ML20198H551
| ML20198H551 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 01/07/1998 |
| From: | Joshua Wilson NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9801130336 | |
| Download: ML20198H551 (38) | |
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Nuclear Safety and Regulatory Analysis ._
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. Nuclear and Advanced Technology Division! G. - "j l Westinghouse Electric Corporation _ '% ,
. . P.O. Box 355 ; . # , ;
' Pittsburgh, PA 15230i ft '
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SUBJECT:
OPEN ITEMS IN THE AP600_ DESIGN CERTIFICATION REVIEW ! l
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Dear'Mr Liparulo:
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. As a result of the Nuclear Regulatory Commission staffs continuing _ review of the AP600 1
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- de' sign certification application, the Quality Assurance and Maintenance Branch has prepared the ' _.
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a enclosed status report on the AP600 Initial Test ProgramnThis report identifies nine open items -
s and one confirmatory item that. require resolution by Westinghouse before the staff can complete : i
-its review. If you have any questions regarding this request, please contact me at s l '(301) 415-3145.
p Sincerely, j Jh,Wils n nibolicy Analyst .
Standardization Project Directorate .
e Division of Reactor Program Management; i, . Office of Nuclear Reactor Regulation - ,
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Docket No 52-003 e U
Enclosure:
As stated- ;
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DATE 01/7 /98 ' M 01/7 /98 P ##--
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,9 s Mr. Nicholas J. Uparuto Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Mr. Russ Bell-Advanced Plant Safety & Ucensing Senior Project Manager, Programs Westinghouse Electric Corporation Nuclear Energy Institute Energy Systems Business Unit 1776 i Street, NW P.O. Box 355 Suite 300 Pittsburgh, PA 15230 Washington, DC 20006 3706 Ms. Cindy L. Haag Ms, Lynn Connor Advanced Plant Safety & Licensing Doc Search Associates Westinghouse Electric Corporation Post Office Box 34 Energy Systems Business Unit Cabin John, MD 20818 Box 355 Pittsburgh, PA 15230 Dr. Craig D. Sawyer, Manager Advanced Reactor Programs Mr. Sterling Franks GE Nuclear Energy U.S. Department of Energy 175 Curtner Avenue, MC-754 NE60 San Jose, CA 95125 19901 Germantown Road Germantown, MD 20874 Mr. Robert H. Buchholz GE Nuclear Energy Mr. Frank A. Ross 175 Curtner Avenue, MC-781 U.S. Department of Energy, NE 42 San Jose, CA 95125 Office of LWR Safety and Technology 19901 Germantown Road Barton Z. Cowan, Esq.
Germantown, MD 20874 Eckert Seamans Cherin & Mellott 600 Grant Street 42nd Floor Mr. Charles Thompson, Nuclear Engineer Pittsburgh, PA 15219 AP600 Certification NE 50 Mr. Ed Rodwell, Manager 19901 Germantown Road PWR Design Certification Germantown, MD 20874 Electric Power Research Institute 3412 Hillview Avenue Mr. Robert Maiers, P.E. Palo Alto, CA 94303 Pennsylvania Department of Environmental Protection Bureau of Radiation Protection Rachel Carson State Office Building P.O. B0x 8469 Harrisburg, PA 17105-8469
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l 260,140F Open items in the initial Test Program j
, Background. By letter dated June 25 i1997, the NRC staff provided Westinghouse with the results of Ms review of the AP600 Standard Safety Analysis Report (SSAR) Chapter 14, initial Test Program (ITP) open hems that are being trac.Md in the AP600 Open item Tracking System
- (OITS). As a result of this review, the status of searal DSER and RAI Open items changed and additional information needed to complete the revow was identified.
I By' letter dated October 23,1997, Westinghouse provided its response to the NRC staff's June 25,1997 letter. Also, by sepa' ate correspondence on October 10,1997 Westinghouse submitted proposed SSAR changes to incorporate AP600 design changes to address post 72 .
hour actions. Both letters described proposed SSAR changes that were subsequently incorporated into Revision 17 of the SSAR, dated October 31,1997. Therefore, this status report encompasses the acceptability of changes incorporated into Chapter 14 as a result of Revision 17 to the SSAR as well as those for which Westinghouse has proposed a solution to be
- . included in a future revision to the SSAR (Confirmatory). -
! OITS 1124/DSER Open item 9.5.1.4-7: In Subsection 9.5.1.4.8, "Prooperational Testing," of the DSER, the staff found that additional Information was required from Westinghouse to establish >
the acceptability of the fire protection system (s) preoperational test program in complying with Section C.4.e of BTP CMEB 9.51. This issue was identified as DSER Open item 9.5.1.4-7.
A in its August 13,1996, response to the NRC, Westinghouse stated that " Subsection 14.2.9.2.8, Fire Protection System Testing, has been revised to state that the system operates as specified in Subsection 9.5.1 and in appropriate design specifications. These documents identify the
! applicable NFPA standards for the testing of individual components in the fire protection system.
Subsection 14.2.9.2.19 and 14.2.9.4.13 describe testing of the plant lighting and communication
, systems, respectively. The breathing apparatus provided at the plant and the use of this equipment will be identified by the COL applicant, as part of the fire protection personnel training."
In Revision 11 to the AP600 SSAR chapter 14, Subsection 14.2.9.2.8,
- Fire Protection System Testing," under " General Test Method and Acceptance Criteria," Westinghouse stated "The following testing demonstrates that the system performs its defense-in-dep'h functions specified
- in Subsection 9.51 and as specified in appropriate design specifications
- The capability of the
- seismic standpipes to supply the required fire water quantity and flow rate is verified."
The NRC staff disagrees with Westinghouse's conclusion that verifying that the seismic standpipes can supply the required fire water quantity and flow rate demonstrates that the fire i
protection system " performs its defense-in-depth functions specified in Subsection 9.5-1 and as
, specified in appropriate design specifications". Westinghouse needs to modify this subsection to
. encompass testing of the AP600 fire protection system in an integrated manner, i.e., fire doors, fire dampers, smoke control systems, automatic fire detection underground fire main, fire pumps, automatic suppression systems, electrical isolation devices for non-safety related equipment in opposite divisional fire areas, and trained fire brigade. This is to insure that a strong and seliable
- fire protection program is available to fight, contain and extinguish any type of fire prior to fuel load. Additionally, Westinghouse needs to incorporate the other subsystems in the Initial Test Program to insure:
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- 1. . Fire detection will be available to detect fires in their incipient stage and alert key personnel of fire conditions.
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. 2. Fire. Barriers, fire walls, fire dampers and smoke control systems limit spread of fire and smoke.
- 3. Fire pumps, underground fire main and water supply system will be available to provide a strong water supply to fight Ares in safety related areas and non safety related areas.
- 4. Fire Brigade will be adequately trained to fight fires including electncal and flammable liquid type fires.
The integration of these systems ensures a strong Defense-in-Depth system that provides an acceptable level of fire protection for the AP600 Advance Reactor, in its May 9,1997, response to the staff, Westinghouse proposed that SSAR Subsection 14.2.9.2.8 be revised to incorporate the NRC comments with the following
- exceptions:
- a. the training of the Fire Brigade is the responsibility of the COL applicant as specified in SSAR subsection 9.5.1.6
- b. the AP600 fire protection system does not contain smoke control equipment Although the changes to Subsection 14.2.9.2.8,
- Fire Protection System Testing," of Chapter 14 that address testing of fire detection, barriers, and fire water supply systems are acceptable, the staff concludes that the Westinghouse response does not address all of the issues raised in the NRC's request. Specifically, Westinghouse has not addressed testing of automatic suppression systems or electrical isolation oevices for non-safety related equipment in opposite divisional fire areas. Therefore, OITS 1124/DSER Open item 9.5.1.4 7 remained open.
In its October 23,1997, response to the NRC,' Westinghouse stated that Subsection 14.2.9.2.8,
" Fire Protection System Testing," would be revised to include testing of sutomatic fire suppression equipment and electrical isolation devices for non safety related equipment in opposite divisional fire areas.
While changes incorporated into Subsection 14.2.9.2.8, Revision 17, addressed most of the NRC staffs concems in this area, the NRC staff does not agree that the AP600 design does not contain smoke control equipment. Specifically, in Subsection 9.5.1.1.1, " Safety Design Basis,"
vVestinghouse specified that the fire protection objective ; include prevention of smoke, hot gases, or fire suppressant from migrating from one fire area to another to the extent that they could adversely affect safe shutdown capabilities, including operator actions. Additionally, in Subsection 9A.3.1.2, " Auxiliary Building-Nonradiologically Controlled Areas, " Westinghouse specified the smoke control of fire areas in the nonradiological controlled portions of the auxiEary building that contain the main Class 1E electrical equipment rooms served by the nuclear island nonradioactive ventilation system (VBS). Accordingly, Westinghouse needs to modify Subsection 14.2.9.2.8, General Test Method and Acceptance Criteria, items a) and c) to include pressure testing for the weismic stendpipe water supply, and testing of the HVAC smoke control and exhaust systems sections, respectively. Therefore, OITS 1124 remains open.
OITS 1162/DSER Open item 10.4.71(RAI 410.263): In the DSER, the staff found that Westinghouse should provide procedures for testing feedwater hammer occurrence.
Westinghouse responded in their August 13,1996 letter that Subsection 14.2.9.1.7, " Expansion,
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. 3- t Vibration and Dynamic Effects Testing," nw revised to include testing to start /stop startup feedwater to the steam generators to vertfy that unsoceptable feedwater hammer does not occur.
Staff review of this subsection determined that it does not provide sufficient information for-
- testing feedwater hammer occurrence. Additionally, Sut,section 14.2.9.2.2, " Main and Startup _
Feedwater System," should be modified to include the following:
- a. perform FW system test and monitor that no effects due to water hammer are detect 4d.
- b. check for water hammer noise and vibration using suitable instrumentation. .
- c. visual inspection indicates that the integrity of FW piping, support, and fooding have not been violated.
Therefore, OITS 1162/DSER Open item 10.4.71 remained open.
l In its May 9,1997, response to the NRC, Westinghouse proposed to include testing of dynamic events (e.g. water hammer) for all applicable systems in subsection 14.2.9.1.7.c (including applicable portions of main and startup feedwater piping) to address the NRC comments. The
. NRC staff confirmed that Westinghouse has revised Subsection 14.2.9.1.7, " Expansion, Vibration, and Dynamic Effects Testing," to specifically address the testing, monitoring and visual
- inspection for the effects of water hammer on the feedwater system as requested. However,
!- these changes will be evaluated in conjunction with Westinghouse's pending response to i RAI 410.263/01410.307F. Therefore, OITS 1162/DSER Open item 10.4.7-1 remains open.
OITS 1244/DSER Open item 14.2.84(RAI 240.24): In the DSER, the NRC 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 -
Subsection 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. Review of this subsection 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, Subsection 14.2.9.1.6 should be modified to include applicability of this testing to subsequent AP600 plants, or Appendix 1A in the SSAR should provida appropriate justification for this exception to RG 1.68, Appendix A, item 1.n.(14)(f).
