ML20071Q726
| ML20071Q726 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 06/07/1983 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| JNF-038-83, JNF-38-83, MFN-099-83, MFN-9-83, MFN-99-83, NUDOCS 8306090236 | |
| Download: ML20071Q726 (125) | |
Text
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GENERAL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 MFN 099-83 (408) 925-5722 M/C 682 JNF 038-83 June 7, 1983 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention: Mr. D.G. Eisenhut Division of Licensing Gentlemen:
SUBJECT:
IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)- DOCKET NO. STN 50-447 DRAFT AMENDMENT PERTAINING TO CP/ML AND SRP RULES Attached please find a draft of GESSAR II Amendment Number 17 pertaining to our response to the CP/ML Rule (10CFR50.34(f)) and SRP Rule (10CFR50.34(g)).
Our responses to these rules are included separately in this transmittal as Attachment Number 1 (CP/ML Rule) and Attachment Number 2 (SRP Rule).
We plan to formally file Amendment Number 17 in mid June 1983.
If there are any questions on the information provided herein, please contact J.N. Fox of my staff at (408) 925-5039.
Sincerely, Glenn G. Sherwood, Manager Nuclear Safety & Licensing Operation Attachments cc: F.J. Miraglia (w/o attachments) C.0. Thomas (w/o attachments)
D.C. Scaletti L.S. Gifford (w/o attachments)
(EG06090236 830607 h
PDR ADOCK 05000 llh
ATTACHMENT NO.1
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DRAFT OF GESSAR II AMENDMENT NO.17 RESPONSE TO CP/ML RULE l
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GESSAR II 22A7007 238 NUCLEAR ISLAND- Rev. 17 I I *,
M APPENDIX 1G l RESPONSE TO CP/ RULE j il 10CFR50.3d V
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APPENDIX 1G CONTENTS b 'ection S Title Page EEE.iiE - - E3 EEE M l; 1G.0 INTRODUCTION i
lG.0-1 I
1G.1 Probabilistic Risk Assessment [ Item (1) (i) ] 1G.1-1 k_
i 1G.2 Auxiliary Feedwater Systems 5 valuation
-i [ Item (1) (ii) ] 1G.2-1 1G.3 Potential-Reactor Coolant Pump Seal Damage.
[ Item (1) (iii)] 1G.3-1 V
1G.4 Small-Break LOCA Caused by Stuck Open Power Relief Valve [ Item (1) (iv) ] 1G.4-1 8' 1G.5 Safety Effectiveness in Separating HPCS/RCIC l d i
Initiation Levels [ Item (1) (v) ] 1G.5-1 f 1G.6 Challenge Reduction of Relief Valves d
[ Item (l>(vi)] 1G.6-1 O
[ 4 1G.7 Optimum Automatic Depressurization System l ,
(ADS) Design [ Item (1) (vii)] 1G.7-1 1G.8 Automatic Restart of HPCS and LPCI
[ Item (1) (viii) ] 1G.8-1
) Y
{ T 1G.9 Space Cooling for RCIC and HPCS 3
[ Item (1) (ix) ] 1G.9-1 M 1G.10 ADS Study with No Credit for Non-Safety Related Equipment [ Item (1) (x) ] 1G.10-1 li d 1G.11 Alternate Depressurization Methods
'j [ Item (1) (xi) ] 1G.ll-1
't j 1G.12 Alternative Hydrogen Control Systems f, [ Item (1) (xii) ] 1G.12-1
! 1G.13 Simulator Capability [ Item (2) (i) ] 1G.13-1 4,
. 1G.14 Program for Improving Plant Procedures
[ Item (2) (ii)] 1G.14-1 l 1G.15 State-of-the-Art Human Factors in Control Room [ Item (2) (iii)] 1G.15-1 1G-i
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GESSAR II 22A7007 p 238 NUCLEAR ISLAND Rav. 17 b APPENDIX 1G I CONTENTS (Continued)
Section Title Page 6 155533 533 5 Eia af
{} 1G.16 Safety Parameter Display System (SPDS) d [ Item (2) (iv)] 1G.16-1 h
N( y\ 1G.17 Safety System Status Monitoring [ Item (2) (v) ] 1G.17-1 1G.18 High Point Venting [ Item (2) (vi) ] 1G.18-1 g 1G.19 Radiation and Shielding Design Review
.j [ Item (2) (vii) ] 1G.19-1 3
f 1G.20 Prompt Analysis of Samples from Reactor
[ Coolant System and Containment [ Item (2) (viii)] 1G.20-1 j 'lG.21 Hydrogen Control System Design [ Item (2) (ix) ] 1G.21-1 1
l 1G.22 Test Program for Qualification of Safety
! Relief Valves (SRVs) [ Item (2) (x)] 1G.22-1
?
$ 1G.23 SRV Position Indication [ Item (2) (xi)] 1G.23-1 i
j 1G.24 Auxiliary Feedwater (AFW) Initiation
[ Item (2) (xii) ] 1G.24-1
?
,I 1G.25 Natural Circulation in Hot Standby t
Conditions [ Item (2) (xiii) ] 1G.25-1
( 1G.26 Containment Isolation Dependability
- [ Item (2) (xiv) ] 1G.26-1 i
.f 1G.27 Containment Purging / Venting [ Item (2) (xv) ] 1G.27-1 i 1G.28 ECCS and RPS Design Criterion [ Item (2) (xvi)] 1G.28-1 i
i 1G.29 Containment Parameters Instrumentation in If Control Room [ Item (2) (xvii) ] 1G.29-1 1G.30 Unambiguous Indication of Inadequate Core l Cooling [ Item (2) (xviii)] 1G.30-1 I
t lG.31 Post-Accident Monitoring [ Item (2) (xix) ] 1G.31-1 1G.32 Pressurizer Instrumentation [ Item (2) (xx)] 1G.32-1 i
i 1G.33 Auxiliary Heat Removal System Design and Procedures (Item (2) (xxi) ] 1G.33-1 IG.34 FMEA of Integrated Control System
[ Item (2) (xxii)] 1G.34-1 I
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, 238 NUCLEAR ISLAND Rav. 17 APPENDIX lG CONTENTS (Continued) 4 Section Title Page Jl m asa JL sm m s --
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1G.35 Anticipatory Reactor Trip [ Item (2) (xxiii) ] 1G.35-1
-lG.36 Post-Accident Water Level Instrumentation
[ Item (2) (xxiv) ]
1G.36-1 1
) 1G.37 Onsite Support Centers [ Item (2) (xxv) ] 1G.37-1
'I i 1G.38 Leakage Control Program for Systems j outside Containment [ Item (2) (xxvi) ] 1G.38-1 t
l 1G.39 Inplant Radiation and Airborne Radioactivity
[ Item (2) (xxvii) ] 1G.39-1 1G.40 Control Room Habitability [ Item (xxviii)] 1G.40-1 1G.41 Administrative Procedures for E a @ ing M Experience [ Item (3) (i) ] 1G.41-1 I 1G.42 Expanded Quality Assurance (QA) List j [ Item (3) (ii) ] 1G.42-1 1G.43 Detailed QA Criteria [Iem (3) (iii) ] 1G.43-1 &
- i. 1G.44 Dedicated Containment Penetrations <
( [ Item (3) (iv)] 1G.44-1 I
i 1G.45 Containment Integrity [ Item (3) (v) ] 1G.45-1 i
1G.46 External Hydrogen Recombiners [ Item (3) (vi)] 1G.46-1
$ 1G.47 Management Planning and Proce e for
- Plant Design and Construction {I m (3) (vii)] 1G.47-1 4 i
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GESSAR II 22A7007 238 NUCLEAR ISLAND REV 17 APPENDIX 16
- RESPONSE TO CP/ML RULE 10CFR 50.34 (f) 4 16.0, INTRODUCTION
' On January 15,1982(47FR2286)theNRCamended10CFR34toinclude_(fl f
' Additional ~TMI-Related Requirements. These additional requirements were
- directed to each applicant for a light-water-reactor Construction Pennit
! or Manufacturing License (CP/ML) whose application was pending as of h February 16, 1982.
