ML20153G963
| ML20153G963 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 04/29/1988 |
| From: | Artigas R GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8805120071 | |
| Download: ML20153G963 (100) | |
Text
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GE Nuclear Energy
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- c. : . , o m . v :oa April 29,1988 MFN No. 42 88 Docket No. STN 50-605 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Lester S. Rubenstein, Director Standardization and Non Power Reactor Project Directorate
Subject:
Submittal of Responses to Additional Information as Requested in an NRC Letter from Dino C. Scaletti, Dated February 22,1988
Dear Mr. Rubenstein:
Enclosed are thirty four (34) copies of the Responses to Additional Information on the StanJard Safety Analysis Report (SSAR) for the Advanced Boiling Water Reactor (ABWR). These responses principally pertain to Chapters 4,5,6 and 15.
It is intended that GE will amend the SSAR with these responses in June 1988.
Sincerely, 3
f
/
Ricardo Artigas, Manager Licensing and Consulting Services (dj i 8805120071 880429 PDR ADOCK 05000605 N DCD
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Document Control Desk U.S. Nuclear Regulatory Commission April 29,1988 Page 2 i l
cc: D. R. Wilkins (GE)
F. A. Ross (DOE)
J. F. Quirk (GE)
D. C. Scaletti (NRC)
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i CHAPTER 20 QUESTION AND RESPONSE GUIDE O
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M 23A6100AT Standard Plant REV.A Q CHAPTER 20 TABLE OF CONTENTS Section T1111 East 20 QUESTION AND RESPONSE GUIDE 20.1 OUEST10N INDEX 20.1 1 20.2 OUESTIONS 20.2-1 20.2.1 Chapter 1 Ouestions 20.2 2 20.2.2 Chapter 2 Questions 20.2 3 20.2.3 Chapter 3 Questions 20.2-4 20.2.4 Chapter 4 Questions 20.2-5 20.2.5 Chapter 5 Questions 20.2 5 20.2.6 Chapter 6 Questions 20.2 10 20.2.7 Chapter 7 0uestions 20.2 13 20.2.8 Chapter 8 Questions 20.2 14 20.2.9 Chapter 9 Questions 20.2-15 20.2.10 Chapter 10 Questions 20.2 16 20.2.11 Chapter 11 Ouestions 20.2 17 20.2.12 Chapter 12 Ouestions 20.2 18 20.2.13 Chapter 13 Questions 20.2-19 20.2.14 Chapter 14 Questions 20.2-20 20.2.15 Chapter 15 Questions 20.2 21 20.2.16 Chapter 16 0uestions 20.2 23 20.2.17 Chapter 17 Questions 20.2 24 20.2.18 Chapter 18 Questions 20.2 25 20.2.19 Chapter 19 Ouestions 20.2-26
- 20. u Amendment 2
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CHAPTER 20 TABLE OF CONTENTS (Continued)
O Section I]Ils Eagt 203 OUESTIONS/ RESPONSES 20 3-1 20 3.1 Response to First RAI Reference 1 20 3-1 20,4 REFERENCES 20.4-1 O
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20-lii O
SECTION 20.1 TABLES Titie O '"b'" ease 20.1-1 Identification Numbers for NRC Resiew Questions 20.1-4 l
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20.1 0 0
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ABWR 23461004r Standard Plant any.A 20.1 QUESTION INDEX This subsection provides an index to each NRC request for additionalinformation (RAI) during its review of the ABWR standard plant. Each NRC question is designated with the NRC branch que.stions ID number (see Table 20.1 1) followed by the number of the question of the review area for that branch.
For example, question number 210.2 designates the second question of the mechanical engineering branch (EMEB). The index below provides an up-to-date listing in numerical order of each question.
NRC* Res!cw Question SSAR Response RAl" Branch Area Number Subsection Subsection letter EMEB Mcchaical 210.1 5.2.1.2 20 3.1 1 Engmeenng 210.2 5.2.1.2 20 3.1 1 EMTB Insenice 250.1 5.2.4.1 20 3.1 1 Inspection 250.2 5.2.4.2 20 3.1 1 250 3 6.6.8 20 3.1 1 Component 251.1 53.1.1 20 3.1 1 Integrity 251.2 53.1.2 20 3.1 1 2513 53.1.4.4 20 3.1 1 53.1.4.5 20 3.1 1 53.1.4.7 20 3.1 1 O 53.1.5.2 53.1.53 20 3.1 20 3.1 1
1 53.2.1.5 20 3.1 1 251.4 53.1.6.1 20 3.1 1 251.5 53.1.63 20 3.1 1 251.6 53.2.1 20 3.1 1 251.7 53.2.1.1 20 3.1 1 t 53.2.1.2 20 3.1 1 53.2.13 20 3.1 1 53.2.1.5 20 3.1 1 251.8 533 20 3.1 1 l 251.9 533.1.1.1 20 3.1 1 251.10 533.2 20 3.1 1 251.11 533.6 20 3.1 1
- See Table 20.11for abbrenations.
" Letter reference ofSection 20.4 O
Amendment 2 20.11
ABWR 23461 mar Standard Plant as NRC* Reslew Question SSAR Response RAl" Branch Area Number Subsection Subsection letter Materials 252.1 4.5.1.1(1) 20 3.1 1 Application 252.2 4.5.1.1(2) 20 3.1 1 2523 4.5.2.2 20 3.1 1 252.4 4.5.23 20 3.1 1 252.5 4.5.2.4 20 3.1 1 252.6 4.5.2.5 20 3.1 1 252.7 5.23.2.2 20 3.1 1 252.8 5.23.23 20 3.1 1 252.9 5.233.1 20 3.1 1 252.8 5.23.23 20 3.1 1 252.10 5.23.4.1.1 20 3.1 1 252.11 5.23.4.23 20 3.1 1 ECEB Chemicai 281.1 5.1 20 3.1 1 Technology 281.2 5.23.2.2 20 3.1 1 2813 5.23.2.2 20 3.1 1 281.4 5.23.2.2 20 3.1 1 2'15 5.23.2.2 20 3.1 1 281.6 5.23.2.2.2 20 3.1 1 281.7 5.23.2.23(4) 20 3.1 1 281.8 5.23.2.23(13) 20 3.1 1 .
281.9 6.4.9.2 20 3.1 1 281.10 Chap.5 20 3.1 1 i PRPB Radiological 470.1 15.5.2 20 3.1 1 Report 470.2 15.6.2 20 3.1 1 470 3 15.6.4.5.1.1 20 3.1 1 470.4 15.6.5.5 20 3.1 1 470.5 15.6.5 20 3.1 1 470.6 15.7.5 20 3.1 1 470.7 15.7 20 3.1 1 470.8 15.7 20 3.1 1 470.9 15.7 l
20 3.1 1 470.10 15.7 20 3.1 1 1
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O Amcnoment 2 2012
MM 23A6100AT Standard Plant REV A TABLE 20,1-1 IDENTIFICATION NUMBERS FOR NRC REVIEW QUESTIONS Question ID Number Rewiew Area Branch Anoticable SRP Sections _
100 Miscellaneous Responsible None Project Directorate 210 Mechanical EMEB 3.2.1, 3.2.2, 3.6.2, Engineering BTP.MEB 3-1, 3.9.1, 3.9.2, 3.93,3.9.4,3.9.5, 3.9.6, 5.2.1.1, 5.2.1.2 220 Structural ESGB 33.1,33.2,3.4.2,3.53, Engineering 3.7.1,3.7.2,3.73,3.7.4, 3.8.1,3.8.2,3.83,3.8.4, 3.8.5 230 Seismology ESGB 2.5.2 231 Geology ESGB 2.5.1,2.53 240 Hydrologic ESGB 2.4.1,2.4.2,2.43,2.4.4, Eogineering 2.4.5, 2.4.6, 2.4.7, 2.4.8, 2.4.9, 2.4.10, 2.4.11, 2.4.12, BTP HGEB-1,2.4.13,12.4.14 241 Geotechnical ESGB 2.5.4, 2.5.5 Engineering 250 Insetsice EMTB 5.2.4, 5.4.2.2, 6.6, Inspection 10.23 1
251 Component EMTB 3.5.13,53.1,53.2, Iotegrity BTP MTEB 52,5.4.1.1, 6.2.7 252 Materials EMTB 4.5.1,4.5.2,5.23, I Application BTP MTEB 52,533,5.4.2.1, 6.1.1, 10 3.6 260 Quality LOAB 17.1, 17.2 Assurance 270 Ensironmental SPLB 3.11 Qualification i
l Amendrnent 2 20,1 3 j
,r MM Standard Pl;nt 23A6100AT REv. A TABLE 20.1 1 O
IDENTIFICATION NUMHERS FOR NRC REVIEW QUESTIONS Question ID Nuq1htr Review Area Haulcli Aeolicable SRP Sections 100 Miscellaneous Responsible None Project Directorate 210 hiechanical EhfEB 3.2.1, 3.2.2, 3.6.2, Engineering BTP MEB 3- 1, 3.9.1, 3.9.2, 3.93,3.9.4,3.9.5, 3.9.6, 5.2.1.1, 5.2.1.2 220 Structural ESGB 33.1,33.2,3.4.2,3.53, Engineering 3.7.1,3.7.2,3.73,3.7.4, 3.8.1,3.8.2,3.83,3.8.4, 3.8.5 230 Seismology ESGB 2.5.2 231 Geology ESGB 2.5.1, 2.5.3 240 Hydrologic Engineering ESGB 2.4.1,2.4.2,2.43,2.4.4, 9 2.4.5, 2.4.6, 2.4.7, 2.4.8, 2.4.9, 2.4.10, 2.4.11, 2.4.12, BTP HGEB-1,2.4.13,12.4.14 241 Geotechnical ESGB 2.5.4, 2.5.5 Engineering 250 Insersice EhiTB 5.2.4, 5.4.2.2, 6.6, inspection 10.23 251 Component Eh(TB 3.5.13,53.1,53.2, Ir.cegrity BTP hlTEB 52,5.4.1.1, 6.2.7 252 hiaterials EhfTB 4.5.1,4.5.2,5.23, Application BTP htTEB 5 2,533,5.4.2.1, 6.1.1, 10 3.6 260 QuaEty LOAB 17.1, 17.2 Assurance 270 Endicamental SPLB 3.11 Qualificalion 9
Arnendinent 2 20.14
MM 23A6100AT Standard Plant REV.A TABLE 20.1-1 IDENTIFICATION NUMBERS FOR NRC REVIEW QUESTIONS (Continued)
Question ID Number Review Area Branch Anolicable SRP Sections 271 Seismic and Dynamic EhfE3 3.10 [
Load Qualification r 280 Fire Protection ECEB 95.1, BTP Chf EB 9.5.1 ,
281 ChemicalTechnology ECEB BTP hfTEB 5 3,5.4.8, !
BTP h1TEB 6-1,6.1.2, ,
93.2,93.4,9.5.1, BTP ChiEB 9.5-1, 10.4.6, !
10.4.8 i
290 Emhonmental ESGB Embonmental Report [
Engineering ,
310 RegionalImpact ESGB Emironmental Report Aralysis i O >>> site ^ >>rsis esoa 211.21.2.2.13 2.2.1 2.2.2,2.23, 3.5.1.5, 3.5.1.6 320 Antitrust and FTSB None Economic Analysis ,
410 Auxiliary Systems SPLB 3.4.1, 3.5.1.1, 3.5.1.2, 3.5.1.4, 3.5.2, '
3.6.1, BTP ASB 31, i 5.2.5, 5.4.11,6.7, [
9.1.1,9.1.2,9.13,9.1.4, :
BTP ASB 91,9.1.5,9.2.1, ;
9.2.2, 9.2.4, !
9.2.5, BTP ASB 9-2, !
9.2.6,93.1,933, ;
9.4.1,9.4.2,9.43,
[
9.4.4, 9.4.5, '
10.4.5, 10.4.7, !
BTP ASB 10-2, 10.4.9 i BTP ASB 10-1
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SRXB 4.6,93.5 ECEB 9.23 l O
Amendment 2 2015 f t
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23AG100AT Standard Plant REV A TABLE 20,1 1 O
IDENTIFICATION NUMilERS FOR NRC REVIEW QUESTIONS (Continued)
Question ID Number Review Areamngh Aeolicable SRP Sectiona 420 Instrumentation SICB 7.1,7.2,73,7.4,7.5, and Control Sptems 7.6, 7.7 430 Power Systems SELB 8.1,8.2,83.1,83.2, 9.53 SICB 9.5.2 SPLB 9.5.4, 9.5.5, 9.5.6, 9.5.7,9.5.8,10.2, 103,10.4.1,10.4.4 440 Reactor Systems SRXB 5.2.2, FTP RSB 5-2, 5.4.6, 5.4.7, BTP RSB 5-1, 5.4.12, 6 3, BTP RSB 61, 15.1. N 5 1.4,15.1.5, 15.2.1-15.2.5, 15.2.6, 15.2.7, 15.2.8, 15 3.1 15 3.2 1533 153.4,15.4.4-154.5 15.4.6, 15.5.1 15.5.2, 15.6.1, h 15.6.5, 15.8 450 Accident Evaluation SPLB 6.4,6.53 PRPB App. A to 15.4.8, Appendix A to 15.4.9, 15.6.2, 15.6 3, 15.6.4, Append'aA to 15.6.5 Appendix B to 15.6.5 Appendix C to 15.6.5 Appendix D to 15.6 5 15.7.4, 15.7.5 ECEB 6.5.2, 6.5.4 SRXB Appendix A to 15.1.5 451 Metcorology PRPB 23.1,23.2,233, 23.4,23.5 O Amendment 2 2014
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\ 23A6100AT Standard Plant REV.A TABLE 20.1-1 0 IDENTIFICATION NUMBERS FOR NRC REVIEW QUESTIONS (Continued)
Question IR ERGdEI Review Area Dranch Anolicable SRP Sections 460 Effluent SPLB 6.5.1,10.4.2,10A 3, Trcatment 11.1, 11.2, 113, 11.4, BTP ETSB 11-3, BTP ETSB 115, 11.5, 15.7 3 470 Radiological PRPB Environmental Report i Impact 471 Radiation Proteetion PRPB 12.1, 12.2, 12 3-12.4,12.5 1 480 Containmeni SPLB 6.2.1, 6.2.1.1.A, Systems 6.2.1.1.B, 6.2.1.1.C, ti.2.1.2, 6.2.13, 6.2.1.4, 6.2.1.5, BTP CSB 6-1,6.2.2,6.23, BTP CSB 6-3,6.2.4 BTP CSB 6-4,6.2.5, BTP CSB 6-2,6.2.6 j
*:90 Fuels SRXB 4.2 I
491 Physics SRXB 43, BTP CPB 43-1, 15 A.1,15.4.2,15.43, 15.4.7, 15.4.8, 15.4.9 492 Thermal- SRXB 4.4 Hydraulics 610 Operator LHFB 13.2.1 Liceusing 620 Human Factors LHFB 18, 18.1, 18.2 Engineering 630 Licensee LPEB 13.1.1, 13.1.2 13.13, j Oualifications 13.4, 13.5.1 LHFB 13.2.2, 640 Procedures and LHFB 13.5.2, 14.2 Systems Review O Amendment 2 20.1 7 i
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MkN 23A6100AT Standard Plant REV A TABLE 20.1 1 O IDENTIFICATION NUMBERS FOR NRC REVIEW QUESTIONS (Continued) Question ID Number Review Area Branch Aeolicable SRP Sections 720 Reliability and PRPB None Risk Assessment 730 Generic Issues None 810 Emergency Planning PEPB 13.3 910 Safeguards RSGB 13.6 I Abbresiations ECEB Cnemical Engineering Branch 3 EhfEB hiechanical Engineering Branch l EhiTB hiaterials Engineering Branch ESGB Structural and Geosciences Branch l l l LilFB liuman Factors Assessment Branch ! LPEB Performance Evaluation Branch l LOAB Quality Assurance Branch l PEPB Emergency Preparedness Branch PRPB Radiation Protection Branch PTSB Policy Development and Technical Support Branch RSGB Safeguards Branch SELB Electrical Sptems Branch SICB Instrumentation and Control Sptems Branch SPLB Plant Systems Branch ERXB Reactor Systems Branch O Amendment 2 20.14
\ 23A6100AT Standard Plant REV.A SECTION 20.2 CONTENTS Section Title P_ age 20.2.1 Chanter 1 Ouestions 20.2-2 i
20.2.2 Chaoter 2 Ouestions 20.2-3 20.23 Chapter 3 Ouestions 20.2-4 20.2.4 Chanter 4 Ouestions 20.2-5 20.2.5 Chaoter 5 Ouestions 20.2-5 20.2.6 Chaoter 6 Ouestions 20.2 10 20.2.7 Chaoter 7 Ouestions 20.2 13 i 20.2.8 Chaoter 8 Ouestions 20.2 14 20.2.9 Chaoter 9 Ouestions 20.2 15 20.2.10 Chaoter 10 Ouestions 20.2-16 i 20.2.11 Chaoter 11 Ouestions 20.2 17 20.2.12 Chaoter 12 Ouestions 20.2 18 20.2.13 Chaof er 13 Ouestions 20.2 19 20.2.14 Chaoter 14 Ouestions 20.2 20 , i 20.2.15 Chaoter 15 Ouestions 20.2 21 20.2.16 Chaoter 16 Ouestions 20.2-23 20.2.17 Chanter 17 Ouestions 20.2 24 20.7 18 Chanter 18 Ouestions 20.2-25 20.2.19 Chapter 19 Ouestlons 20.2 26 4
- O 20.2-il Ame ndment 2 i
MM 23A6100AT Standard Plant REV.A l 20.2 QUESTIONS l O This subsection provides an up to date chapter wise listing of the NRC questions. Subsections are l I numbered (e.g.,20.2.x) in accordance with the questions received for specific chapters. l 1 i l l I I i 4 1 l l 1 1 O i l 3 l 4 1 1 J t t 3 . O , j I a Amendment 2 20.2 1
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ABWR 22A6toorr Standard Plant REV.A 20.2.1 Chapter 1 Questions None to date. O 1 l O 1 Amendment 2 2022 l
NN 23AMMA f Standard Plant REV,A 20.2.2 Chapter 2 Questions None to date. O . l l l l O Amendment 2 g 3,3
ABWR 23A6100AT Standard Plant ma 20.2.3 Chapter 3 Questions Noac to date. g O O Amendment 2 20.2-4
ABM 23AM00AT ! Standard Plant REV A < 20.2.4 Chapter 4 Questions ! O 152.1 Subsection 4.5.1.1 (1) should state: *The properties of the materials selected for the control rod drive mechanism must be equivalent to those given in Appendix I to Section III of the ASME Code, or parts A and B of Section II of the ASME Code, or are included in Regulatory Guide 1.85, except that cold worked austenitic stain less steels should have a 0.2% offset yield strength no greater than 90,000 psi.' 252.2 e Subsection 4.5.1.1 (2) should state: 'All materials for use in this system must be selected for i their compatibility with the reactor coolant as described in Articles NB 2160 and NB-3120 of the ASME Code." 252.3 Subsection 4.5.2.2: The first sentence should read: "Core tuppor: stru:tures are fabricated in accordance with the requirements of ASME Code, Section III, Subsection NG-4000, and the examination and acceptance criteria shown in NG 5000.* 252.4 Subsection 4.5.2.3: The following statement should be added to the last sentence of the first paragraph: The examination will satisfy the requirements of NG 5300.' 252.5 Subsection 4.5.2.4 should state: ' Furnace sensitized material should not be allowed.' 252.6 Subsection 4.5.2.5 should state: 'All materials used for reactor internals will be selected for their compatibility with the reactor coolant as shown in ASME Code Section III, NG 2160 and NG 3120. The fabrication and cleaning centrols will preclude contamination c,f nickel based alloys by chloride ions, nuoride ions, or lead.' 20.2.5 Chapter 5 Questions 210.1 In Subsection 5.2.1.2, the statement is made that Section 50.55a of 10CFR50 requires NRC staff I approval of ASME code cases only for Class 1 components. Revise this statement to be consistent with the current (1987) edition of 10CFR50.55a, which requires staff approval of code cases for ASME Class 1,2, and 3 components. l 1 l 1 0 ! Amendment 2 20.2 3 l l I
ABM ua61ooar Standard Plant nrv. A 210.2 Revise Table 5.21 or provide additional tables in Subsection 5.2.1.2 which identify all ash 1E code cases that will be used in the construction and in plant operation of als ash 1E Class 1,2, and 3 components in the ABWR. All code cases in these tables should be identified by code case number, revision, and title. These tables should include those applicable code cases that are listed either as acceptable or conditionally acceptable in Regulatory Guides 1.84,1.85, and 1.147. For those code cases listed as conditionally acceptable, verify that the construction of all applicable components will be in compliance with the additional Regulatory Guide conditions. 250.1 Subsection 5.2.4.1 should state that the system boundary includes all pressure vessels, piping, pumps, and valves which are part of the reactor coolant system, or connected to the reactor systems, up to and including-(1) The outermost containment isolation valve in system piping that penetrates the primary reactor containment. (2) The second of two valves normally closed during normal reactor operation in system piping that does not penetrate primary reactor containment. (3) The reactor coolant system and relief vahrs. 250.2 Subsection 5.2.4.2 should satisfy the requirements in AShiE Code IWA 1500. 251.1 O Subsection 5.3.1.1 should state that the materials will comply with the provisions of the ash 1E Code, Section 111, Appendix 1, and meet the specification requirements of 10CFR50, Appendh G. 251.2 Subsection 5.3.1.2 should state the specific subsection NB of AShiE Code to which the manufacturing and fabrication specifications were alluded. 251.3 Subsections 53.1.4.4 and 53.1.4.5 should be rewritten; the cross reference is unacceptable. Subseetions 5.3.1.4.7, 5.3.1.5.2, 5.3.1.6.3, aod 5.3.2.1.5: Revision 2 of Regulatory G uide 1.99 should be added in these subsections. 251.4 Subsection 5.3.1.6.1: the third capsule of the vessel surveillance program is designated as a standby; however, according to ASThi 185 82, the capsule should be withdrawn at the end of life. Provide justification for this deviation. O Amudmut2 2o.2-6
I ABM axe,ooar Standard Plant REV A 2513 Subsection 5.3.1.6.3 states that according to estimates of worst-case irradiation effects, the adjusted reference temperature at end of life is less than 100'F, and the end of. life upper shelf energy exceeds 50 ft Ib. Proside the calculation and analysis associated with the estimate. _ r 251.6 i Subsection 5.3.2.1 should clarify where Reference 2 is located. Has the NRC staff reviewed and approved Reference 27 If not, the staff needs to resiew Reference 2 in order to complete the resiew of this subsection. , 251.7 Subsections 5.3.2.1.1, 5.3.2.1.2, 5.3.2.1.3, and 5.3.2.1.5 need to be rewritten. The level of detail must be comparabic to that of Standard Review Plan 53.2 and Branch Technical Position MTEB 52. 252.8 Subsection 533 cited three GE documents: (1) GE quality assurance program (2)
- Approved
- inspection procedures, and (3) NEDO 10029. ,
Has the NRC staff reviewed and approved the above documents? The staff cannot satisfactorily review inis subsection without reviewing the above three documents, i 251.9 Sub<ection 5.3.3.1.1.1 discusses the 60 year life of the ABWR reactor vessel. The NRC requirements and calculations on the fracture toughness and material properties are based on a 10 year life. Provide justificatioa for the applicability of NRC's requirements on the 60 year life reactor vessel. 251.10 Subsection 53.3.2 should include the following information: neutron fluence, shift in reference temperature RTNDT and upper shelf energy. The staff needs this information to compare to that of predicated values using Regulatory Guide 1.99. 251.11 Subsection 5.3.3.6 should indicate that operating conditions should satisfy the. pressure. temperature limits prescribed in Subsection 53.2.