OlTS 1244/DSER Open item 14.2.8 6 remained open.
The AP600 does not provide active, safety-related HVAC for the main control room, l&C equipment rooms, and class 1E de equipment rooms. The habitability of the main control rooms t is provided by operation of the MCR emergency habitability system, and by the passive heat g- sinks associated with the main control room structure. Likewise, the environmental conditions that the qualified l&C equipment and class 1E equipment will be exposed to are based on the
. passive heat sinks associated with the building and structures that house this equipment. In its May 9,1997, response to the staff, Westinghouse stated the following:
"In the AP600, a design basis hostup analysis of the main control room, l&C equipment rooms, and class 1E de equipment rooms is performed, and 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 Sc, and item 8c in
. Table 2.2.5-4) specifies that an evaluation will be performed using as built information and heat i
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loads from installed equ'pment for the 1) MCR,11) l&C equipment rooms, iii) Class _1E de equipment rooms. In addition, this evaluation considers the as-built passive heat sinks associated with these rooms, as specified in Certifed Design Material Section 3.3, Nuclear Island
! Building Structures. The acceptance criteria for this heat sink capacity analysis results are that: I
- l} the temperature rise for the MCR is less than or equal to 15' F for the 72-hour period; li) the i maximum temperature for the 72-hour period for the l&C rooms is less than 125' F; iii) the ' I maximum temperature for the 72-hour period for the Class 1E dc equipment rooms is less than l l or equal to 125' F.
1 This evaluation ensures that the as-built information for the pertinent buildings and structures, as l well as the as-built MCR equipment and l&C and class 1E heat loads are less than that assumed i in the design basis heatup analysis for the MCR and the safety-related equipment rooms noted. ,
- i The first plant only test specified in 14.2.g.1.6 is a test of the long term heatup characteristics of the main control room, l&C and class 1E equipment rooms. It is performed to demonstrate the heatup characteristics of these rooms when they are subjected to a known heat load. This 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 requirements specified in the Certifed Design Material. As a passive heatup of these rooms is not dependent 1
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."
Proposed AP400 asAR Changes i
that states this test needs to be performed for the first plant only provided the design basis heat loads used as assumptions in the heat sink capacity analysis bound the actual as-built !
information and heat loads.
The staff is evaluating Westinghouse's proposal to perform first-plant only testing of long term performance of the main control room habitability system on the basis that requirements in the certified design material (CDM) provide adequate assurance that future plants will comply with the design and performance requirements of the first plant. However, during pmliminary-4 discussions on this issue, Westinghouse was informed that a "first plant-only" test approach is
- unacceptable. While the staff agrees that "The ability of the habitability system to maintain the main control room envircament as well as temperatums in the protection and safety monitoring 4- system cabinet and emergency switchgear rooms during a long term loss of the nuclear island nonradioactive ventilation system may be wnTsed with a limited duration fest [ emphasis added)."
The staff does not agree that such testing: (1) is so impractical or burdensome that it should be
- - performed on the first plant oniv. (2) should not include specific verification and duration acceptance criteria. Therefore, OITS 1244 (RAI 260.26) remained open.
i in its October 23,1997 response to the staff, Westinghouse stated that Subsection 14.2.g.1.6,
" Main Control Room Emergency Habitability System Testing," would be revised to require testing i of the main control room habitability (item e) as part of the preoperational testing for all plants. in
( addition, SSAR Section 6.4 would be revised to specify the criteria for air quality and 5
temperature. This test would be specified to be performed for a sufficient duration to verify that criteria are met at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Testing to verify that protection and safety monitoring system ,
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l cabinet and emergency switchgear rooms heatup at a rate consistent with the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I temperature limit criteria (item f), would be revised to be performed as a preoperational test for 1 all plants. <
' The NRC staff finds that the soceptance criteria changes to subsection 14.2.g.1.6, [ items (e) and :
(f)), in SSAR Revision 17, do not establish measurable parameters for verifying that the l protection and safety monitoring system cabinet and emergency switchgear rooms heatup at a <
rate consistent with the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> temperature limit criteria. Westinghouse has modeled the Main Control Room (MCR), the protection and safety system cabinets, and the emergency switchgear rooms heatup rate. Therefore, Westinghouse should be able to establish a more specific time duration for the test (i.e. the time-frame which is necessary to validate the model).
Absent an objective validation of the heaW rate model, a complete 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> test would be -
required. Additionally, Westinghouse should remove the exception to RG 1.68, Appendix A, ;
Item 1.n.(14)(f) from Appendix 1A of the SSAR. Therefore, OITS 1244/DSER Open item 14.2.8 6 (RAI 260.26) remains open.
OlTS 1254/DSER Open item 14.2.8-18: Closure of this is, was contingent upon the 1 satisfactory resolution of OITS 1255/DSER Open item 14.2.t.J 1, below. Since the staff finds the response to OITS 1255 acceptable, OITS 1254 is closed. ,
01781255/DSER Open item 14.2.8.31:. In the DSER, the NRC staff found that the preoperational and startup test phase descriptions in Section 14.2.8 of the SSAR did not provide assurance that the operability of several of the systems and components listed in Appendix A of RG 1.68 (Rev. 2, August 1978) will be demonstrated. The test abstracts of Section 14.2.8 of the SSAR should be expanded to address the following items identified in Appendix A to RG 1.68, or Appendix 1 A of the SSAR should be revised to provide technicaljustification for any exceptions i taken, Preoperational Testino h 1.a.(2)(l) pressurizer safety valves 1.b.(1) control rod withdrawal inhibit and rod runback functions 1.c diverse actuation system, which protects the facility from anticipated transients without a scram (ATWS) i.e.(4) steam generator pressure safety valves 1.e.(10) feedwater heaters and drains
, 1.f.(2) - cooling towers and associated auxiliaries 1.J.(7) leak detection systems used to detect failures in the emergency core cooling system (ECCS) and containment recirculation systems located outside i
containment (for example, potential leakage in the normal residual heat removal (RHR) system or the post-accident sampling systems that could be used to recirculate reactor coolant outside containment after an accident)
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- 1.J.(8) automatic reactor power control system and pdmary T-average control system 1.J.(13) excois neutron instrumentation 1.J.(17) foodwater heater temperature,' level, and bypass controls 1.J.(20) instrumentation used to detect extemal and infomal flooding condNions 1.J.(22) instrumentation used to track the course of postulated socidents such as containment wide range pressure indicators, reactor vessel water level monitors, containment sump level monitors, high radiation detectors, and humidity monitors
- 1.J.(23) post accident hydrogen monitors 1.J.(24) annunciators for reactor control and engineered safety features 1.k.(2) personnel monitors and radiation survey instruments (As t% calibration program -
applied to these devices will be site specific, it nuld be appropriate to identify this as a COL action item.)
1.k.(3) laboratory equipment used to analyze or measure radiation levels and radioactivity concentrations 1.l.(5) isolation featuius for condenser offgas systems 1.m.(4) static load testing at 125 percent rated load of cranes, hoists, and associated lifting and rigging equipment 1.n.(5) secondary sampling systems 1.n.(g) drain systems and pumping systems serving essential areas 1.n.(12) boron recovery system 1.n.(13) communications systems relating to offsite emergency notification
- 1.n.(14)(c) Class 1E electrical room heating, ventilating, and air conditioning i 1.n.(14)(f) mein control room (including proper operation of smoke and toxic chen,ical detection systems and ventilation shutdown devices, including leak tightness of ducts).
, 1.n.(15) shield cooling systems
, 1.o.(1) dynamic and static load tests of reactor components handling system cranes, hoists, and associated lifting and rigging equipment l
1.o.(2) protective devices and interiocks of reactor components handling system
- i. equipment l
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7 1.o.(3) safety devices for reactor components handling systems equipment kr.jfal Fuel Loadino and Procritical Tests 2.f reactor core and other major components differential pressure and vibration testing after fuelloading Low Power Testina 4.1 control rod biock and inhibit functions Power Ascension Tests 5.m reactor core and major reactor coolant system components differential pressure 5.r process computer and control room computer 5.t pressurizer safety valves and secondary system safety valves 5.c.c gaseous and liquid radioactive waste processing, storage, and release systems (operating in accordance with design) 4 5.g.g design features to prevent or mitigate anticipated transients without scram l (ATWS)
, 5.k.k dynamic response of the plant for loss of feedwater heaters or bypassing feedwater heaters These issues were previously identified by the staff in Q260.30 and were subsequently identified in the DSER as Open item 14.2.8.3-1.
In its August 13,1996 response to the NRC, Westinghouse stated that " Subsection 14.2.9 has been revised to include test abstracts for appropriate AP600 systems and components as specified in RG 1.68, Revision 2, Appendix A."
In its November 8,1996 response to Westinghouse, the NRC staff found that Westinghouse had not satisfactorily revised test abstracts to demonstrate the requested items. A detailed review of the SSAR will be conducted to determine whether the test abstracts accurately reflect suitable test methods under the appropriate plant conditions. Therefore, DSER Open Item 14.2.8.31 remained open. Nonetheless, the following items were provided to Westinghouse as initial comments derived from a limited review of these items:
. Appendix A to RG 1.68, Section (d) identifies steam line atmospheric dump valves and relief valves to be included in the preoperational tetting. In Attachment 3 to the letter of July 16,1996, Westinghouse listed these veives to be included in SSAR Chapter 14 Subsections 14.2.9.2.1 and 14.2.9.1.2 respectively. However, the NRC staff could not find the testing of these valves in the above two SSAR subsections. Westinghouse is requested to add these valves according to Attachment 3.
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+ Appendix A to RG 1.68, Section (e) identifies steam generator pressure relief valves, turt>ine control and intercept valves, and main condenser hotwell level control system to be included in the preoperational testing. In Attachment 3 to the letter of July 16,1996, Westinghouse listed these items to be included in SSAR Chapter 14 Subsections 14.2.9.1.2,14.2.9.2.1, or 14.2.9.4.1. However, the NRC staff could not find the testing of SG pressure relief valves, turbine control and intercept valves in the above SSAR subsections. Westin9 house is requested to add these items according to Attachment 3.
- Appendix A to RO 1.68, Section (f) identifies cooling towers and associated auxiliaries, and raw water and service water cooling towers to be included in the preoperational testing, in Attachment 3 to the letter of July 16,1996, Westinghouse listed these items to be included in SSAR Chapter 14 Subsection 14.2.9.4.6. However, the NRC staff could not find the testing of cooling towers and associated auxillaries, and raw water and service water cooling towers in the above SSAR subsection. Westinghouse is requested to add these items according to Attachment 3.
In its December 6,1996 response Westinghouse stated the following:
"Section 14.2.9.1.2 Item a) commits to tests of safety related valves in the SGs which includes the SG Power operated Relief (atmospheric dump) Valves. This section will be revised to delineate these valves specifically under item a).
Section 14.2.9.2.1 lists th other va,lves mentioned (with the appropriate AP600-specific name).