( In its' Proposed Consission Policy Statement on Severe ~ Accidents and Related f ~ Views on Nuclear' Reactor' Regulation'(7590-01) , the NRC proposed to extend i
its policy such that future CP applications or reactivations of CP appli-r cations previously docketed also comply with the CP/ML rule .
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This appendix IG reports GE's responses for the 238 Nuclear. Island to the NRC positions taken regarding the " Licensing Requirements for Pending l Applications for Construction Permits and Manufacturing License" as
- referenced in NUREG-0718. Rev 2. These responses have developed as the NRC positions have evolved and been clarified by the issuance of subsequent documentation by the NRC.
The responses demonstrate that the NRC requirements are satisfactorily i
fulfilled for the 238 Nuclear Island. For each item a stamary of the NRC position is given and followed by a response. The response clarifies the issue as it pertains to the 238 Nuclear Island and/or provides a listing of applicable GESSAR II sections, relevant correspondence, or
, other necessary doctanentation that may be referenced for complete clarification of our postion. Where a particular requirement is not
> applicable to the 238 Nuclear Island, a statement to that effect is
, provided in the response.
?
GESSAR II 22A7007 b 238 NUCLEAR ISLAND REV 17 1
16.0 INTRODUCTION
(Cont'd) j For NRC positions that affect equipment outside the scope of the Nuclear Island design or utility operations and procedures, the response indicates that the subject.will be addressed by the applicant. Otherwise, this
, Appendix IG is complete in that all of the " Additional TMI-Related Require-ments" approved for implementation by the NRC as listed in 10CFR 50.34 (f) have been favorably addressed where they apply'to the 238 Nuclear Island.
The bracketed item nebers at the end of each title correspond with the I subsectionsin10CFR50.34(f). Alphan meric designations at the end of
) each "NRC Position" statement correspond to the related action plan items in NUREG-0718 and NUREG-0660. They are provided in 10CFR 50.34 (f) for
! infonnation only.
3 Table 1G.0-1 is provided as a convenient cross-reference which consolidates 9 pertinent information associated with each of the 47 requirements. This includes the 10CFR34 (f) subsection, the TMI Action Plan numbers, the j GESSAR II Section n eber where each NRC position and response is given, the item title as given in NUREG-0718; and the GESSAR 11 reference.
detailing resolution.
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TABLE R \ G . O - t j GESSAR II ADHERENCE TO CP RULE 10 CFR 50.34(f)
GESSAR II gpggggE f SECTION ACTION PLAN SECTION ITEM GESSAR II {m3GliiT!=/";TATJ
[ (1) (1) II.B.8 IG.1 Probabilistic Risk Assessment Appendix 15D (11) II.E.1.1 1G.2 Auxilary Feedwater System Evaluation (PWR Only)
(iii) II.K.2.16 & 1G.3 Effect of loss of alternating . Section 1A.46 & 1A.66 II.K.3.25 Current on pump seals.
(iv) II.K.3.2 1G.4 Report on overall Safety Effect (PWR Only)
], of PORY Isolation System (v) II.K.3.13 1G.5 Separation of HPCS'and RCIC System -
. Section 1A.58 Initiation Levels i
l1 (vi) II.K.3.16 1G.6 Reduction'on of Challenges and Section 1A.60
) failures of safety relfef valves-j feasibility study & system modification 1
1$ (vii) II.K.3.18 1G.7 Modification of ADS logic-feasibility Section 1A.62 i study & modification for-increased diversity'of some event sequences (viii) II.K.3.21 1G.8 Restart of core spray and LPCI systems Section 1A.63 on low level-design and modification (ix) II.K.3.24 IG.9 Confinn adequacy df space cooling- study Section 1A.65
- for HPCS and RCIC (x) II.K.3.28 1G.10 Verify qualification of accumulators Section 1A.68 on ADS valves (xi) II.K.3.45 1G.11 Evaluate depressurization with other Section IA.72 than full ADS (xii) - 1G.12 Evaluation of alternative hydrogen l p % ^s y ' / j g g g 1 control systems
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SECTION ACTION PLAN E ON ITEM GESSAR II T 'T"J'I -
(2) (1) I.A.4.2 1G.13 Long-Tenn' Training. Upgrade Section 18.2 (ii) I.C.9 1G.14 .Long-Term Program of Upgrading Applicant Responsibility of procedures (iii) I . D.1 1G.15 Control Room Design Reviews Section 18.4 (iv) I . D.2 1G.16 Plant Safety Parameter Display Console Appendix 188
.)
