPANY: NRC l
LOCATION:
Cover + Pages = 1+6 REMOVE ALL STAPLES PENCIL WILL NOT TRANSMIT USE BLACK PEN PLEASE MAKE COPlES OF TWO-SIDED PAGES l
Comments:
l Bill, Attached are revised versions of the SSAR markups sent to you by Earl Novendstern on April 28. The proposed SSAR changes have been revised following Westinghouse intemal review comments and comments provided by Jim Lazevnick in our telephone call on May 1. I have also attached a road map to the changes between the April 28 version and this version and the Westinghouse responses to the three qu stions raised in the phone call.
Regards, M
cc: B. McIntyre (NRC Informal Correspondence), E. Carlin, E. Novendstern 1
i Phone Number of Receiving Equipment:
FAX. dot
I ROAD MAP TO PROPOSED SSAR REVISIONS l l
The following 4 pages of proposed SSAR revisions were previously sent to the NRC for review on April 28,1997. The proposed revisions have been changed based on ;
comments received from intemal Westinghouse review and NRC review. The l changes are summarized as follows:
I Table 1.8-1, item 8.3 --
l l Revise transient voltage requirement to require the voltage to meet safety
! analysis assumptions.
Add requirement for switchyarri breakers to be set with consideration given to preserving the plant grid conriection follow!ng a turbine trip.
Section 8.2.2 -- !
Third paragraph revised to incorporate review comments and add the time delay requirement on reverse-power trip.
i The sixth (last) paragraph from the April 28 version has been incorporated into
! the third paragraph.
i Section 8.2.3 --
Minor editorial changes.
Section 8.3.1.1.1 --
Minor editorial changes.
l l
9 Y
. _ _ . .. . - _ . _ ~ _ . ~ . . . _ _ . - _ - _ - _ _ _ . _ _ . . _ . . . _ _ _ __ _ , _
Westinghouse Response to Questions Raised May 1,1997 Regarding Grid Connection l
l .
1 Is there a time delay which inhibits the reverse power generator trip following a turbine trip?
W RESPONSE: Yes. The time delay is shown in SSAR Figure 7.2-1, Sheet 14. I The delay is specified as 15-60 seconds. No reverse power trip is allowed during this time period. The COL will determine the final delay time (between
" 15 and 60 seconds) based on grid concems. A description of this time delay l has been added to SSAR 8.2.2 (third paragraph).
- 2. A COL item is needed to specify that the switchyard breakers are set with consideration given to preserving the grid connection following a turbine trip.
W RESPONSE: Agree. This item has been added to the proposed changes to ,
Table 1.8-1 and the discussion in Section 8.2.2 (third paragraph). I
- 3. ' Will a voltage at the undervoltage trip point keep the RCPs pumping at the assumed flow rate for safety analyses?
i W RESPONSE: No, the COL requirement has been revised to specify that the-transient voltage must remain above the voltage required to maintain the flow assumed in the safety analyses. Proposed revisions to Table 1.8-1 and Section 8.2.2 reflect this requirement.
l i
1
\
4
e Tabb 1.8-1 (Sheet 3 of 7)
SUMMARY
OF AP600 PLANT INTERFACES WITH REMAINDER OF PLANT Matching Section item Interface ~ ofSub.
No. Interface Interface Type item section 6.1 Inservice inspection requirements for the Requirement of Combined License 6.2.1 -
containment AP600 applicant program 6.2 Off site environmental conditiens assumed for AP600 Interface . Site specific 6.4 Main Control Room and technical support parameter center habitability design i
7,1 Listing of all design criteria applied to the Not an Interface N/A 7 design of the I&C systems
! 7.2 Power required for site service water NNS and Not an N/A 7 instrumentation Interface
, 7.3 Other provisions for site service water NNS and Not an N/A 7 instrumentation Interface
, 8.1 Listing of design enteria applied to the design NNS Combined License 8 l of the offsite power system applicant coordination
- 8.2 Offsite ac requirements NNS Combined License 8 l Steady-state load applicant l
Inrush kVA for motors coordination Nominal voltage Allowable voltage regulation l Nominal frequency 3
l Allowable frequency j fluctuation 1 Maximum frequency decay rate Limiting under frequency value for RCP 8.3 Offsite transmission system analysis: NNS Combined License 8.2 l Loss of AP600 or largest unit applicant analysis
! J
- Voltage operating range I Transient stability must be maintained and I the RCP bus voltage must remain above i l the voltage required to maintain the flow !
