ML20211J429
| ML20211J429 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 10/03/1997 |
| From: | Joshua Wilson NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9710080121 | |
| Download: ML20211J429 (9) | |
Text
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October 3, 1997 Mr. Nicholas J. Liparuto, Manager Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230
SUBJECT:
AP600 INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE CRIT;...iA
Dear Mr. Liparulo:
The enclosure to this letter contains requests for additional information (RAls) or corrections concoming Revision 3 of the AP600 Tier 1 information including the inspections, tests, analyses, and acceptance criteria (ITAAC).- These requests were provided by the Civil Engineering and Geosciences Branen. I have also enclosed markups of pages 2.5.13 and 2.5.218 from the Tier 1 information. The revisions to these pages were provided by the Instrumentation and Control Systems Branch.
Your review of Revision 3 should include a verification that all Tier 1 information was extracted from Tier 2 and that Tier 1 is consistent with the information in Tier 2 (SSAR). If you have any questions regarding this matter, you may phone me at (301) 415 3145.
Sinceroty, original signed by:
Jerry N. Wilson, Senior Policy Analyst Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 52 003
Enclosure:
At stated cc w/ encl: See next page DISTRIBUTION:
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Mr. Nicholas J. Liparuto Docket Nn. 52 003 Westinghouse Electric Corporation AP600 cc:
Mr. B. A. McIntyre -
Mr. Russ Bell Advanced Plant Safety & Licensing Senior Project Manager, Programs Westinghouse Electric Corporation Nuclear Energy institute Energy Systems Business Unit 1776 l Street, NW P.O. Box 355.
Suite 300 Pittsburgh,l'A 15230 Washington, DC 20006 3706 Mr. 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. Storting Franks GE Nuclear Energy U.S. Department of Energy 175 Curtner Avenue, MC-754 NE-50 San Jose, CA 95125 19901 Germantown Road Idaho Falls,10 83415 Germantown, MD 20874 Mr. Robert H. Buchholz Mr. Frank A. Ross GE Nuclear Energy U.S Dtpartment of Energy, NE 42 _
175 Curtner Avenue, MC 781 Office of LWR Safety and Technology San Jose, CA 95125 19901 Germantown Road Germantown, MD 20874 darton Z. Cowan, Esq.
Eckert Seamans Cherin & Mellott Mr. Charles Thompson, Nuclear Engineer 600 Giant Street 42nd Floor AP600 Certification Pittsburgh, PA 15219 NE-50 19901 Germantown Road Mr. Ed Rodwell, Manager Germantown, MD 20874 PWR Design Certification Electric Power Research Institute 3412 Hillview Avenue Pato Alto, CA C4303
+
RAls for AP600 lTAAC 640.136 2.1.3 Reactor System l
In Table 2.1.3 3, add an ITAAC to verify the structural integrity of reactor intomals of both prototype and non pro'otype AP600 plants. The following information t
L should be added:
Desion Commitmsat
'The reactor pressure vessel (RPV) intomals will withstand the effects of flow-induced vibration."
- Inspection. Tests. Analvses
"(a) A vibration type test will be conducted on the AP600 prototype plant RPV Intemals."
"(b) A flow test and post-test inspection will be conducted on the as built RPV intemals of all non-protatype AP600 plants."
Acceptance Criteria
"(a) A vibration type test report exists and concludes that the AP600 prototype RPV intemals have no damage or loose parts as a result of the vibration type test."
"(b) A report exists which documents that the testing and inspection results of the non-prototype AP600 plants demonstrate tM the as-built RPV intemals have no damage or loose parts."
640.137 2.2.1 - Containment System in Figure 2.2.1-1, containment penetration piping is shown for the Liquid Rad-waste System, the Domineralized Water Transfer and Storage System, and the Compressed and instrument Air bystem. However, the line name and number for
~ these systems are not listed in Table 2.2.1-2. Since Table 2.2.1-2 is referenced in the ITAAC (Table 2.2.13), either odd these systems to Table 2.2.1-2 or provide a basis for not including these systems in the ITAAC.
3.3 Buildinas 640.138 The information in this ITAAC related to protection from the dynamic effects of-postulated pipe breaks might not be appropriate as a Tier 1 commitment, and may be accomplished as a SSAR (Tier 2) commitment. The staff suggests the following changes:
a.