(v) I.D.3 1G.17 Safety System Status Monitoring Not Applicable (GESSAR II complies with R.G.1.47)
(vi) II.B.1 1G.18 Reactor Coolant System Vents Section 1A.19 1 (vii) II.B.2 1G.19 Plant shielding to pfoyide access to Section 1A.20 Nh*ko!t0!chdWopera$f$n#"
- (viii) II.B.3 IG.20 Post-Accident sampling Section 1A.21 l
l- (ix) II.B.8 1G.21 Hydrogen Control System Pre- /9 f4p(sf/
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(x) II . D.1 1G.22 Testing Requirements #
Section 1A.23 l
l (xi) II .D.3 -
1G.23 Relief and Safety Valve Position Section 1A.24 l- Indication (Xii) II.E.1.2 1G.24 Auxiliary feedwater system automatic (PWR Only)/
initiative and flow indicator (xiii) I.E.3.1 1G.25 Relability of pane supplies for (PWR Only) natural circulation
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, SECTION ACTION PLAN SECTION ITEM GESSAR II K ,R"T!"/ STAT"",-
(2) (xiv) II.E.4.2 1G.26 Isolation Dependability Section 1A.29 (xy) II.E.4.4 1G.27 Purging i Fui sc T o n . . . L e ,-- \ G . 2.,7
- (xvi) II.E.5.1 1G.28 Design Evaluator B & W Only (xvii) II . F.1 1G.29 Additional Accident Monitoring Appendix 1D Instrumentation (xviii) II . F.2 1G.30 Identification of and Recovery from- Section 1A.31
,a Conditions leading to Inadequate Core (xix) II . F. 3 1G 31 Coolin9 Appendix ID Instrumentation for Monitoring Accident Conditions (Reg. Guide 1.97) f (xx) II.G.1 1G.32 Power supplies for pressurizer (PWR Only%J i Relief Valves, Block Valves & Level j Indication (xxi) II.K.1.22 1G.33 he)bgAuggtg r
gMg}A igfor- Section 1A.38 Removal Systems when FW System not operable 1
(xxii) II.K.2.9 1G.34 Analysis. bf.Up' grading of Inter- (B&WOnly) grated Control System (xxiii) II.K.2.10 1G.35 Hand-wired Safety-grade Anticipatory (B & W Only)
Reactor trips (xxiv) II.K.3.23 IG.36 Centrol WatFL vel Rec 6rding Section 1A.39 i
(xxv) III.A.1.2 1G.37 Upgrade License Emergency Support Applicant Responsibility Facility l (xxvi) II I . D.1.1 1G.38 Primary Coolant Sources Outside the Section 1A.77 Containment Structure (xxvii) III .D.3.3 1G.39 In-Plant Radiation Monitorino Applicant Responsibility
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j (2) (xxviii) III . D. 3.4 1G.40 Control Room Habitility Section IA.79 4
(3) (1) I.C.5 1G.41 Procedures for Feedback of Applicant Responsibility j1 Operating Design and Construction l Experience (11) I . F.1 1G.42 Expand QA List C;i:::tte. 13.5.17.; I G . 4 'L-(iii) I . F.2 1G.43 Develop More Detailed QA Criteria 1.eelf ; t .t;er.;Ibil: 6T l G- 4 3
' (iv) II.B.8 1G.44 " esid; er.; er .T.e. [ Dedicated h e 02" re ..uia6iuns l G.4 4 Containment Penetrations, Equivalent
, to a Single 3-foot diameter opening j (v) II.B.8 1G.45 Containment Integrity gg ' iced , cuc> ign 6u ouvumuda ti ggjg i' 45 p;ig :t Sc.-sice Le el C.
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1 (vi) II.E.4.1 1G.46 Dedicated Penetration Section IA.28 4
] (vii) II.J.3.1 1G.47 Organization and Staffing to Oversee f.pplicent .t;;;--ibili+y g g,47
.'j Design and Construction l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 17 1G.1 PROBABILISTIC RISK ASSESSMENT (Item (1) (i)]
NRC Position Perform a plant / site specific probabilistic risk assessment, the aim of which is to seek such. improvements in the reliability of core 'nda containment heat removal systems as are significant and '
practical and do not impact excessively on the plant. (II.B.8)
Response
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238.: NUCLEAR ISLAND Rav. 17
, 1G.2 AUXILIARY FEEDWATER SYSTEM AFWS EVALUATION
- [ Item (1) (ii)]
1
'I NRC Position i e Perform an evaluation of the proposed auxiliary feedwater system
., (AFWS), to include (applicable to PWR's only) (II.E.1.1):
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(.A) A simplified AFWS reliability analysis using event-tree i
j and fault-tree logic techniques.
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(B) A design review of AFWS.
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i (C) An evaluation of AFWS flow design bases and criteria.
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This requirement iq not applicable to the 238 Nuclear Island.
, It applies only to PWR-type reactors.
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GESSAR II 22A7007 r 238 NUCLEAR ISLAND Rsv. 17 -
EHPAcT OF RCP SEAL. D AM ACS Fot Lowbac S n Avt-645AL- ,
1G.3 Ku m.n i W ""* 'mn T'.?.::7 T *J".T ""?.' 07J' a GE ituem '1;(ii4 L O C A we Lo s a of o $ag h _P% [, q4, Mp g g)( NRC Position l l Perform an evaluation of the potential for Ang npact of reac- ,
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tor coolant pump seal damage following small reak LOCA with w' loss of offsite power. If damage cannot be precluded, provide an j analysis of the limiting small-break loss-of-coolant accident with '} subsequent reactor coolant pump seal damage. (II.K.2.16 and , lf
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GESSAR II 2MEYWUY f ' Rsv. 17, 23B !!UCLEAP. ISLAUL I RGPog:T 09 OWu. SAW Ef=pset w PORV n omod Wmu ' e-1G.4 S".?.LL Lnsan LGCA Cnvasu os oivCE Uria 70;22 ALLIL- s 4NWWWD '[ Item (1) (iv) ] , , I l NRC Position Perform an analysis of the probability of a small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve (PORV). If this probability is a significant contri-1 butor to the probability of small-break LOCA's from all causes, provide a description and evaluation of the effect on small-break LOCA probability of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls after the PORV has opened. (Applicable to PWR's only) .- (II.K. 3.2) y
Response
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.his requirement is not applicable to the 238 Nuclear Island. It I applies only to PWR-type reactors.