I assumed in the Chapter 15 analyses for a l minimum of three (3) seconds following a l
l tuttMne trip.
l the protective devices controlling the ;
I switchyard breakers are set with '
I consuleration given to preserving the plant l
I grid connection followmg a turbine trip.
4 ,
L8-5
l . I
('
F 1 1
5 I 8. Dectric Pocer L
l l 8.2.1.2 Transformer Area The transformer area contains the main stepup transformers, the unit auxiliary transformers, and the reserve auxiliary L isformer. Protective relaying and metering required for this equipment is located in the turbine building. The necessary power sources ,
I (480 Vac.120 Vac. and 125 Vde) to the equipment are supplied from the turbine building. !
l See subsection 9.5.1 for a discussion of fire protection associated with plant transformers.
I I l One feeder connects the transformer area with the switchyard to supply power to/from the l main stepup transformers for the unit. An arrangement is shown in Figure 8.3.1-1. j i
l l 8.2.2 Grid Stability 4 L
1 The AP600 is designed with passive safety-related systems for core cooling and containment I i integrity and therefore, does not depend on the electric power grid for safe operation. This ;
l ! feature of the AP600 significantly reduces the importance of the grid connection and the i requirement for grid stability. The AP600 safety analyses assume that the reactor coolant i i pumps (RCPs) can receive power from either the main generator or the grid for a minimum 1 of 3 seconds following a turbine trip.
I l ! The AP600 main generator is connected to the generator bus through the generator circuit l t breaker. The grid is connected to the generator bus through the main step-up transformers ,
l and the grid breakers. The RCPs are connected to the generator bus through the RCP '
l breakers, the 4.16 kV switchgear, and the unit auxiliary transformers. During normal plant ,
I l operation the main generator supplies power to the generator bus. Some of this power is l
l used by the plant auxiliary systems (including the RCPs); the rest of the power is supplied i to the grid.
I , 4 If. during power operation of the plant a turbine trip occurs the motive power (steam) to
! ! the turbine will be removed. The generator will attempt to keep the shaft rotating at l '
synchronous speed (governed by the grid frequency) by acting like a synchronous motor.
1 The reverse-power relay monitoring generator power will sense this condition and. after a
- time delay of at least 15 seconds. open the generator breaker. During this delay time the I generator will be able to provide voltage support to the grid if needed. The RCPs will receive I power from the grid for at least 3 seconds following the turbine trip. The COL will perform a grid stability analysis to show the grid will remain stable and the RCP bus voltage will i
remain above the voltage required to maintain the flow assumed in the Chapter 15 analyses i i for a minimum of three (3) seconds following a turbine trip. The COL will set the protective !
devices controlling the switchyard breakers with consideration given to preserving the plant
, grid connection following a turbine irip. '
if the turbine trip occurs when the grid is not connected (generator supplying plant house loads only), the main turbine-generator shaft will begin to slow down as the energy stored
! in the rotational inertia of the shaft is used to supply the house loads (including RCPs). The i system will coast down until the generator exciter can no longer maintain generator terminal voltage and the generator breaker is tripped on either generator under-voltage or Revision: t O Apiii 30.1930 ^ 8.2-2 T Westinghouse
4
- 8. Dectric Pocer I.....
' l exciter over-current. This coast doc will la .t at least 3 seconds before the generator breaker trips.
i
- The sequence of events following a loss-of-offsite-power event is the same as those i described for grid-disconnected operation. 7 i 8.2.3 Conformance to Criteria i
The offsite sources are not Class IE. Commercial equipment is manufactured to the !
industrial standards listed in subsection 8.2.5. The design meets General Design Criterion ;
- 1. Unit trips occur at the generator breaker and do not cause the loss of the preferred '
power source to the plant electrical systems. The AP600 does not require ac power sources l I
for mitigating design basis events; Chapter 15.0 describes the design bases assumptions I utilized for analysis of these events. The AP600 meets the intent of General design Criteria 17 as outlined in Section 3.1.