Test 14.2.9.4.6 does not specifically mention cooling towers for the following reasons:
the circulating water system cooling tower is not within the scope of the AP600 design certification heat removal of an ultimate heat sink (such as a cooling tower) can not be tested during preops due to the absence of core power- commitments are made in 14.2.9.4.6 to test the t'timate heat sir,k (cooling tower or other) during hot functionals as appropriate The service water cooling towers are tested as specified in 14.2.9.2.6."
The NRC staff finds Revision 11 to the AP600 SSAR still does not address allissues identified under this open item. Specifically, the test abstracts of Sections 14.2.9 and 14.2.10 should be expanded to address the following items identified in Appendix A to RG 1.68, or Section 1.9 of the SSAR should be revised to provide technicaljustification for any exceptions taken.
- 1. PreooerationalTestinr1 1.J.(7) Leak dete; tion systems used to detect failures in normal RHR system, the post accident s ampling system, or other systems that could be used to recirculate reactor cot lant outside containment
- 2. Initial Fuel Loadino and Precritical Tests
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2.b Testing of control rod withdrawal and insert speeds 2.c Final functional testing of tb rsector protection restem 2.d Final test of the RCS to verify that system leak rates are acceptable 2.e Measurements of water que!ity.
- 4. Low Power T@
4.e Determination of flux distribution to vet.fy proper oore loading ar.o fuel enrichments 4.1 Control rod Nock and inhibit functkms 4.1 Operability and response time tests of main steam isoletion valves and their bypass valves at rated temp 3rature and pressure conditions 4.r Operability of RCS purification and cleanup systems 4u Operability of pressurizer pressure and level control systems
- 5. Power Ascension Tests 5.1 Design capability of rt.sidual or decay heat removal systems including turt>lne bypass valves, simospheric dump valves, normal residual heat removal, and feedwater systems, including demonstration that excessive flow instabilities will not occur 5.o Operabiiity of RCS leak detection systems 5.r Process computer and control room computer.
5.s Operability of pressurizer pressure and level control systems 5.u Operability and response time tests of main steam isolatior valves and their bypass valves 5.c.c Demonstrate that gaseous and liquid radioactive waste processing, storage, and release systems operate in accordance with design 5.k.k Dynamic response of the plant for loss of feedwater heaters or bypassing feedwater heaters Therefore, OITS 1255/DSER Open item 14.2.8.3-1 remained open.
In its May 9,1997, letter to the staff, Westinghouse responded with the following:
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- 1. Preocerational Testing 1.J.(7) The AP600 does not have a dedicated system to detect leakage outside containment for the purpose of detecting leakage in the reormal RHR system, post-accident sampling lines or other systems that penetrate containment and may circulate reactor coolant outside containment post-accident. This function is O performed by the Radioactive Waste Drain System which is tested as discussed in Subsection 14.2.g.3.4.
- 2. Initial Fuul Leadina and Procrttical Tests 2.b The Performance Criterion section of SSAd Subsection 14.2.10.1.13 has been modified to 'nclude rod withdrawal and insert speeos.
2.c Final functional testing is accomplished by the procritical setpoint verification (subsection 14.2.10.1.10) and the nuclear instrumentation procritical testing (subsection 14.2.10.1.g). These tests, in combination with the other procritical tests (such as rod control system test, rod drop time measurement, etc.) as well as the preoperational testing of the reactor protection system testing of system logic and operability of trip breakers and safety-related valves is sufficient to meet this requirement. Titis is consistent with the approach taken in procritical testing in current operating plants, 2.d RCS leak testing is performed during the preoperational testing as specified in subsection 14.2.g.1.1, Once fuel load is accomplished, the plant must conform to -
the plant Technical Specifications including the RCS leakage. Therefore, no additional test is required, 2.e Measurements of water quality at 0 percent power are performed and are discussed in subsection 14.7.10.4.8. Ti.is subsection has been modified to provide the applicable SSAR references for water chemistry requirements.
- 4. Low Power Testino 4.e This testing is performed during the ascension to power tests (as ackr.owledged in Reg. Guide 1,68 Appendix A section 4.e) as specified in 14.2.10.4.2. The objectives of this test has been modified to include determination of flux distribution to verify proper core loading and fuel enrichments.
4.1 Test 14.2.10.1.11 has been modified to include testing of control rod block and inhibit functions.
4.1 Operability and response time tests of main steam isolation valves and tholt bypass
- valves at rated temperature and pressure conditions is peiformed during the preoperational test of the steam generator system as discussed in Subsection 14.2.g.1.2. This subsection has been revised to explicitly include tnese valves.
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Operability of the RCS purification and cleanup systems is oorformed during the i preoperational test of the chemical and volume control system discussed in SSAR -
Subsection 14.2.9.2.3 as revised.-
4.u operability of pressurizer pressure and level control systems is performed during ,
the preoperational test of the reactor coolant system (subsection 14.2.9,1,1) and during the testing of the pressurizer spray capability (14.2.10.1.19) included in the pre critical testing phase of the ITP. -
- 5. Power Ascension Tests 5.1 Design capability of residual or decay heat removal systems including turbine bypass valves, atmospheric dump valves, normal residual heat removal, and feedwater systems, including demonstration that excessive flow instabilities will not ,
l occur are all verified during the applicable portions of the various preoperational testing of these systems or subsystems and during the dynamic event testing specified in 14.2.9.1.7 and is also performed as required during the Dynamic Response test specified in 14.2.10.4.18 under the ascension to power test phase.
5.o RCS leak testing is performed during the preoperational testing as specifed in subsection 14.2.9.1.1, Once fuel load is accomplished, the plant must conform to the plant Technical Specifications including the RCS leakage. Therefore, no additional test is required.
5r This testing is performed during the preoperational testing of the plant control system (14.2.g.2.12), the data display and processing system (14.2.9.2.13) with the exception of the nuclear instrumentation. The nuclear instrumentation is tested during the ascension to power testing discussed in Subsections 14.2.10.4.2-4.
5.s Operability of pressurizer pressure and level control systems is performed during the preoperational test of the reactor coolant system (Subsection 14.2.9.1.1) and during the testing of the pressurizer spray capability (14.2.10.1.19) included in the pre-critical testing phase of the ITP. In addition, the performance of the pressurizer
. pressure and level control are verified during the Thermal Power Measurement and 8
- Stetspoint Data Collection test (14.2.10.4.17). This test has been modited to ,
include this verification.
5.u Operability and response time tests of main steam isolation valves and their bypass valves at rated temperature and pressure conditions is performed during the preoperational test of the steam generator system as (scussed in Subsection 14.2.9.1.2. This subsection has been revised to explicitly include these valves. ,
5.c.c ~ Demonstration that the gaseous and liquM radioactive waste processing, storage, and release systems operate in accordance with design is performed during the preoperational testing of these systems (14.2.9.3) and is supplemented by the Radiation and Effluent Monitoring System test (14.2.10.4.14) during the ascenum
- to power phase of the ITP.
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5.k.k A test of the dynamic response of the plant for the loss of a feedwater heater has been added to the ascension to power test as Subsection 14.2.10.4.27, l
The staff has reviewed the Westinghouse response to each of the Regulatory Guide 1.68 items and finds the response acceptable with the following exceptions:
- 4. Low Power Testina 4.1 While the proposed changes to Subsection 14.2.9.1.2(a) clarify the type of valves to be tested, they do not address the regulatory position in this parograph of 4 Regulatory Guide 1.68 that testing be performed at rated temperature and J
pressure. SSAR Sections 3.9.6 and 10.3 are referenced in the test abstract as sources of additional critoria, but review of these sections did not lead to the conclusion that testing would be performed at rated temperature and pressure.
Westinghouse should specifically state the test conditions for these valves.
- 5. Power Ascensi .23313 4
5.1 Testing of the heat removal capability of the power operated relief valves (atmospheric dump valves) could not be located. The last sentence in Sub7ction 14.2.9.1.2 states that the heat transfer performance of the steam generator system
- is verified as part of reactor coolant system testing. While it is agreed that ADV testing could be performed as part of RCS testing, it is not evident from the existing test abstracts for the RCS, that they also involve verification of proper heat removal through the ADVs.
5.o Subsection 14.2.9.1,1 does not specify how leakage testing / verification is to be i performed. Technical Specification compliance following fuelload relies on proper
- operation of plant leakage detections systems. GDC 30 requires that means for 3 detection and identification of RCS leakage be provided. Leak detection is also t necessary to support the application of L.eak Before Break to high energy piping as described in Section 3.6 of the SSAR. Proper operation and calibration of the leakage detection systems, as described in SSAR Section 5.2.5, should be verified, and the ITP should be revised accordingly, i 5.u - See response to 4.1 above.
5.k.k The Regulatory Guide states that the dynamic response of the plant should be demonstrated for the most severe case of feedwster temperature reduction -
resulting from a credible loss of a feedwater heater (s) due to single failure or operator error. The controlled removal of feedwater heaters from service, as
- described in Subsection 14.2,10.4.27, does not meet the intent of the Regulatory Guide. Additionally, the Regulatory Guide states that this test is to be performed at 50 percent and 90 percent of rated power. The test prerequisites only require that power be "no less than 50 percent".
On these bases, OITS 1255/DSER Open item 14.2.8.3-1 remained open.
i in its October 23,1997, response to the staff, Westinghouse that the following revisions would be made to address the Staff's comments:
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Subsection 14.2.9.1.2, " Steam Generator System Testing," would be revised to clearty specify that valves be tested at nominal operating pressure and temperature.
Subsection 14.2.9.1.2 would also be revised to include verification of the capability of the steam 9enerator power Operated relief valves to provide the required heat removal.
Subsection 14.2.9.1.1, " Reactor Coolant System Testing," item d, would be revised to specify theat instrumentation related to reactor ooolant system leak detection be properly calibrated and their operation verified.
. Subsection 14.2.10.4.27, "Feedwater Hester Out of Service Test," would be revised to include testing simulating the loss of one of two main feedwater heaters with the reactor operating at 50 percent power and at 90 percent power.
The staff finds the incorporation of these changes into the ITP, SSAR Revision 17, acceptable and, therefore, OITS 1255/DSER Open item 14.2 S.3-1 is closed.
OITS 1256/DSER Open item 14.2.8.41: In the DSER, the NRC staff found that the preoperational and startup test phsse descriptions in Section 14.2.8 of the SSAR did not provide assurance that the operability of several of the systems and components listed in the following RGs would be demonstrated. The test abstracts of Section 14.2.8 of the SSAR should be expanded to address the following items, or Appendix 1 A of the SSAR should be revised to provide technical justification for any exceptions taken.
- RG 1.68.2, Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water Cooled Nuclear Power Plants"- Preoperational test abstract 14.2.8.1.94, " Remote Shutdown, does not provide sufficient detail to verify conformance with the following Regulatory Positions (RP) of RG 1.68.2:
Hot Standby DemonstrStion (RP C.3), including the following:
With initial conditions of the reactor at a moderate power level (10 to 25 percent),
demonstrate that plant systems are in the normal configuration with the turbine generator in operation and with the minimum shift crew Using only credited remote shutdown equipment, demonstrate the capability to achieve hot standby status, and maintain stable hot standby conditions for at least 30 minutes.