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. 238 NUCLEAR ISLA,ND Rsv. 17
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Pe'rform an evaluation of the safety effectiveness of providing for
. separation of high pressure coolant injection (HPCI) and reactor h core isolation cooling (RCIC) system initiation levels so that the
,f RCIC system initiates at a higher water level than the HPCI system, I and of providing that both systems restart on low water level. [ (For plants with high pressure core spray systems in lieu of high f pressure coolant injection systems, substitute the words, "high h pressure core spray" for "high pressure coolant injection" and
$ ("HPCS" for "HPCI") (Applicable to BWR's only). (II.K.3.13)
. Response
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 17
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op sue asts af.vntve I- pcanos usw my e' Aw0 A VMM WooTtsi m op [,t.l. w (L)(v[,)] NRC Position Il l Perform a study to identify practicable system modifications that would reduce challenges and failures of relief valves, without compromising the performance of the valves or other systems. j (Applicable to bWR's only) . (II.K.3.16) 1 [ j Response
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j GESSAR II 22A7007 p 238 NUCLEAR' ISLAND Rav. 17 1 MODL A WW of: ADS LOGtC -csAJsG lWTy' iwDY AND i 1G.7 TT!".U:' AUT^:2."'!C EErnuaduRIZATION b X d 1 I.cl gAUC) "c"" (_.-- O tca- ' ' 'ii: P e oD\ f5t CAT t0 0 FOS 1N CS.oAIM D1WR. sit 7 (1 j o t* 28MW sWkT g1ll;C2,vsuCe J C T 4cm ( t ) ( v ~ C )] NRC Position Perform a feasibility and risk assessment study to determine the optimum automatic depressurization system (ADS) design modifica-tion that would eliminate the need for manual activation to ensure f adequate core cooling. (Applicable to BWR's only) . (II.K.3.18) I i l
Response
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GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rnv. 17 l $ESTART $5 COME SfRAY AMP LfCl SYSTEMS ON LN LEWCL-ggy y AMO M#fff/C4MhV 1G.8 HJTC."_'. TIC TATA ,. OT !!PCC ."?D LPC. [ Item (1) (viii)] NRC Position 4 Perform a study of the effect on all core-cooling modes under accident conditions of designing the core spray and low pressure coolant injection systems to ensure that the systems will auto-matically restart on loss of water level, after having been manually stopped, if an initiation signal is still present. (Applicable to BWR's only). (II.K.3.21)
Response
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l GESSAR II 22A7007 .j 238 NUCLEAR ISLAND Rav. 17 'l . Confl4M { $T99Y fed A N!CS9EQUACY OF $9Act CocLING. AN9 Mcic
.; 1G. 9 -S".'.CC CGGLI;;C. rOn 20IC A;;D I;FD[ Item (1) (ix) ] '
~ NRC Position Perform a study to determine the need for additional space cooling to ensure reliable long-term operation of the reactor core isola-
, tion cooling (RCIC) and high-pressure coolant injection (HPCI) j systems, following a complete loss of offsite power to the plant k
1 for at least two (2) hours. (For plants with high pressure core i spray systems in lieu of high pressure coolant injection systems, p substitute the words, "high pressure core spray" for "high f pressure coolant injection" and "HPCS" for "HPCI") (Applicable to {- BWR's only). (II.K.3.24) i 9
- Response h fh 4
- b'kNd i y'g, pm - -
At & &y +' " p s ecs d Reic & "" W ~ p W W '^' W ' k l . I 1G.9-1/lG.9-2
===;===.g, .m.-.-,.-- _
- n. -~~~ ~ ~ ~ . - -
3
1 GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rav. 17 y! RIFF 80At.rficf770W Ot' A CCVPlul41MS CN A PS VAL 1G.10 AL, STUO'1 " }... M0 CALL 11 Fun nun oArr.u RELalEL e fiO9fPMEN9htem (1) (x) ] NRC Position y Perform a study to ensure that the Automatic Depressurization 3 System, valves, accumulators, and associated equipment and instrumentation will be capable of performing their intended 1 functions during and following an accident situation, taking no 2 credit for non-safety related equipment or instrumentatio.., and accounting for normal expected air (or nitrogen) leakage through i valves. (Applicable to BWR's only). (II.K.3.28) Response ,
$ k Ek k /lCf4 5 9 tt' -
k M /A.tv. - l l l t l i f l l I l l l 1 l l 1G.10-1/lG.10-2 1
GESSAR II 22A7007 238 NUCLEAR ISLAND R2v. 17 EVALVATE OE!AESSort1 Art:N w177f oftfER Tif4( fyLL APS x 1G.ll -Af2E"E.'.TI: DE""ESS"nI"ATICI; i-iElliGL& [ Item (1) (xi) ] C 1 i NRC Position i l
! Pr, ovide an evaluation of depressurization methods, other than by full actuation of the automatic depressurization system, that f would reduce the possibility of exceeding vessel integrity limits f during rapid cooldown. (Applicable to BWR's only) (II.K.3.45)
I Response
&v& b $14 $h sI&l5 i n.n. or g M 24t ~ av~ 4-sud" k l , 4h a sy m ade, Mcc rc~~ ~ < , p i, WuCEG 0q79, f+pt 6-y2. .
1 6 3 0 e i 1
's 1G.11-1/1G.11-2
*~
- - - . -- ~ ~- &--
_. . _ n ;.-- ma _,_,.a__.:--.. , __
i j ,
' GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 I /
/
EVALUATidM df [ Item (1) (xii) ] 1G.12 4 ALTERNATIVE HYDROGEN CONTROL SYSTEMS I l NRC Position . j Perform an. evaluation of alternative hydrogen control systems that t l would a sgghg u ements of paragraph (f) (2) (ix) of 44ns
- a h .p As a minimum . include consideration of a hydrogen ignition
- and post-accident inerting system. The evaluation shall include:
I k (A) A comparison of costs and benefits of the alternative systems considered. (B) For the selected system, analyses and test data to verify i compliance with the requirements of (f) (2) (ix) of t+rrs-h 10 CF R SO-3* h (C) For the selected system, preliminary design descriptions of equipment, function, anc layout.
Response
(A) C.o M =<.5 % of Q.. ,amJ % ,_Ekr d Mc s d4endu sp% conn;te,e4 JIt be pwdH hh N<. 1G.12-1/-10.12-2 4
**d *7***w
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- 8 ,*
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! t i !
I ' i
, I .
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, I i ! I I I
t.SL.S m b u lh, ! o7 ' ( o C p ft.- M o . 3 4 ( t)(5V ! I E I : I i - i I I : i. I I i ! l 5 l (C) O C.s I 1.L 0 .sk,c,s k o b ! 2_ . e-ri l I J J ' i'
! . F i b og A c__ e i s V o. b ~~
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 i: ' ' 1 LeMcr-TERM TRAtHIN6- U96Rh0E
,j 1G.13 cI_" " 'a i UR CAFABILIA'A [ Item (2) (i) ] c__
NRC Position I Provide simulator capability that correctly models the control room and includes the capability to simulate small-break LOCA's. (Applicable to construction permit applicants only) (I.A.4.2) t j Response *
) W % , $$& bo ~4 .-
a
*'*a A " W G & reJG-rneem An hebr:
l & fM a~2 da y MEW 182, i % M m & (p , e n . 3
; cS4 & .& M nam bS, fn 2L i 238 %)ar DM .sfapt as M k t 4 4.