Conformance with General Design Criterion 18 is provided by the test' and inspection capability of the system.
8.2.4 Standards and Guides !
In addition to the General Design Criteria the industry guides and standards listed as !
Reference 2 through 4 are used as guides in the design and procurement of the offsite power '
system.
I 8.2.5 Combined License Information for Offsite Electrical Power i
Combined License applicants referencing the AP600 certified design will address the design of the ac power transmission system and its testing and inspection plan.
The Combined License applicant will address the technical interfaces for this nonsafety-related system listed in Table 1.8-1. These technicalinterfaces include those for ac power requirements from offsite and the analysis of the offsite transmission system.
, 8.2.6 References
- 1. ANSI C2-1990. National Electric Safety Code.
- 2. ANSI C37.010-1972. Application Guide for ac High Voltage Circuit Breakers.
- 3. ANSI C37.90-1989. IEEE Standard for Relays and Relay Systems Associated with Electric Power Apparatus.
l 4. ANSI C57.12.00-1973. General Requirements for Distribution. Power. Regulating Transformers, and Shunt Reactors.
[ Revision:-?O ,
l T Westinghouse 8.2-3 Apd! 20.1000 ,
l
- 8. Electric Pooer buses tagged with even numbers (ES2. ES4. etc.) are connected to the other unit auxtliary transformer. These 4.16 kV buses are provided with an access to the maintenance source through normally open circuit breakers connecting the bus to the reserve auxiliary transformer. Bus transfer to the maintenance source is manual The arrangement of the 4,16 kV buses permits feeding functionally redundant pumps or group of loads from separate buses and enhances the plant operational flexibility. The 4.16 kV switchgear powers large motors, and the load center transformers. There are four switchgear (ESI. ES2. ES5 and ES6) located in the annex building, and four (ES3, ES4. ES7 and ES8)in the turbine building.
l l
The main stepup transformers have protective devices for sudden pressure, neutral i
overcurrent, and differential current. The unit auxiliary transformers have protective ;
I devices for sudden pressure, overcurrent. differential current. and neutral overcurrent. If I i
these devices sense a fault condition the following actions will be automatically taken: trip I
high-side (grid) breaker. trip generator breaker, trip exciter field breaker, trip the 4.16 kVl ,
I buses connected to the faulted transformer. The reserve auxiliary transformer has l l protective devices for sudden pressure. overcurrent, and differential current. The reserve I
auxiliary transformer protective devices trip the reserve supply breaker and any 4.16 kV I buses connected to the reserve auxiliary transformer The onsite standby power system powered by the two onsite standby diesel generators supplies power to selected loads in the event of loss of normal, and preferred ac power supplies. Those loads that are priority loids for defense-in-depth function based on their specific functions (permanent nonsafety toads) are assigned to buses ESl and ES2. These plant (degree ofpermanent redundancy for eachnonsafety loads load is described are divided in the sections in two for the respective functiona systems .
Each load group is connected to either bus ESI or ES2. Each bus is backed by a non-Class lE onsite standby diesel generator. In the event of a loss of voltage on these buscs, the diesel generators are automatically started and connected to the respective buses. The source incoming breakers on switchgear ESI and ES2 are interlocked to prevent inadvertent connection of the onsite standby diesel generator and preferred / maintenance ac power sources to the 4.16 kV buses at the same time. The diesel generator however,is capable of being manually paralleled with the preferred power supply for periodic testing. Design provisions protect the diesel generators from excessive loading beyond the design maximum rating, should the preferred power be lost during periodic testing. The control scheme, while protecting the diesel generators from excessive loading, does not compromise the onsite power supply capabilities to support the defense-in-depth loads. See subsection 8.3.1.1.2 for starting and load sequencing of standby diesel generators.
Two separate 4.16 kV switchgear buses ES5 and ES6 located in the annex building power four reactor coolant pumps. Each pump is powered through two Class IE circuit breakers i
connected in series. These are the only Class IE circuit breakers used in the main ac power system for the specific purpose of satisfying the safety-related tripping requirement of these pumps. The reactor coolant pumps connected to a common steam generator are powered i
Revision: M) l .ApnL3&fJ97 83-2 W Westinghouse l
l
- . i i
1 .