)
O ! i 1 Amendment . 20.2-7 I l l
ABWR 2346iocar Standard Plant REv ^ 252.7 Subsection 5.2.3.2.2 is mostly an academic discussion of BWR water chemistry effect on intergranular stress corrosion cracking (IGSCC) in sensitized stainless steels. The subsection should discuss the actual ABWR water chemistry effects on the IGSCC. The subsection is vague about specific remedies or preventive measures to avoid IGSCC in ABWR, For example, the sub>ection failed to discuss how much hydrogen is needed for injection into the feedwater system or how the ' tight conductisity control
- would be implemented.
Also provide references for the ' Laboratory studies..." and 'available evidence...* that were mentioned in this subsection. 252.8 Subsection 5.2.3.2.3 should state that the requirements of GDC 4, relative to the compatibility of components with environmental conditions are met by compliance with the applicable provisions of the ASME Code and by compliance with the recommendation of Regulatory Guide 1.44. Specify the 'very low limits
- of the contaminants in the reactor coolant.
252.9 Subsection 5.233.1 should clarify where and how was the 45 ft lb Charpy V value obtained. The ferritic material used for piping, pumps, and valves should comply with Appendix G, Section G 3100, of ASME Code Section 111. This subsection should indicate that
- calibration of instruments and equipment shall meet the requirements of the code, Section III, Paragraph NB 23/0.*
h 252.10 Subsection 5.2.3.4.1.1 should be rewritten to include more detailed discussion on avoidance of significant sensitization and on how the ABWR design complies with the NRC regulatory requirements. 252.11 Subsection 5.2.3.4.2.3 states that the ABWR design meets the intent of this Regulatory Guide (1.71) by utilizing the alternate approach given in Section 1.8. We cannot review this subsection because we have not received Section 1.S. In addition, this subsection should be rewritten because it lacks detailed discussion about welder qualification. 281.1 in Section 5.1 (page 5.12) the function of the reactor cleanup system filter demineralizer should include the removal of radioactive corrosion and fission products in addition to particulate and dissohed impurities. Oll j Amendment 2 20 2-8 l l
ABM ux6iootr SigpJard Plant REV A 281.2 p v i i in Subsection 5.2.3.2.2 (page 5.2 7) irradiation assisted stress corrosion cracking (IASCC) of reactor i.;ternal components and its mitigation are not discussed. Present laboratory data and plant experience has shown that IASCC can be initiated even at low conductivity (< 0.3pS/cm) after long . exposure to radiation.
, 281.3 1
a in Subsection 5.2.3.2.2 (pages 5.2 7 and 8) the ABWR Standard Plant design does clearly incorporate hydrogen water chemistry to mitigate IGSCC Since the plant design life is .0 years, ' j hydrogen water chemistry may be of greater importance in reducing reactor coolant electrochemical corrosion potential to prevent IGSCC as well as IASCC If hydrogen water chemistry is the referenced ; ABWR standard design, the following documents should be cited: EPRI NP 5283-SR A, Guidelines for Permanent BWR Hydrogen Water Chemistry Installations 1987 Resision. EPRI NP-4947 SR.LD, BilR Hydrogen Water Chemistry Guidelines 1987 Revision (to be published). 281.4 i , j in Subsection 5.2.3.2.2 (page 5.2 9) the utilization of the General Electric zine injection passivation (GEZIP) process for radiation buildup control for the ABWR is not discussed. GEZIP was identified as a required design feature in the ABWR presentation to NRC staff. l 281J l O In Subsection 5.2.3.2,2 (page 5.2 9) prefilming of stainless steel appears to be a promisin
- method to reduce the buildup rate of activated corrosion products during subsequent plant operation.
1 SIL No. 428 recommends preoperational testing of the recirculation system conducted at temperatures , i 230 F be done with the dissolved oxygen level controlled to between 200 and 400 ppb. Is cc,ntrol I ) of radiation buildup through preoperational oxygen control being considered for the BWR Standard I Plant? Are mechanical polishing and electropolishing of piping internal surfaces also being considered for reducing radiation buildup? 1 1 281.6 In Subsection 5.2.3.2.2.2 (page 5.2 9) cobalt 60 is identified as the principle contributor to shutdown radiation levels, especially the recirculation piping system of BWRs. Stellite contributes
- about 90% of the total cobalt 59 input to the reactor water (EPRI NP 2263, BWR Cobalt Source 1 Identification, February 1982). Since irradiation of cobalt 59 yields cobalt 60, reduction in the l
source of cobalt 59 is needed to reduce the buildup of shutdown radiation levels. Indicate Stellite surface areas (square feet) in nuclear steam supply system and balance of plant. Provide the i criteria for selecting Stellite plant materials for the designed application. Provide evaluation of_ noncobalt containlug materials waose properties are adequate to replace Stellite in plant. j applications, l i ! i i O l Amendment 2 20.2.g l 1
MN 23A61ooAT Standard Plant REV.A 281.7 Subsection 5.2.3.2.2.3(4) (page 5.210) states that control of reactor water oxygen during O startup/ hot standby may be accomplished by utilizing the de aeration capabilities of the condenser. In addition, this section states that independent control of control rod drive (CRD) cooling water oxygen concentrations of < 50 ppb during power operation is desirable to protect against IGSCC of CRD materials. Are either one or both of the above dissolved oxygen controls incorporated in the ABWR Standard Plant design? 281.8 In Subsection 5.2.3.2.2.3(13) (page 5.211) it states thC the main steam line radiation monitor indicates an excessive amount of hydrogen being injected. An explanation of this occurrence should be discussed. 281.10 in the October 1987 ABWR presentation to the NRC staff the design features and/or requirements to improve water chemistry for GE ABWR were specified. Address each one of these design features and/or requirements listed in Table I in the ABWR Standard Safety Analysis Report. TABLEI Comparison of requirements in ABWR standard safety analyses report and ABWR presentation to NRC staff (October 21 and 22,1987) ABWR Presentation ABWR Standard Safety O to NRC Staff Analysis Report 1- Selection oflow cobalt Required Design Feature Not discussed in materials to minimize Subsection 5.23. radiation buildup 2- Hydrogen water chemistry Required Design Feature Subsection 5.23.2.2 to suppress IGSCC references normal water chemistry. 3- Zine injection to mini- Required Design Feature Not discussed in mize radiation buildop Subsection 5.23.2.2.2. 4- Full flow deep bed Required Design Feature Not discussed in condensate system Subsection 5.23.2.23. to reduce feedwater impurities O Acendment 2 20.2 10
ABM 23461004r Standard Plant any.A TABLEI A U Comparison of requirements in ABWR standard safety analyses report and ABWR presentation to NRC staff (October 21 and 22,1987) (continued), , ABWR Presentation ABWR Standard Safety to NRC Staff Analysis Report 5- Improved online ton chromatography, Only electrochemical . monitoring instrumen- electrochemical corrosion corrosion potential tation to assure water potential, and crack arrest discussed in Subsec-quality verification system tion 5.23.2.23. required design features 6- Improved corrosion- Required Design Feature Not discussed in resistant materials for Subsection 5.23.2.23. steam extraction piping to minimize feedwater impurities 7- Highly corrosion- Required Design Feature Not discussed in resistant condenser Subsection 5.23.2.23. tubes to minimize leakage into condensate system V 8- Maintain electrochemical Required Design Feature Not listed in corrosion potential Table 5.2-5.
< 0.23 V to suppress IGSCC 9- Erosion / corrosion. Design Feature Not discussed in resistant materials Subsection 5.4.9.
in steam extraction and drain lines to minimize failures 10 - Ease oflead detection Design Feature May be in Subsection 10.4.1 in and repair of the which has not been main condenser submitted yet. 11 - 2% Reactor water cleanup Design Feature Not discussed in system to improw water Subsection 5.23.2.2. quality and occupational radiation exposure i 12 - Full flow recirculation Design Feature Not discussed in to main condenser from Subsection 5.23.2.23. cleanup outlet to reduce l feedwater impurities ' O l Amendment 2 20.2.11 l
ABWR uA610aAr Standard Plant RIN. A 20.2.6 Chapter 6 Questions O 250 3 Subsection 6.6.8 should discuss the augmented inservice inspection for those portions of high energy piping enclosed in guard pipes. 252.12 Subsection 6.1.1.1 should discuss ferritic steel welding in detail. It should also discuss the control of ferrite content in stainless steel weld metal similar to that of Regulatory Guide 131, 252.13 Subsections 6.1.1.1.3.1, 6.1.1.1.3.2, a n d 6.1.1.1.3.5 should be rewritten because the cross reference is unacceptable. 281.9 Subsection 6.4.4.2 (page 6.4 6) discusses personnel respirator use in the event of toxic gas intrusion into the control room. Ilowever, the chlorine detection system is not discussed. Also, any control functions that are automatically triggered by a chlorine detector alarm (closing intake dampers, energizing controf room HVAC system recirculation) should be identified. O O Amendment 2 20212
ABWR 234610041 Standard Plant REV.A 20.2.7 Chapter 7 Questions None to date. l I O 1 l l l l l O Amendment 2 20213
ABWR - ummar Standard Plant RLY. A 20.2.8 Chapter 8 Questions None to date. g O l l l l 9 Amendment 2
ABWR ux61 mar Standard Plant REV A 20.2.9 Chapter 9 Questions O NODC to date. O I l l O Amendment 2 3213
ABM 23461oon1 Standard Plant REV.$ 20.2.10 Chapter 10 Questions None to date, h i l O 1 l 9 Amendment 2 20.2.l6
I MN 23A6100AT Standard Plant RIN A 20.2.11 Chapter 11 Questions O None to date. O 1 O Amendment 2 y217
Mkk 23A0100AT Standard Plant REV.A 20.2.12 Chapter 12 Questions None to date. I l l l l l O O Amendment 2 20.2 18 i
MM MA61%AT Standard Plant arv A 20.2.13 Chapter 13 Questions O None to date. O l , O Amendment 2 M2-19 _ = , . _ - _
ABWR - 2W100AT Standard Plant REV A 20.2.14 Chapter 14 Questions None to date. Ol' l I l l l O' l I l l l l O Amendment 2 20.2 3
4 4 ABM 2346ioaar Standard Plant REV A 20.2.15 Chapter 15 Questions 470.1 Subsection 15.6.2 of the ABWR FSAR prosides your analysis for the radiological consequences of a
- failure of smalllines carrying primary coolant outside of containment. This analysis only considers j the failure of an instrument line with a 1/4 inch flow restricting orifice. Show that this failure 1
scenario provides the most severe radioactive releases of any postulated failure of a smallline. Your evaluation should include lines that meet GDC $$ as well as small lines exempt from GDC 55. j 470.2 Proside a justification for your assumption that the plant continues to operate (and therefore no iodine peaking is experienced) during a small line break outside containment (Subsection 15.6.2) accident scenario. Also provide the basis for the assumption that the release duration is only two hours. 470.3 Subsection 15.6.4.5.1.1 of the FSAR gives the iodine source term (concentration and isotopic mix) used to analyze the steam line break outside of containment accident. The noble gas source term, however, is not addressed. Provide the noble gas source term used. Also, the table in Subsection 15.6.4.5.1.1 seems heavily weighted to the shorter lived activities (i.e., (I 134). Provide the , bases for the isotopic mix used in your analysis (iodine and noble gas). ' 470.4 O s#d tie 25.6 5.5 ' t 's t 's i>>is is 6 ae -etie vre'4a d i= a 8 t terr o "ia-1.3 except where noted. For all assumptions (e.g., release assumed to occur one hour after accident i initiation, the chemical species fractions for iodine, the temporal decrease in primary containment leakage rates, credit for condenser leakage rates, and dose conversion factors) which devive from NRC guidance such as regulatory guides and ICRP2, provide a detailed description of the jus:ification , for the deviation or a reference to another section of the SSAR where the deviations are diset.ssed in i detail. Provide a comparison of the dose estimates using these assumptions versus those which would
- result from using the NRC guidance.
1 470.5 l Provide a discussion of, or reference to, the analysis of the radiological consequences of leakage
! from enginected safety feature components after a design basis LOCA.
4 470.6 For the spent fuel cask drop accident, what is the assumed period for decay from the stated power ] condition? What is the justification for that assumption? l 470.7 1 j Tbc tables in Chapter 15 should be checked and revised as appropriate. In several cases the footnotes contain typographical errors related to defining the scientific notation. Table 15.712 ! also appears to contain inappropriate references to Table 15.716, rather that Table 15.713. !O 1 Amendment 2 2o.2 21
ABWR maar Standard Plant REV.A J70.8 It is stated that Regulatory Guides 1.3 and 1.45 were used in the calculations of X/O values. O Based on the values presented, it appears as though a Pasquill stability Class F and one meter per second wind speed were assumed, with adjustment for meander per Figure 3 of Regulatory Guide 1.145. If this is not the case, describe the assumptions and justification used in calcule. ting the X/O values which are used in Chapter 15 dose assessments. 470.9 The SGTS filter efficiencies of 99% for inorganic and organic iodine are higher than the 90% and 70% values, respectively, assumed in Regulatory Guide 1.25 if it can be shown that the building atmosphere is exhausted through adsorbers designed to remove iodine. Provide a justification for the use of the higher values. 470.10 Dose related factors such as breathing rates, iodine conversion factors and finite versus infinite cloud assumptions for calculating the whole body dose are not stated explicitly, ahhough reference is made to Regulatory Guide 1.25 and another document. State these assumptions explicitly and justify use of any values which deviate from Regulatory Guide 1.25. O O Amendment 2 gym
ABWR 2346 w ,1 SlaDJi ard Plant Rev.A 20.2.16 Chapter 16 Questions O None to date. 1 I O O Amendment 2
ABWR ux6imar Standard Plant REV.A 20.2.17 Chapter 17 Questions g Nonc to date. l l l l l l i e:i 1 l l l l O Amendrntet 2 20.2 24
MN MA6100AT Standard Plant guy A 20.2.18 Chapter 18 Questions None to date. O O Amendment 2 37,3
ABWR - ua6im41 Standard Plant REV A 20.2.19 Chapter 19 Questions None to date. O O i l I i Amendment 2 20.2 26 l 1
ABM 234 iocar Standard Plant REV.A SECTION 20.3 !O i CONTENTS
- Section lillt Eagt
] 203.1 Resnonne to First RAI Reference 1 203 1 i TABLES 1 j Table 11112 East 203 1 Comparison of Requirements in ABWR Standard Safety I Analyses Report and ABWR Presentation to b1C 5taff l. (October 21 and 22,1987) 203 13 I 20 3-2 Sensitivity Study of Parameters for LOCA Analysis 203 22 i ) 1 iO t l O 20 x Ameadment 2
ABM 23^6100Ar Standard Plant REV A , 4 20.3 QUESTIONS / RESPONSES - This subsection provides the responses for each of the NRC questions identified in Sections 20.1 and 20.2. For convenience, each question is repeated here before its corresponding response. These questions / responses are provided in groups corresponding to the NRC Requests for AdditionalInforma- , tion (RAI) referenced in Section 20.4. Within each group, the questions / responses are presented in ' the numerical order of the question numbers. ' 20.3.1 Response to First RAI. Reference 1 QUESTION 210.1 I in Subsection 5.2.1.2, the statement is made that Section 50.55a of 10CFR50 requires NRC staff approval of AShf E Code Cases only for Class I components. Revise this statement to be consistent with the current (1987) edition of 10CFR50.55a which requires staff approval of Code Cases for ash 1E Class I,II, and 111 components. ; RESPONSE 210.1 i Response to this question is prosided in resised Subsection 5.2.1.2. QUESTION 210.2 Revise Table 5.21 or provide additional tables in Subsection 5.2.1.2 which identifies all AShtE Code Cases that will be used in the construction and in. plant operations of all ash!E Class I, II, and ill components in the ABWR, All Code Cases in these tables should be identified by Code Case number, revision and title. These '.,bles should include those applicable Code Cases that are listed either ; as acceptable or conditionally acceptable in Regulatory Guides 1.84,1.85 and 1.147. For those Code Cases listed as conditionally acceptable, verify that the construction of all applicable components will be in compliance with the additional Regulatory Guide conditions. RESPONSE 210.2 1 Response to this question is prosided in revised Subsection 5.2.1.2 and Table 5.21.
- QUESTl0N 250.1 Subsection 5.2.4.1 should state that the system boundary includes all pressure sessels, piping, l pumps, and valves which are part of the reactor coolant system, or connected to the reactor coolant sptems, up to and including (A) The outermost containment isolation valve in system piping that penetrates the primary reactor containment. I (B) The second of two valves normally closed during normal reactor operation in system piping that does not penetrate primary reactor containment.
l l (C) The reactor coolant system and relief vahes.