M- AuA L ocg5 m j adactd (p, &. 17c w i .24 i
,i aimAtoh
+3w mXJ . A .4 M <w-E' M 5
.,esa 97 ab<M 6e O S wW9e .a A WM
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{ r. A v. 2. q Aiu1>e s o gir. 'h 4 t T lt t l l 1G.13-1/lG.13-2
~
I
"~
. . - - . - - - - , _ - ,.,- _ _ _ , _ , . , . _f. ,_ _ ,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 i LONb'*TW of UP&Rh0tMit' 0F f " 1G.14 APROGRAM FOR-4MPROV4N64 PLANT PROCEDURES [(Item (2) (ii)] NRC Position 1 Establish a program, to begin during construction and follow into 1 operation, for integrating and expanding current efforts to im-
, ~ prove plant procedures. The scope of the' program shall include f emergency procedures, reliability analyses, human factors engi-neering, crisis management, operator training, and coordination }
J with INPO and other industry effortc. (Applicable to construction 1 i permit applicants only) (I.C.9)
Response
}
fY dAC hy eff*b O" D h
% %p Rvvelav 2 d l" (FFYs) i pf g&, a pd q 2L. 13 7 7% dw GM &
M sapw A yopwssek,W 18me.) '
& fAh &ps (Em . d :sfwwC"_
^
j s
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p & MT %Iw Oc 4'P
- i i
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- 1 I
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.I 1
1* I ti I b e l l 1G.14-l/lG.14-2 i . I k 4.___.*__....___
~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
.k CcNMel. ReeM PESW $6 VIES 3 [pe,a (s)Qii)]
.; 1G.15 ET?.T O r -- Tis z T. R T M "."J.!: r;.CTORS I;; CO!; TROL R001-1 c
,} { It= (2) (iiili NRC Position 1
g Provide,.for Commission review, a control room design that reflects state-of-the-art human factor principles prior to committing to lj fabrication or revision of fabricated control room panels and layouts. (I.D.1) i{n !
Response
- 1. I
'1 CA~p% /7 pA e s % 9Acem neuas, .. i udwt uatf 2tr ) Nugtfr 0800 ; W !!- M% P6m(ser)Eag W PA Er& - ?$ /R d W (4ec % (7 z,iT-3) x w n ' \1 l *
& w pouf asjn:b/ go sat. bWW d au~y~ena. J
% ( w e.g)
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- MA h dM EActci ii & //e %O Gy' % &E6) Q N
,\ pc +4 & .#pt L Aunt avf pp ~/ an
- j s i t 4 ;tti h % ! A &q x ^co"' & A ^%j OM i g, p % e .x4,d%'4~. n ~N -
i g sa & 2 4 k m f X< p ~ u +1e +d i ,;, guns & ogoo , % Mc w & ~*'A N l w ;, 7t, &esne it & fen) Ev=LA' Ayd I (popg(,-o999). d2e W ^ # ^*'*"f ? " . \ ~~ I. . p& } ' j{"y _om&e4 r am& s its te M *a "
, now t
t i i f 1G.15-1/lG.15-2
.c . -e- r ,.7 - y, - - , - _-
.-.w _,y- . - . - - .,o s , . . - - - . , , . __- _ _ _ _ _
GESSAR II 22A7007 i 238 NUCLEAR ISLAND R3v. 17 r- ' fLANr 8 1G.16 A SAFETYPARAMETER[DISPLAYSYSTEM (SPDS) [ Item (2) (iv) ) . NRC Position
. I Provide a plant safety parameter display console that will display !.
- p. to operators a minimum set of parameters defining the safety status -
'k 3 of the plant, capable of displaying a full ran'ge of important plant
, parameters and data trends on demand, and capable of indicating I when process limits are being approached or exceeded. (I.D.2) i I
I. i - Response . j { t l } R 'J-3 7 T/ A 9 M j f ,, Q & p I \l %M #w 9%&& (sms) & ' l W W jlb BCS da JaevA'd .L,
$ NRE6- 09m, pw I. p.2, i l gg 7yy. .I .
b I-i f . i 1.. u 5 l? 1 1i 1 P 1i i: !i 1,
, 1G.16-1/lG.16-2 f
\ -- v+y. w . . - - - . - - - _m ww, w ,, . . .
GESSAR II 22A7007 l-l 238 NUCLEAR ISLAND Rav. 17
}
; 1G.17 SAFETY SYSTEM STATUS MONITORING [ Item (2) (v) ]
NRC Position
'I Jc Provide for automatic 3.ndication of the bypassed and inoperable *
-a status of safety systems. (I.D.3) .j Response 1 fL 137 %ck 9$4d & @ <cylw Q y, AvA I.41 (Jee M 173.9./4). -: - 'h y 0 92 *hO 1 l 4 4 p z && MJ L A %- i v i i l i ( r
?
i i 'i t h ' .i i l 't 't I - i: Il I 't li
- Assumed " operable" incorrect; changed to " inoperable."
,l 1G.17-1/lG.17-2 -t '
- 1 1
P.- . . . . . .p , _ -. r :__-,,_._ - - ~ , - - - - , . . - . .
GESSAR II 22A7007
. 238 NUCLEAR ISLAND Rev. 17 puerog coowr svsren vexts
- 1G.18 H MH-POftPP-VEN MNG4[ Item (2) (vi) ] g-NRC Position
- b l . Provide'the capability of high point venting of noncondensible I
5 gases from the reactor coolant system, and other systems that t i may be required to maintain adequate core cooling. Systems to
't ,
- achieve this. capability shall be capable of being operated from i
; th'e control room and their operation shall not lead to an unaccept-i able increase in the probability of loss-of-coolant accident or g
i an unacceptable challenge to containment integrity. (II.B.1) l t a [
Response
rur t M
& ae 4r l ! M *1 W' A w 9 m ceny4 A W a J w i ; WN 1***A M A /A t7 j MC M6 W "l'k & weeror,1,p ,s.-b.
i' t o, l\ li I) i 'k lI
\
1G.18-1/lG.18-2 i l5 'it i I.._...._ _ _ . . . . . _ _ . . - - . . .
~
GESSAR II 22A7007
- ( 238 NUCLEAR ISLAND Rev. 17 i
PLAMi' $NIEL9/N& 70 fnoVtfE ACCESS TD YlTA1. MEAS & PA*TFcr SAFETY EddI/ MENT FOR Pest- AcetNNr O!!QTHW & ,. j 1G.19 "'r!?. TION ?."C C!!!ELCINC OECIGN-4EVfBWg[ Item (2) (vii) ]
- 4 NRC Position ;
'l: l
'I Perform radiation and shielding design reviews of spaces around l systems that may, as a result of an accident, contain TID 14844 i source term radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety 2 equipment from the radiation environment. (II.B.2) t i . I g sponse h .gfj tsAls k & A2Afn ACcLNs> stM YW } p & t A R d @ /A - cla Adv&/ + Me i RW&; &w . , (A 20, me &&4 a e f I P I N i 1G.19-1/lG.19-2 > r- ~ : ~ Y *! M: .n .,. . - - . . , , - , - . , , . .
*k..-.. .