Cold Shutdown Demonstration (RP C.4), including the follow'.ng:
- with the plant at hot standby conditions;
- with the procedurally designated crew positions; using only credited remote shutdown equipment, demonstrate the capability to perform a partial cooldown by performing the following actions:
lower reactor coolant pressure and temperature sufficiently to permit operation of the residual heat removal (RHR) system
14 -
initiate and control operation of the RHR system :
establish a heat transfer path to the ultimate heat sink reduce reactor coolant temperature approximately 50 F using the RHR system
- RG 1.68.3, "Prooperational Testing of instrument and Control Air Systems"- -
Prooperational test abstract 14.2.8.1.6, " Compressed and instrument Air Systems," does not provide sufficient detail to verify conformance with the fotowing RPs of RG 1.68.3:
After noolers, oil separators, air receivers, and pressure-reducing stations (RP C.2)
Flow, temperature, and pressure meet design specifications (RP C.4)
Total air demand with leakage meets design (RP C.5)
- Single failure criterion (RP C.7)
Budden and gradualloss of system pressure and appropriate response of air power equipment (RP C,8)
Functional test for increase in the air supply system pressure does not cause loss of operability (RP C.11)
RG 1.140 Prooperational test abstracts 14.2.8.1.28," Containment Air Filtration System,"
, 14.2.8.1.29, " Radiologically Controlled Area Ventilation Test," and 14.2.8.1.88, "High-Efficiency Particulate Air Filters and Charcoal Absorbers" do not provide sufficient detail i to verify conformance with the following RP of RG 1,140.
" heaters (RP C 3.a)
- profilters (RP C 3.m)
HEPA filters DOP tests (RPs C.3.b and C.5.c) ductwork (RP C.3.f) fans and motors mounting and ductwork (RP C.3.1)
- dampers (RP C.3.1) adsorber sections / cells and activated charcoal (RPs C.3.h and C.5.d)
These issues were previously identified by the staff in Q260.31. This was identified in the DSER as Open item 14.2.8.4-1.
In its August 13,1996 response to the NRC, Westinghouse stated the following:
Subsection 14.2.g.1.12 has been revised to include testing to verify the ability to initiate actuation signals to the systems / components required for reactor shutdown from the remote shutdown workstation. Note that the AP600 remote shutdown workstation-provides the operator with the same capability to maintain the plant at hot shutdown conditions, or to cool the plant down; as is provided from the main control room.
Therefore, the operator does not need to perform manual actions or operate equipment from local control panels. in addition, test abstracts for the instrument and compressed c.ir system and appropriate HVAC systems have been revised.
b L . , i 16 . _
in its November 8,1996 response to Westinghouse, the NRC staff concluded that Westinghouse -
had not satisfactorily revised test abstracts to demonstrate the requested items, in general, the
! - revised test abstracts provide less detail than did their predecessors.: A detailed review of the
- SSAR will be_ conducted to determine whether the test abstracts soeurately reflect appropriate - '
. _ test conditions. Therefore, DSER Open item 14.2.8.4-1 remained openf 1
in its December 6,1996 response Westinghouse stated the following:
" Westinghouse would appreciate specific comments from the staff on the ap ' yriate test abstracts so that we can address the staffs concems in these areas more readily, )
i l For the instrument and con *rol air systems, and the containment air filtration system, it i 4-should be noted that these are non safety systems in the AP600 and therefore may not '
require as explicit details for testing these systems."
]
During its review of Revision 11 of the SSAR, the NRC staff concluded that Westinghouse has j satisfactorily address the staffs concems related to RG 1,140, However, the following issues '
RG 1.68,2 I Section 1.g. Appendix 1 A, states that an exception has been taken regarding testing of the )
- AP600 remote shutdown workstation in accordance w;tn Regulatory Guide 1.68.2. The basis for this exception is the similarity of the remote shutdown station (RSS; to the main control room H workstations, the testing of plant control capability from the main control room, and the testing of )
- the RSS controls and indications during pre operational testing. !
I - The RSS testing in the ITP is described in Subsections 14.2.9.1.12 and 14.2.9.2.12.
Subsection 14.2.9.1.12. " Protection and Safety Monitoring System Testing," tests, in part, i manual reactor trip capability from the RSS, and also tests the processing of manual actuation
- commands from the RSS to the protection logic cabinets through simulated command inputs to )
i _ the logic cabinets and simulated logic cabinet outputs on component status to the RSS.
. Subsection 14.2.9.2.12. " Plant Control System Testing," provides testing of RSS control
, functions based on simulated inputs at the RSS and verification of proper output through contact operation, component actusilon, or electrical test. ;
While similarity of the RSS workstations to those in the main control room, and successful testing -
p of the main control room workstations and individual RSS process signals can provide a certain 4
- 4 avel of corfidence with regard to proper RSS operation, they do not suffice as a replacement for
} - Integrated t, 'rol system testing of the RSS. In addition, although the control room and RSS l- vorkstatiort .,tay be similar, the working environment is different to the operator from that of the cetrol room which is the normal workspace. The operators should demonstrate the ability to 1
perform plent control in an abnormal work environment with the minimum not of controls and
. Indications available under postulated control room evacuation scenarios. The Section 14 test apstracts should therefore be modified to demonstrate the remote shutdown capability of the
- plant in accordance with RG 1.68.2,
)
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L RG 1.68.3 During the March 21,1995 meeting, Westinghouse committed to resolving RAI 410.161 (item No. 244) by including pre-operational testing as described in RG 1.68.3, "Prooperational Testing of Instrument and Control Air Systems"in Subsection 14.2.9.4.10. " Compressed and instrument Air System Testing." Specifically, the following information still needs to be added to the test -
abstract.
. s. All safety-related pneumatically operated valves should be verified to fail in the position specified in SSAR Table 9.3.1 1 upon a complete and sudden loss of instrument air pressure and a gradualloss of instrument air pressure, b, The instrument air system should be' functionally tuted to ensure credible failures resulting in an increase in instrument air system pressure will not cause loss of
, operability
- c. The instnJment air system air quality should be tested to meet ANSI /ISA 67.3, " Quality Standard for instrument Air."
- d. While at instrument air system normal steady state conditions, if practical, simultaneously operate those plant components requiring large quantities of instrument air for operation, to verify pressure transients in the distribution system do not exceed acceptable values,
- e. Verify that the total air demand at normal steady state conditions, including leakage from the systems, is in accordance with design.
- f. Additionally, the test abstract should include the following statements:
3
+ " Demonstrate the operability of the air compressor dryers and filters, intercoolers, aftercoolers, moisture separators, and air receivers."
. " Verify appropriate differential pressures (e.g., detta P across profilters and afterfilters)."
+ " Verify relief valve settings."
Therefore, OITS 1256/DSER Open item 14.2.8.4-1 remained open, in its May 9,1997, response to the staff, Westinghouse stated the following:
1.68.2 A test of the remote shutdown workstation has been added as Subsection 14.2.10.4.28.
1.68.3 Westinghouse has modified the Compressed and Instrument Air System preoperational test (subsection 14.2.9.4.10) to provide sufficient detail to show conformance to the applicable portions of RG 1.68.3.
The NRC staff finds Revision 13 to SSAR Section 14.2.10.4.28," Remote Shutdown Workstation," acceptable. However, the exception ;o RG 1.68.2 in Appendix 1 A of the SSAR should be deleted based on the new test for the remote shutdown station as described in Subsection 14.2.10.4.28 of the ITP, 1-
I Revision 13 to SSAR Subsection 14.2.9.4.10 adequately incorporated the NRC staffs comments on the test abstract with the exception to a specific reference to ANSt/ISA S7.3, " Quality Standard for instrument Air." Nevertheless, the test abstract also states that testing will verify system functions as described in SSAR Section 9.3.1. Subsection 9.3.1.4, " Tests and inspections,' dies the standards for air quality, including ANSI /ISA S7.3 On this basis, l OITS 1256 remained open pending revision to the exemption in Appendix 1A of the SSAR for l Regulatory Guide 1.68.2.
In its October 23,1997, response to the NRC, Westinghouse stated that the exception in Appendix 1A of the SSAR for RG 1.68.2 would be deleted based on the previously revised testing for the remote shutdown station as described in Subsection 14.2.10.4.28. The NRC staff confirmed that Westinghouse has deleted the exception to RG 1.68, Appendix A.5, regarding the remote shutdown station from SSAR Section 1.9, Appendix 1 A, but inadvertently (apparently) left intact the same exception to RG 1.68.2.
While Westinghouse has addressed the staffs concems with respect to SSAR Subsection 14.2.9.4.10 as described above, Westinghouse has not satisfactorily addressed the staffs concems identified in RAI 410.308. Specifically, Westinghouse's changes to Subsection 14.2.9.4.10 only clarify the reference to SSAR Section 9.3.1 and do not address the !
inconsistencies in the testing criteria between Sections 3.9.6,9.3.1.4, and 14.2.9.4.10, as described in RAI 410.308.
Section 14.2.9.4.10 references 9.3.1.4 which specifies air operated valve testing in accordance with RG 1.68.3. As noted in 410.308, Section 14.2.9.4.10 states that air-operated valves in safety systems w41 be tested as part of the test progcam for the individual system, and the test abstract for the individual system typleally references Section 3.9.6 for valve testing requirements, which does nut address RG 1.68.3 testing.
The issue identified in 410.308, therefore, remains unaddressed. Westinghouse may resolve this issue in several ways: (1) Section 3.9.6 could be revised to address air-operated valve testing in accordance with RG 1.68.3, (2) the individual test abstracts for systems with safety ralated air operated valves could be revised to reference Subsection 9.3.1.4 for air operated valve testing requirements, or (3) Subsection 14.2.9.4.10, General Test Method and Acceptance Criteria d), could be revised to state the following:
"d. Testing is performed to verify the fail safe positioning of safety-related air operated valves for sudden loss of instrument air or gradual loss of pressure as described in SSAR Section 9.3.1.4."
Based on the above, OITS 1256/DSER Open item 14.2.8.41 remains open pending (1) deletion of Westinghouse's exception to RG 1.68.2 fror. SSAR Section 1.9, Appendix 1A, and (2) resolution of RAI 410.308.
OITS 1257/DSER Open item 14.2.5-1: In the DSER, the NRC staff recommended that Section 14.2.9 of the SSAR be retitled as ' COL License Information-Initial Test F-rogram." This title would more accurately reflect the purpose of this secticn within the SSAR (i.e., to identify the information to be supplied to the NRC by COL applicants referencing the AP600 design). In addition, the content of Section 14.2.9 of the SSAR should be revised to include " site specific aspects of the plant," such as T following systems that may require testing "to satisfy certain AP600 interface requirements":
1 i .
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. is.
- olectrical sw".chyard equipment
. ' site security plan equ:pment }
. personnel monNors and radiation survey instruments t
.
- automatic dispatcher control system (if applicable) {
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) This Nem corresponds to Q260.32 and was identified in the DSER as Open hem 14.2.91. !
in its August 13,1996 response to the NRC, Westinghouse stated that 'Section 14.3 provides t l reference to COL information items to vertfy she spoolfic aspects of the plarc. that may require !