) RESPONSE 250.1 I
1 Response to this question is prosided in resised Subsection 5.2.4.1. i O
. Amendment 2 203 1 I
j
ABWR mmwr Standard Plant nov ^ QUESTION 250.2 Subsection 5.2.4.2 should satisfy the requirements in AShlE Code, IWA 1500. 9 RESPONSE 250.2 Response to this question is prosided in resised Subsection 5.2.4.2. QUESTION 2503 Subsection 6.6.8 should discuss the augmented inservice inspection for those portions of high energy piping enclosed in guard pipes. RESPONSE 250.3 Augmented inservice inspection is not required for the ABWR design since there are no guard pipes erdosing high energy piping between the containment isolation valves. QUESTION 251.1 Subsection 5.3.1.1 should state that the material will comply with the provisions of the AShtE Code, Section ill, Appendix 1, and meet the specification requirements of 10CFR50, Appendit G. RESPONSE 251.1 Response to this question is prosided in revised Subsection $3.1.1 QUESTION 251.2 h Subsection 5.3.1.2 should state the speciGe subsection NB of AShtE Code to which the manufacturing and fabrication specifications were alluded. RESPONSE 251.2 Respon e to this question is prosided in revised Subsection 53.1.2. QUESTION 251.3 Subsections 53.1.4.4 and 53.1.4.5 should be rewritten; the cross reference is unacceptable. Subsections 5.3.1.4.7. 5.3.1.5.2, 5.3.1.6.3, and 5.3.2.1.5; Revision 2 of Regulatory Guide 1.99 should be added in these subsections. RESPONSE 251.3 Response to the first part of this question is provided in revised Subsections 5.3.1.4.4 and 53.1.4.5. The GE ABWR Licensing Resiew Bases issued by the NRC on August 7,1987 specifies a SRP cffectisity date of htarch 30,1987. Thus, the Regulatory Guides in effect as of that date are applicable to the ABWR. Ilowever, rather than providing the specine revision of each Regulatory Guide each time it is noted in the test a ne SSAR, GE has chosen to provide the applicable revisions of the Regulatory Guides in SSAR Subsection 1.8.2, w hich will be presided by June 30,1988, g Amer 4 ment 2 2012
ABM u^6too^r Standard Plant strv. A - QUESTION 251.4 O Subsection 5.3.1.6.1: the third capsule of the vessel surveillance program is designated as a standby; however, according to ASTM 185 82, the capsule should be withdrawn at the end of life. Provide . justification for this desiation. i RESPONSE 251.4 Response to this question is prosided in revised Subsection 5.3.1.6.1 f QUESTION 251J . Subsection 5.3.1.6.3 states that according to estimates of worst case irradiation effects, the ' 2 adjusted reference temperature at end.of life is less that 100'F, and the end of life upper shelf i energy ex:ceds 50ft lb. Provide the calculation and analysis associated with the estimate. 1 RESPONSE 2513 f The calculation and analysis associated with the estimate is prosided below-l l Calculate RTNDTShift in Vessel Material : Ref.: February 1986 draft of Regulatory Guide 1.99 A. I Weld Metal )
. l ! Assume the following maximum values: l
? 1 0 P = 0.020%, V = 0.05% Cu = 0.08% 1 Ni = 1.20% (Max Ni value considered in Regulatory Guide)
- I I A RTNDT surface = (CF) f (0.28 0.10 log f)
Chemistry factor CF = 108'F Fluenee: 4.0 x 1017 n/em2 j f = 4.0 x 1017*19 = 4.0 x 10-2 .I a RTNDT = 108 x (4.0 x 10 2) (0.28 0.10 log 0.04) i
= 22 3 'F
- 11 Plate: Cu = 0.05%, P = 0.015%, Ni = 0.73% (max) .
l CF = 31'F, Fluence 4.0 x 1017 n 2 a RT 2 x 27.96 = 8.03*/cm F l NDT = 108 l III Forging: Cu = 0.05%, P = 0.015%, Ni = 1.0% (max) 3 CF = 31'F, Fluence 4.0 x 1017 n/cm2 1 a RT 3 x 27.% = 8.03*F ] NDT = 108 i i !O Amendment 2 ll0.3 3 i
.----,,-.------.,-,,--.-e.. . - - , - - . . .,-v ,.n.. . . - - - - - . . . , ,.,e-.-- - ---., . -., ,-e -- -~ -
l ABWR . momr Standard Plant REV.A II. For fluence 6.0 x 1017 n 2 after 60 > tars f = 6.0 x 10(1719) 0.06 =/cm g; I Weld Metal CF = 108'F A RTNDT = 108 x (0.05) (0.23 0.10 log 0.06)
= M.83'F II,III Forcine and Plate CF = 31*F aRTNDT = M.83 F x.3 - 10'F 108 UNDT requirements rer Vessel Snecifications (Maximum Specified Values)
Shell courses, 20* F and nonles Weld 20* F MNDT Shifts ner new Reculatory Guide 139 (February 10S6 Draft) Calculated Shift AllWR ART at surface Final Materal Initial RTNDT 4x1017 n/cm2 after 40 )ms Margin = 2 / og2 og2 RTNDT (ART) h Weld 20"F 27.%'F 27.96* F 36'F Plate or 20'F 8,03' F 8.0' F 4* F Forging Final RTNDT = ART = Initial RTNDT + ART + Margin I i The above projections are for the 40 year full power basis. The corresponding . final RTNDT for 60 years would be 50 F for welds and O'F plates and I forgings. ; 1 l l O -. -_ o m.
i ABM 2346i004r . l Standard Plant EV A DROP IN UPPER SHELF ENERGY !
.! O Ref: Regulatory Guide 1.99 Rev 2 l
Material % Cu % Drop per Fig 2 of RG 1.99' l Weld .08 max 14 i g Base Metal .05 max 11
- Basedon cutoffluence of1018 thTTIAL VALUE 75 FT.Ib Final Values 1; ,
" i Weld 75 x 0.S6 = 65 Base 75 x 0.83 = 67 QUESTION 251.6 l Subsection 5 3.2.1 should clarify where ' Reference 2* is located. Has the NRC staff reviewed and approved Reference 27 If not the staff needs to review Reference 2 in order to complete the review of this subsection. - RESPONSE 251.6 ; Reference 2, Transient Pressure Rises Affecting Fracture Toughness Requirements for Boiling Water i Reactors, January 1979, (NEDO 21778.A), is an NRC staff approved licensing topical report. This , topical report was approved by letter to GE, dated November 13,1978 according to NUREG.0390 Vol.7, No. l 2 (October 15, 1984). i 1 j QUEST 10N 251.7 1
! Subsections 5.3.2.1.1, 5.3.2.1.2, 5.3.2.1.3, and 5.3.2.1.5 need to be rewritten. The level of i detail must be comparable to that of Standard Review Plan 53.2 and Branch Technical Position MTEB 5-2. i RESPONSE 251.7 i Response to this question is provided in resised Subsections 53.2.1,53.2.1.1, and 53.2.1.5.
1 QUESTION 251.8 l ) Subsection 533 cited three GE documents: (1) GE quality assurance program, , I (2)
- Approved
- Inspection procedures, and
- (3) NEDO.10029 l 2
l Has the NRC staff reviewed and approved the above documents? The staff cannot satisfactorily resiew j this subsection without resiewing the above three documents. ' l 1 i I Amendmeat 2 20.3-5 i i I
ABWR mm.m Standard Plant R13' A RESPONSE 251.8 The GE quality assurance program is contained in topical report NEDO.11029 04A, GE Bil'R Quality O Assurance Progrant, Resision 7, which has been approsed by the NRC staff (May 1987).
' Approved inspection procedures
- refers to GE approved inspection procedures which govern the manufacturing, fabrication, and testing operations of the reactor vessel fabrication process. These inspection procedures are originated at the time the reactor vessel fabricator is selected, and, as has been the case in the past, the NRC staff will have review opportunities in accordance with 10CFR$0 Appendix B.
NEDO 10029, An Analytical Study on Brittle Fracture of GE.Bil'R l'essel Subject to the Drsign Basis Accident, July 1969, was also referenced in Subsection $33 of GESSAR, Docket No. STN 50 447. This information applica equally well to the ABWR. QUESTION 251.9 Subsection 5.33.1.1.1 discusses the (4) ear life of the ABWR reactor vessel. The NRC requirements and calculations on the fracture toughness and material properties are based on a 40 year life. Provide justification for the applicability of the NRC's requirements on the 60. year life reactor vessel. RESPONSE 251.9 Response to this question is provided in resised Subsection 533.1.1.1. QUESTION 251.10 Subsection 5.3.3.2 should include the following information: neutron fluence, shift in reference temperature RTNDT, and upper shelf energy. The staff needs this information to compare to that of predicted values using Regulatory Guide 1,99. RESPONSE 251.10 Response to this question is prosided in revised Subsection 533.2. QUESTION 251.11 Subsection 533.6 should indicate that operating conditions should satisfy the pressure. temperature limits prescribed in Subsection 53.2. RES PONSE 251.11 Response to this question is provided in resised Subsection $33.6. QUESTION 252.1 Subsection 4.5.1.1 (1) should state: 'The properties of the materials selected for the control rod drive mechanism must be equivalent to those gisen in Appendit I to Section ill of the ASME Code or parts A and B of Section 11 of the ASME Code or are included in Regulatory Guide 1.85, except that cold worked austenitic stainless steels should have a 0.2% offset yield strength no greater than 90,000 psi.' O Anoendment 2 :o u
ABWR ua61oorr Standard Plant REV.A RESPONSE 252.1 b Response to this question is provided in revised Subsection 4.5.1.1 (1). QUESTION 252.2 Subsection 4.5.1.1 (2) should state: "All materials for use in this system must be selected for their compatibility with the reactor coolant as described in Articles NB-2160 and NB 3120 of the ASME Code."
. SPONSE 252.2 sponse to this question is provided in revised Subsection 4.5.1.1 (2). t' 4
'ESTION 252.3 Subsection 4.5.2.2: The first sentence should read, "Core support structures are fabricated in accordance with the requirements of ASME Code, Section III, Subsection NG-4000, and the examination and acceptance criteria shown in NG 5000."
RESPONSE 2523 Response to this question is provided in resised Subsection 4.5.2.2. QUESTION 252.4 Subsection 4.5.2.3: The following statement should be added to the last sentence of the first I O earasraph. The ex mination w4ti satisfv the resetrements ef No-5300-i RESPONSE 252.4 . Response to this question is provided in resised Subsection 4.5.23. QUESTION 252.5 Subsection 4.5.2.4 should state:
- Furnace sensitized material should not be allowed."
RESPONSE 252.5 Response to this question is provided in revised Subsection 4.5.2.4. QUESTION 252.6 Subsection 4.5.2.5 should state: 'All materials used for reactor internals will be selected for their compatibility with the reactor coolant as shown in ASME Code Section III, NG 2160 and NG 3120. The fabrication and cleaning controls will preclude contamination of nickel base alloys by chloride ions, Duoride ions, or lead.' RESPONSE 252.6 Response to this question is prosided in revised Subsection 4.5.2.5. O Amendment 2 20.3 7 _ _ . _ _ _ -_ _ - . _ _ _ _ _,_~ _
ABM 23AMMAT Standard Plant REV.A QUESTION 252.7 Subsection 5.23.2.2 is mostly an academic discussion of BWR water chemistry effect on intergranular stress corrosion cracking (IGSCC) in sensitized stainless steels. The subsection should discuss the actual ABWR water chemistry effects on IGSCC. The subsection is vague about specific remedies or preventive measures to avoid IGSCC in ABWR. For example, the subsection failed to discuss how much hydrogen is needed for injection into the feedwater system or how the "tight conductivity controP would be implemented. Also, provide references for the "Laboratory studies...' and "available evidence. .* that were mentioned in this subsection. RESPONSE 252.7 Response to this question is provided in resised Subsection 5.23.2.1. QUESTION 252.8 Subsection 5.2.3.23 should state that the requirements of GDC 4, relative to the compatibility of components with environmental conditions, are met by compliance with the applicable provisions of the ASME Code and by compliance with the recommendation of Regulatory Guide 1.44. Specify the "very low limits" of the contaminants 'n the reactor coolant. RESPONSE 252.8 Response to this question is provided in resised Subsection 5.23.23. QUESTION 252.9 O Subsection 5.233.1 shold clarify where and how the 45 ftJb Charpy V value was obtained. The ferritic material used for piping, pumps, and valves should comply with Appendix G, Section G 3100, of ASME Code Section III. This subsection should indicate: ' calibration of instruments and equipment shall meet the requirements of the code, Section Ill, Paragraph NB 2360.* RESPONSE 252.9 Response to this question is provided in resised Subsection 5.233.1.
- QUESTION 252.10 1
Subsection 5.2.3.4.1.1 should be rewritten to include more detailed discussion on avoidance of significant sensitization and on how the ABWR design complies with the NRC regulatory requirements. RESPONSE 252.10 Response to this question is prosided in resised Subsection 5.23.4.1.1. O Amendment 2 20.3-8 l t l l l l
ABM 23A6100AT Standard Plant REV.A QUESTION 252.11 o) V Subsection 5.2.3.4.2.3 states that the ABWR design meets the intent of this Regulstory Guide (1.71) by utilizing the alternate approach given in Section 1.8;'We cannot review this subsection because we have not received Section 1.8. In addition, this subsection should be rewritten because it lacks detailed discussion about welder qualification. RESPONSE 252.11 Response to this question is provided in resised Subsection 5.23.4.23. QUESTION 281.1 In Section 5.1 (page 5.12) the function of the reactor cleanup system filter demineralizer should include the removal of radioactive corrosion and fission products in addition to particulate and dissolved impurities. RESPONSE 281.1 Response to this question is provided in resised Subsection Y.t.T.Y.Y - QUESTION 281.2 In Subsection 5.2.3.2.2 (page 5.2-7) irradiation assisted stress corrosion cracking (IASCC) of reactor internal components and its mitigation are not discussed. Present laboratory data and plant experience has shown that IASCC can be initiated even at low conductivity (< 03pS/cm) after long exposure to radiation. \) RESPONSE 281.2 Response to this question is provided in the new Subsection 5.23.2.4,IGSCC Considerations. QUESTION 2813 + In Subsection 5.2.3.2.2 (pages 5.2 7 and 8) the ABWR standard plant design does not clearly incorporate hydrogen water chemistry to mitigate IGSCC. Since the plant design life is 60 years, hydrogen water chemistry may be of greater importance in reducing reactor coolant electrochemical corrosion potential to prevent IGSCC as well as IASCC If hydrogen water chemistry is the referenced ABWR standard design, the following documents should be cited: EPRI NP 5283-SR-A, Guidelines for Permanent BlVR Hydrogen IVater Chemistry Installations - 1987 Revision. EPRI NP-4947-SR LD, BilR Hydrogen IVater Otemistry Guidelines 1987 Revision (to be published). RESPONSE 2813 Subsection 5.2.3.2.2 will be modified by September 30,1988 to more clearly discuss hydrogen water chemistry as part of the ABWR standard plant design. O Amendment 2 20.3-9
l M _ 23A6100AT Standard Plant REV. A QUESTION 281.4 In Subsection 5.2.3.2.2 (page 5.2 9) the utilization of the General Electric zine injection passivation (GEZIP) process for radiation buildup control for the ABWR is not discussed. GEZIP was identified as a required design feature in the ABWR presentation to NRC staff. RESPONSE 281.4 The General Electric zine injection passivation process (GEZIP) is not in the Nuclear Island scope. However, an interface requirement has been added (see new Subsection 5.7.6) that requires the remainder of the plant to meet the water quality requirements of Table 5.2 5. QUESTION 281.5 In Subsection 5.2.3.2.2 (page 5.2 9) prefilming of stainless steel appears to be a promising method to reduce the buildup rate of activated corrosion products during subsequent plant operation. SIL No. 428 recommends preoperational testing of the recirculation system conducted at temperatures 230 F be done with the dissolved oxygen level controlled to between 200 and 400 ppb. Is control of radiation buildup through preoperational oxygen control being considered for the BWR standard plant? Are mechanical polishing and electropolishing of piping internal surfaces also being considered for reducing radiation buildup? RESPONSE 281.5 Since the recirculation system piping has been eliminated from the ABWR design, SIL No. 428 does not apply. Preoxidation, mechanical polishing, and electropolishing are not being considered for other ABWR components at this time. However, these methods are available as promising techniques to reduce radiation buildup on all internal stainless steel surfaces. g QUESTION 281.6 In Subsection 5.2.3.2.2.2 (page 5.2 9) cobalt 60 is identified as the principle contributor to shutdown radiation levels, especially the recirculation piping system of BWRs. Stellite contributes about 90% of the total cobalt 59 input to the reactor water (EPRI NP-2263, BWR Cobalt Source Identification, February 1982). Since irradiation of cobalt 59 yields cobalt 60, reduction in the source of cobalt 59 is needed to reduce the buildup of shutdown radiation levels. Indicate Stellite surface areas (square feet) in nuclear steam supply system and balance of plant. Provide the criteria for selecting Stellite plant materials for the designed application. Provide evaluation of noncobalt containing materials whose properties are adequate to replace Stellite in plant applications. RESPONSE 281.6 (1) Stellite Surface Area for BWR/6: l Total Nuclear Steam Supply System: 74.39 Sq. Ft. Total Balance of Plant: 138.0 Sq. Ft. For ABWR design, the above numbers are greatly reduced. Cobalt based alloys have been climinated from fuel assemblies, and control rod blades and drives. Amendment 2 203 10 l i
}
ABM MA6100AT Standard Plant REV.A (2) Criteria for Selecting Stellite Materials: ( 1. Wear resistance
- 2. Weldability
- 3. Experience and senice history
- 4. Radiation levelin area of application (3) Evaluation of Noncobalt-containing Material to Replace Stellite:
The major source of cobalt from the reactor core has been Haynes 25 and Stellite 3 (cobalt based alloys) for pins and rollers, respectively, in BWR control rods. Replacement of the cobalt alloy pins and rollers with noncobalt alloys has been extensively investigated under a joint GE-EPRI program (Project 1331-1). The results of this investigation are documented in the report, EPRI NP-2329, Project 13311, Final Report, March 1982. The current design noncobalt l materials are alloy X-750 for control rod rollers and 13-8 PH for the pins. QUESTION 281.7 Subsection 5.2.3.2.2.3(4) (page 5.2-10) states that control of reactor water oxygen during startup/ hot standby may be accomplished by utilizing the de aeration capabilities of the condenser. ; in addition, this section states that independent control of control rod drive (CRD) cooling water l oxygen concentrations of < 50 ppb during power operation is desirable to protect against IGSCC of CRD' materials. Are either one or both of the above dissolved oxygen controls incorporated in the ABWRA standard plant design? " j RESPONSE 281.7 In Subsection 5.2.3.2.2.3, control of reactor water oxygen by using the condenser and control of control rod drive water were mentioned as dissolved oxygen control methods. These two plant features are not in the Nuclear Island scope. However, an interface requirement has been added (see new Subsection 5.2.5) that requires the remainder of the plant to meet the water quality requirements of Table 5.2 5. QUESTION 281.8 In Subsection 5.2.3.2.2.3(13) (page 5.211) it states that the main steam line radiation monitor indicates an excessive amount of hydrogen being injected. An explanation of this occurrence should be discussed. RESPONSE 281.8 I Subsection 5.2.3.2.2.3(13) will be revised by September 30, 1988 to discuss the effects of excessive hydrogen injection upon the main steam line radiation monitor. QUESTION 281.9 Subsection 6.4.4.2 (page 6.4-6) discusses personnel respirator use in the event of toxic gai-intrusion into the control room. However, the chlorine detection system is not discussed. Also, any-control functions that are automatically triggered by a chlorine detector alarm (closing intake dampers, energizing control room HVAC system recirculation) should be identified. O Amendment 2 20.511
33A3100AT l Standard Plant REV A 1 RESPONSE 281.9 l Response to this question will be provided upon submittal of control room HVAC system scheduled for December 31,1988. l QUESTION 281.10 In the October 1987 ABWR presentation to the NRC staff the design features and/or requirements to improve water chemistry for GE ABWR were specified. Address each one of these design features and/or requirements listed in Table I in the ABWR Standard Safety Analysis Report. TABLEI Comparison of requirements in ABWR standard safety analyses REPORT and ABWR presentation to NRC staff (October 21 and 22,1987) ABWR Presentation ABWR Standard Safety to NRC Staff Analysis Report 1- Selection oflow cobalt Required Design Feature Not discussed in materials to minimize Subsection 5.23. radiation buildup l 2- Hydrogen water chemistry Required Design Feature Subsection 5.23.2.2 ; to suppress IGSCC references normal I water chemistry. 3- Zine injection to mini- Required Design Feature Not discussed in mize radiation buildup Subsection 5.23.2.2.2. 4- Full flow deep bed Required Design Feature Not discussed in I condensate system Subsection 5.23.2.23. j to reduce feedwater impurities 5- Improved online Ion chromatography, Only electrochemical monitoring instrumen. electrochemical corrosion corrosion potential tation to assure water potential, and crack arrest discussed in Subsec-quality verification system tion 5.23.2.23. required design features 6- Improved corrosion- Required Design Feature Not discussed in resistant materials for Subsection 5.23.2.23. steam extraction piping to minimize feedwater impurities O Amendment 2 20.3 12
l 23A6100AT Standard Plant REV.A O Q TABLEI l Comparison of requirements in ABWR standard safety analyses REPORT l and ABWR pasentation to NRC staff (October 21 and 22,1987) (continued) ' AB%R Presentation ABWR Standard Safety to NRC Staff Analysis Report 7- Highly corrosion- Required Design Feature Not discussed in resistant condenser Subsection 5.23.2.23. tubes to minimize leakage into condensate system 8- Maintain electrochemical Required Design Feature Not listed in corrosion potential Table 5.2-5.