" ^
GESSAR.II 22A7007 238 NUCLEAR ISLAND Rnv. 17 [ , POST"- A CCI96Mr SAMPLINV & IG.20 PPC".rT A"ALi'0IC Or C.*_"PLEC rnO" EACTOR COOLAin sisu,ri AN9-GONTAINMENTA[ Item . (2) (viii) ] l i l NRC Position i 'i Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID V 14844 source term radioactive materials without radiation exposures to any individual exceeding 5 rem to the whole-body or 75 rem to } the extremities. Materials to be analyzed and quantified include j certain radionuclides that are indicators of the degree of core j damage (e.g., noble gases, iodines and cesiums, and non-volatile ! isotopes), hydrogen in the containment atmosphere, dissolved gases, e i chloride, and boron concentrations. (II.B.3) 0 1
Response
h fh ' d/C f des *j3h Y Y
, p ;, ,s e 4 M M zi %d '
l 4mpp 616 )G 4 pen 4r M "I [ kune c- o 999 w i + p) ,,gig.3.umg n. a 4 m q k s- @ 9 3:"2 " ? 4 4. gb Ar au .& an N & Y' ""?! ** k i i I e 5 I i 1G.20-1/1G.20-2 i -
- --= - - -
, 3 g ,, JJU NUCLEAR ISLAND M;V. 1/
h PR,Gt.g M td A W
//7 1G.21 HYDROGEN CONTROL SYSTEMgDESIGN [Itcm (2) (ix))
NRC Position
'sj?.*
jr . Provide a system for hydrogen control that can safely accommodate hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction. Preliminary design information on the tentatively
. pre f erred system option hose being evaluated in paragraph (1) (xii) of A. .s..p;p;;;;;sa so so.3 4 s sufficient at the construction permit stage. The hydrogen control' system and associated systems shall provide, with reasonable assurance, that: (II.B.8)
(A) Uniformly distributed hydrogen concentrations in the containment do not exceed 10% during and following an accident that releases an equivalent amount of hydrogen as would.be generated from a 100% fuel clad metalewater reaction, or that the post-accident atmosphere will not support hydrogen combustion. (B) Combustible c.oncentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appro-priate mitigating features. l (C) Equipment necessary for achieving and maintaining safe I shutdown of the plant and maintaining containment integrity will perform its safety function during and after being exposed to the environmental conditions attendant with the release of hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction l including the environmental conditions created by activation of the hydrogen control system. 1G.21-1 4e e M e*
**"**'H,_ , .--- -
,,,_ , y a n. ,, _ ,
238 NUCLEAR ISLAND Rw. 17 [*
/I PRma^ N ARY 1.F? IG.21 HYDROGEN CONTROL SYST N DESIGN [ Item (2) (ix)] (Continusd) ,
s
.t . ~}
(D) If the method chosen for hydrogen control is a post- E accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation. Rasponse i
, Appl s caN
! The 4h4NMP will provide a hydrogen control system capable of . ! handling ydrogen generated by the equivalent of a 100% active fuel-clad metal water reaction. The hyd-a;=m e mtr:1 cret: aill uvu.4., ;f i;r!*er= di=&rth"*=d thr ;heut the dryr:11, the uvui.i..;;nt, r-d 2ny 10:21 ;;; - whi h h== 6'e petential ef pr:Metir; hydreger. The-ignitere eill hurn hy d ra; n se it ir ;;n;;;t;d and will-terrerrhly 3--n-- *hme untra--1;-dietrihet;d hyd;;;rn_ cene:nt::ticn: rill aet ;;cc;d 100 f;;ing :nf fellerin; : 11 rid;at. T following criteria will be used to design the hyd gen ign er system: I a. urning of the hydrogen generated by the e ivalent of a 100% active fuel-clad tal water rea on such that: (1) Un ormly-distributed hydr en conc trations will not ceed 10% during and fo wing the accid t. (2) Local pocke g of drogen in the drywell, the containmen local areas will be prevented.
- b. The system will b ingle ac ve failure proof.
- c. Operation of e hydrogen ignitio system will not adversely a ect the safe shutdown the jplant.
- d. The sys a will be protected from tornad nd exter missile hazards. %
N.
- e. T system will not compromise the, containment N Ns esign.
1c.21-2 P l ~'""*
- mm= g y- e-, ,_ % ,_ ,_ ,
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, I 1
l i O l l , l y 7kg I /gp h$l S4f *1' A D/~chG 5 W2 N 02 tty
- M 4* * *
'l I - I l I I i
& A) Uniformly distributed hydrogen concentrations in the containment U do n(ot exceed 10% during and following an accident that releas lent amount of hydrogen as would be generated from a 100% fuel clad j
metal. water hydrogen reaction, or that the post-accident atmosphere will not support combustion. l j j B Combustible concentrations of hydrogen will not collect in areas j . wher(e u)nintended combustion or detonation could caus integrity or loss of appropriate mitigating features. i (C) Equipment necessary for achieving and maintaining safe shutdown I of the plant and maintaining containment integrity will perform its safety j , function during and after being exposed to the environmental conditions attendant with the release of hydrogen generated by the equivalent of a j j 100% fuel. clad metal water reaction including the environmental conditions
, created by activation of the hydrogen control system. ,
; I l * ,
l 1 t I '
^^
a -.u- . .u_. C lb tiq CdlLF / N k N h-- -- bh&9 -WW - -. . __(. ._,5yoh ! i i 1 : I i l . _ _ ~ _ _ _ _ _ icnecw,u l z % . n{ 2 %l i i l h t j ; I . Q-. h $ . _ b _ - I \ aHed fhe seE .sidownif_t/a ,o/ ant . _ - - . . : -- I j l i ! i l i -- I a._Irhehs<Annk.phiedy
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 TeSTsNtr REGditEMEMTS 1G.22 . _ _ _ _ - _ - _ _ . _ . _ _ _ _ _ _ _ _. . . . ... . SAr2TY RELIZr VAL'ICO ,
'027:} A[ Item (2) (x)]
NRC Position Provide a test program and associated model development and con-duct tests to qualify reactor coolant system relief and safety valves and, for PWR's, PORV block valves, for all fluid conditions expected under operating conditions, transients and accidents. Consideration of anticipated transient without scram (ATWS) conditions shall be included in the test program. Actual testing under ATWS conditions need not be carried out until subsequent phases of the test program are developed. (II.D.1)
Response
g W Q me . La bem &N b b? P
& aux g '< o w g s A : d C J m. Ztr -
M /A 2 3- 'A A PM"' ' "d f Wm ' Wg g ft, u oNtc p-nantt a5fgf ( Eee NMEG O M i . N 143' g & y c: dw y dMev'd"^r C d*r P4 A 7+M su N &n y y op @e ~ h k# u cM ^
" f .Q h & delam M pr tw 4 p /5; F.
s. 1G.22-1/lG.22-2
,m. an ogs =
- m
, - ,---w - -- ye
GESSAR II 22A700.7 238 NUCLEAR ISLAND Rev. 17 RELIEF ANP SAFETY VALVE #
. 1G.23 GRV 4 POSITION INDICATION [ Item (2) (xi)]
N L NRC Position I l
. Provide direct indication of relief and safety valve position
{ i (open or closed) in'the control room. (II.D.3) Res'ponse
$ f C Yd*l
;,, cv m y M M .19- b>Cmu~t wd pr Mec l w M h NMES 0973,&<aM ,
323 bd% ' Ts/ @ % GESS M E 5 W h} qW< l l l .. 1G.23-1/1G.23-2 n... = .. : = .. ... -- - - - _:--
v.