1
,i no i
bl j 10 ', i 'VTD10 h 4
i bf. Off Cg&k (Vldf Cups h S.D.2 and 5.3 or Jim .
$n'l l W4 netC, -D bisO)ss (\ls i
crnh en S.J.
We's aft of 3.bo ,
~3 morrow .
l
.,[
es .
wma css a
i i
! s . aP 8
! l i
1 i
3 1
l* __ Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 g nizations Organizations shall be established for unit operation and corporate management, respectively. The organizations shall include the positions for activities affecting safety of the nuclear power plant.
- a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, ar.d all operating organization positions.
These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job
- descriptions for key personnel. positions, or in equivalent foms of documentation. These requirements shall be documented in the (FSAR];
- b. The (Plant Manager] shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;
. c. The (a specified corporate executive position] shall have corporate responsibility for overall plant nuclear safety and shall take.any measures needed to ensure acceptable performance of the staff in operating, maintaining, and l providing technical support to th'e plant to ensure nuclear l safety; and '
t i d. The individual,s'who train the operating staff,~ carry out healt.h physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operation pressures.
~
e,4 5.2.2 Unit Staff REddcE S2.c.4MJ 5'.Z.2 M Td /ASIAT
}
Note: The final Unit St -
~aing the minim'm u
shift crew comp i ermined based upon the verification an '
ion of the process described within SSAR , Human Factors Engineer .
N (continued) 3 h AP600 5.0-2 08/96 Amendment 0 i
organization 5.2 5.2 Organization 5.2.2 Enit Staff (continued) ggr4 M wmf/N5 Fir i
The unit staff organization shall include the following:
- a. A licensed or senior reactor operator shall be ass _ipned to each reactor containing fuel. This operator may afso fill the quirements of section 5.1.2 and other 5.2.2 se ions,
- b. At- ast one licensed Reactor Operator (RO) sh 1 be present in th control room when fuel is in the reac r. In addition, l while t e unit is in MODE 1, 2, 3, or 4, at east one licensed Senior R ctor Operator (SRO) shall be pre ent in the control room.
i c. Shift Crew C position may be less th the minimum -
requirement o the following table fpt a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> order to accommodate unexpected absence of on-duty shift er members provid imediate action is .taken
' to restore the shi crew compos ion to within the minimum requirements. -
Senior Plant Reactor Reactor
- l. Condit. Operator Operator MODE 1,d3,oh4 (2) [2]
'M00[5and6 \ (1] [1]
- d. Administrative pro dures shall be veloped and implemented tolimitthewor$4nghoursofunitst f who perform safety related functions (e.g., licensed SR0s, licensed R0s, auxiliary oper ors, and key maintenance ersonnel).
Adequate sh t covera'ge shall be maintained ithout routine heavy use pf overtime. The objective shall by to have !
operating /versonnel work a (12] hour day, nomihal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week while t unit is operating. However, in the eknt that ,
unfore en problems require substantial amounts O' overtime to i be us ,orduringextendedperiodsofshutdownf(oyrefueling, maj maintenance, or major plant modification, on temporary ba s the following guidelines shall be followed:
l . An individual should not be permitted to work more an 16 l hours straight, excluding shift turnover time; l
l l (continued) h AP600 5.0-3 08/96 Amendment O' l
l
I' organ uouon 5.2 5.2 Organization b 1
< l 5.2.2 Unit Staff (continued) ggft,g.E. Wi rW INTE47- !
. An individual should not be permitted to work more tha 6 j hours in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 da riod, all excluding shift turnover time;
- 3. A bre of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> should be allowed etween work i periods including shift turnover time; i
- 4. Except dur g extended shutdown periods the use of overtime sho ld be considered on an in vidual basis and )
not for the en ire staff on a shift. '
I Any deviation for the bove guideline shall be authorized in i advance by the [ Plant M ager) of h designee, in accordance
, with approved administrat e proc ures, or by higher. levels l
- of management, in accordanc wi established procedures and
- with documentation of the ba for granting the deviation.