Delete Table 3.3-4 and add all of the detailed information under" Room Description," " Essential Target Description," and " Hazard Source" in that table to SSAR Table 3.6-3.
Enclosure f
2-l b.
Revise item 7 under " Design Description" to read as follows:
" Structures, systems, and components required for safe shutdown are protected from the dynamk, effects of postulated pipe breaks using pipe break mitigation features."
c.
Revise item 7 in Table 3.3 5 to read as follows:
Deslan Commitment
" Structures, systems, and components requirer! for safe shutdown are protected from the dynamic effects of postulated pipe bwks using pipe break mitigation features."
inspection. Tests. Anatvses "An inspection will be performed of the as built high and moderate energy pipe break mitigation features."
Acceptance Criteria "An as-built Pipe Rupture Hazard Analysis Report exists that includes documenta-tion of the results of the high and moderate energy pipe break mitigetion features, and concludes that structures, systems, and components required for safe a
shutdown can withstand the effects of postulated pipe rupture without loss of the required safsty function."
640.139 Design Description To be consistent with the SSAR, the first sentence of the first paragraph a.
should read "The nuclear island (NI) structures include the containment building (the steel containment vessel and the containment intemal struc-tyres), the shield building, and auxilitry building."
b.
Descriptions (such as construction materials and primery functions) should be provided for the containment intemal structures, the shield building, and the auxiliary building, c.
The last phrase of Sentence i.a should read "without loss of structural inteority and safety function."
d.
A description should be provided for the PCCWS tank and the shield building roof structures.
in NRC letter, dated March 4,1997 (RAI 640.5), the staff requested Westing-e.
house to provide key dimensions in the AP600 CDM. However, only the
3 thickness of walls and floor slabs was given in CDM Table 3.3-1. Westing-house should also provide SSAR Table 3.7.1 16, elevations as well as the distance between column lines, and between column lines and the edge of the foundation met in the CDM (Tier 1).
f.
Footnote #1 in Table 3.3-1 states that the applicable column lines, elevation levels, and NI basemat reinforcement are identifed and included on Figures 3.3-1 through 3.3 20. Kowever, only Figures 3.31 through 3.315 are provided in Tier 1. Clarification is needed.
g.
Table 3.3-1 on Page 3.3-10, For modular walls (Walls 1 and 2, and M 2 Wall), nominal reinforcement is provided in the vertical direction but not in the horizontal direction. Clarification is needed for (1) what kind of reinforce-ment is used for modular walls, and (2) why no reinforcement is provided in the horizontal direction?
h.
The code boundary should be defined in the Design Description and shown in the figures, l.
As documented in the Tier 1 information for the ABWR and System 80+
designs, the structural design basis loads should be clearly defined.
640.140 ITAAC a.
AP600 ITAAC should commit to the basic configuration of the nuclear island structures as shown in Figure 3.31 and to inspect the as-built structures against this basic configuration.
b, Westinghouse should provide key dimensions (such as overall dimensions, the distance between column lines, the distance between the center of the reactor vessel and column lines, elevations, total embedmont depth, etc.) as the acceptance criteria for verifying the as-built conditions.
c, What is the basis for acceptance cnteria #g (leak rate of 100 pal /hr or smaller for the PCCWS tank).-
d.
For verifying that the NI structures will withstand the structural design basis-loads, Westinghouse should require the existence of a structural analysis report in the ITAAC, which concludes that the as-built Ni structures will withstand the structural design basis loads. Also, a description of the contents of a structural analysis report must be provided in the SSAR (Tier 2).
640.141 in Table 3.3-5, the acceptance criterion refers to an as-built Pipe Rupture Hazards Analysis Report. However, a Pipe Rupturs Hazards Analysis Report is not discussed in the SSAR. A pipe rupture hazards analysis is discussed in SSAR Section 3.6.2.5. It is recommended that the following statement be added to SSAR Section 3.6.4.1:
4 The as-built pipe rupture hazards analysis will be documented in an as built Pipe Rupture Hazards Analysis Report.
I 640.142 Similaity, for leak-before-break, the ITAAC for those systems in which LBB is used refers to an LBB evaluation report. However, an LBB evaluation report is not discussed in the SSAR. It is recommended that the following statement be added -
to the SSAR Section 3.6.4.2: 'The leak-before break evaluation will be docu-mented in an LBB evaluation report."