~
GESSAR II 22A7007 I$ 238 NUCLEAR ISLAND Rev. 17
$4. STEM AUTeMnrtC ANP Flo w /NDICAr/cg d 1G.24 AUXILIARYFEEDWATER4Af'WtgINITIATIONf[ Item (2) (xii) ] A NRC-Position t
Provide automatic and manual auxiliary feedwater (AFW) system initiation, and provide auxiliary feedwater system flow indication in the control room. (Applicable to PWR's only) (II.E.1.2)
Response
I ! This requirement is not applicable to the 238 Nuclear Island. It applies only to PWR-type reactors. l l l I l i l i 1G.24-1/lG.24-2
- T
- g *9 ewee go
-ew+++e I
GESSAR II 22A7007
.G 238 NUCLEAR ISLAND Rav. 17 i
RC).IA6fLITY ef POWER SoPfLIES FoA ^!M'WAL C /At0LAffgg 1G.25 m""J.L CIRCULT.T 05: !!! ::0T OTIJ;Dai CLiiDITIO::r- e q [ Item (2) (xiii))
- ,j J NRC Position Prbvide pressurizer heater power supply and associated motive and control power interfaces sufficient to establish and maintain natural circulation in hot standby conditions with only onsite q power available. (Applicable to PWR's only) (II.E.3.1)
J
Response
t*
- j This requirement is not applicable to the 238 Nuclear Island. It l applies only to PWR-type reactors.
i s 1
?-
i e ~ J t i e p P r r 1G.25-1/lG.25-2
*W 6 *
- e - _ _ -. ogy w e g- +
6
GESSAR II 22A7007 238 NUCLEAR ISLAND' Rav. l'7 1G.26#CC:: TAI;;;"":T ISOLATION DEPENDABILITY [ Item (2) (xiv)] NRC Position Provide containment isolation systems that: (II.E.4.2) (A) Ensure all non-essential systems are isolated auto-matically by the containment isolation system, s (B) For each non-essential penetration (exct:pt instrument lines) have two isolation barriers in series, ? s . e (C) Do not result in reopening of the containmenh isolation valves on resetting of the isolation si p al, (D) Utilize a containment set point pr sure for initiating containment isolation as low as is compatibleswith normal operation, s 3 (E) Include automatic closing on a high radiation signal for s all systems that provide a path to the environs. M Response -
's -
4t. Nw s AM # b, ya A innt& -0937r, kWE.:2. s 2- .' 2c
).a&+ /A19 a~d 4 W U W M 0!Y& W A-e h (c) d - 4 4 % 4 t w u WJ e W In c S . MAC 47g opprnsl' .
b& h nac- om, W c.f4 (tg s-n). s
< y s
'Ni s s 1G.26-l/lG.26-2
,- - e<,- , -
,m. , MdMADRI 27 A700,7
! ti
\
e i - 238 NUCLEAR ISLAND Rav. 17 (., ,' . 1G'.27 N spURGINGM [ Item (2) (xv)) NRC Position , Provide a capability for containment purging / venting designed to minimize the purging time consietent with ALARA principles for occupational exposure. Provide and demonstrate high as.surance that the purge system will reliably isolate under accident conditions. (II.E.4.4) Response \
\
- 5 s
, m.
The general safety concern.over containment purging stems from the presumptio'n that the purge line provides a path
'for accident releanr:s p;ior to isolation, and further,dhat the dynamic effects.of the' accident may inter'fere with effective isolation of the purge line; o& O M 5.3AC K These presumpti s are'not-cirectly applicable to the Mark III containment design. The'/eactor foolant fystem piping
- , is enclosed in the drywell Ubich'-cpmmunicates with the 1., containment only through the suppression pool. Releases s
from the primary system are subjected to the quenching and scrubbing action of the suppressionipool before entering the containment, so the purge systen does not provide a s path for primary system releases in i the same sense as other containment designs. Even so, special care is being taken
,[, -
in the purge system design, specifically for valve s' op r bility assurance.C t
~ '
M i l ise 5-' VMv*g au Pv4W ~.2 d ? 5-im:L F W k sk P*nb 'MbjLA mb} - Q Thq spec (ific points of are addressed below:
}- o 4k s dts N %
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n- . . 1> ., 1LnQ $ Y"v t4t)r1_ Y / '+ g gg (f 8O
- The basis for the purge syste design is justified in
'the} response to W Question 002.00 ..C ; fi.~..d aeeed LL 1.. LJ i- A ') 't ^# "r r h ; ' t """ *""""-
OM*P- g 't - Q___ ,7 5 ble $ The present 4esign provides for continuous [ urging of the contairapent during power operation at f000 cfm l through an 5t" line to reduce airborne radionuclide concent. rations to a level which permits continuous i access. This is in keeping with occupational ALARA consi'derations'<. because extensive containment access for* routine masntena.rre is required. h Psefhrmant'epPyr%\,,,dfG-.Whnh,,,WNreyMdf3sF8cp9I@
'l ~&~]^
l'; w -r p p s* j. The performance ofKee purge isolation valves has' been i evaluated and meets the requirements of PTP-CSB'6-ig g N for isolation and dependabilfty under-g). , v Rdu M l pa s% .s . j g k 4 h i N w p p a=v6. m w dbead mh- ws d' .v-- 4c w ,r e e e, a % m + A u w eq
<= ,l
- A i
<,a,p i y k A ,g -r i
l cLS 4 4 \ b.a M 0 t c~ i n W I ph ofssp.:tgE e .- w Cr C4k . ! 1 i _ ML oWm-. . a. - a - u D,_ L4 _A A__ cP 4 Y6 s+-e. L rLa ? - Jo O~ c F 4 L l- 4 d
- d d vG ~vswal h Mm34Ui U:
l l w%. L MU L6_ ._ ' A MS ~ h6 tb Auh
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3 i
e i - s0 - } GESSAR II
'I 22A7007 238 NUCLEAR ISLAND Rsv.