' testing are wNhin the certifloation envelope?
l In its November 8,1996 response to Westinghouse, the NRC staff ciertfied that in its July 22, 4 1994 letter to the NRC, and in response to Q200.32. Westinghouse had agreed to the NRC r l staff s proposed revisions and recommendations. However, Revision 9 to the SSAR has !
~
relocated such information to Section 14.3, " Certified Design Matettal? In Hs August 13, i994 response to this open hem, Westinghouse stated that Section 14.3 *provides reference to COL ;.
information Nems to vertfy site specific aspects of the plant that may require testing are within the l
, [ design) certification envelope?
i 4 Based on the above, the NRC staff requested that Westinghouse identify which subsection of I Section 14.3, " Certified Design Material,' designates " site specific aspects of the plant" that may l require testir,g by the COL applicant to satisfy certain AP600 interface requirements, such as l those identified in Q260.32. DSER Open item 14.2.91 remained open.
a i
j in Ns December 6,1996 response, Westinghouse stated the following' I
" Interface requirements as defined by 10 CFR Part 52 47 (a)(1)(vil) are discussed in
?
Section 14.3, fourth bullet, it is not necessary to provide a list of possible systems that ;
- may or may not require testing, as this determination will be made by the NRC at the time i of the COL application."
l The NRC staff disagrees with Westinghouse's interpretation of $52.47(a)(1)(vil). Westinghouse ,
1' needs to specifically identify the structures and systems that ere wholly or partially outside the design scope and specify the interface req @emsms for those systems, including testing to be
! performed by the COL applicant. Westinghouse should address this issue in conjunction with
! 0640.52. OITS 1257/DSER Open hem 14.2.91 remained open.
l In its May 9,1997, response to the NRC, Westinghouse stated thsM able 1.61 provides a e summary of the AP600 plant interfaces with the remainder of the plant (outside of design :
certification). Westinghouse proposed that subsection 14.3.4 be revised to state that verified
. testing of interfacing systems is the responsibility of the combinted license applicant, in addition, i Subsection 14.4.5, interface Documents, would be added to include a list of the these plant j- interfaces that may require testing to be performed by the combined license applicant. The lists ;
would consist of the following:
j.
4 storm drains (item 2.10) site selsmic sensors (item 3.3) - i offsite (switchyard) ac power systems (item 6.2) e b . - circulating water heat sink (item 9.3) .
=. raw and sanitary water systems (item 9.5)
- 2-4 4
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- e equipment assoolated with fire protection program (Mem 9.7) e personnel monitors and portable radiation survey instruments (item 12.1) i The NRC staff finds that OITS 1267/DSER Open item 14.2.91 remains open pending ,
satisfactory resolution of Q640.82.
0178 2647/Q200.39: ITP Test Abstract 14.2.8.1.30, Foodwater Control System: The Test Method subsection should be revised to incorporate vertfloation that automatically irJtiated valve <
l opervolosure cycling and timing meets the system design basis requirements.
In its Au9ust 13,1998, response to the NRC, Westinghouse stated that "The test abstract for the steam 9enerator system in Subsection 14.2.9.1.2, spoolfles that the proper operation of the main ,
j and startup feedwater valves is verifed, including autometic opervolose valve operation and a
timing. Additional testing of the main feedwater valves is specifed with the reactor at power ;
- during the startup testing described in Subsection 14.2.10.1.22."
~
l l in its November 8,1996, response to Westinghouse the NRC staff concluded that Subsection 14.2.9.1.2 does not specify that the proper operation of main and startup feedwater valves is vertred as noted. Therefore, OITS 2547/Q260.39 remained open. i
! In its December 6,1996, response to the NRC, Westinghouse stated the following:
J
" Subsection 14.2.9.1.2 bullet (a) verifies proper operation of safety related valve functions and includes the main feedwater SG isolation valves.
[ Subsection 14.2.9.2.2 bullet (s) tests the defense in-depth valve functions associated with the FWS to verify their proper operation. This subsection is revised to include verification ,
of the proper functioning of the main feedwater pump and control valves."
While the NRC staff agreed with Westinghouse with regards to the content of Subsections 14.2.9.1.2 and 14.2.9.2.2, during its review of Revision 11 of the SSAR, Chapter 14, the staff ,
i found that Subsection 14.2.10.1.20, *Feedwater Valve Stroke Test,'should be modified to i provide an acceptable reference to system design basis requirements for allowable closure and cycling times. Therefore, OITS 2547/Q260.39 remained open, in its May 9,1997, response to the NRC, Westinghouse stated that the performance criteria for this test would be revised to reference the appropriste SSAR subsection (7.7.1.8) and appropriate design specifications. The NRC staff found that Revision 13 to SSAR l Subsection 7.7.1.8 does not specify criteria directly related to the itsue of allowable closure and
' cycling times. Therefore, OITS 2547/Q260.39 remained open.
l in its October 23,1997, response to the NRC, Westinghouse stated that Subsection l 14.2.10.1.20, "Feedwater Valve Stroke Test," would be revised to delete references to i
Subsection 7.7.1.8.- Appropriate valve stroke and response times would be specified in applicable design specifications. The NRC staff finds the incorporation of changes in
- Subsection 14.2.10.1.20, SSAR Revision 17, requiring that valve timing be tested in accordance
- with design specifications acceptable. OITS 2547/Q260.30 is closed s
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OITS 2483/RAI 260.44: in RAI 260.44, the NRC staff requested that various reactor coolant l
! system flow measurements be ver6Aed, including baseline RCS pressure drops. Westinghouse j l responded in their August 13,1996 letter that Subsection 14.2.9.1.1, " Reactor Coolant System j
- Testing," has boon revised to include appropriate flow measurement testing. NRC review of this ;
4 subsection dolermined that baseline RCS pressure drop testing is not specified. Subsedion !
! 14.2.9.1,1 " Reactor Coolant System Testing,should be revised to include the establishment of, ;
j and appropriate acceptanos criteria for, baseline RCS pressure drops. OITS 2662 (RAI 260.44) ,
i remained open. !
I i in its May 9,1997, response to the NRC, Westinghouse stated that it would not require baseline RCS pressure drops to be recorded for individual RCS components. Overall system flow rete information is obtained during preoperational testing and startup testing of the RCS. The NRC j
. staff found that if determination of RCS pressure drops is not necessary for the APS00, then exooptions to the provisions in Regulatory Guide 1.68 Appendix A, paragraphs 2.f and 5.m ;
, should be included in SSAR Section 1.9. Therefore, OITS 2SS2/RAI 260.44 remained open. i in its October 23,1997, response to the NRC, Westinghouse stated that Subsection 14.2.9.1.1, l
!. " Reactor Coolant System Testing," item r) would be added to include measurement of the !
j pressure drop across major components and verification that the pressure drops are in ;
accordance with appropriate design specifications. l t :
i The NRC staff finds the incorporation of changes in Subsection 14.2.9.1.1, SSAR Revision 17, i requiring that pressure drops be tested in accordance with design specifications acceptable.
OITS 2562/Q200.44 is closed. '
l OITS 2864/RAI 240.50: In RAI 260.50, the NRC staff requested that, among other items, the 4
initial fuel loading test sequence should outline all systems required for the initial fuel loading.
Westinghouse responded in their August 13,1996 letter that Subsections 14.2.10.1.1 and 14.2.10.1.5 had been modified accordingly NRC review of these subsections determined that the revision did not completely address the noted issues. The prerequisites of either Subsection 14.2.10.1.1, "Fuol Loading Prerequisites and Periodic Chocks," or Subsection 14.2.10.1.5, " initial ,
Fuel Loading,'should be modified to outline all systems required for initial fuel loading, and n should additionally address criteria or prerequisites for minimum count rate, instrumentation
! signal to noise ratios, criticality predictions, and any special procedural actions. Rod withdrawal sequences should also be specified to be the same as for a normal startup. (RG 1.68, App. A, Section 3). OITS 2558/RAI 260.50 remained open.
l ;
In its Mv/ 9,1997, response to the NRC, Westinghouse stated that the minimum conditions for i initial fuelloading and initial critica!ity are provided in section 14.2.7. Section 14.2.7 would be ,
- - revised to state that the systems and conditions necessary to bring the plant into compliance with
, the Technical Specifications must be in place prior to initial fuelload. Minimum count rate, .
Instrumentation signal to noise ratios, criticality predictions, and rod withdrawal sequences would be addressed in the initial Criticality and Nuclear Instrumentation During Criticality tests (subsections 14.2.10.2.2 and 14.2.10.2.3 respectively).
The NRC staff foune) that the minimum count rate and signal to noise ratios as discussed in l . Regulatory GuHe 1.68, Appeodd A, Section 3, were not specified in the cited tests. Minimum cSunt rats and signal to noine ratio criteria should be included in the prerequisites of 14.2.10.2.2,
~
14.2.10.2.3 or other tests as appropriate. Therefore, OITS 2556/RAI 260.50 remained open.
t l
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r
o
- 21-In its October 23,1997, response to the NRC, Westinghouse stated that Subsection 14.2.10.2.7,
' Nuclear instrumentation System Verification During criticality," would be revised to specify that minimum neutron count rate and noise to signal ratio are within design specifications, as a performance criteria. This verification is a prerequisite for the initial criticality in Subsection 14.2.10.2.2. The NRC staff finds the changes incorporated into Subsections 14.2.10.2.2 and 14.2.10.2.3, SSAR Revision 17, acceptable. Therefore, OITS 2558/Q260.50 is closed.
OITS 2559/RAI 240.51: In RAI 260.51, the NRC staff requested that certain test abstracts be revised to provide specific acceptance critoria or oosign basis functional requirements traceable to the appropriate SSAR sections. Westinghouse responded in their August 13,1996 letter that the noted test abstracts have been revised. NRC rsview of the revised Chapter 14 submittal deterrr.)ned that a number of additional subsections require more detailed acceptance criteria.
The following subsections should be revised to provide or reference specific acceptance criteria or design basis functional requirements traceable to specific subsections or numbered paragraphs of the SSAR, the plant Technical Specifications, or other appropriate references that contain the detailed structure, system, or component design / performance criteria that is being verified by the testing:
14.2.10.1.7 14.2.10.1.8 14.2.10.1.9 14.2.10.1.10 t 4.2.10.1.11 14.2.10.1.12 14.2.10.1.15 14.2.10.1.16 14.2.10.1.19 14.2.10.1.20 14.2.10.2.3 14.2.10.2.4 14.2.10.3.2 14.2.10.3.3 14.2.10.3.5 14.2.10.3.6 14.2.10.4.3 14.2.10.4.5 14.2.10.4.7 14.2,10.4.10 14.2.10.4.13 14.2.10.4.15 14.2.10.4.20 Therefore, OITS 2559/RAI 260.51 remained open, in its May 9,1997, response to the NRC staff, Westinghouse provided the following table to outline the additional performance criteria references that would be added to the noted subsections of the SSAR.
l !
l o e f i
4 !
SSAR Sub6ectien Additional Reforence Pro ided !
\
i j 14.2.10.1.7 4.4.8:7.2 l 14.2.10.1.8 4.4.6: 7.2: 7.3 !