Controls shall be included n th procedures such that l 1
individual overtime shall e revi ed monthly by the [ Plant l l
Manager) or his designe to ensure at excessive hours have i not been assigned. R tirie deviation or the above guidelines 1.. is not authorized. l l e. The [ Operations nager) shall hold an SR license. i i f. The Shift Tec ical Advisor (STA) shall prov e advisory -
! technical s port to the Shift Supervisor (SS) n the areas of j thermal hy raulics, reactor engineering', and pla t analysis i with re d to the safe, operation of the un.it.. I addition, i the ST shall meet the qualifications specified by he j Comm sion Policy Statement on Engineering Expertise f Shift.
4 - .
On of the SR0s on shift may perform the functions of e STA i . p vided that this individual has the specified enginee ng l xpertise.
f h AP600 5.0-4 08/96 Amendment 0 m,um aeosmecom r
t l
/ 9 5) Q 'r 6, 2. 2. l 5.2.2 Unit Staff .
f
\ Reviewer's Note: Dete,mination of the unit staffpositions, numbers, and l qualifications are the responsibility of the COL applicant. Input provided in WCAP.
14694 will be used in the determination. Each of the following paragraphs may i
need to be corrected to specify the plant specific staffing requirements.
~
The unit staff organization shall include the followmg:
- a. A non licensed operator shall be assigned to each reactor containing l
fuel and an additional non licensed operator shall be assigned for i
each control room from which a reactor is operating in MODES 1,2,
- 3, or 4.
Two unit sites with both units shutdown or defueled .
require a total of three non licensed operators for the two units.
l l
i
- b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuelis in the reactor. In addition, while the unit is in MODE 1,2,3, or 4, at least one licensed Senior Reactor l ,
Operator (SRO) shall be present in the control room.
! c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.g for a period of time
!j
' not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
- d. A [ Health Physics Technician) shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpec'ted absence, provided immediate action .
is taken to fill the required position.
- s. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licen' sed ROs, health ~ physicists, auxiliary operators, and key maintenance personnel).
Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an [8 or 12] hour day, nominal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week while the unit is i operating. However, in the event that unforeseen problems require i
substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:
- 1. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time: -
l* / Anrstr 6.2.7_ c w T-f 5.2.2 Unit Staff (continued) l 2. An individual should not be permitted to work more than l 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any l- 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period, all excluding shiR turnover time;
~
I 3. A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> should i= allowed between work t periods, including shiR turnover time;
- 4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized in advance by the [ Plant Superintendent) or his designee, in accordance with j approved administrative procedures, or by higher levels of management, in accordance with established procedures and'with documentstion of the basis for granting the deviation. .
Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the [ Plant Superintendent] or his designee to ensure that excessive hours have not been assigned. !
Routine deviation from the above guidelines is not authorized.
i DE w
The amount of overtime worked by unit staff members performing safety related functions shall be limited and controlled in accordance with the NRC Policy Statement on working hours (Generic Letter 8212).
- f. The [ Operations Manager or Assistant Operations Manager] shall hold an SRO license.
- g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engirieering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the quali6 cations specified by the Commission Policy Statement on Engineering Expertise on Shift.
6 l
l
- Unit Staff Qualifications 5.3 l
l 5.0 ADMINISTRATIVE CONTR01.S . :
l 1
5.3 Unit Staff Qualifications 86/4M W 'M /MSEAT 6.3.)
97 -
/
5.3.1 Each member bi 6;.e =i+ ctaffA11 meet. or exceed th'e minimum ,
qua tory Revision.2,1987).
1 i
l l
l l
, . l
! 1 l
l 1
i l
l L
l 1
l l
I i
i 1
f b AP600 5.0-5 08/96 Amendment 0 mi-a mm m..
. . . . . . . = . . - . - - . . .- . . - - - . . -. . _ - - -.
4 s /AGER T S.3.t Reviewer's Note: Minimum qualifications for members of the unit staff shall be '
specified by use of an overall qualification statement referencing an ANSI Standard
\ acceptable to the NRC staff or by specifying individual position qualifications.
Generally, the first method is preferable; howtwr, the second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures.
5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of [ Regulatory Guide 1.8, Revision 2,1987, or more recent revisions, or ANSI Standard acceptable to the NRC staff]. The staff not covered by [ Regulatory Guide 1.8) shall meet or exceed the minimum qualifications of [ Regulations, Regulatory Guides, or ANSI Standards l
acceptable to NRC staff).
- O h
t 4
l 4
w