< 0.23 V to suppress IGSCC 9- Erosion / corrosion- Design Feature Not discussed in resistant materials Subsection 5.4.9.
in steam extraction and drain lines to minimize failures 10 Ease oflead detection Design Feature May be in Subsection 10.4.1 in and repair of the which has not been main condenser submitted yet. 11 2% Reactor water cleanup Design Feature Not discussed in sptem to improve water Subsection 5.23.2.2. quality and occupational radiation exposure 12 - Full flow recirculation Design Feature Not discussed in to main condenser from Subsection 5.23.2.23. cleanup outlet to reduce feedwater impurities O l Amendmeat 2 20.3 13
- - _ _ - - ---__ _ ~. _ - _ _E ___ ,
ABWR 2346toorr Standard Plant REV.A RESPONSE 281.10 ltem l O After the first paragraph in Subsection 5.23.2.2.2, add the following as a second paragraph: A! a means to reduce cobalt, GE has reduced cobalt content in alloys to be used in high fluence areas such as fuel assemblies and control rods. In addition, cobalt base alloys used for pins and rollers in control rods have been replaced with noncobalt alloys. Item 2 Subsection 5.23.2.2 will be revised by September 30,1988 to reference the EPRI guidelines for hydrogen water chemistry and for installation of the facilities. Item 3 Information is being obtained and evaluated from operating plants with GEZIP. However, this feature is not in the Nuclear Island scope. Item 4 This feature is not in the Nuclear Island scope. However, an interface requirement has been added (see new Subsection 5.2.6) that requires the remainder of the plant to meet the water quality requirements in Table 5.2-5. Item 5 New and improved water quality monitoring instrumentation is being constantly developed and introduced for use in BWR plants. Several useful instruments base been developed and introduced within the past few years. GE will evaluate the state of the art when a BWR is undergoing detailed design and will incorporate such instruments that are necessary to assure proper water quality. Item 6 Response to Item 6 of this question is prosided in resised Subsection 5.23.2.23. Item 7 Response to item 7 of this question is prosided in revised Subsection 5.23.2.23. Item 8 Table 5.2 5 will be resised by September 30,1988 to include control of ECP. Item 9 Response to Item 9 of this question is provided in revised Subsection 5.23.2.23. l l An eadment 2 gow l l l J
ABM 2346200xr Standard Plant RD/ A Item 10 Response to Item 10 of this question is provided in revised Subsection 5.23.2.23. Item 11 ' In the ABWR standard plant design, a 2% reactor water cleanup system is provided. By September; 30,1988 Subsection 5.23.2.2 will be changed to discuss this. Item 12 This design feature is not in the Nuclear Island scope. However, an interface requirement has been added (see new Subsection 5.2.6) that requires the remainder of the plant to meet the water quality requirements in Table 5.2 5. QUESTION 470.1 Subsection 15.6.2 of the ABWR FSAR provides your analysis for the radiological consequences of a failure of smalllines carrying primary coolant outside of containment. This analysis only considers the failure of an instrument line with a 1/4 inch flow restricting orifice. Show that this failure scenario provides the most severe radioactive releases of any postulated failure of a small line. Your evaluation should include lines that meet GDC 55 as well as smalllines exempt from GDC 55. RESPONSE 470.1 The analysis for failure of a small line carrying primary coolant was conservatively analyzed as a failure of an instrument line with full flow for a period of two hours. This analysis is deemed O ce s >v <>ve rer ts re se sive seie . Of all the lines carrying coolant penetrating the primary containment wall, only the instrument. lines are exempt from GDC 55. All other lines use some form of check valve / motor-operated valve s combination to stop the flow of primary coolant in the event of a line break. Typically, the motor operated valves close at the rate of two inches per ten seconds. Considering a two-inch line and assuming that a flow of 175 pounds per second would result in operator action within 60 seconds, the total mass released over the 70 second period would be approximately 12,000 pounds or about one half of the assumed release over two hours from the instrument line. Using this logic and these simplified calculations,it is found that a two-hour instrument line break bounds releases for small lines. QUESTION 470.2 Provide a justification for your assumption that the plant continues to operate (and therefore no iodine peaking is experienced) during a small line break outside containment (Subsection 15.6.2) accident scenario. Also provide the basis for the assumption that the release duration is only two hours. O Amendment 2 20.3 13
- - - - . _ _ ~ - . __
ABM _ 23A6100AT Standard Plant REV.A RESPONSE 470.2 The analysis for failure of a small line carrying primary coolant was based upon considering the plant remaining at full power for a period of two hours at which time flow was stopped. For conservative purposes, the release was considered instantaneous in the actual computations. These parameters were chosen for conservatism and ease of computation. The actual case of the rupture of an instrument line is described in Chapter 8 of NEDO 211431 (Reference 2 of SSAR Subsection 15.6.7) and results in full flow for approximately ten minutes following operator action and gradual depressurization over a five-hour period. The total mass of liquid released is approximately 12,000 pounds or one half of the assumed release analysis. In addition, iodine spiking is considered on a release per fuel bundle basis. With the spiking term, which is estimated as a 15% initial release following release of the remaining 85% proportional to the depressurization,it is found that the results are similar to those analyzed in Section 15.6 but slightly less conservative. QUESTION 4703 Subsection 15.6.4.5.1.1 of the FSAR gives the iodine source term (concentration and isotopic mix) used to analyze the steam line break outside of containment accident. The noble gas source term, however, is not addressed. Provide the noble gas source term used. Also the table in Subsection 15.6.4.5.1.1 seems heavily weighted to the shorter lived activities (i.e., (I 134). Provide the bases for the isotopic mix used in your analysis (iodine and noble gas). RESPONSE 4703 Subsection 15.6.4.5.1.1 states that for case 1 the noble gas source term used was equivalent to an offgas release of 50,000 microCuries per second and 300,000 microCuries per second for case 2. In both cases, the source term is referenced to a 30 minute decay time. The isotopic distribution for such source terms are relatively standard throughout the industry and can be found in Table 2-2 of NU R EG-0016. For the iodine isotopes the concentrations are technical specification limits of 0.2 g microCurica per gram (case 1) and 4 microCuries per gram (case 2) dose equivalent to I-131. The isotopic breakdown is based upon evaluations of BWR iodine chemistry in the early 1970's and is given in Reference 2 of SSAR Subsection 15.6.7. The breakdown is as follows, and is similar to that found in Table 2 2 of NUREG-0016: 1131 0.073 I-13 2 0.71 1 133 0.5 1 134 1.4 I 135 0.73 I QUESTION 470.4 ! { Subsection 15.6.5.5 states that the analysis is based on assumptions provided in Regulatory Guide 1.3 except where noted. For all assumptions (e.g., release assumed to occur one hour after accident initiation, the chemical species fractions for iodine, the temporal decrease in primary containment leakage rates, credit for condenser leakage rates, and dose conversion factors) which deviate from l NRC guidance such as regulatory guides and ICRP2, provide a detailed description of the justification I for the deviation or a reference to another section of the SSA.R where the deviations are discussed in detail. Provide a comparison of the dose estimates using these assumptions versus those which result from using the NRC guidance. O Arnendment 2 gg
1 1 23A6100AT Standard Plant REV A RESPONSE 470.4 V The evaluation of the loss of coolant accident (LOCA) involved several assumptions which differ 3 for those outlined in Regulatory Guide 1.3 and SRP 15.6.5. Each assumption is shown in Table 20.12 with an associated explanatory paragraph below. In addition, the estimated dose for the two hour site boundary dose at 300 meters and the LPZ 30-day dose at 800 meters is given in Table 20.3-1 for each assumption. (1) 1 Hour Release Followine Scram. The ABWR incorporates a redundant emergency core cooling system 1 (ECCS) to supply makeup water in the event of a LOCA. The ECCS is sized so that in such an j event sufficient water is supplied to insure that core uncovery does not occur. Therefore, the' I assumptions as to fission product release under Regulatory Guide 1.3 for a LOCA with proper l operation of the ECCS are not justified. However, given a potential spectrum of failure of l equipment or operator error in conjunction with a LOCA, core uncovery is justified ranging on a l time scale of a few tens of minutes for total failure of all systems, to several days for gradual deterioration of equipment. Based upon evaluations of ECCS responses to a wide variety , of conditions, it is reasonable to assume that core uncovery would not proceed for a minimum of i one hour given the single failure proof design of the system. (2) Primary Containment Leakace. Following a LOCA case, Regulatory Guide 1.3 stipulated the l containtnent leakage should remain constant for 30 days. Regulatory Guide 1.4 (PWR) permits a reduction by a factor of two 24 hours after a LOCA, Containment leakage is proportional to ! containment pressure assuming that design leakage is not significantly exceeded. The analysis ! of containment pressure given in Section 6.2 and long term studies under a variety of ; conservative assumptions show that the ABWR primary containment pressure is a factor of twos ; below design pressure within 12 hours following a LOCA and decreasing slowly after that. Based' ! q upon this type of evaluation, the reduction in leakage by a factor of two 24 hours after a LOCA' V is justifiable. r (3) Iodine Release Fractions. The release of substantial quantities (>10%) of iodine from the core' of a nuclear reactor predicates significant damage to the fuel and the associated fuel assemblies. The only means by which such damage might be sustained is extensive high' temperatures leading to fuel melt. Such damage, even though partial such as at TMI, will result in core conditions resulting in the evolution of Csl rather than the 12 assumed in the regulatory guide (Reference 1). The formation of organic iodides is based upon the release of 12 and adequate concentrations with organic constituents to form organic iodides (References 2 and 3). Such conditions cannot be reasonably expected since the iodine will be bound as Cs!. Therefore, it has been assumed that the formation of iodine species will result in primarily Csl with a minor fraction as is estimated in Reference 1 of organic iodides. Two other points need to be considered. The first is production of organic forms by radiolysis in the suppression pool. Based upon Reference 4, with pH levels in the wetwell greater than 9, the evolution of iodine species is not expected. The second is consideration of accident situations leading to only minor fuel damage resulting in primarily a fuel gap inventory release. Such a release duel primarily to low temperature can be expected to consist of 12 gas and result in some organic iodide formation. However, such releases are considered under the smallline break accident. case and control rod drop accident cases. O Amendment 2 20.3 17
MM 23A6100AT Standard Plant REV.A (4) Suooression Pool Scrubbine. The ability of the suppression pool in the BWR to remove particulate material and elemental iodine has in the past been prohibited under Regulatory Guide g 1.3, due, it is .aought, to a lack of adequate understanding of the phenomena involved. The W ABWR is designed with safety / relief valves and horizontal downcomers integrated into the building to insure that any release of fission product material will be subject to transport to the wetwell via the suppression pool. Over the last several years a preponderance of both empirical and theoretical evidence has been gathered which adequately states the case for suppression pool scrubbing. This has culminated in the development of the GE DECON computer code for evaluation of suppression pool scrubbing and which, when evaluated against the empirical evidence, accurately predicts the empirical results better than any current simulation. Using the DECON code on the conditions expected in the ABWR under LOCA simulation results in overall decontamination factors far in excess of the 100 assumed in the analysis. Therefore, it was considered reasonable in light of the current knowledge to assume a conservative overall decontamination factor for the pool of 100. (5) MSIV Leakace. The evaluation of potentialleakage during a LOCA from the main steam lines has centered on the leakage of the MSIVs and potential for direct release to the environment. This has in the past resulted in the issuance of Regulatory Impact issue C 8 by the NRC and the use of main steam leakage control systems. Over the past several years, considerable effort has been expended on this subject by the BWR Owners Group and the NRC and has resulted in a series of reports and maintenance procedures for utilities. The ABWR technical specification of MSIV performance recognized potential seating and leakage problems and therefore uses a graduated leakage performance criteria shown in Figure 15.6-2. The evaluation of radionuclide leakage from these valves were then made in accordance with the procedure given in Reference 5, except as noted below. (a) The primary containment served as a single large repository for fission products from which leakage was derived for the pathways via the reactor building and the MSIVs. Material directly injected from the pressure vessel to the drywell were assumed over a short period g of time to cycle into the wetwell. Following pressure suppression, the wetwell airspace would then be considered linked to the dr>well airspace via the vacuum breakers. Therefore, a multiple flow path in primary containmect was not evaluated. (b) Flow through each steamline was considered independently at that line flow rate (see Figure 15.6-2). The transport time down each line was considered at a rate three times the plug flow rate specified in Reference 5. This value is a rule of thumb derived from experience in flow through large pipes, and when compared to the results of Reference 6,is similar. (c) Plateout in the steamlines was not considered for the first 48 hours after the LOCA to allow for line cooling. This was adopted as conservative, based upon the arguments found in Reference 6 (page F-3). The plateout modelis found in Reference 5. (d) The primary controlling factor for MSIV leakage is condenser plateout and leakage. For the condenser, a single mixed volume equal to one half of the free air volume in the condenser was assumed. Leakage from the condenser assumed that the totalin leakage from the steam lines was non condensible plus an additional leakage of 100 ft3 per hour based upon barometric pressure changes from Reference 6 (page F 3). Such a leakage is considered conservative since the leakage to the condenser would primarily te condensible and the barometric pressure change required to cause a 100 ft 3 er p hour change could be extremely large (a hurricane). The condenser plateout model used was that found in Reference 5. (e) Following release from the condenser, the meteorology and the health effects model were those assumed in Regulatory Guide 1.3 and are described in SSAR References 2 through 4. Amendment 2 20.3 13
23A6100AT Standard Plant REV.A l c References for Response 470.4 ! () ~
- 1. Technical Basis for Estimating Fission Product Behavior During LWR Accidents, NUREG-0772, June.
i i 1981.
- 2. Postma, A.K. and Zavadoski, R.W., Review of Organic lodide Formation Under Accident Conditions in Water-Cooled Reactors, WASH 1233, October 1972.
- 3. Malinauskas, A.P. and Bell, J.T., "The Chemistry of Fission Product Iodine Under Nuclear Reactor Accident Conditions,* Nuclear Safety, Vol. 28, No. 4, Oct Dec 1987, ,
- 4. Lin, C.C.,
- Chemical Effects of Gamma Radiation on Iodine in Aqueous Solutions, fournal of inorganic and Nuclear Chemistry, Vol. 42 (1980) 1101-1110.
- 5. Careway, H.A. et al, A Technique for Evaluation of BWR MSIV Leakage Contribution to Radiological Dose Rate Calculations, NEDO-30259, GE, Sept.1985.
- 6. Ridgely, J.N. and Wohl, M.L., Resolutions of Generic Issue C-8, NUREG-1169, Aug.1986.
QUESTION 470.5 Provide a discussion of, or reference to, the analysis of the radiological consequences of leakage, from engineered safety feature components after a design basis LOCA. < RESPONSE 470.5 A Leakage from engineered safety features are not specifically analyzed. The totalleakage from the,, V primary containment is restricted to 0.5% per day for allleakage except that through the main steam; line isolation valves. Leakage from engineered safety features is then included in the 0.5% per day , such that all leakage from equipment external to the primary containment shall not result in an. airborne release which when combined with the containment leakage shall result in an equivalent release greater than 0.5% per day. QUESTION 470.6 For the spent fuel cask drop accident, what is the assumed period for decay from the stated power condition? Wht is the justification for that assumption? RESPONSE 470.6 Table 15.7-12 has been corrected by changing ' core
- to
- storage
- under item I.E. The cask drop l accident assumes a 1000 day exposure prior to removal from the core with a radial peaking factor of 1.5. Decay time upon removal from the core is 120 days prior to the accident. This 120 day period.
was conservatively estimated at one third of a year, since based upon current practice the minimum ! time to ship fuel to a long term storage facility is one year (in the case of the GE Morris facility) and 10 years in the case of government storage facilities. QUESTION 470.7 The tables in Chapter 15 should be checked and revised as appropriate. In several cases the footnotes contain typographical errors related to defining the scientific notation. Table 15.712, also appears to contain inappropriate references to Table 15.7-16, rather than Table 15.713. Amendment 2 20.3 19
Mkb 23A6100AT i Standard Plant REV.A l l RESPONSE 470.7 ' The response to this question is provided in revised Tables 15.7-10,15.712, and 15.7-13. Oi , 1 QUESTION 470.8 l It is stated that Regulatory Guides 1.3 and 1.145 were used in the calculations of X/O values. Based on the values presented, it appears as though a Pasquill stability Class F and one meter per I second wind speed were assumed, with adjustment for meander per Figure 3 of Regulatory Guide 1.145. If this is not the case, describe the assumptions and justification used in calculating the X/O values which are used in the Chapter 15 dose assessments. RESPONSE 470.8 1 All meteorological calculations in Chapter 15 were made based upon the equations given in Regulatory Guide 1.145 and the tables in Regulatory Guide 1.3.C.2.g(3). Calculations have been j encoded into a computer program for routine use and are detailed in report NEDO-20804, Atmospheric i Dispersion CHIQUO2 Function, Feb.1979. In all cases, a ground level release was assumed, and where l permitted by regulatory guide or SRP, building wake and plume meander accounted for. In addition, j the basis for meteorological calculations is found in Appendix B of Reference 2, SSDR Subsection 15.6.7. I QUESTION 470.9 l The SGTS filter efficiencies of 99% for inorganic and organic iodine are higher than the 90% and l 70% values, respectively, assumed in Regulatory Guide 1.25 if it can be shown that the building atmosphere is exhausted through adsorbers designed to remove iodine. Provide a justification for the a use of the higher values. W RESPONSE 470.9 The ABWR incorporates a 6 inch charcoal bed in the SGTS filter train, and in accordance with Table I 2 of Regulatory Guide 1.52 is permitted a removal efficiency for both elemental and organic forms of ; iodine of 99%. QUESTION 470.10 Dose related factors such as breathing rates, iodine conversion factors and finite vs. infinite cloud assumptions for calculating the whole body dose are not stated explicitly, although reference i is made to Regulatory Guide 1.25 and another document. State these assumptions explicitly and I justify the use of any values which desiate from Regulatory Guide 1.25. l RESPONSE 470.10 In all cases except the control room evaluation, a semi infinite cloud model was used to calculate dose conversion factors. This model was based upon Regulatory Guide 1.3 and Slade's Mercorolog and Atomic Energy - 1968. A detailed explanation of the model with related factors is found in Appendix C in Reference 2 of SSAR Subsection 15.6.7. In the case of the control room dose, the dose model was a finite cloud model to account for the limited size of the control room and is given in Section 2.5 of Reference 3 of SSAR Subsection 15.6.7. O Amendment 2 20.3-20
ABM 23A6100AT j Standard Flant REV.A l TABLE 20.31 U, SENSITIVITY STUDY OF PARAMETERS FOR LOCA ANALYSIS Site Boundary 24 Hr. LPZ Dose for 30 Days ! Dose at 300 m (REAf) at 800 m (REht) ' l Th n oid Whole Body Thyroid % hole Body 4
- 1. LOCA Results 1.5 0.62 22. 12,
- 2. No Initial 1 Hr. Hold up 1.5 0.90 22 13
- 3. No Pressure Reduction NC NC 22 13
@ 24 Hrs
- 4. lodine Species Consistent 10.0 0.64 1700 13 with Regulatory Guide 13
- 5. No Suppression Pool 140. 0.92 930 13 Scrubbing
- 6. No Steamline Platcout 1.5 0.62 23 12 No Steamline Plateout p)
( 7. or Hold up 1.5 0.64 23 12 --
- 8. No Condenser Plateout 23 0.62 340 12
- 9. No Condenser Plateout 280 41 1300 70 t or Hold up NOTE:
All evaluations are made independently of each other. l l l O Amendment 2 20.3 21
ABM 2346imio Standard Plant REV.D 4.5 REACTOR MATERIALS Guide Roller Stellite No.3 q l V 4.5.1 Control Rod Drive System Structural Guide Roller Pin Haynes Alloy No. 25 Materials Guide Shaft Stellite No. 6 4.5.1.1 Material Specifications Guide Shaft Bushing Stellite No.12 (1) Material List Separation Spring loconel X-750 The following material listing applies to the control rod drive mechanism supplied for this Separation Magnet Alnico No. 5 application. The position indicator and minor