- 14.2.10.1.9 4.4.6 l J !
- 14.2.10.1.10 Technical Specifications :
T i 14.2.10.1.11 7.7.1.2 j 14.2.10.1.12 7.7.1.3 i 14.2.10.1.15 7.7.1.10 l 14.2.10.1.16 7.7.1.1 l i !
14.2.10.1.19 7.7.1.6 .
- 14.2.10.1.20 7.7.1.8 i
! 14.2.10.2.3 4.4.6 i 14.2.10.2.4 4.3.2.6 3
14.2.10.3.2 4.3.2.2.8 14.2.10.3.3 Technical Specifications
~
14.2.10.3.5 4.3.2.5 14.2.10.3.6 Design Specifications 14.2.10.4.3 4.4.6 14.2.10.4.5 7.7.1.1 14.2.10.4.7 7.7.1.1
} ,
L ;
14.2.10.4.10 4.3.2.4.16
=
14.2.10.4.13 7.7.1.8
=
14.2.10.4.15 9.4.6:9.4.1 w.
1 A2.10.4.20 7.7.1.1
. t j The NRC staff finds the added references acceptable with the folbwing exceptions:
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23 I i
SSAR Additional Reference Provided Comment i Subsection ,
I 14.2.10.1.7 4.4.6; 7.2 it is not clear what design requirements in l Section 7.2 are applicable to the incore !
instrumentation, and specifically the thermal couples, as stated in the revised ,
test. !
14.2.10.1.8 4.4.6; 7.2; 7.3 SSAR Subsection 4.4.6 does not describe the calibration procedure as indicated in the revised test abstract.
i 14.2.10.1.16 7.7.1.1 The referenced subsection does not provide nor define design requirements or ,
i limits for the RCS temperature variables listed in the test abstract performance criterion. ,
14.2.10.1.20 7.7.1.8 The specified criteria in the revised test were not found in Subsection 7.7.1.8.
14.2.10.2.4 4.3.2.6 SSAR Subsection 4.3.2.6 discusses criticality control during refueling (which merely references other SSAR sections),
and for fuel handling outside the reactor, and does not appear to be the appropriate reference.
14.2.10.3.2 4.3.2.2.8 SSAR Subsection 4.3.2.2.8 references Chapter 14 for the acceptance criteria.
14.2.10.4.5 7.7.1.1 SSAR Subsection 7.7.1.1 does not describe the enteria as stated in the revised test.
14.2.10.4.10 4.3.2.4.16 SSAR Subsection 4.3.2.4.16 does not appear to be the appropriate reference for design requirements related to limits on RCS temperatures as described in the revised abstract.
Based on the remaining issues that require resolution as noted above, OITS 2559/RAI 2ti0.51 remained open. '
in its October 23,1997, response to the NRC, Westinghouse stated that these references will be modified to include a reference to 'applicab!e design specifications." The following changes would be made to the identified test abstracts to resolve the remaining issues with regards to specific acceptance criteria or design basis functional requirements traceable to appropriate '
SSAR sections:
o *
. 24 !
e Subsection 14.2.10.1.7, *lncore instrumentation System Procritical Vert 6 cation,* would be revised to odd references to Section 7.6 and applicable design spoolfications.
e subsection 14.2.10.1.8, *Rosistance Temperature Detectore incore Thermocouple Cross .
Calibration,* would be revised to add references to Tables 7.21 and 7.H in sections 7.2 i and 7.3 tot,pectively.
l
- Subsection 14.2.10.1.18, *Proce4.s instrumentation Alignment,* would be revised to ;
delete the reference to subsection 7.7.1.1 and add reference to Tables 7.21 and 7.M in ;
Sections 7.2 and 7.3 ree;+#;@. ;
Subscotion 14.2.10.1.20, Teodwater Valve Stroke Test,* would be revised to delete Subsection 7.7.1.8 and only referonos applicable design 2;+2xtss.
+ Subsection 14.2.10.2.4,
- Post CrHiool Reactivity Computer Checkoul,* would be revised ,
to delete reference to subsection 4.3.2.6 and add reference to Section 7.7.
- Subsection 14.2.10.3.2,
- Determination of Physics Testing Range,* would be revised to delete reference to subsection 4.3.2.2.8. No reference performance orHeria is provided ,
since the aero power testing range is the result of the test as determined by the stated test method. This is consistent with previous tests of this nature as performed on current i operating plants. ,
- Subsection 14.2.10.4.5,
- Stariup Adjustments of Reactor Control Systems,* would be ;
revised to delete reference to subsection 7.7.1.1 and add reference to Section 5.1. ;
- Subsection 14.2.10.4.10,
- Process instrumentation Alignment at Power Constions,*
would be revised to delete reference to subsection 4.3.2.4.16 and add refereries to Section 5.1.
i Based on the incorporation of these changes in the ITP, SSAR Revision 17, the NRC staff finds >
the Westinghouse response acceptable. OITS 2559/RAI 260.51 is closed.
OlTS 2547/RAI 260.89: In RAI 260.59, the NRC staff requested that the load swing test be l revised to address a number of specific concems including providing acceptable ranges of key plant parameters. Westinghouse responded in their August 13,1996 letter that Subsection i 14.2,10.4.20, " Load Swing Test,* acceptance criteria has been expanded to include a review of plant response and adjustment of control systems,if necessary. Staff review of this subsection l determined that is has not been expanded from the earlier submittal. The performance ortlerion !
subsection should be modified to specify or reference the acceptable ranges of the evaluated parameters, OITS 2567/RAI 260.59 remained open, in its May 9,1997, response to the NRC, Westinghouse proposed that the performance criteria for this test be modified to reference 7.7.1.1. The NRC staff found that the reference to SSAR Subsection 7.7,1,1 is not sufricient to pmvide criteria for the measured parameters specified in ;
i the test abstract. Additional criteria or leferences specific to the measured parameters should be ,
provided, Therefore, OITS 2567/RAI 260.59 remained open.
l I in its October 23,1997, response to the NRC, Westinghouse stated that Subsection j 14.2.10,4.20,
- Load Swing Test,* would be revised to delete reference to subsection 7.7.1.1. l l
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i 26 (
i The response of the control system would be reviewed against a spec 6Ac control system setpoint !
and performance analysis that is performed prior to plant preoperational testing and is not part of ;
the SSAR. The NRC staff Ands the incorporation of this change into Subsection 14.2.10.4.20, i SSAR Revision 17, acceptable. OITS 2067/NAl 260.69 is closed. ;
- Of78 2871tRAI 240.83: in RAI 260.63, the NRC staff requested that preoperational testing
, should be provided for the 460V non Caess 1E transportable AC generator and its distribution panel. Westinghouse regrI: in their August 13,1996 letter that Subsection 14.2.9.1.16, :
'Long Term Safety Related System Support Testing," has been revised to include vert # cation of -t the proper operation of the generator. Staff review determined that this subsection only tests the !
abilNy to power post sooident monitoring instrumentation from the por'able generator. Table !
8.3.14 of the SSAR lists several other design loads, including onttain flued and portable HVAC {
systems, and Control Room / Remote Shutdown Station lighting, that are additioneNy potentia #y [
poworod from the ancillary generator. The portable units should be tested to ensure that they j can supply design load to these various systems. OITS 2671/RAI 260.63 remained open, j In its May 9,1997, response to the NRC, Westinghouse proposed that Subsection 14.2.9.1.16, !
"Long Term Safety Related System Support Testing" be revised to incorporate testing of the j r design loads of the various systems requiring electr6 cal power including post-accident instrumentation, control room lighting and ventilation, l&C toom ventilation, passive containment cooling system pumps, ancillary generator room lights, and ancillery' generator fuel heaters. In addition, this subsection would be revised to reflect recent design changes made to the AP600 in ,
response to SECY g6128 including: '
- Deletion of the testing of main control room compressed air supply which is no longer !
used forlong term alr supply, :
- The use of ancillary fans to provide ventilation cooling to the main control room and the ;
post accident monitoring instrumentation equipment rooms as described in SSAR ;
Section 9.4.1.
The addition of the passive containment cooling system water storage tank as a source of '
makeup water to the spent fuel pool, f The NRC staff finds that Revision 17 to SSAR Subsection 14.2.9.1.1 is consistent with Table 8.3.14 of the SSAR, Therefore, OITS 2571/RAI 260.63 is closed.
OITS 2842/Q240.48: Subsecteon 14.2.8.1.61, operations and Control Center System: This test abstract does not ieflect the design and configuration of the AP600 Operations and Control Center System. Specifically, the primary plant control system operator interface is a set of ' soft" control units that replace conventional switch / light or potentiometer / meter assemblies used for ,
operator interface with control systems. The function based test analysis serves as the basis for determining the alarms, displays, controls, and procedures in the main control area. The Test ;
Methods and Performance Criterion subsections of this abstract need to be revised to :
demonstrate acceptable performance of, and to encompass, these unique design features. !
I
'In tsaa Chapter s.3.1.1.F. 'stenty AC Power stoply * )
- non Class it diesel seneratore deelsnetten from 'trensporteble'sevision to 'encillery.'II, Westtrghause revloed the esovec l
)
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l 26 In hs August 13,1996 response to the NRC, Westinghouse stated that "The test abstract for the plant control system in subsection 14.2.9.2.12 has been revised to reflect the use of ' soft' controls and function-based analysis for alarms, displays, controls, and proosJuros used in the AP600.*
in its November 8,1996 respones to Westinghouse, the NRC staff found that the general test methods and soceptance artierla should include the use of " soft" controls and function based l
- analysis for alarms, displays, controls, and procedures used in the AP600. Thorofore,01T8 l 2642/Q260.68 remained open. '
In hs December 6,1998 response to the NRC, Westinghouse requested that the staff *peovide !
more specific information regarding the comment to include Was of ' son' controls in this test l abstract. While the term ' soft"is not used in this abstract, the test methods do include the use of i
- soft" controls during testing of the plant control system hardware and software " Upon further !
review, the NRC staff agrees with Westinghouse that the term
- operator interface features
- es L used in Subsection 14.2.9.2.12 encompasses the exercise of " soft" controls during the execution !