640.143
- Because the final seismic amplified response spectra has not been used in the preliminary piping design, Westinghouse proposed to add a COL action item to have the COL applicant verify the final as-built piping design by using the final seismic input loadings.- The staff has determined that a COL action item is not appropriate for this verification, and that the verification must be accomplished through ITAAC. Westinghouse is requested to develop ITAAC for those applica-ble piping systems where the final seismic input loadings will need to be verified by the COL applicant.
640.144 Because Westinghouse has not completed the design of the reactor coolant loop piping by using time history seismic analyses that accounts for 115 percent peak broadening effects, Westinghouse proposed to add a COL action item to have the COL applicant verify the time-history analyses of the RCL piping by using a time-history analysis that varies the time-history loading by *15 percent. The staff has determined that a COL action item is not appropriate for this verification, and that the verification must be accomplished through ITAAC. Westinghouse is re-quested to develop ITAAC for the verification of the RCL piping by a time-history analysis that varies the time-history loading by *15 percent.
640.145 Westinghouse is not planning to complete the fatigue analysis for the ASME Code Class 1 piping prior to design certification. Westinghouse proposed to add a COL action item to have the COL applicant venfy through final analysis that the Class i fatigue limits have been satisfied. The staff has determined that a COL action item is not appropriate for this verification, and that the verification must be accomplished through ITAAC. Westhghouse is requested to develop ITAAC for the COL applicant to verify that the ASME Code fatigue requirements have been met for the ASME Code Class 1 piping. ITAAC are needed for each system that contain ASME Code Class 1 piping.
640.146-Westinghouse is not planning to complete the stress analysis for small bore piping
- (less than 3 inches nominal pipe size). Westinghouse proposed to add a COL action item to have the COL applicant complete the small-bore piping stress analysis. The staff has determined that a COL action item is not appropriate to complete this effort, and that the effort must be completed through ITAAC.
Westinghouse is requested to develop ITAAC for the COL applicant to complete the design of the small-bore piping stress analysis.
640.147 Westinghouse is planning to complete the pipe break hazards analysis as de-scribed in SSAR Section 3.6.2.5 except for the design of protective hardware and
5-the responsibility of the COL applicant to complete the design of protective hardware and reconciiation of the as-built condition. An additional ITAAC is not needed because the ITAAC in Table 3.3-5 (#7) already addresses this effort.
640.148 The qualifying comment on soft soll sites in ITAAC #10 needs to be clarified and incorporated into the Design Commitment, if appropriate.
640.14g Pages 3.3-31 and 3.342 are missing, Verify that no pages are missing from Revision 3 of the AP600 Tier 1 information.
640.150 2.5.g - Seismic Monitoring System Add the following Tier 1 information to ITAAC 2.5.g. The seismic instrumentation is designed to provide for the collection of seismic data in digital format, analysis of seismic data, notification of the operatorif the ground motion exceeds a threshold value, and notification of the operator if the ground motion is of such a level that the plant must be shut down. The seismic instrumentation shall consist of at least four triaxial accelerometery, and appropriate analysis, recording and playback, and timing systems. The system is capable of analyzing, recording, and playing back data at a sampling rate of 200 samples por second. The system initiation value is adjustable from 0.002g to 0.02g. Pre-event recording time is adjustable from 1.2 to 15.0 seconds. The system will have a dynamic range of 0.001g to 1.0g and a frequency range of 0.2 to 50 Hertz.
1 30/ 0 Certified Design Material 4gN pcW f;W I4 JIVERSE ACTUATION SYSTEM
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Reactor Coolant Pump Trip and Turbine Trip on Low Pressurizer Water Level 3.
Passive Residual Heat Rem.wal (PRHR) Actuation on Low Wide range Steam Generator Water Level or on High Hot Leg Temperature 4 Core Makeup Tank (CMT) Actuation on Low Pressurizer Water Level 5.
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PRHR Actuation
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First stage Automatic Depressurization System (ADS) Valve Actuation 5.
Second-stage ADS Valve Actuation 6.
Third-stage ADS Valve Actuation
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