, non-structuralitems are omitted. (c) Buffer Mechanism The properties of the materials selected for Buffer Spring Inconel X 750 the control rod drive mechanism shall be equivalent to those given in Appendix I to Buffer Sleeve 316L (Hardsurfaced Section 111 of the ASME Code or parts A and B of with Colmonoy No. 6) y Section 11 of the ASME Code or are included in o Regulatory Guide 1.85 except that cold worked Guide Roller Stellite No. 3 austenitic stainless steels shall have a 0.2%
offset yield strength no greater than 90,000 psi. Guide Roller Pin Stellite No. 25 (a) Spool Piece Assembly Buffer Cone 316L (Hardsurfaced with Stellite No. 6) Spool Piece Housing ASME 182 Grade P34tL O Se > aensin8 ^ Sue 182 o< de r3a L Piston Tube XM 19 Drive Shaft ASME 479 Grade XM 19 (Hardsurfaced with Piston Head 316L (Hardsurfaced Colmonoy No. 6) with Stellite No. 3) Ball Bearings 440C Latch Inconel X-750 Gland Packing Asbestos Latch Spring Inconel X 750 Gland Packing Spring Inconel X-750 Coupling Spud Inconel X 750 (b) BallSpindle (e) Guide Tube Ball Screw Shaft ASTM A 564 TP630 ; (17-4) Guide Tube 316L Condition H 1100 (f) Outer Tube Assembly Ball Nut ASTM A 564 TP630 (17-4) Outer Tube XM 19 Condition H 1100 Flange ASME SA182 Grade Balls 440C F304L l i O l Amendment 2 l l
ABWR mmo Standard Plant REV. D (g) hiiscellaneous Parts jects selected 300 Series stainless steel compo-nents to temperatures in the sensitization g Ball for Check Valve Haynes Alloy range. The drive shaft, buffer sleeve, piston W bead and buffer are hard surfaced with Colmonoy 0 Ring Seal (Between 321SS Coated with 6 (or its equivalent). Colmonoy (or its equiva-CRD Housing and Teflon lent) hard surfaced components have performed CRD) successfully for the past 15 to 20 years in drive mechanisms. It is normal practice to CRD Installation AShfE SA193 remove some CRDs at each refueling outage. At Bolts Grade B7 this time, the Colmonoy (or its equivalent) hard-surfaced parts are accessible for visual (2) Special hiaterials examination. This inspection program is ade-quate to detect any incipient defects before The coupling spud, latch and latch spring, they could become seri ous enough to cause separation spring and gland packing spring are operating problems. The degree of conformance fabricated from Alloy X 750 in the annealed or to Regulatory Guide 1.44 is presented in equalized condition, and aged 20 hours at Subsection 4.5.2.4. 13000F to produce a tensile of 165,000 psi minimum, yield of 105,000 psi minimum, and (2) Control of Delta Ferrite Content elongation of 20% minimum. The ball screw shaft and ball nut are ASThi A-564, TP 630 (17-4) (or Discussion of this subject and the degree of its equivalent) in condition H-1100 (aged 4 hours conformance to Regulatory Guide 1.31 is at 11000F), with a tensile of 140,000 psi presented in Sul.sectioa 4.5.2.4. minimum, yield of 115,000 psi minimum, and elongation of 15% minimum. 4.5.1.3 Other Materials These are widely used materials, whose proper- These are pretented in Subsection 4.5.1.1(2) ties are well known. The parts are readily accessible for inspection and replaceable if 4.5.1.4 Cleaning and Cleanliness Control g necessary. All the CRD pris listed in Subsection All materials for use in this system shall be 4.5.1.1 are fabricated under a process y selected for their compatibility with the reactor specification which limits contaminants in g coolant as described in Articles NB 2160 and NB- cutting, grinding and tapping coolants and 3120 0f the ASME Code. lubricants. It also restricts all other processing materials (marking inks, tape etc.) All materials, except SA479 or SA249 Grade to those which are completely removable by the Xht 19, have been successfully used for the past applied c':aning process. All contaminants are
-15 to 20 years in similar drive mechanisms, then required to be removed by the appropriate Extensive laboratory tests have demonstrated that cleaning process prior to any of the following:
AShfE SA479 or SA249 Grade Xht 19 are suitable materials and that they are resistant to stress (1) Any processing which increases part corrosion in a BWR ensironment. temperature above 2000F, No cold worked austenitic stainless steels (2) Assembly which results in decrease of with a yield strength greater than 90,000 psi are accessiblity for cleaning. employed in the Control Rod Drive (CRD) system. (3) Release of parts for shipment. 4.5.1.2 Austenitic Stainless Steel Components The specification for packaging and shipping (1) Processes, Inspections and Tests the Convol Rod Drive provides the following: There is a special process employed which sub- The drive is rinsed in hot deionized water Amendment 2 43-2
ABM 23461oorn Standard Plant Rev n and dried in preparation for shipment. The ends or SA 358, SA-312, or SA 249 (Type 304L or 316L) h of the drive are then covered with a vapor tight ot ASME SA 351 Type CF3M (Type 316L). barrier with dessicant. Packaging is designed to protect the drive and prevent damage to the vapor Orificed fuel support. ASME SA 351 Type CF3 barrier. Audits have indicated satisfactory (Type 304L) or CF3M (Type 316L). protection. Materials employed in shroud head and Semiannual examination of the humidity separator assembly and steam dryer assembly: indicators of ten percent of the units is required to verify that the units are dry and in All materials are 304L or 316L stainless satisfactory condition. This inspection shall be steel: performed with a GE-Engineering designated representative present. The position indicator Plate, Sheet--ASTM A240 Type 34tL or 316L probes are not subject to this inspection. and Strip Site or warehouse storage specifications Forgings--ASTM A182 Grade 304L require inside heated storage comparable to level B of ANSI N45.2.2. Bars--ASTM A276 Type 316L The degree of surface cleanliness obtained by Pipe ASTM A312 Grade TP-3G4L these procedures meets the requirements of Regulatory Guide 137. Tube--ASTM A269 Grade TP-304L 4.5.2 Reactor Internal Materials Castings ASTM A351 Grade CF8 4.5.2.1 Material Specifications All core support structures are fabricated . from ASME specified materials, and designed Materials used for the Core Support Structure: accordance with requirements of ASME Code,; Section III, Subsection NG. The other reactor Shroud Support Nickel Chrome Iron Alloy, internals are noncoded, and they are fabricated ASME SB166 or SB168. from ASTM or ASME specification materials. , Shroud, core plate, and grid ASME SA240, 4.5.2.2 Controls on Welding SA182, SA479, SA312, SA249, or SA213 (all Type 3G4L or 316L). Core support structures are fabricated in ac-cordance with requirements of ASME Code Section Peripheral fuel supports - ASME SA312 Grade Ill, Subsection NG 4000 and the examination and 2 Type-344L or 316L acceptance criteria shown in NG 5000. Other $ , internals are not required to meet ASME Code re- ! Core plate and top guide studs, nuts, and quirements. ASME Section IX BPV code require-sleeves. ASME SA 479 (Type 304,316, or XM 19) ments are followed in fabrication of core. (all parts); or SA 193 Grade BS (studs); or SA- support structures, j 194 Grade 8 (Type 304) (nuts); or SA 479 (Type 304L or 316L), SA 182 (Grade F304L or F316L), 4.5.2.3 Nondestructive Examination of Wrought , SA 213 (Type 304L,316 or 316L), SA 249 (Type Seamless Tubular Products j 34tL,316, or 316L) (sleeves). i Wrought seamless tubular products for CRD l Control rod drive housing. ASME SA 312 Grade housings, and peripheral fuel supports, are sup- l TP304L or 316L SA 182 Grade F304L or F316L, and plied in accordance with ASME Section III, Class ASME SA 351 Type CF3 (Type 304L) or Type CF3M CS, which requires examination of the tubular I (Type 316L). products by radiographic and/or ultrasonic me-thods according to paragraph NG 2550, the exami- Z Control rod guide tube. ASME SA 351 Type CF3, nation will satisfy the requirements of NG 5000. @ Amendment 2 U3
23A6100AD Standard Plant RP/ B Wrought seamless tubular products for other materials which contract stainless steel during internals were supplied in accordance with the manufacture and construction. Any inadvertent g applicable ASTM or ASME material specifica. surface contamination is removed to avoid W tions. These specifications require a potential detrimental effects. hydrostatic test on each length of tubing. Special care is exercised to insure removal 4.5.2.4 Fabrication and Processing of of surface contaminants prior to any heating Austenitic Stainless Steel Regulatory Guide operation. Water quality for rinsing, flushing, Conformance and testing is controlled and monitored. Cold worked stainless steels are not used in The degree of cleanliness obtained by these the reactor internals except for vanes in the procedures meets the requirements of Regulatory g steem dryers. Furnance sensitized material shall Guide l.37. a not be allowed. The detta ferrite content for 0 weld materials used in welding austenitic stain- 4.5.2.5 Other Materials l less steel assemblies is verified on undiluted 1 weld deposits for each heat or lot of filler Hardenable martensitic stainless steel and l metal and electrodes. The delta ferrite content precipitation hardening stainless steels are not ) is defined for weld materials as 5.0 Ferrite used in the reactor internals. Number (FN) minimum and 8.0 FN average. This ferrite content is considered adequate to prevent Materials, other than Type 300 stainless any micro fissuring (Hot Cracking) in austenitic steel, employed in reactor internals are: l l stainless steel welds. This procedure complies with the requirements of Regulatory Guide 131. (1) SA479 Type XM 19 stainless steel; Proper solution anonealing of the 300 series (2) SB166,167, and 168, Nickel Chrome tron ; austenitic stainless steel is verified by testing (Alloy 600); and per ASTM.A262, "Recommended Practices for Detecting Susceptibility to Intergranular Attack (3) SA637 Grade 688 Alloy X-750. hlj j in Stainless Steels." Welding of austenitic stainless steel parts is performed in accordance Alloy 600 tubing, plate, and sheet are used with Section IX (Welding and Brazing in the annealed condition. Bar may be in the Qualification) and Section 11 Part C (Welding Rod annealed or cold drawn condition. , Electrode and Filler Metals) of the ASME Beiler ! and Pressure Vessel Code. Welded austenitic Alloy X-750 components are fabricated in the stainless steel assemblies require solution annealed or equalized condition and aged when annealing to minimize the possibility of the required. sensitizing. However, welded assemblies are dispensed frorn this requirement when there is Stellite 6 (or its equivalent) bard surfacing documentation that welds are not subject to is applied to some austenitic stainless steel significant sustained loads and assemblies have castings using the gcs tungsten arc welding or been free of service failure. Other reasons,in plasma are surfacing processes, line with Regulatory Guide 1.44, for dispensing with the solution annealing are that assemblies All materials used for reactor internals shall are exposed to reactor coolant during normal be selected for their compatibility with the operation service which is below 2000F reactor coolant as shown in ASME Code Section ,e temperature or assemblies are of material of low Ill, NG 2160 and NG-3120. The fabrication and N carbon content (less that 0.025%). These cleaning controls will preclude contamination of 0 controls are employed in order to comply with the nickel base alloys by chloride ions, fluoride intent of the Regulatory Guide 1.44 ions or lead. Exposure to contaminant is avoided by All materials, except SA479 Grade XM 19, have carefully controlling all cleaning and processing been successfully used for the past 15 to 20 Amendment 2 4.H
MM 23A6100AB Standard Plant REV.B years in BWR applications. Extensive laboratory Q tests have demonstrated that XM 19 is a suitable material and that it is resistant to stress corrosion in a BWR emironment. l I L l
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23A6100AB Standard Plant nry n /3 The main steamline flow restrictors of the a portion of reactor coolant through a filter-V venturi type are installed in each main steam demineralizer to remove particulate and dis-nozzle on the reactor vessel inside the primary solved impurities with their associated corro-containment. The restrictors are designed to sion and fission products from the reactor cool- A limit the loss of coolant resulting from a main ant. It also removes excess coolant from the steamline break inside or outside the primary reactor system under controlled conditions, containment. The coolant loss is limited so that reactor vessel water level remains above the top 5.1.1 Schematic Flow Diagrams of the core during the time required for the main steamline isolation valves to close. This action Schematic flow diagrams (Figures 5.11 and protects the fuel barrier. 5.1-2) of the RCS show major components, principal pressures, temperatures, flow rates, Two isolation valves are installed on each and coolant volumes for normal steady state main steamline. One is located inside, and the operating conditions at rated power. other is located outside the primary containment. If a main steamline break occurs 5.1.2 Piping and Instrumentation Diagrams inside the containment, closure of the isolation valve outside the primary containment seals the Piping and instrumentation diagrams covering primary containment itself. The main steamline the systems included within RCS and connected isolation valves automatically isolate the RCPB systems are presented as follows: when a pipe break occurs outside containment. This action limits the loss of coolant and the (1) the nuclear boiler system (Figure 5.13); release of radioactive materials from the nuclear system. (2) main steam (Figure 5.13a &b); The RCIC system provides makeup water to the (3) feedwater (Figure 5.13c); , core during a reactor shutdown in which feedwater flow is not available. The system is started (4) recirculation system (Figure 5.4-4); automatically upon receipt of a low reactor water level signal or manually by the operator. Water (5) reactor co:e isolation cooling system is pumped to the core by a turbine pump driven by (Figure 5.4-8); reactor steam. (6) residual heat removal system (Figure The residual heat removal (RHR) system 5.410); and includes a number of pumps and heat exchangers that can be used to cool the nuclear system under (7) reactor water cleanup system (Figure 5.4-12) a variety of situations. During normal shutdown and reactor servicing, the RHR system removes 5.1.3 Elevation Drawings residual and decay heat. The RHR system allows decay heat to be removed whenever the main heat An elevation drawing (Figure 5.14) shows sink (main condenser) is not available (i.e., hot the principal dimensions of the reactor and standby). One mode of RHR operation allows the connecting systems in relation to the removal of heat from the primary containment containment. following a LOCA. Another operational mode of the RHR system is low pressure flooder (LPFL). The LPFL is an engineered safety feature for use during a postulated LOCA. Operation of the LPFL is presented in Section 6.3. The reactor water cleanup system recirculates O Amendment 2 5.12
ABM 2mine Standard Plant REV.B-secTroN s.2 O CONTENTS (Continued) Section Ill].g Egge ., 5.23.2.23 Sources ofImpurities 5.2-9 i 5.23.2.2.4 IGSCC Considerations 5.2 12 281.2 5.23.23 Compatibility of Construction hiaterials with Reactor Coolant 5.2 12 5.23.2.4 Compatibility of Construction , hiaterials with External Insulation 5.2-12 5.233 Fabrication and Processing of Ferritic hiaterials 5.2-12 5.233.1 Fracture Toughness 5.2-12 5.23 3.2 Control of Welding 5.2 13 l 5.233.2.1 Regulatory Guide 1.50: Control ; of Preheat Temperature Employed ; for Welding of 1.ow-Alloy Steel 5.2 13 5.233.2.2 Regulatory Guide 1.M: Control E ectroslag Weld Properties 5.2-13 i 5.23 3.23 Regulatory Guide 1.71: Welder Qualification for Areas of Limited Accusibility 5.2 13 , 5.2333 Regulatory Guide 1.66: Nondestrvtive Examination of Tubular Produc* 5.2-13 , 5.233.4 hioisture Control for Low Hydrogen, Covered Arc Welding Electrodes 5.2 14 5.23.4 Fabrication and Processing of Austenitic Stainless Steels 5.2-14 i 5.23.4.1 Avoidance of Stress / Corrosion Cracking 5.2 14 i 5.23.4.1.1 Avoidance of Significant Sensitization 5.2 14 5.2iv l Amcodment 2 f
l 23A6100AD Standard Plant anv. s j SECTION 5.2 , CONTENTS (Continued) Section Et!g Page i l 5.23.4.1.2 Process Controls to Minimize l Exposure to Contaminants 5.2-14 5.23.4.13 Cold Worked Austenitic Stainless Steels 5.2-15 1 5.23.4.2 Control of Welding 5.2 15 j i 5.23.4.2.1 Avoidance of Hot Cracking 5.2-15 5.23.4.2.2 Regulatory Guide 134: Electroslag Welds 5.2 15 5.23.4.23 Regulatory Guide 1.71: Welder Qualification for Areas of I Limited Accessibility 5.2 15 ! l 5.23.43 Regulatory Guide 1.66: Non destructive ' Examination of Tubular Products 5.215a 5.2.4 In-Senice Insocetion and Testinc of Reactor Coolant Pressure Boundarv 5.2-15a 5.2.4.1 System Boundary Subject to ; Inspection 5.2-15a ; I 5.2.4.2 Provisions for Access to the Reactor Coolant Pressure l Boundary 5.215a 5.2.4.2.1 Design and Arrangement of Reactor Coolant Boundary Components 5.2-15a 5.2.4.2.2 Reactor Pressure Vessel 5.2-15a 5.2.4.23 Pipe, Pumps, and Valves 5.2 16 5.2.43 Examination Tech tiques and Procedures 5.2 16 5.2.43.1 Equipment for Insenice Inspection 5.2 16 5.2.43.2 Visual Examination 5.2 16 9' 5.2 v
ABM ux61oors Standard Plant REV.A , SECTION 5.2 O CONTENTS (continued) Section Title East 5.2.433 Magnetic Particle and Liquid Penetrant Examination 5.2 17 l I 5143.4 Volumetrie Ultrasonic Direct Examination 5.2 17 l 5.2.43.5 Recording and Comparing Data 5.217 s i 5.2.4.4 Inspection Intervals 5.2 18 514.5 Inservice Inspection Program and Categories and Requirements 5.2-18 514.6 Evaluation of Examination Results 5.2-18 514.7 System Pressure Tests 5.2-18 5.2.5 Etartor Coolant Pressute Boundary and Core Coollne Systems Irakane Detection 5.2-18 515.1 1.cakage Detection Methods 5.2 18 O. 5.25.11 Detectien ef teais8e witsin Drywell 5.2 18 5.2.5.1.2 Detection of Leakage External to Drywell 5.2 19 515.2 Leak Detection Instrumentation and Monitoring 5.2 20 5.2.5.2.1 Leak Detection Instrumentation and Monitoring Inside the Drywell 5.2 21 51512 Leak Duection Instrumentation and Monitoring External to Drywell 5.223 5.2.5.23 Summary 5.225 5153 Indication in the Control Room 5.2-26 515.4 Limits for Reactor Coolant Leakage 5.226 , l O l i 5.2vi i
ABM - ux61ooin Standard Plant Rev. n SECTION 5.2 g CONTENTS (Continued) Section Illig Page 5.2.5.4.1 Total Leakage Rate 5.2-26 5.2.5.4.2 Identified Leakage Inside Drywell 5.2 26 5.2.5.5 Unidentified leakage Inside Drywell 5.2 26 5.2.5.5.1 Unidentified Leakage Rate 5.2 26 5.2.5.5.2 Margins of S1fety 5.2-27 5.2.5.5.3 Criteria to Evaluate the Adequacy and Margin of the leak Detection System 5.2 27 5.2.5.4 Differentiation between Identified and Unidentified Leaks 5.2 27 5.2.5.5 Sensitisity and Operability Tests 5.2-27 g 5.2.5.6 Testing and Calibration 5.2-27 5.2.5.7 Regulatory Guide 1.45: Compliance 5.2-27 5.2.6 Interfaces 5.2 28 l 281.7 2814 5.2.7 References 5.2-28 :s1.10 item 4 & 11 i i O 5.2-sii Amendment 2 I
ABM 23A6100AB Standard Plant REV.B TABLES O rasi. ra1, eES, 5.2-1 Applicable Codes Cases 5.2-29 l 210.1 l 5.22 Systems Which May Initiate During Overpressure Event - 5.2 30 5.2 3 Nuclear System Safety / Relief Valve Setpoints 5.2 31 5.2-4 Reactor Coolant Pressure Boundary Materials 5.2-32 5.25 BWR Water Chemistry 5.2.34 5.2-6 Summary ofIsolation/ Alarm of System Monitored and Leak Detection Methods Used (Summary of Isolation Signals and Alarms System Isolated Versus Variable Monitored) 5.2 35 5.2-7 Summary of Isolation / Alarm of System Monitored and Leak Detection Methods Used (Summary of Variable Trip Alarms- leakage Source Versus Generated Variables) 5.2 36 ILLUSTRATIONS Figure Ihlt East O 5.21 Safety Action Valve Lift Characteristics 5.2-37 5.2-2 MSIV Closure with Flux Scram and Installed Safety / Relief Valve Capacity 5.2 38 5.23 Safety / Relief Valve Schematic Elevation 5.2 39 , 3 5.2-4 Safety / Relief Valve and Steamline Schematic 5.2-40 5.2-5 Nuclear Boiler System - P&lD Data 5.2-41 i 5.2-6 Schematic of Safety / Relief Valve with j Auxiliary Activating Desice 5.2-42 5.2-7 Examples of the Minimum Length of Spool Pieces 5.2-43 5.28 Leak Detection System Instrument Electrical Diagram (IED) 5.2 44 i I O 5.24ili Amendment 2
23A6100AD Standard Plant RIN.B 5.2 INTEGRITY OF REACTOR COOLANT formance with 10CFR50, Appendix A, General De-
- ('j PRESSURE BOUNDARY sign Criterion 15. Preoperational and startup instructions are given in Chapter 14.
This section discusses measures employed to provide and maintain the integrity of the reactor 5.2.2.1.1 Safety Design Bases coolant pressure boundary (RCPB) for the plant design lifetime. The nuclear pressure relief system has been designed to: 5.2.1 Compliance with Codes and Code Cases (1) prevent overpressurization of the nuclear system that could lead to the failure of the 5.2.1.1 Compliance with 10CFR50, Section 50.55a RCPB; Table 3.2-4 shows the compliance with the (2) provide automatic depressurization for small rules of 10CFR50, Codes and Standards. Code edi- breaks in the nucicar system occurring with tion, applicable addenda, and component dates maloperation of both the reactor core will be in accordance with 10CFR50.55a. isolation cooling (RCIC) system and the high pressure core spray (HPCS) system so that 5.2.1.2 Applicable Code Cases the low pressure flooder (LPFL) mode of the residual heat removal (RHR) system can The reactor pressure vessel and appurtenances operate to protect the fuel barrier; and the RCPB piping, pumps, and valves will be designed, fabricated, and tested in accordance (3) permit verification ofits operability, and with the applicable edition of the ASME Code, in-ciuding addenda that were mandatory at the order (4) withstand adverse combinations of loadings date for the applicable components. Section and forces resulting from normal, upset,
,e ; 50.55a of 10CFR50 requires code case approval for emergency, or faulted conditions.