- of testing described therein. Therefore, OITS 2642/Q260.68 is closed. 2 OITS 2446/Q240.72: Subsection 14.2.8.1.81, Pressuriser Pressure and Level Control: The Test j Method subsection does not include testitig of signal selector and isolation devices, j Westinghouse should revise this subsection to encompass testing of these devices or should -
Identify the test abstract that encompasses such testing, in its August 13,1996 response to the NRC, Westinghouse stated that "The test abstract for the reactor coolant system in subsection 14.2.9.1.1 specifies that the proper operation of the pressuriser pressure and level controlis verified. Additional testing is also performed during the '
startup testing. Detailed methods for performing this verification, including signal selector and <
isolation devices, are to be included in the actual test procedures developed by the COL !
applicant."
in its November 8,1996 response to Westinghouse, the NRC staff found that the RAl's concem on testing of signal selector and isolation devices was not addressed in subsection 14.2.9.1.1 or ,
any other startup testing subsections. The NRC staff requested that Westinghouse either specify how the applicant can develop test procedures to cover those components, or modify the ;
appropriate test abstracts to reflect these tests. Therefore, OITS 2646 remained open. t in its December 6,1996 response to the NRC, Westinghouse stated that " Consolidated system level tests encompass multiple functions provided by integrated system assemblies, it is the intent that subsections (a) and (d) of subsection 14.2.9.2.12, ' Plant Control System. . ' include
- testing of the signal selector, distributed controllers, process bus multiplexers, etc. as related to Pressuriser Pressure and Level Control as well as other significant PLS functions.'
in a November 19,1997, letter to the NRC documenting the results of a November 14,1997, teleconference, Westinghouse proposed to revise item d) of the General Test Methods one :
Acceptance Criteria of subsection 14.2.9.2.12 to include testing of the signal selector processing l function. The NRC staff agreed that this change would resolve the staff's coricems with this test abstract. Therefore, OITS 2646/0260.72 is confirmatory pending the incorporation of this -
change in a future revision of the SSAR.
e.-- en e~ vw-
27 OITS 3644N3240.74: Subsection 14.2.8.2.44, Plant Control System: The scope of this test !
should be expanded to encompass all other Plant Control System subsystems as identi6ed in 1
- SSAR Chapter 7.1. Ahematively, Westinghouse should identify the test abstracts that currently j oncompass such subsystems. ,
, i in its August 13,1996 response to the NRC, Westinghouse stated that 'The test abstreet for the !
plant control system in Subsection 14.2.9.2.12 has boon revised to include the control functions !
speci6ed in SSAR Sootion 7.1" r
l In its NoYember 8,1996 response to Westin 50use, the NRC staff found that Suboeotion i 14.2.9.2.12 had not addressed all the control functions spooined in the SSAR. Therefore,01T8 !
2648/0280.72 remained open, t i
in hs December 6,1998 response to the NRC, Westinghouse stated that 'The plant control i systems functions to be tested are delineated in the two bullets under the subsection 14.2.9.2.12 !
labeled ' Purpose' and coincide with the functions listed in SSAR section 7.1.3, first paragraph.
While each function is not specifically mentioned in the General Test Methods and Acceptance
- Criteria of subsection 14.2.9.2.12, the general test methods of paragraphs a), b), c), and d) apply I to each function described above." Upon further review, the staff agrees that testing described in subsection 14.2.9.2.12 encompasses all control functions speci6ed in the SSAR. Therefore, OITS 2648/Q260.74 is closed. ;
OITS 8314/Q240.138: All the recombiner plates should be tested under [14.2.9.1.11) test b) of :
the General Test Acceptance Criteria and Methods unless it can be established that they are :
from the same batch or manufacturing lot. Westinghouse also needs to spoolfy how many plates will be tested once traceability to the same batch or manufacturing lot has been established.
In its June 11,1997, response to the NRC, Westinghouse stated that a response to this item was still being prepared. Therefore, pending receipt and evaluation of a response, OITS
$316/Q260.138 remained open.
In its October 23,1997, response to the NRC staff, Westinghouse stated that the performance of the autocatalytic recombiner plates (or cartridges) would be tested by the manufacturer for each lot or batch of catalyst material. The number of plates tested would be based on the guidance provided in ANSl/ASQC Z1.41993, " Sampling Procedures and Tables for inspection by ;
Attributes,' (formeriy Military Standard 105), required to achieve inspection Level lli quality level. l This results in a maximum number of non-conformities of 0.01 per hundred plates. The in-plant i test of plates (or cartridges) would be limited to one plate per batch of catalyst material in each recombiner to verffy that the as installed plates perform in accordance with the as-manufactured ,
plates. The NRC staff finds 'Nestinghouse's response acceptable. However, Westinghouse needs to revise subsection 6.2.4.5, as appropriate, to include reference to ANSI /ASQC Z1.4 -
1993. Therefore, OITS 5316/Q260.138 remains open. ;
I OITS 8317/Q240,139: Section 6.2.4 of the SSAR does not support the determination of a specified plate temperature as described in [14.2.9.1.11) test b) of the General Test Acceptance ,
Criteria and Methods. Temperatures within the PAR cartridge can vary greatly and are
, dependent on a number of factors such as location and mounting of the thermocouple, and the proximity of the thermocouple to the hydrogen source. This information is needed because during a test the temperature within a plate may be below or above the specifed acceptance temperature depending on the location of the thermocouple,
- k. -
- )
- , + , . . . - - -- -- . . . . - - - - - - -- - - _ _ - - - -- - - ------ - - .-. ~
i
. 28 .
In its June 11,1997, response to the NRC, Westinghouse stated that a response to this item is !
still being prepared. Therefore, pending receipt and evaluation of a response, OITS ;
S317/0260,139 remained open. i I in its October 23, igg 7, response to the NRC, Westinghouse stated that Subsection 14.2.g.1.11 ,
would be modiflod to refer to SSAR subsection 8.2.4.5.1 in order to provide additional guidance i
on the location of the thermocouple (s) in me testing device. This reference to the manufacturer's guidance for plate (or cartridge) testing assures consistency with plant testing and ,
manufacturer's testing. This modification has been included in the response for DITS
. 5318/Q260.134. Based on the Westinghouse response and the changes incorporated into Subsection 14.2.g.1.11. SSAR Revision 17,0178 6317/Q200,13g is closed.
OITS 832170260.ga: Section 14.2.5, " Utilization of Reactor Operating and Testing Experience in !
the Development of Test Program,' states that "special tests" used to establish a unique l - performance parameter of the AP600 design that will not change from plant to plant are to be
- performed on the first plant only. Westinghouse should revise this section to include the ;
l following:
i s. Selection (screening) criteria used by Westinghouse to identify such special tests l 1 i
- b. Provisions or programmatic controls that will be utilized by Westinghouse and/or the COL applicant to establish that system configuration or design engineering changes do not invalidate previous special test results.
in its May g,1997, response to the NRC, Westinghouse addressed these lasue as follows:
L a. The screening criteria that was used by Westinghouse as presented in Section 14.2.5 has been revised as follows:
Special tests to further establish a unique phenomenological performance parameter of the AP600 design features beyond the testing performed for Design Certification and that will not change from plant to plant, are performed for the first plant only. Because of the standardization of the AP600 design, these special tests (designated as first plant only tests) are not required on follow plants. These first <
plant only tests are identified in the individual test descriptions. (See subsections '
14.2.g and 14.2.10.)
4
- b. This item is addressed by the provisions 10 CFR Part 52.47 which stipulates that changes to Tier i documentation requires new rulemaking, and that changes to Tier 2 will j require a 50.5g type evaluation that will assess the impact to the SSAR including :
Cleapter 14. Changes identified that impact these special tests will be identifed, and tests may be modified or repeated as appropriate.
i Based on this response, the NRC staff found the following:
- a. The revised criteria is not responsive to the NRC staffs request for a ' screening" or -
selection criteria. Westinghouse should include clear, concise, and objective criteria that establishes the basis for designating first plant only tests. Altematively, given that the number of first plant only tests is limited, Westinghouse could identify the individual tests in Section 14.2.5 and the associated justification for the first plant-only designation. This
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approach would allow the staff to review and approve the justifications, as opposed to l generic selection erheria that may be difficuN to develop given the diversity of the systems invoiv.d .nd m. pn rs b.ing i.si.d. ;
i
- b. While the NRC staff apress a change to Tier 1 information would require rulemaking in ;
1 accordance wHh the requirements of 10 CFR Part 62.47, *spoolal tests" or first plant only ,
j lest abstracts are, int definition, not part of the AP600 ITAAC, i.e., not Tier 1 information. ;
! Therefore, Westinghouse should desertbo in Section 14.2.5 of the SSAR how the control !
j mechanisms for T6er 2 information would establish that system ocMlquretion or design i
- engineering changes do not invalidate previous special test results. The issue here is not !
i only control of changes, but vertfication that the design, construction, fabrication, and '
operation of future plant systems are monsistent with those of the prototype plard for I which first plard only testing is proposed. Westinghouse should describe the general .
i approach that is to be followed to ensure that future plant configurations are consistent !
i with the design, construction, operation, and testing performed for the first plant. On .
these bases, OITS 6321/Q260.g3 remained open. '
In Hs October 23,1997, response to the NRC, Westinghouse addressed these issues as follows:
- a. Sectior 14.2.5 has been modified to include a list of first plant only tests. The general l criteria for designation of certain tests as first plant only is provided in SSAR :
Section 14.2.5. The following is the detailed justification by Westinghouse for designating these tests as first plant only.
IRWST Heatup Test (14.2.9.1.3 item (h)) ;
During preoperational testing of the passive core cooling system, a natural circulation test of the passive residual heat removal (PRHR) heat exchanger is conducted (hem f). For the first plant only, thermocouples are placed in the IRWST to observe the thermal profile developed during the heatup of the IRWST water during PRHR heat exchanger operation.
This test will be useful in confirming the resuHs of the AP600 Design certification Program PRHR tests with regards to IRWST mixing, and is useful in quantifying the conservatism in the Chapter 15 transient analyses.
Due to the standardization of the AP600, the heatup and thermal stratification characteristics of the IRWST will not vary from plant to plant. The PRHR heat exchanger design, and the size and configuration of the IRWST are standardized, such that the ;
heatup characteristics will not significantly change from plant to plant.
. Therefore, since the phenomenon to be tested (i.e. heatup and mixing characteristics of the IRWST) will not very significantly from plant to plant due to standardization, a first
- plant only test of the IRWST heatup characteristics is justified.
t Core Makeup Tank Heated Recirculation Tests (14.2.9.1.3 Items (k) and (w)) .
During preoperational testing of the passive core cooling system, tests of the core makeup tanks are performed to verify the CMT inlet and inlet piping resistances, in ,
addition, cold draining tests of the CMTs are conducted that verify proper operation of the i
3
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i 30 !
t CMTs. For the Arst plant only, two additional CMT tests are condueled during hot i functional testing of the RCS. These tests are a natural circulation hostup of the CMTs :
followed by a test to verify the sollity of the CMTs to transition from a recirculation mode
, to a draindown mode while at elevated temperature and pressure.
l i
Operation of the CMTs under natural circulation is conducted on the Hrot plant only for the l 4 follow 6ng reasons: :
- Natural circulation of the CMTs we not very from plant to plant, provided that the
, other vertfloations discussed above are performed as speciRed.
- Natural circulation testing of the CMTs was er'.ensively tested in the various Design Certification Tests including the CMT separate affects test and the SPES2 and l l OSU integral tests, j
- Performance of this test results in significant thermal transients on Class 1 components including the CMTs and the detect vessel ir(ection nozzles. ;
l ADS Blowdown Test (14.2.9.1.3 learn (s))
l During preoperational testing of the passive core cooling system, the resistance of the automatic depressurization system Stage 1,2,3 flowpath(s)is verified. For the first plant i only, en automatic depressurization blow down test is performed. This test is performed .
during hot functional testing of the RCS, and results in a significant blowdown of the RCS
,. Into the IRWST. This tests verifies proper operation of the ADS velves, and !
demonstrates the proper operation of the ADS spargers to limit the hydrodynamic loads in containment to less than de61gn limits. This test is performed on the first plant only for the followirg reasons: ,
- The operation of the ADS, and the resultant hydrodynamic loads will not very from plant to plant.