5 A Class 1. 2, and 3 components. These code cases contain requirements or special rules which mas 5.2.2.1.2 Power Generution Design Bases be used for the construction of pressure retain. ing components of Quality Group Classification A, The nuclear pressure relief system safety /
; B, and C. The various ASME code cases that may relief valves (SRVs) have been designed to meet A be applied to components are listed in Table the following power generation bases:
5.21. (1) discharge to the containment suppression Regulatory Guides 1.84 and 1.85 provide a list p 1 4and of ASME Design and Fabrication code cases that have been generically approved by the Regulatory (2) correctly reclose following operation so Staff. Code Cases on this list may, for design that maximum operational continuity is purposes, be used until appropriately annulled. obtained. Annulled cases are considered active for equip'nent that has been contractually committed 5.2.2.13 Discussion to fabrication prior to the annulment. The ASME Boiler and Pressure Vessel Code 2l 4 5.2.2 Overpressure Protection requires that each vessel designed to meet Settion 111 be protected from overpressure under This sub;ection evaluates systems that' protect upcet conditions as discussed in Subsection the RCPB from overpressurization. S.2.3 of Reference 1. 5.2J.1 Design Basis Overpressure protection is provided in con. Arnendtnent 2 5.21
ABM 2 moors Standard Plant REV.A The SRV setpoints are listed in Table 5.2 3 dump valve of the turbine control valve and satisfy the AShiE code specifications for hydraulie actuation system. The position safety valves because all valves open at less switches are actuated when the respective valves g than the nuclear system design p. essure of 1250 are closing and following 15% travel of full psig. stroke. The pressure switches are actuated when a fast closure of the turbine isntrol valves is The automatie depressurization capability of initiated. Credit is not taken for the the nuclear system pressure relief system is power operated mode. Credit is only taken for evaluated in Section 6.3 and Section 7.3. the safety / relief valve capacity which opens by the spring mode of operation direct from inlet The following criteria are used in selection pressure. of SRVs: The rated capacity of the pressure-relieving (1) must meet requirements of AShiE Code. Section devices shall be r.ufficieat to prevent a rise in III; pressure within the protected vessel of more than 110% of the design pressure (1.10 x 1250 (2) must qualify for 100% of nameplate capacity psig = 1375 psig) for events defined in Section credit for the overpressure protection 15.2. function; and Full account is taken of the pressure drop on (3) must meet other performance requirements both the inlet and discharge sides of the such as response time, etc., as necessary to valves. All combination safety / relief valves provide relief functions, discharge into the suppression pool through a discharge pipe from each valve which is designed The SRV discharge piping is designed, to achieve sonic flow conditions through the installed, and tested in accordance with AShf E valve, thus providing flow independence to Code, Section 111. discharge piping losses. 5.2.2.1.4 Safety / Relief Valve Capacity Table 5.2 2 lists the systems which could initiate during the design basis overpressure SRV capacity of this plant is adequate to event. limit the primary system pressure, including transients, to the requirements of AShfE Boiler 5.2.2.2 Design Evaluation and Pressure Vessel Code Section III, Nuclear Power Plant Components, up to and including 5.2.2.2.1 blethod of Analysis applicable addenda. The essential AShfE require-ments which are met by this analysis follow. See Appendix A, Subsection A.5.2.2.2.1 of Reference 1. It is recognized that the protection of vessels in a nuclear power plant is dependent 5.2.2.2.2 Spterr. Design upon many protective systems to relieve or terminate pressure transients. Installation of A paramctric study was conducted to i pressure. relieving devices may not independently determine the required steam flow capacity of proside complete protection. The safety valve the SRVs based on the following assumptions. sizing evaluation gives credit for operation of the scram protective system which roay be tripped 5.2.2.2.2.1 Operating Conditions by either one of two sources: s direct or a flux trip signal. The direet seram trip signal is (1) operating power = 4005 hfWt (102% of nuclear derived from position switches mounted on the boiler rated power); main steamline isolation valves, the turbine stop valves, or from pressure switches mounted on the (2) vessel dome pressure .sloto psig; and O 312 l l
ABWR - ua6ioorn Standard Plant RF'V D material adjacent to welds in Type 304 and Type manipulate the corrosion potentialin laborato9 316 stainless steel piping qstems has occurred tests) (Reference 10). in the past. Substantial research and develop- [ I h ment programs have been undertaken to understand As the corrosion potential is reduced oclow Q the IGSCC phenomenec <nd develop remedial mea- the range typical of normal BWR power optration A sures. For the ABWR, JiSCC resistance has been (+ 50 to 50 mVSHE), a region of immurJty to l achieved through the use of IGSCC resistant mate- IGSCC appears at 230 mVSHE. It is apparent Q rials such as Type 316 Nuclear Grade stainless that a combination of corrosic.: potential (whid M steel and stabilized nickel base Alloy 600hi and can be achieved in a BWR by injecting usually < ', 182hi. 1 ppm hydrogen into the feedwater) plus tight 'y conductivity control (0.2 pS/cm) sLould permit hiuch of the early remedy development work fo. BWRs to operate in a regime where sensitized cused on alternative materials or local stress stainless st:els arc immune to lGSCC. ' reduction, but recently the effects of water che-mistry parameters on the IGSCC process have rece. Since the ABWR has no sensitized stainless ived increasing attention. hiany important fea- steel, IGSCC control by hydrogen inje i tures of the relationship between BWR water chem- not required. However, irradiation ;; istry and IGSCC of sensitized .tainless steels stress corrosion cracking (IASCC) can o ar in g have been identified. highly irradiated annealed stainless steel and A
, nickel base alloys. Preliminary inaeactor and
$l Laboratory studies (References 3 and 4) have laboratory studies (Reference 11) have indicated shown that although IGSCC can occur in simulated that HWC will be useful in mitigating lASCC.
BWR startup endronments, most IGSCC damage pro-bably occurs during power operation. The normal In. reactor and laboratory evidente also indi-BWR environment during power operation is -280 cates that carbon and low alloy steels ala teod OC water containing dissolved oxygen, hydro- to show improved resinance to environcentally gen and small concentrations of ionic and non- assisted cracking with both increasing water pu-
, ionic impurities (conductivity generally below rity and decreasing corrosion potentia!
d 0.3 pS/cm at 250C). It has been well docu. (Reference 12). mented that some ionic impurities (notably sul- } fate and chloride) aggravate IGSCC, and a number 52.3.2.2.1 Fuel Performance Considerations of studies have been made of the effects of indi-vidual impurity species on IGSCC initiation and Nuclear fuel is contained in Zirca;oy tubes D growth rates (References 3 thru 7). This work Nl that constitute the first boundary or primary clearly shows that IGSCC can occur in water at containment for the highly radioactive species 2800C with 200 ppb of dissolved oxygen, even at generated by the fission process; therefore, the low conductivity (Iow impurity levels), but the integri:y of the tubes must be ensured. Zirca-rate of cracking decreases with decreasing impu- loy interacts with the coolant water and some rity content. Although BWR water chemistry coolant impurities. This results in oxidation guidelines for reactor water cannot prevent by the water, increased hydrogen content in the IGSCC, maintaining the lowest practically achiev- Zircatoy (hydriding), and, often, buildup of a able impurity levels will minimize its rate of layer of crud on the outside of the tube. Ex-N progression (References 5 and 9). cessive oxidation, hydriding, or crud deposition A may lead to a breach of the cladding wall. Stress corrosion cracking of ductile materials in aqueous environments often is restricteu to hietallic impurities can result in neutron specific ranges of corrosion potential *, so a losses and associated economic penalties which number of studies of impurity effects on IGSCC increase in proportion to the amount being have been made as a function of either corrosion introduced into the reactor and deposited on the potential or dissolved oxygen content (the fuel. With respect to iron oxide type crud dissolved oxygen content is the major chemical deposits, it can be concluded that operation variable in BWR type water that can be used to k 'Also called electrochemical corrosas potential or ECP, see Reference 9. Amendment 2 52-8
MM 23A6100AB Standard Plant REY. B to have an important innuence on IGSCC ini. (14) Constant Extension Rate Test tlation times for smooth stainless steel O' specimens in laboratory tests. In addition, Constant extension rate tests (CERTs) are pH can serve as a useful diagnostic parame- accelerated tests that can be completed in e ter for interpreting severe water chemistry a few days, for the determination of the' transients and pH measurements are recom. susceptibility to IGSCC. It is useful for. mended for this purpose, verifying IGSCC suppression during initial. Implementation of hydrogen water chemistry (10) Electrochemical Corrosion Potential (HWC) or following plant outages that could' ' have had an impact on system chemistry The electrochemical corrosion potential (e.g., condenser repairs during refueling). (ECP) of a metal is the potentialit attains , when immersed in a water environment. The (15) Continuous Crack Growth Monitorina Test ECP is controlled by various oxidizing agents including copper and radiolysis pro- This test employs a reversing DC potential ducts. At low reactor water conductivities, drop technique to detect changes in crack the ECP of stainless steel should be below length in IGSCC test specimens. The crack 0.23 VSHE to suppress lGSCC. growth test can be used for a variety of purposes, including the following: (11) Feedwater Hydrocen Addition Rate , (a) Initial verification of IGSCC suppresi i A direct measurement of the feedwater hydro- sion following HWC implementation.
"I gen addition rate can be made using the by-drogen addition system flow measurement de- (b) Quantitative assessment of water che . ;
vice and is used to establish the plant spe- mistry transients. l l cific hydrogen flow requirements required to ' 7 satisfy the limit for the ECP of stainless (c) Long term quantification of the success ' steel (Paragraph 10). Subsequently, the ad-of the HWC program. ,
- dition rate measurements can be used to help '
diagnose the origin of unexpected ECP The major impurities in various parts of a l changes. BWR under certain operating conditions are, listed in Table 5.2 5. The plant systems have (12) Recirculation System Water Dissolved been designed to achieve these limits at least Hydrocen 90% of the time. The plant operators are
.. encouraged to achieve better water quality by A direct measurement of the dissolved hydro- using good operating practice.
gen content in the reactor water serves as a
- cross check against the hydrogen gas flow Water quality specifications require that meter in the injection system to confirm the erosion corrosion resistant low alloy stects are actual presence and magnitude of the to be used in susceptible steam extraction and hydrogen addition rate. drain lines. Stainless steels are considered e for baffles, shields, or other areas of severe o3 (13) Main Steam Line Radiation Level duty Provisions are made to add nitrogen gas' *2 E j to extraction steamlines, feedwater heater' M' The rnain steam line radiation monitor read. shells, heater drain tanks, and drain piping to' ing indicates an excessive amouet of hydro- minimize corrosion during layup. Alternatively,, !
gen being injected. Likewise, a decrease in the system may be designed to drain while hot so , this parameter would be a quick indication that drylayup can be achieved, of a decrease or stoppage of hydrogen , ,, injection. Condenser tubes and tubesheet are required to ' *: e be made of titanium alloys. M5 lO 4 1 Amendment 2 1 5.2 11
ABWR m-n Standard Plant nu Water quality specifications for the AllWR re. (3) carbon steel and low alloy steel; quire that the condenser is to be designed and erected to minimize tube leakage and to facili- (4) some 400 series martensitic stainless steel h tale maintecance. Appropriate features are in. (all tempered at a minimum of 110@f); 2 ,j corporated to detect leakage and segregate the Ei source. The valves controlling the cooling water (5) Colmonoy and Stellite hardfacing material to the condenser sections are required to be (or cquivalent) operable from the control room so that a leaking section cao be scaled off quickly. All of dise construction materials are resistant to stress corrosion in the llWR cool. 5.2.3.2.2.4 IGSCC Considerations ant. General corrosion on all materials except carbon and low alloy steel, is negligible. Plant experience and laboratory tests indicate Conservative corrosion a!!owances are provided that !GSCC can be initiated in solution annealed for all exposed surfaces of carbon and low alloy stainless steel above certain stress levels after steels. losing exposure to radiation. The requirements of GDC 4 relative to compat. Extensive tests have also shown that IGSCC has abihty of components with environmental condi- , not occurred at fluence levels below ~5x1020 tions are met by compliance with the applicabic y n/cm2 (E>1MeV) even at high stress levels. Ex- provisions of the ASME Code by compliance with periments indicate that as fluence increases the recommendations of Regulatory Guide l.44. above this threshold of 5x1020 n/cm 2, there is a decreasing threshold of sustained stress be- Contaminants in the reactor coolant are low which IGSCC has not occurred. (Examination controlled to sery low limits. These controls of top guides in two operating plants which have are implemented by limiting containment levels 3 cresiced designs has not revealed any IGSCC ) of elements (such as halogens, S, Pb) to as low , 71 as possible in miscelleaneous mater!als used d Reactor core structural components are design-during fabrication and installation. These materials (such as tapes, penentrants) are h ed to be below these thresholds of exposure and/ completely removed and cleanliness is assumed. ( or stress to avoid IASCC. In addition, crevices No detrimental effects will occur on any of the have been climinated from the top guide design in materials from allowable contaminant levels in order to prevent the synergistic interaction with the high purity reactor coolant. Expected IASCC. radiolytic products in the llWR coolant hase no in areas where the 5x1020 n/cm2 threshold of irradiation is not practicall, avoided, the 5.2.3.2.4 Compatibility of Construction stress level is maintained below the stress Materials with Esternal lnsulation threshold. liigh purity grades of materials are used in control rods to extend their life. Also All non metallic insulation applied to auste-IlWC introduced in the plant design to control nitic stainless steel meets Regulatory Guide IGSCC may also be beneficial in avoiding lASCC, 136 5.2.3.2.3 Compatibility of Construction 5.2.3.3 Fabrication and IYocessing of Ferritle Materials with Reactor Coolant Materials The construction materials exposed to the 5.2.3.3.1 Fructure Toughness reactor coolant consist of the following: Compliance with Code requirements shall be in ; (1) solution annealed austenitic stainless accordance with the following: ' steels (both wrought and cast), Types 304, 3NL, and 316L; (1) The ferritic materials used for piping, i pumps, W valves of the reactor coolant (2) nickel based alloy and alloy steel; pressure boundary are usually 21/2 inches g! l AmenJment 2 32'I2
Mkh 11A6100AB Standard Plant REV. B or less in thickness. Impact testing is boundary are fabricated-from carbon steel performed in accordance with NB 2332 for materials, e thicknesses of 21/2 inches or less. The j materials comply with Appendix G. Section G 3100 of ASME Code Section 111. Preheat temperature employed for welding of low alloy steel meet or exceed the recommenda. tions of ASME Code Section 111, Subsection NA.- (2) Materials for bolting with nominal diameters Components are either held for an extended time exceeding one inch are required to meet both at preheat temperature to assure removal of hy. the 25 mils lateral expansion specified in drogen, or preheat is maintained until post weld NB 2333 and the 45 ft lb Charpy V value spe- heat treatment. The minimum preheat and maximum cified in 10CFR50, Appendix G. The 45 ft lb interpass temperatures are specified and requirement stems from the ASME Code where monitored. It applies to bolts over 4 inches in diame- . e ter, starting Summer 1973 Addenda. Prior to Acceptance Criterion II.3.b(1)(a) of SRP Sec-j this, the Code referred to only 2 sizes of tion 5.2.3 for control of preheat temperature bolts (.<_1 inch and > 1 inch). GE continued requires that minimum and rnaximum interpass the two size categories, and added the 45 temperature be specified. While the ABWR ft !b as a more conservative requirement. control of low hydrogen electrodes to prevent hydrogen cracking (provided in Subsection (3) The reactor vessel complies with the requi- 5.2.3.3.4) does not explicitly meet this rements of NB.2331. The reference tempera- requirement, the ABWR control will assure that ture (RTNDT) is established for all cracking of components made from low alloy required pressure retaining materials used steels does not occur during fabricationc in the construction of Class 1 vessels. Further, the ABWR control minimizes the This includes plates, forgings, weld possibility of subsequent cracking resulting-material, and heat affected zone. The from hydrogen being retained in the weldment. , RTNDT differs from the nil ductility
! ( temperature (NDT) in that in addition to All welds were nondestructively examined by passing the drop test, three Charpy V Notch radiographic methods. In addition, a supple.
I specimens (traverse) must exhibit 50 ft lb mental ultrasonic examination was performed. r absorbed energy and 35 mil lateral expansion , at 600F above the RTNDT. The core 5.2JJ.2.2 Regulatory Guide l.34: Control of . beltline material must meet 75 ft lb Electrostag Weld Properties ' absorbed upper shelf energy. No electroslag welding is performed on BWR o (4) Calibration of instrument and equipment components. 2 shall meet the requirements of the ASME A Code, Section III, paragraph NB 23M. 5.233.2.3 Regulatory Guide 1.71: Welder Quallnestion for Areas of Limited Accessibility 5.233.2 Control of Welding Welder qualification for areas of limited 5.2.3.3.2.1 Regulatory Guide 1.50: Control of accessibility is discussed in Subsection Prebest Temperature Employed for Welding of 5.23.4.23. Im Alloy Steel 5.2.3.33 Regulatory Guide 1.66: Nondestive. Regulatory Guide 1.50 delineates prcheat tem- tise Examinatloa of Tubular Products , perature control requirements and welding proce-dure qualifications supplementing those in ASME Regulatory Guide 1.66 describes a meicd of Sections III and LX. implementing requirements acceptable to NRC re- , garding nondestructive examination requirements The use of low alloy steel is restricted to of tubular products used in RCPB. This Regula-l the reactor pressure vessel. Other ferritic tory Guide was withdrawn on September 28,1977, 4 comporents in the reactor coolant pressure by the NRC because the additional requirements i Amenomeat 2 $~2-1)
I ABWR m-n Standard Plant REV B imposed by the guide were satisfied by the ASNIE with the guide lines of NUREG 0313, to avoid Code, significant sensitization. g Wrought tubular products were supplied in Process control are exercised during all accordance with applicable ASThi/AShlE raaterial stages of component manufacturing and construc-specifications. Additionally, the specification tion to minimize contaminants. Cleanliness con-for the tubular products used for CRD housings trols are applied prior to any elevated temper. specified ultrasonic examination to paragraph ature treatment. For applications where stain. NB 2.550 0f AShlE Code Section ll!. less steel surfaces are exposed to water at tem. peratures above 200'F low carbon (<0.03%) These RCPB components meet 10CFR$0 Appendix B grade materials are used. For critical applica-requirements and the AShtE Code requirements thus tions, nuclear grade materials (carbon content assuring adequate control of quality for the .s 0.02%) are used. All materials are supplied products. in the solution heat treated condition. Special o sensitization tests are applied to assure that Z 5.2.3.3.4 hfolsture Control for im flydrogen, the materialis in the annealed cordition. N Cosertd Arc Welding Electrodes During fabrication, any heating operation All low hydrogen covered welding electrodes (except welding) between 800 1800'F rre are stored in controlled storage areas and only avoided, unless followed by solution heat authorized persons are permitted to release and treatment. During welding, heat input is distribute electrodes. Electrodes are received controlled. The interpass temperature is also in hermetically sealed canisters. After removal controlled. Where practical, shop welds are from the sealed contain:rs, electrodes which are solution heat treated. In general, weld filler not immediately used are placed in storage ovens material used for austenitic stainless steel which are maintained at about 250oF (generally base metals is Type 30SL/3161/3091 with an 2000F minimum). average of 8% (of Fn) ferrite content. Electrodes are distributed from scaled con-tainers or ovens as required. At the end of each 5.2.3.4.1.2 Process Controls to blin!mlie O Exposure to Contaminants work shift, unused electrodes are returned to the storage ovens. Electrodes which are da maged, Exposure to contaminants capable of causing w::, or contaminated are discarded. If any stress / corrosion cracking of austenitic stain-electroae ~ Nadvertently left out of the less steel components was avoided by carefully ovens for a w n one shift, they are discard- controlling all cleaning and processing mate. ed or recon 'in ned in accordance with rials which contact the stainless steel during manufacturer instrut ions. manufacture, construction, and installation. 5.2.3.4 Fabrication an'I Processing of Special care was exercised to insure removal Austenitic Stainless F'.cels of surface contaminants prior to any heating operations. Water quality for cleaning, 5.2J.4.1 ASoldance of Stress / Corrosion rinsing, flushing, and testing was controlled Cracking and monitored. Suitable protective packaging was provided for components to maintain 5.2.3.4.1.1 Asoldance of Significant cicanliness during shipping and storage. Sensitization The degree of surface cleanliness obtained by When austenitic stainless steels are heated in these procedures meets the requirements of the temperature range 800 1800 F, they Regulatory Guides 137 and l.44, 2 are considered to become ' sensitized" or U susceptible to intergranular corrosion. The ABWR For commitment and revision number, see design complies with Regulatory Guide 1,44 and Section 1.8. O Amendmnt 2 12-14
ABM MAH00AD Standard Plant REV B j 5.23.4.1J Cold Worked Austenitic Stainless high alloy steels or other materials such as O Steels static e.nd centrifugal castings and bimetallic
, V joints should comply with fabrication require-4 Cold work controls are applied for components ments of Sections 111 and IX of the ASME Boiler
' made of austenitic stainless steel. These mate. and Pressure Vessel Code. it also requires rials are used in the cast condition. During additional performance qualifications fo'r fabrication cold work is controlled by applying welding in areas oflimited access, limits in hardness, bend radii and surface finish on ground surfaces. All ASME Section 111 welds are fabricated in accordance with the requirements of Sections 111 5.23.4.2 Control of Welding and IX of the ASME Boiler and Pressure Vessel Code. There are few restrictive welds involved 5.23.4.2.1 Avoidance of Hot Cracking in the fabrication of BWR components. Welder qualification for wc!ds with the most restrie-Regulatory Guide 131 describes the acceptable tive access is accomplished by mockup welding.
method c' implementing requirements with regard Mock up is examined sectioning and radiography = to the control of welding when fabricating and (or UT). y joining austenitic stainless steel components and systems. The Acceptance Criterion II.3.b.(3) of SRP Section 5.2.3 is based on Regulatory Guide Written welding procedures which are approved 1.71. The ABWR design meets the intent of this by GE are required for all primary pressure boun. regulatory guide by utilizing the alternate dary welds. These procedures comply with the approach as follows: N requirements of Sections til and IX of the ASME l Boiler Pressure Vessel Code and applicable NRC When access to a non volumetrically examined Regulatory Guides. ASME Section III production weld (1) is less than 12 inches in any direction and (2) allows O ^ii i'ic i i i i > ia riii - - iai=> rre- e== direciie o ix.==ch materials were required by specification to have weld and repairs to welds in wrought and cast a minimum delta ferrite content of 5 FN (ferrite low alloy steels, austenitic stainless steels number) determined on undiluted weld pads by and high nickel alloys and in any combination of magnetic measuring instruments calibrated in these materials shall comply with the fabrica. accordance with AWS specification A4.2 74, tion requirements specificd in ASME Boiler and Pressure Vessel Code Section til and with the Delta ferrite measurements are not made on requirements of Section IX invoked by Section qualification welds. Both the ASME Boiler and III, supplemented by the following requirements: Pressure Vessel Code and Regulatory Guide 131 specify that ferrite measurements be performed on (1) The welder performance qualification test undiluted weld filler material pads when magnetic assembly required by ASME Section IX shall ' instruments are used. There are no requirements be welded under simulated access condi. = l for ferrite measurement on qualification welds. tions. An acceptable test assembly will N ! provide both a Section IX welder , 5.23.4.2.2 Regulatory Gulde 134: Electroslag performance qualification required by this l Welds Regulatory guide. 1 Electroslag welding was not employed for if the test assembly weld is to be judged
, reactor coolant pressure boundary components. by bend tests, a test specimen shall be , removed from the location least favorable 5.23.4.23 Regulatory Guide 1.71: Welder for the welder. If this test specimen Qualification or Areas of umited Accessibility cannot be removed from a location prescribed by Section IX, an additional l Regulatory Guide 1.71 requires that weld bend test specimen will be required. If ;
- fabrication and repair for wrought low-alloy and the test assembly weld is to be judged by Amendment 2 S.2 15 i
ABWR mman Standard Plant RN H radiography or UT, the length of weld to be system, or connected to the reactor coolant examined shall include the location least favorable for the welder. sptems, up to and including: g (1) The outermost containment isolation valve Records of the results obtained in welder in system piping that penentrates the accessibility qualification shall be as primary reactor containment, certified by the manufacturer or installer, shall be maintained and shall be made (2) The second of two valves normally closed accessible to authorized personnel. during normal reactor operation in system piping that does not penentrate primary Socket weld with a 2 in. nominal pipe sire reactor containment. and under are excluded from the above requirements. (3) The reactor coolant system and relief vahrs.
- (2) (a) For accessibility, when more restricted y access conditions than qualified will 5.2.4.2 Prmisions for Access to the Reactor obscure the welder's line of sight to Coolant Pressure Boundary the extent that production welding will require the use of visual aids such as 5.2.4.2.1 Iksign and Arrangement of Reactor mirrors. The requalification test as- Coolant Pressure Boundary Comp <ments sembly shall be welded under the more restricted access conditions using the Accessibility in accordance with AShtE Code visual aid required for production Section XI, lWA 1500 is provided as described in welding. the following paragraphs. 6" (b) GE complies with AshtE Section IX. 5.2.4.2.2 Reactor Pressure Vessel (3) Surveillance of accessibility qualification requirements will be performed along with Access for examination of the RPV has been provided through provisions incorporated into h
normal surveillance of AShtE Section IX the design of the vessel, shield wall, and performance qualification requirements. vessel insulation as followt 5.2J.4J Regulatory Guide 1.66: (1) The shield wall an.:1 vessel insulation be-Nondestructhe Examination of Tubular Products hind the shield wall are spaced away from the RPV outside surface. Access ports are For discussion of compliance with Regulatory located at each reactor pressure vessel Guide 1.66, see Subsection 5.2333, norile. The annular space between the reactor vessel outside surface and insula-5.2.4 Inservice Inspection and Testing tion inside surface permits insertion of of Reactor Coolant Pressure lloundary remotely operated ultrasonic devices for examination of vessel longitudinal and This subsection discusses the inservice in- circumferential welds. Access for ! spection and testing program for the NRC Ouality insertion of the automated devices is Group A components; i.e., ASME Boiler and Pres- provided through removable insulalation sure Vessel Code Section 111, Class 1, compo- panels at the top of the shield wall. nents, it will show how the program meets requirements of Section XI of the AShtE Code. (2) Aceess to the reactor pressure circum- ; ferential, longitudinal, and nozzle- l 5.2.4,1 Sptem noundary Subject to lnspection to vessel welds above the shield wall is I provided through use of removable j The system boundary subject to inspection insulation panels. Either manual or j includes all pressure vessels, piping, pumps, and valves which are part of the reactor coolant automated examination methods may be employed. gl Amendment 2 3 2 tSa
ABM MA6100AB Standard Plant REV B ; i i gallon per minute, thus meeting Position C.2 and testing is prosided. requirements. O By monitoring (1) floor drain sump fillup and These satisfy Position C.8 requirements. pumpout rate, (2) airborne particulates, and (3) Limiting unidentified leakage to the range of air coolers condensate flow rate, Position C.3 is 1 to 5 gpm and identified to 25 gpm satisfies. satisfied. Position C.9. hionitoring of the reactor building cooling 5 2,6 Interfaces i i water heat exchanger coolant return lines for : radiation due to leaks within the RHR, RIP and The remainder of plant will meet the water RWCS heat exchangers (and the fuel pool cooling chemistry requirements given in Table 5.2 5. - system heat exchangers) satisfies Position C.4. For system detail, see Subsection 7.6.1.2. 5 2.7 References The floor drain sump monitoring, ait particu. 1. General Electric Standard Application for lates monitoring, and air cooler condensate moni. Reactor fuel, (NEDE 24011 P A, latest app-toring are designed to detect leakage rates of roved version). one gpm within one hour, thus meeting Position C.5 requirements. 2. BWR Normal Water Chemistry Guidelines: 1986 Revision, EPRI NP 4946 SR, Final The fission products monitoring subsystem is Draft, October 17,19S6 (To be published). - qualified for SSE. The containment floor drain sump monitor, air cooler, and condensate flow 3. D.A. Itale, The Effect of BWR Startup En- l meter are qualified for OBE, thus meeting vironments on Crack Growth in Structural 4 Position C.6 requirements. Alloys, Trans, of AShf E, vol 108, January. 1986.
. Leak detection indicators and alarms are provided in the main control room. This 4. F.P. Ford and bl. J. Povich, The Effect of satisfies Position C.7 requirements. Procedures Oxygen / Temperature Combinations on the and graphs will be provided by the applicant to Stress Corrosion Susceptibility of Sensi, plant operators for converting the various tized T 304 Stainless Steel in High Purity indicators to a common leakage equivalent, when Water, Paper 94 presented at Corrosion 79, necessary, thus satisfying the remainder of Atlanta, GA, hiarch 1979. 31.4 Position C.7. The leakage detection system is 31.7 equipped with provisions to permit testing for 5. BilR Normal Water Chemistry Guidelines: 1986 3 1.10 operability and calibration during the plant Revision, EPRI NP-4946-SR, October 1987. Item 4 & 11 operation using the following methods:
i (1) simulation of signals into trip units; Oxygen on the Stress Corrosion Cracking of Stainless Steels: Review of Literature, (2) comparing channel A to channel B of the same hiaterial Performance, NACE, Vol.19, No. 4,
- leak detection method (i.e., area tempera. April 1980. ;
ture monitoring); ,
- 7. W.J. Shack, et al, Environmentally Assist. 1 (3) opetability checked by comparing one method ed Cracking in Light Water Reactors: Annual versus another (i.e., sump fillup rate ver. Report, October 1983 September 1984, sus pumpout rate and particulate monitoring NUREG/CR-42S7, AN1-85-33, June 1985.
j on air cooler condensate flow versus sump fillup rate); and 8. D.A.11 ale, et al, BWR Coolant impurities Program, EPRI, Palo Alto, CA, Final Report (4) continuous monitoring of floor drain sump on RP2293 2, to be published. j level and a source of water for calibration 4 j Amendment 2 gy 4 1 I
ABWR maae l Standard Plant ativ n
- 9. K.S. Brown and G.M. Gordon, Effects of BlVR g Coolant Chemistry on the Propensity ofIGSCC W Initiation and Growth in Creviced Reactor Internals Components, paper presented at the Third laternational Symposium of Envi-ronmental Degradation of Materials in Nucle-at Power Systems, ANS NACE TMS/AIME, Traverse City, Michigan, September 1987.
- 10. B.M. Gordon et al, EAC Resistance of BlVR Materials in HIVC, Proceeding of Second .
International Symposium Environmental Degration of Materials in Nuclear Power Systems, ANS, LaGrange Park, ILL 1986.
- 11. B.M. Gordon, Corrosion and Corrosion Controlin BliRs, NEDE 30637, December 1934.
- 12. B.M. Gordon et al, Halogen lYater Chemistry for BlVRs - Materials Behavior, EPRI NP 5080, Palo Alto, CA, March 1987.
O O Amendment 2
- l 23A6100AB Standard Plant arv. e i Table 5.21 O REACTOR COOLANT PRESSURE BOUNDARY COMPONENTS APPLICABLE CODE CASES Number I111g Annlicable Eaulement Remarks N 7115 (1) Component Support Accepted per RG 1.85 N 122 (2) Piping Accepted per RG 1.84 N 247 (3) Component Support Accepted per RG 1.84 N 249-9 (4) Component Support Conditionally Accepted per RG 1.85 N 309-1 (5) Component Support Accepted per RG 1.M N 313 (6) Piping Accepted per RG 1.84 N 316 (7) Piping Accepted per RG 1.84 N 318 3 (8) Piping Conditionally Accepted pcr RG 1.84 $ ,
N 319 (9) Piping Accepted per RG 1.84 i N 391 (10) Piping Accepted per RG 1.84 N 392 (11) Piping Accepted per RG 1.84 N 393 (12) Piping Accepted per RG 1.84 - N-411 1 (13) Piping Conditionally Accepted ; per RG 1.84 N-414 (14) Component Support NEW Not yet listed N-430 (15) Component Support NEW Not yet listed I N-433 (16) Component Support NEW Not yet listed N-451 (17) Piping NEW Not yet listed 'I 1 O Amendment 2 5.2 29 I
ABWR mamn Standard Plant REV.D Table 5.2 1 REACTOR COOLANT PRESSURE HOUNDARY COMPONENTS h APPLICABLE CODE CASES (Continued) (1) Additional Afaterials for Subsection NF, (13) Alternative Damping Values for Seismic Classes 1, 2, 3 and AfC Component Supports A nalysis of Classes 1, 2, 3 Piping Fabricated by if'elding, Section 111, Division Sections, Section lil, Dishion 1. 1. (14) Tack if' elds for Class I, 2, 3 and AfC (2) Stress indices for Structure Attachments, Components and Piping Supports. Class 1, Section 111, Dishion 1. (15) Requirements for il'elding il'orkmanship and (3) Certified Design Report Summary for Com. Visual Acceptance Criteria for Class 1, 2, ponent Standard Supports, Section ill, 3 and MC linear Type and Standard Suppons. Dishion 1, Class 1, 2, 3 and MC. (16) Non Threaded Fasteners for Section Ill, (4) Additional blaterial for Subsection NF, Division 1, Class 1,2, and 3 Component and Uasses 1, 2,3 and AlC Compo. Piping Suppons. ($) Identification of Afaterials for Component (17) Alternatise Rules for Analysis of Piping Suppons, Section 111, Disbion 1. Under Seismic Loading Class 1. (6) Alternate Rules for ifalf-Coupling Branch Connections, Section 111, Dishion 1. (7) Alternate Rules for Fillet it' eld Dimensions for Socket IVelded Fittings, Section 111, Q Dishion 1, Class 1,2,3. (8) P:ocedure for Evaluation of the Design of Rectangular Cross Section Attachments on Class 2 or 3 Piping, Section ill, Division 1. (9) Alternate Procedure for Evaluation of Stress in Butt IVeld Elbows in Class 1 Piping, Section 111, Dishion 1. (10) Procedure for Evaluation of the Design of Hollow Circular Cross Section il'elded Attach-ments on Class 1 Piping. Section 111, Dishion 1. (11) Procedure for Evaluation of the Design of ifollow Circular Cross Section lYelded Attachments on Classes 2 and 3 Piping, Section 111, Dishion 1. (12) Repair IVelding Structural Steel Rolled Shapes and Plates for Component Supports, Section 111, Dishion 1. O Amendment 2
$22%
ABM uraow Standard Plant nev. h 5.3 REACTOR VESSEL these vessel components is in accordance with procedures qualified per ASME Section III and O-5.3.1 Reactor Vessel Materials IX requirements. Weld test samples are requir. ed for each procedure for major vessel full. 53.1.1 Materials Specifications penetration welds. Tensile and impact tests are performed to determine the properties of The materials used in the reactor pressure the base metal, beat affected zone, and weld vessel and appurtenances are shown in Table 5.2 4 r-etal. together with the applicable specifications. Submerged are and manual suck electrode The RPV materials shall comply with the welding processes are etnployed. Electroslag provisions of the ASME Code Section III, Appendix welding is not applied. Preheat and interpass A I and meet the specification requirements of temperatures employed for welding of low alloy 10CFR50, Appendix G. steel meet or exceed the values given in ASME, Section III, Appendix D. Post weld heat treat-5.3.1.2 Special Procedures Used for Manufactur- ment at 11000F minimum is applied to all ing and Fabrication low alloy steel welds. The reactor pressure vessel is primarily con. Radiographic examination is performed on all structed from low alloy, high strength steel plate pressure containing welds in accordance with re- , and forgings. Plates are ordered to ASME SA 533, quirements of ASME, Section 111, Subsection NB TYPE B, Class 1, and forgings to ASME SA 508, 5320. In addition, all welds are given a Class 3. These materials are melted to fine grain supplemental ultrasonic examination, l , practice and are supplied in the quenched and tem- ; pered condition. Further restrictions include a The materials, fabrication procedures, and ' requirement for vacuum degassing to lower the hy- testing methods used in the construction of BWR drogen level and improve the cleanliness of the reactor pressure vessels meet or execed require-O ie* iie> i i - " i ri i a i th beltline region also specify limits of 0.05% max-e>- - i er^ sue s tie >>> ci se i vess i>. l Imum copper and 0.015% maximum phosphorous con- 53.1.3 Special Methods for Nondestructhe tent in the base materials and a 0.08% maximum Examination I copper and 0.020% maximum phosphorous content in I weld materials. The materials and welds on the reactor pres; sure vessel are examined in accordance with Studs, nuts, and washers for the main closure methods prescribed and meet the acceptance re-flange are ordered to ASME SA 540, Grade B23 or quirements specified by ASME, Section 111. In Grade B24. Welding electrodes for low alloy steel addition, the pressure retaining welds are ul-are low hydrogen type ordered to ASME SFA 5.5. trasonically examined using manual techniques. The ultrasonic examination method, including All plate, forgings, and bolting are 100% ultra- calibration, instrumentation, scanning sensitiv-sonically tested and surface examined by magnetic ity, and coverage, is based on the requirements particle methods or liquid penetrant methods in ac- imposed by ASME, Section XI, Appendix 1. Accep-cordance with ASME Section !!!, Division 1. tance standards are equivalent or more restrie-
~ tive than required by ASME, Section XI.
5 Fracture toughness properties are also measured l 53.1.4 Special Controls for Ferritic and l and controlled in accordance with Disision 1.Austenitic Stainless Steels All fabrication of the reactor pressure vessel is performed in accordance with GE approved draw- 53.1.4.1 Regulatory Guide 1J1: Control of ings, fabrication procedures, and test procc- Stainless Steel Welding dures. The shells and vessel heads are made from formed plates or forgings, and the flanges and Controls on stainless steel welding are dis-nozzles from forgings. Welding performed to join cussed in Subsection 5.23.4.2.1. 2 O Amendment 2
ABM - utsiooxo Standard Plant REV D 53.1.4.2 Regulatory Guide 134: Control of of low hydrogen electrodes to prevent hydrogen Electroslag Weld Properties cracking (provided in Subsection 5.2.3.3.4) does not explicitly meet this requirements the ABWR h Electroslag welding is not employed for the control will assure that cracking of components reactor pressure vessel fabrication, made from low alloy steels does not occur during fabrication. Further, the ABWR control minimi-53.1.4.3 Regulatory guide 1.43: Control of zes the possibility of subsequent cracking re. Stainless Steel Weld Cladding of low Alloy Steel sulting from hydrogen being retained in the Components weldment. Reactor pressure vessel specifications require All welds are nondestructively examined by that all low alloy steel be produced to fine radiographic methods. In addition, a supplemen-grain practice. The requirements of this regula- tal ultrasonic examination is performed. tory guide are not applicable to BWR vessels. 5.3.1.4.6 Regulatory Guide 1.71: Welder 53.1.4.4 Regulatory Guide 1.44: Control of Qualification for Areas of1Jmited Accessibility the Use of Sensitized Stainless Steel Qualification for areas of limited accessi-Sensitization of stainless steel is controlled bility is discussed in Subsection 5.2.3.4.2.3. by the use of service proven materials and by use of appropriate design and processing steps in- 53.1.4.7 Regulatory Guide 1.99: Effects of y cluding solution heat treatment, corrosion resis. Residual Elements on Predicted Radiation Damage 3 tant cladding, control of welding heat input, to Reactor Pressurt Vessel Materials control of heat treatment during fabrication and control of stresses. Predictions for changes in transition tem-perature and upper shelf energy are made in ac. 53.1.4.5 Regulatory Guide 1.50: Control of cordance with the requirements of Regulatory Preheat Temperature For Welding IA* Alloy Steel Guide 1.99. Regulatory Guide 1.50 delineates preheat tem- 5.3.1.5 Fracture Toughness perature control requirements and welding proce-dure qualifications supplementing those in ASME 53.1.5.1 Compilance with 10CFR50, Appendix G Sections III and IX. Appendix G of 10CFR50 is interpreted for ; The use of low alloy steel is restricted to Class 1 primary coolant pressure boundary compo- I the reactor pressure vessel. Other ferritic com. nent of the ABWR reactor design and complied l ponents in the reactor coolant pressure boundary with as discussed in Subsections 5.3.1.5.2 and l are fabricated from carbon steel materials. 5.3.2. The specific temperature limits op- l eration of the reactor when the core is critical Preheat temperature employed for welding of are based on 10CFR50, Appendix G, Paragraph IV, low alloy steel meet or exceed the recommenda- A3. , tions of ASME Code Section 111, Appendix D. g Components are either held for an extended time 5.3.1.5.2 Methods of Compliance at preheat temperature to assure removal of hydrogen, or preheat is maintained until post- The following items are the interpretations weld heat treatment. The minimum preheat and and methods used to comply with 10CFR50, Ap-maximum interpass temperatures are specified and pendix G. moaltored. (1) Material Test Coupons and Test Specimens Acceptance Criterion II.3.b(1)(a) of SRP (GIII A) Section 5.2.3 for control of preheat temperature requires that minimum and maximum interpass Test coupons are removed from the loca-temperature be specined. While the ABWR control tion in each product form as specified g Amendment 2 2
21%61M 4 Standard Plant nev. s j in subarticle NB 2220 of the ASME Code, i Section III. The heat treatment of the ; test coupons is performed in accordance with subarticle NB 2210. 1 It is understood that separately pro-duced test coupons per Subparagraph , NB 2223.3 may be used for forgings. ; (2) Location and Circulation of Test Specimens (GIII A) w The test specimens are located and ori- $ ented per ASME, Section III, Paragraph NB 2322. Transverse Charpy V impact specimens are used for the testing of plate and forged material other than bolting and bars. Longitudinal specimens are used for bolting and bars. Both longitudinal and transverse specimens are used to determine the re- h quired minimum upper shelf energy level 3 ; of the core belt line materials. In regard to 10CFR50, Appendix H, the surveillance test material is selected , O. ee is 6 t er t6 a tr - t er^sT" E185 82 and Regulatory Guide 1.99 to
- provide a conservative adjusted refir- t ence temperature for the beltline materials. The weld test plate for the t i surveillance program specimens has the principal working direction parallel to the weld seam to. assure that <
L heat affected zone specimens are trans-l verse to the principal working direc-l tion. ; q i (3) Records and Procedures for Impact ; Testing (G III C) ; l ) 1 l t .