- Full scale automatic depressurlastion testing was performed in the AP600 Design Certification Program. Testing was conducted to conservatively bound ADS flow l rates and resultant hydrodynamic loads that will be experienced by the plant during l ADS operation.
- Performance of this test results in significant thermal transients on Class 1 '
components including the primary components, it also results in hydrodynamic loads in containment including the IRWST.
Pressurtaer Surge Line Stratification Evaluation (14.2.9.1.7 Item (d))
As part of the AP600 conformance to NRC Bulletin 8811 a monitoring program will be
- implemented by the Licensee at the first AP600 to record temperature distributions and thermal displacements of the surge line piping during hot functional testing and during the first fuel cycle, as discussed in SSAR section 3.g.3.
I Reactor Vessel hitemals Vibration Testing (14.2.9.1.9)
J
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j 31 The preoperational vibration test program for the reactor intemals of the AP600 conducted on the first AP600 is consistent with the guidelines of Regulatory Guide 1.20 for a comprehensive vibration assessment program. This progrsm is discussed in SSAR Section 3.9.2.
Natural Circulation Teek (14.2.10.3.5,14.2.10.3.7)
Natural circulation tests using the steam generators and the passive residual host removal host exchangers are performed at low core power during the startup test phase of the initial test program for the first AP600 plant. This testing of the host removal i systems meets the intent of the requirement to perform natural circulation testing and the -
results of this testing is factored into the operator training. Justification for performing these natural circulation tests for the first plant only is provided in SSAR Section 1.9.4.
, Load Follow Demonstration (14.2.10.4.22)
A load follow demonstration test is not required by Regulatory Guide 1.68. However, the AP600 performs load follow with grey rods, as opposed to outront Westinghouse PWRs which manipulate RCS boron concentration to perform load follow operations. Therefore, Westinghouse has included a proof of principle load follow demonstration for the first APC00 plant, to demonstrate the ability of the AP600 plant to follow a design basis daily load follow cycle,
- b. In response to item (b), Section 14.2.5 has been modified to include a description of the general approach that is to be followed such that future plant configurations are consistent with the design, construction, operation, and testing performed for the first plant, as recommended in part (b) of the RAl.
Based on this response, the NRC staff finds the following:
- a. Westinghouse's response to item (a) is acceptable with respect to SSAR Subsections 14.2.9.1.3 ltem (h),14.2.g.1.7 ltem (d),14.2.10.3.6,14.2.10.3.7, and 14.2.10.4.22.
However, the justifications for the first time only testing should be included in Chapter 14, section 14.2.5, of the SSAR. For subsection 14.2.9.1.9, Westinghouse needs to clarify which portion (s) of the test are to be conducted on the first plant only.
With respect to subsections 14.2.91.3 llems (k) and (w), ad f 4.2.9.1.3 item (s), the staff does not share Westinghouse's confidence that other testing and verifications will, in all ;
instances, prevent minor engineering or construct!on variances from affecting test results obtained on the first plant, or from introducing uncertainties into such results. Absent some empirical data to support such conclusions, the NRC staff is unable to conclude that these tests need not be repeated on subsequent plants. .
- b. The NRC staff finds that the changes to Section 14.2.5 as described in item (b),
unacceptable. The changes to Section 14.2.5 describe the approach to address design changes that could affect the applicability of first plant only tests. The staff interprets this change to address only planned design changes, and not the potential variances that can occur in design, construction, fabrication, and operation from one plant to another. The T
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- 32 f
4 approach described in Section 14.2.5 should also address potential variances discovered !
during the engineering and construction process, i.e., between the as bulh plant and the certified design.
On these bases, OITS 6321/Q260.93 remains open.
t 0178 8322/Q260.84: in its June 25,1997, response to Westinghouse, the NRC staff stated that l
- OITS 6322/Q200.84 would remain open pending satisfactory resolution of OITS 1266/DSER ,
Open Hem 14.2.8.31, above. The NRC staff has reviewed the Westinghouse response to OlTS 1266/DSER Open item 14.2.8.31 and finds N acceptable. On this basis, OITS 6322 is closed.
0178 8323/0280.95: in its June 25 ,1997 6esponse , to Westinghouse, the NRC staff stated that l OITS 5323/Q260.95 would remain open pending satisfactory resolution of OITS 1266/DSER l Open item 14.2.8.31, above. The NRC staff has reviewed the Westinghouse response to 01T8 1255/DSER Open item 14.2.8.31 and finds it acceptable. On this basis, OITS 5323 is closed. 1 0178 8334/Q290.108: Subsection 14.2.9.2.17, ' Diesel Generator Testing,'should be modifed to !
address the following:
- a. The abstract should include a reference to Section 9.4.10 for appropriate design criteria related to the diesel ventilation systems (GTM&AC ltem f). ,
- b. The abstract should reinstate that the proper automatic restart of the diesel is to be tested immediately following the load test (GTM&AC ltem k) per RG 1.108, C.2.a.(5).
- c. - The abstract should reinstate appropriate start tests per RG 1,108. C.2.a.(9).
- d. The abstract should include the statement: " Demonstrate by performing a loaded run of the diesel generator with he day tank filled to its low level alarm point, that the day tank provides sufficient fuel for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of diesel generator operation with the diesel generator supplying its expected power requirements."
In its May 9,1997, response to the NRC, Westinghouse addressed these issues as follows:
- a. The reference to section 9,4.10 has been added as requested, ,
- b. Regulatory Guide 1.108 provides requirements for testing safety-related emergency diesel generators. The AP600 diesel generators are not safety related, and are not -
1 required to be tested as specifed.
- c. Same as b., above.
I d. This test is not required to be performed and is not necessary to verify the proper
- operation of the diesel generators.
While the diesel generators are classified as defense-in-depth systems in the AP600, the NRC staff agrees that RG 1.108 is not applicable to them. Therefore, Westinghouse exception to RG 1,108 in Appendix 1A to the SSAR is acceptable. OITS 5336/Q260.108 is closed. '
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- 33 0178 8337/Q260,109: Subsection 14.2.9.2.19, " Plant Lighting System Testing
- is limited to !
i testing of the main conti01 room and remote shutdown station emergency lighting sy6tems. As i described in SSAR Section 9.6.3, emergency lighting is also provided for emergency ingress and ogress, equipment areas assoolated with power recovery actions (e.g., diesel building), as well !
as for manuel equipment actions in the event of a fire with loss of normal lighting. These self- -!
contained lighting units support safe shutdown operations and should be included in the pre- .
operational testing to vertfy performance and proper positioning. l In its May 9,1997, response to the NRC, Westinghouse stated that Subsection 14.2.9.2.19 would be revised to incorporate this commerd.
The NRC staff found that Revision is to subsection 14.2.9.2.10 did not address testing of self- l contained emergency lighting units thet provide lighting for emergency operations outside the control room such as power recovery and manual / local operation of equipmerd daing a fire, as described in SSAR Subsection 9.5.3.2.2. Therefore, OITS 6337/Q260,109 remained open. ;
- i in Hs October 23,1997, response to the NRC, Westinghouse stated that subsection 14.2.9.2.19, e
Plant Lighting System Testing,' would be revised to clearty specify that all self contained emergency lighting units be tested. The NRC staff finds the changes incorporated into l Subsection 14.2.9.2.19, SSAR Revision 17, acceptable. OITS 6337/Q260.109 is closed.
01T8 8339/Q260.111: Subsection 14.2.9.3.3,
- Solid Radweste System Testing,* should be modified to verify that no free liquids are present in the packaged wastes per RG 1.68, l Appendix A,1.1.(3).
In its May 9,1997, response to the NRC, Westinghouse stated that packaging of solid redweste :
la the responsibility of the COL applicant and is identified in SSAR Table 1.61 under item 11.3. i The NRC staff found this response unacceptable. While the NRC staff acknowledges that ,
packaging of solid redweste is the responsibility of the COL applicant, such responsibility does !
not exempt the COL applicant or its designee from having to conduct testing as described in SSAR Subsection 14.2.9.3.3 in accordance with RG 1.68. Therefore, Westinghouse should i modify subsection 14.2.9.3.3 to include verification that no free liquids are present in the _
packaged wastes per RG 1.68, Appendix A,1.1.(3), or provide an acceptab's : 'smative. On this basis, OITS $339/Q260,111 remained open.
In its October 23,1997, response to the NRC, Westinghouse stated that Subsection 14.2.9.3.3,
' Solid Radweste System Testing," would be revised to specify the verification that free liquids not be present in solid packaged waste. Sased on the changes incorporated into Subsection 14.2.9.3.3, SSAR Revision 17, OITS 5339/260.111 is closed.
OITS 8340/Q260.112: Subsectic 14.2.9.3.4, " Radioactive Weste Drain System Testing," should be revised to address the following:
I
- a. Section 9.3.5 should be added as a refsrence in the Purpose and General Test -
Acceptance Criteria and Methods subsections, c ;
- b. The following statement should be added:
- Flow water in each drain path to verify that the drains discharge to their designated destination and that system segregation is maintained?
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- 34 In its May 9,1397, response to the NRC, Westinghouse stated that Subsection 14.2.9.3.4, t
- Radoactive Waste Drain System Testing,* would be revised as appropriate. Westinghouse also l noted that system segregation is not a design feature of the AP600 Radioactive Weste Drain i System. ,
The NRC staff found the proposed revisions incorporated in Revision 13 to SSAR
- subsection 14.2.9.3.4 acceptable. However, Westinghouse needs to clarWy the statement that {
the redosotive waste drain system segregation is not en AP900 design feature as M is !
inconsistent with SSAR Subsection 9.3.6,
- Equipment and Floor Drainage Systems.' 3 Spoolfloally, SSAR Subsection 0.3.6 states, in part,
- Equipment and floor drainage is j eegregated [ emphasis added) accordnD to the type of waste. Liquid wastes are classified and segregated for collection as follows:
- Redonative liquid weste
- Nonredioactive liquid weste l
- Chemical and detergerd liquid waste }
- Oily liquid waste * ;
.t Therefore, this portion of OITS $340/Q260,112 remains open pendng revision of Subsection
. 14.2.9.3.4, General Test Method and Acceptance CrNoria, Nom d) to incorporate verification !
'that system segregation is maintained.' i 1 1
- OITS 83421Q200,114: Subsection 14.2.9.4.6, " Circulating Water System Testing," should be <
modified to include testing of cooling towers and associated auxiliaries as docussed in l Attachment 3 to the Westinghouse letter of July 16,1996 per RG 1.88, App. A,1.f.(2), i In hs May 9,1997, response to the NRC, Westinghouse stated that the circulating water system ;
cooling towers are not part of the scope of the AP600 Design Certification submittal. They have ;
been addressed in the response to OITS 1267/DSER Open item 14.2.91. Sased on the above, l OITS 5342/Q260,114 remains open pending satisfactory resolution of OITS 1257/DSER Open !
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