's i
l I l l I. O Amendment 2
$12a
ABWR m aoaxa Standard Plant REV,D completed vessel. Each in reactor surveillance These estimates show that the adjusted reference (] V capsule contains 36 Charpy V notch and 6 tensile temperature at end of life is less than specimens. The capsule loading consists of 12 100 0F, and the end of life upper shelf energy Charpy V Specimens each of base metal, weld exceeds 50 ft lb. (See response to Question y metal, heat affected zone material, and 3 tensile 251.5 for the calculation and analysis associ- g specimens each from base metal and weld metal. A ated with this estimate). set of out of reactor baseline Charpy V-notch specimens, tensile specimens, and archive mater- 53.1.6.4 Positioning of Sunelliance Capsules ial are provided with the surveillance test and hiethods of Attachment (Appendix 11.11 B (2)) specimens. Neutron dosimeters and temperature monitors will be located within the capsules as Surveillance specimen capsules are located at required by ASTM E 185 82. three azimuths at a common elevation in the core beltline region. The sealed capsules are not at-Three capsule are provided in accordance with tached to the vessel but are in welded capsule requirements of 10dR50, Appendix 11, since the holders. The capsule holders are mechanically predicted end of the adjusted reference tempera- retained by capsule holder brackets welded to ture of the reactor vessel steel is less than 100 the vessel cladding. Since reactor vessel spe-0 F, cifications require that all low alloy steel pressure vessel boundary materials be produced The following proposed withdrawal schedule is to fine grain practice, underclad cracking is of in accordance with ASThi E 185 82. no concern. The capsule holder brackets allow the removal and reinsertion of capsule holders. First capsule: After 6 effective full power Although not code parts, these brackets are de-years signed, fabricated, and analyzed to the require-Second capsule: After 15 effective full power ments of AShiE Code Section 111. A positive years spring loaded locking device is provided to re- ,3 Third capsule: Schedule determined based on tain the capsules in position throughout any an-i, ') 3 results of first two capsules per ASThi E 185 82, ticipated event during the lifetime of the
'~~
A Paragraph 7.6.2. vessel. Fracture toughness testing of irradiated cap- In areas where brackets (such as the surveil-sule specimens will be in accordance with require- lance specimen bolder brackets) are located, ad-ments of ASThi E 185 82 n called out for by ditional nondestructive examinations are per-10CFR50, Appendix 11. formed on the vessel base metal and stainless steel weld deposited cladding or weld buildup 5.3.1.6.2 Neutron Flut and Fluence Calculations pads during vessel manufacture. The base metal is ultrasonically examined by straight beam tech-A description of the methods of analysis is niques to a depth at least equal to the thick-contained in Subsections 4.1.4.5 and 4.3.2.8. ness of the bracket being joined. The asea exam- ; ined is the area of width equal to at least half l 53.1.6.3 Predicted Irradiation Effects on the thickness of the part joined. The required Belt!!ne blaterials stainless steel weld deposited cladding is simi- i larly examined. The full penetration welds are l Transition temperature changes and changes in liquid penetrant examined. Cladding thickness l upper shelf energy shall be calculated in accor- is required to be at least 1/8 inch. These re-dance with the rules of Regulatory Guide 1.99. quirements have been successfully applied to a l Reference temperatures shall be established in ac- variety of bracket designs which are attached to cordance with 10CFR50, Appendix G, and NB 2330 of weld deposited stainless steel cladding or weld the AShf E Code, buildups in many operating BWR reactor pressure vessels. Since weld material chemistry and fracture toughness data are not available at this time, inservice inspection examinations of core o the limits in the purchase specification were beltline pressure retaining welds are performed C used to estimate worst case irradiation effects, from the outside surface of the reactor pressure Amendment 2
ABM 2mioaru Standard Plant REV B 5 3 2 Pressure / Temperature Limits Reactor Operation 5.3.2.1 Umit Curves Curve C in Figure 5.31 specifies limits ap. plicable for operation whenever the core is The pressure / temperature limit curves in critical except for low level physics tests. Figure 5.31 are based on the requirements of 10CFR50, Appendix G. 53.2.1A ReactorVessel Aenealing All the vessel shcIl and head areas remote Inplace annealing of the reactor vessel, from discontinuities plus the feedwater nozzles because of radiation embrittlement, is not an-were evaluated, and the operating limit curves ticipated to be necessary, are based on the limiting location. The boltup limits for the flange and adjacent shell region 53.2.13 Predicted Shift in RTNDT and are based on a minimum metal temperature of Drop in Upper Shelf Energy (Appendix G IV B) [ RTNDT +600F. The maximum throughwall tem. - perature gradient from continuous heating or For design purposes the adjusted reference cooling at 1000F per hour was considered. The nil ductility temperature and drop in the safety factors applied were as specified in AShtE upper shelf energy for BWR vessels is predicted , Code, Appendix G, and Reference 2. using the procedures in Regulatory Guide 1.99. The material for the vessel will be provided The calculations (see response to Question : with the following requirements of RTNDT as 251.5) are based on the specified limits on O determined in accordance with Branch Technical Phosphorous (0.020%), Vanadium (0.05%), Copper A Position hiTEB 5 2: shell and flanges 20*F; (0.08%) and Nickel (1.2%) in the weld material. nozzles 20'F and welds 20 F. In plate material, the limits are Copper (0.05%) and Nickel (0.73%) Forgings will have the same 53.2.1.1 Temperature Limits for Boltup chemitry as plate but the nickellimit is 1%. hiinimum closure flange and fastener tem- N A surveillance program in accordance with i peratures of RTNDT + 600F are required for ASThi E 185 82 will be used. The surveillance
, tensioning at preload condition and during program willinclude samples of base metal, weld ,
g d e t e nsioning. Thus, the limit is 60* F + metal and heat affected zone material. Subsec- t ( 20 F) = 50
- F. tion 5.3.1.6 provides added detail on the sur-veillance program.
5.3.2.1.2 Temperature umits for ISI Hydro-static and leak Pressure Tests 53.2.2 Operating Procedures Pressure (measured in the top head) versus tem. A comparison of the pressure versus tem-perature (minimum vessel shell and head metal tem- perature limit in Subsection 5.3.2.1 with in-perature) limits to be observed for the test and tended normal operation procedures of the most operating conditions are specified in Figure severe upset transient shows that those limits 5.3 1. Temperature limits for preservice and will not be exceeded during any foreseeable inservice tests are shown in Curve A of Figure upset condition. Reactor operating procedures 5.31. have been established so that actual:; & ents l will not be more severe than those for which the j 5.3.2.1.3 Oper3 ting Limits During IIcatup, vessel design adequacy has been demonstrated. Cooldown, and Core Operation Of the design transients, the upset condition j producing the most adverse temperature and pres. licatup and Cooldown, sure condition anywhere in the vessel head and/ or shell areas yields a minimum fluid tempera-Curve B in Figure 5.31 specifies limits for ture of $23 0F and a maximum peak pressure of non nuclear heatup and cooldown following a 1215 psig. Scram automatically occurs as a nuclear shutdown. result of this c<ent prior to a possible reduc-l l 5" Amw m2 l l l
ABM - ux6ioara Standard Plant REV B tion in fluid temperature to 2500F at a pres. l sure of 930 psig. Per Figure 5.31, both the 1215 psig vessel pressure at 528 0F (Curve C) 1 and the 930 psig at 250 0F (Curve B) are within the calculated margin against nonductile failure. 5.3.3 Reactor Vessel Integrity The reactor vessel material, equipment, and services associated with the reactor vessels and appurtenances would conform to the requirements of the subject purchase documents. Measures to ensure conformance included provisions for source evaluation and selection, objective evidence of quality furnished, inspection at the vendor source and examination of the completed reactor l vessels. GE provides inspection surveillance of the I 1 0
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l l 1 i l O Amendment 2 SMe
4 ABM ux61oorn Standard Plant arv. s Design of the reactor vessel and its support by vertical stilt legs from the bottom head. system meets Seismic Category I equipment require. This support is designed to carry the weight of v ments. The materials used in the reactor pres- peripheral fuel elements, neutron sources, core sure vessel are listed in Table 5.2-4. plate, top guide and the steam separators and to laterally support the fuel assemblies and the .
- The cylindrical shell and top and bottom heads pump diffusers. Design of the shroud support of the reactor vessel are fabricated of low alloy also accounts for pressure differentials across steel, the interior of which is clad with stain- the shroud support plate, for the restraining '
less steel weld overlay except for the top head effect of components attached to the support, and nozzle weld zones, and for earthquake loadings. The shroud support design is specified to meet appropriate ASME Inplace annealing of the reactor vesselis not Code stress limits. necessary because shifts in transition tempera-ture caused by irradiation during the 60 year 5JJ.1.1J Protection of Closure Studs life can be accommodated by raising the minimum pressurization temperature, and the predicted The BWRs do not use borated water for reactiv-value of adjusted reference temperature does not ity control during normal operation. This sub-exceed 200 0F Radiation embrittlement is not section is therefore not applicable. a problem outside of the vessel beltline region because the irradiation in those areas is less 5J3.1.2 Safety Design Basis than 1 X 10 *
- nyt with neutron energies in excess of 1 McV. The use of existing methods of The design of the reactor vessel and appurte.
predicting embrittlement and operating limits nances meets the following safety design bases. ) which are based on a 40 year life are considered i to be applicable to a 60 year life because the (1) The reactor sessel and appurtenances will 3 age degrading mechanism is irradiation and fati- withstand adverse combinations of loading
; n gue duty which are calculated for the 60 year and forces resulting from operation under ab.
life. Time / temperature effects will either not normal and accident conditions, have any effect or will produce a small beneficial co. annealing. (2) To minimize the possibility of brittle frac-ture of the nuclear system process barrier, Quality control methods used during the fabri- the following are required: cation and assembly of the reactor vessel and ap-1 purtenances assure that design specifications are (a) impact properties at temperatures rela-1 met, ted to vessel operation have been specified ' i for materials used in the reactor vessel; ; j The vessel top head is secured to the reactor i vessel by studs and nuts. These nuts are tight. (b) expected shifts in transition tempera- 1 1 ened with a stud tensioner. The vessel flanges ture during design life as a result of envi- ! are sealed with two concentric metal seal rings ronmental conditions, such as neutron flux, , designed to permit no detectable leakag: through are considered in the design and operational ! j the inner or outer seal at any operating condi- limitations assure that NDT temperature ; i tion, including heating to operating pressure and shifts are accounted for in reactor opera- i temperature at a maximum rate of 100 0F in any tion; and f one hour period. To detect seal failure, a vent ; i tap is located between the two scal rings. A mo- (c) operational margins to be observed with i nitor line is attached to the tap to provide an regard to the transition temperature are ( indication of leakage from the inner seal ring specified for each mode of operation. I seal.
; $33.13 Power Generstloa Design Bases l SJJ.1.1.2 Shroud Support l' The design of the reactor vessel and appurte.
The shroud support is a circular plate welded nances meets the following poster generation j to the vessel wall and to a cylinder supported design bases: t
- AmeMmeat 2 m
il
hh MA6100AD Standard Plant REV.B which provide for case of installation and remo- 53.3.2 Materials of Construction val for vessel inservice inspection and mainte-nance operation. Each insulation unit has lift. All material used in the construction of the ing fittings attached to facilitate removal. In- reactor pressure vessel conform to the require-sulation units attached to the shield wall are ments of ASME Code, Section 11 materials. The not required to be readily removable except vessel heads, shells, flanges, and nozzles are around penetrations. fabricated from low alloy steel plate and forg-ings purchased in accordance with ASME Specifica-At operating conditions, the insulation on the tions SA 533 Type B, Class 1 and SA 508 Class shield wall and around the refueling bellows has 3. Special requirements for the low alloy steel an average maximum beat transfer rate of 65 Btu plate employed on the interior surfaces of the per hour per square foot of outside insulation vessel consists of austenitic stainless steel surface. The maximum heat transfer rate for insu. weld overlay, lation on the top head is 60 Btu per hour per square foot. Operating conditions are $500F These materials of construction were selected for the outside temperature of the reactor vessel because they provide adequate strength, fracture and 1350F for the drywell air. The maximum toughness, fabricability, and compatibility with air temperature is 1500F, except for the head the BWR environment. Their suitability has been area above the bulkhead and refueling seal which demonstrated by long term successful operating has a maximum allowable temperature of 2000F. experience in reactor sersice. 5.3.3.1 A5 Reactor Vessel Nozzles The expected peak neutron fluence at the 1/4 t location used for evalution is less than 4 x All piping connected to the reactor vessel 1017 nyt for 60 years, the calculated shift in
- nozzles has been designed not to exceed the allow- RTNDT is 28'F for weld metal and 8'F 3 able loads on any nozzle. The vessel top head for base metal and the drop in upper shelf ener- A 1 (
nozzle is prosided with flanges with small groove gy is 10 ft lbs for welds and 8 ft lbs for base , facings. The drain nozzle is of the full penetra- metal. tion weld design. The feedwater inlet nozzles, core spray inlet nozzles, and ECCS flooding noz. 53.33 Fabrication Methods t zles have thermal sleeves. Nozzles connecting to ] stainless steel piping have safe ends or exten- The reactor pressure vessel is a vertical cy. , sions made of stainless steel. These safe ends lindrical pressure vessel of welded construction l or extensions were welded to the nozzles after fabricated in accordance with ASME Code, Section
; the pressure vessel was heat treated to avoid Ill, Class 1, requirements. All fabrication of furnace sensitization of the stainless steel, the reactor pressure vessel was performed in ac- l Tbc material used is compatible with the material cordance with GE approved drawings, fabrication of the mating pipe, procedures, and test procedures. The shell and vessel head were made from formed low alloy 5.3.3.1.4.6 Materials and Inspections steel plates or forgings and the flanges and nozzles from low alloy steel forgings. Welding
{ The reactor vessel was designed and fabricated performed to join these vessel components was in l
~
in accordance with the applicable ASME Boiler and accordance with procedures qualified to ASME, Pressure Vessel Code as defined in Subsection Section III and IX requirements. Weld test 5.2.1. Table 5.2 4 defines the materials and samples were required for each procedure for specifications. Subsection 51.1.6 defines the major vessel full penetration welds.
! compliance with reactor vessel material surveil-lance program requirements. Submerged are and manual stick electrode j i welding processes were employed. Electroslag 5.33.1.4.7 Reactor Yessel Schematic welding was not applied. Preheat and interpass j temperatures empic*yed for welding of low alloy The reactor vessel schematic is shown in steel met or exceeded the requirements of ASME I Figure 5.3 2. Section III, Appendix D. Post weld heat treat. I Amendment 2 l
1 1 1
ABWR men Standard Plant nry n n ment of 11000F minimum was applied to all (2) if the coolant temperature difference be-() low alloy steel welds, tween the dome (inferred from P (sat)) and the bottom head drain exceeds All previous BWR pressure vessels have 1000F, neither reactor power level not employed similar fabrication methods. These recirculation pump flow shall be vessels have operated for an extensive number of increased. years and their service history is rated excellent. The limit regarding the normal rate of heatup and cooldown (Item 1) assures that the vessel 5JJA Inspection Requirements closure, closure studs, vessel support skirt, control rod drive housing, and stub tube All plates, forgings, and bolting were 100% ul- stresses and usage remain within acceptable trasonically tested and surface examined by mag- limits. Vessel temperature limit on recircu-netic particle methods or liquid penetrant rne- lating pump operation and power level increase thods in accordance with ASSIE Code, Section 111. restriction (Item 2) augments the item 1 limit Welds on the reactor pressure vessel were exam- in further detail by assuring that the vessel ined in accordance with methods prescribed and bottom head region will not be warmed at an ex-meet the acceptance requirements specified by cessive rate caused by rapid sweep out of cold AShtE Code, Section 111. In addition, the pres- coolant in the vessel lower head region by sure retaining welds were ultrasonically examined recirculating pump operation or natural circula-using acceptance standards which are required by tion (cold coolant can accumulate as a result of ash {E Code, Section XI. control drive inleakage and/or low recirculation flow rate during startup or hot standby). 53J.5 Shipment and Installation These operational lim:ts when maintained en-The cornpleted reactor vessel is given a thor- sure that the stress limits within the reactor 3 ough cleaning and examination prior to shipment. vessel and its components are within the thermal (V The vessel is tightly scaled for shipment to pre- limits to which the vessel was Jesigned for nor-vent entry of dirt or moisture. Preparations for mal operating conditions. To maintain the integ-shipment are in accordance with detailed written rity of the vessel in the event that these op-procedures. erational limits are exceeded, the reactor ves-set has been designed to withstand a limited On arrival at the reactor site tbc reactor number of transients caused by operator error, vesselis examined for evidence of any contamina. Also, for abnormal operating conditions where tion as a result of damage to shipping covers, safety systems or controls provide an automatic hicasures are taken during installation to assure temperature and pressure response in the reactor , that vessel integrity is maintained; for example, vessel, the reactor vessel integrity is rnain- ' access controls are applied to personnel entering tained since the severest anticipated transients i the vessel, weather protection is provided, and have been included in the design conditions. i periodic cleanings are performed. Therefore, it is concluded that the vessel integ- i rity will be maintained during the most sescre l 5JJ.6 Operating Conditions postulated transients since all such transients are evaluated in the design of the reactor l Procedural controls on plant operation are vessel. l implemented to hold thermal stresses within ac- j
- ceptat>le ranges and to meet the pressure /tempe- 533 7 Insenice Suncillance
$ rature limits of Subsection 5.3.2. The restrictions on coolant temperature are ss lunvice inspection of the rucer pressure follows: .cose; wil; be in accori- ace with . - aquire-
'nu of the ihtE ILier and Pri ' Vessel (1) the average rate of change of reae lectten AL The vessel will , e.ine d coolant ternperature during normal he o to startup to satidy the s .rera.
g\ ( and cooldown shall not exceed if 1 ments of IWBCGM cl AShui Code, ' during any one hour period; subsequent ir.scr\ ice insNction Amendment 2
ABWR - 2nsioorn Standard Plant RIN R I will be scheduled and performed in accordance 1 with the requirements of 10CFR50.55a, subpara-graph (g) as described in Subsection 5.2.4. h j i The materials surveillance program will l l i l l l l l l l 1 l i 1 9 l I
- 9 Amendment 2 Salta
i MM 23A61ooAs Standard Plant REV.B i Q Table 15.710 - ISOTOPIC RELEASE TO ENVIRONMENT IN CURIES ISOTOPE 1 MIN 10 MIN 1 HOUR 2 HOUR I131 832E-02a 7.45E-01 2.61E + 00 3.19E+ 00 . 1132 1.07E-01 939E-01 3.02E + 00 3.52E + 00 l I133 8f4E-02 7.69E 01 2.67E + 00 3.25E + 00 I 134 4.67E 09 3.95E-08 1.12E-07 1.23E-07 1135 1.41E-02 1.26E-01 4.27E-01 5.13E 01 TOTALI 2.90E-01 2.58E + 00 8.74E + 00 1.05E + 01 KR 83M 4.43E-01 3.86E + 00 1.22E + 01 1.40E + 01 KR 85M 5.65E + 00 5.01E + 01 1.68E + 02 2.00E + 02 ! KR 85 3.0SE + 01 2.76E + 02 9.68E + 02 1.18E + 03 KR-87 8.75E-M 7.54E-03 2.27E-02 2.57E-02 KR SS 1.63E + 00 1.44E + 01 4.71E + 01 5.53E + 01 XE131M 537E + 00 4.81E + 01 1.69E + O2 2.06E + 02 XE133M 7.11E + 01 637E + 02 2.23E + 03 2.72E + 03 XE 133 1.81E + 03 1.62E + 04 5.70E + M 6.%E + N , XE135M 2,12E + 01 1.58E + 02 3.11E + O2 3.16E + O2 i XE 135 4.15E + 02 3.70E + 03 1.27E + 04 1.53E + M ! TOTAL NG 237E + 03 2.11E + 04 736E+N 8.%E + N a 832E-02 = 832 x 10 h I 1 i i l l l
.I j
i l 1 O Amendment 2 117 jg
ABM 234sioosa Standard Plant m 1 - Table 15.712 4 O FUEL CASK DROP ACCIDENT PARAMETERS f f I Data and assumptions used to estimate source terms i 4 A. Power level of reactor while fuel was la core 4005 M Wt B. Radial Peaking Factor while fuel was in core 1.5 L C. Fuel Bundles in Cask 18 , D. FuelDamaged 1116 rods
- E. Minimum time of fuelin storage prior to accident 120 days 6 F. Peak linear power density 13.4 kW/ft G. Average burn up of fuel 32,000 MWV/t H. Maximum Fuelcenterline temperature 3315 F i 1. Fraction of actisity released 10% of allisotopes except 30% Kr 85 J. Time Period for Reactor Building Release 2 hour l K. Iodine Filter Efficiency 99 %
l !! Dispersion and Dose Data i
; A. Meteorology Table 15.713 ,
j B. Boundary and LPZ distances Table 15.713 i C. Method of Dose Calculation Reference 1
- i D. Dose conversion Assumptions Reference 1 and RG 1.109 6 1 E. Activityloventory/ releases Table 15.713 !
; Table 15.713 l l F. Dose Evaluations Table 15.713 !
i i i i ! I f 1 ; l 1 l t l , 1 ) ! f lO ; Amendment 2
}
l [ j ! i I
ABWR m6im^n Standard Plant REV. B Table 15.713 CASK DROP ACCIDENT RADIOLOGICAL RESULTS O INVENTORY IN SHIPPING CASK RELEASED TO ENVIRONMENT IN CURIES RELEASE TO ISOTOPE REACTOR BUILDING ENVIRONh1EhT I.131 1.08E + 01a 1.0SE-01 KR-85 1.10E + N 1.10E + M XE131 5.49E + 00 5.48E + 00 m XE.133 1.25E-01 1.25E-01 ly METEOROLOGY AND DOSE RESULTS x/Q TIIYROID WilOLE BODY 3 (SEC/51 ) (REht) (REht) DISTANCE (51) 300 1.18E-03 6.57E-02 7.24E-03 500 4.83E44 2.69E-02 2.% E-03 800 2.19E-M 1.22E-02 1.34E 03 1000 1.77E44 9.85E-03 1.09E-03 1200 1.48E44 8.27E-03 9.10E-N 1500 1.19E44 6f6E 03 733E44 2000 9.01E-05 5.02E-03 5 53E44 2500 7.22E-05 4.03E-03 4.44E C4 3000 6.02E-05 136E-03 3.70E44 3500 5.16E-05 2.8SE-03 3.17E44 a 1.05E + 01 = 1.M x 10 +
- O Amendment 2 15.7 21}}