l]
./
l 19.3.6.28 QUESTION / RESPONSE 6.28 (480.23)
, ~ p
. QUESTION 6.28 EMSGD S 63909 M j
j You state on pages 1.8-171 and 9.4-62 of your FSAR that the i containment purge valves are open during normal operation and I I that containment pressure is controlled through these valves. l However, we state our position on this matter in branch j technical position CSB6-4 that the use of large purge and t' vent lines should be restricted to cold shutdown conditions and refueling outages. Provide.your basic for purging the containe.nt_ < nnHnunusly._in light.of our position on this matter: (6. 2.4 ) RESPONSE 6.28 1 Continuous purging of the primary containment outside of the drywell during reactor operation is required for access, I 1 inspection, and maintenance associated with the control rod drive - hydraulic control units (one per CRD); safety related instrument calibrations, water sampling of reacto'r water, suppression pool, and upper containment pools, RWCU system and feedwater., l These activities involve several operating personnel occupying the primary containment during a significant portion of each shift. The ventilation rate of 5,000 cfm provides an air change in the containment only every 3 hours and 45 minutes. This is minimal for controlling humidity, odors, and dilution of potential airborne radioactivity released from a small number of safety / relief valve vents, RWCU System filter-demineralizer maintenance, and upper containment pool walls at the wet-dry l' line. 19.3.6.28-1
b 4 GESSAR II 22A7007 s6 - 238 NUCLEAR ISLAND Rev. g [ g* 19.3.6.28 QUESTION / RESPONSE 6.28 (480.23) (Continued) D e w3 No. 44LC Sdv*VW fef Pakwdway D4.s\3w Aypnv .k,' a W_en Sr ncL M ..ic 1 "crition CCD 0 0 = initirlly irrer". PM the containmentge;...'ilatica penetrations were modified to
- reduce their size for normal operating contin ous purge from .
~a42-in.diametertoan18-inkdiameter.490se((-in. pene-g trationsesse*[ pen only during reactor shutdown and refueling to allow for higher ventilation flow rates when more operating personnel would be present and potential airborne radiation levels could be higher than during normal operation.
- . ... . a. . . = . . n.m ... . . . = . . .
- a v e.: +. :-
Fast _ closing ~is~olatii~on' valves ~M t.'Or:gprovided to { . i close on LOCA signal or whenever the exhaust radiation f sensors detected radiation levels high enough to exceed plant
- operating limits. During normal reactor operations with the 4
- 5,000 cgm ventilation rate, the airborne radiation levels must be less than 10CFR20 limits inside the primary containment but outside the drywell. The drywell purge vents are closed 1
during reactor operation. nd 0.1 is ..m...:.1 f00 11 U. "' . t ? L The radiation monitors located in the exhaust duct are { installed far enough away from the primary containment isolation valves so these valves can close before airborne radiation is released from the primary containment. As a further precaution, radiation monitors are also located near f the upper containment pool surface so early detection of Potential radioactivity can be detected during refueling when i the reactor is open. 4 . fwMgv- AMo~w e st)* Mv's e 19-M db , .{. g . 9 _ m . d t n e h
- v d o r W e m a l o y a.b Skww.*J L pq % J A s. a.s m..-u d F & Nn-c.sb F /
%vw M e v- Ftwd N A17eAdf PW 'b ., /
t k t s s.3 vs h (GSsspCII s .L a%% 4L g p e A * ~ b v p . pa4+ k- ' ~ ^ ^ " > b~M{ M Tuhdhow.c L y __
- My bvs op W M -
- - na w ovi i J a-f cA 9 e bJ j -hk 19.3.6.28-2 j i ( -
d A' GESSAR II 22A7007 y*
- t 238 NUCLEAR ISLAND Rav. M 19.3.6.28 QUESTION / RESPONSE 6.28 (480.23) (Continued) g% suppression pool clean p system hee Lmmu s volded fut L' ;
"X ',, M;.,;'a I : : ts ensuregthat radioactivity in the pool }
water can be kept low and to reduce the amount of airborne 3I radioactivity during abnormal plant operating events. I ts This same topic g also n issue contained in NUREG-0737, Item II.E.4.2, which weegaddressed in Section lA.29 " Containment Isolation Dependability." Studies of suppression pool
; scrubbing action have been found to retain considerable amounts of sodine .i_n ,the ;eventgof.a.. major.. pip.e br.eak- and -
y failure of fuel cladding. This same capability also applies [ to normal and abnormal reactor operations when potential I safety / relief valve simmering occurs, and RCIC testing releases condensed reactor steam to the suppression pool.
, All of these plant operating conditions have been studied and 3
(' discussed with the staff during the various reviews and licensing prccc& Tar - of GSSS AftTE i j Periodic isolation valve testing is required during reactor j operation to ensure that the Ventilation Isolation System is ever ready to function. This includes Appendix "J" leakage
- testing and radiation sensor test and calibration.
) t a v i y w [ . f 19.3.6.28-3 d .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 17 05SIM EVALUATOR - y 1G.28 3 [ Item (2) (xvi) ] NRC-Position Establish a design criterion for the allowable number of actuation cycles of the emergency core cooling system and reactor protection system consistent with the expected occurrence rates of severe overcooling events (considering both anticipated transients and accidents). (Applicable to B&W designs only).' (II.E.5.1)
Response
This requirement is not applicable to the 238 Nuclear Island. It applies only to PWR-type (B&W designed) reactors. l 1G.28-1/lG.28-2
~
e-M- a g e ~se M y -
.e. .w-, .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 A 00ITleNAl- A CCIPENT MJN/19%N6 IN3TRuhtSNTAr/M p 1G.29 CC::TA !=!"' r.'.rs2 TZRs I;;0TnUIC:;TATIO!: Ir CO::TRCL 200".? [ Item (2) (xvii) ] NRC Position Pr' ovide instrumentation to measure, record and readout in the control room: (A) containment pressure, (B) containment water level, (C) containment hydrogen concentration, (D) containment radiation intensity (high level), and (E) noble gas effluents at all potential, accident release points. Provide for continuous sampling of radioactive iodines.and particulates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples. (II.F.1) _ Response p& f .eaek e} &x la k QW'ye { p. g_ i I l i 1G.29-1/lG.29-2
.. * , . en .m _ _ . . , ~
.-.....~..-n.. ,.m~...n.. .- . .
GESSAR II 22A7007 238 NUCLEAR ISLANp Rev. 17 E9BMTIf! CAM 4W 0F AN/ $ECdVinf fbM CON 91 MONS V 1G.30 N.$ "^ S I TI ' b n".OCOUbCORE COOLING [ Item (2) (xviii)] NRC Position P vide instruments that provide in the control room an unambigu-
.ous indication of inadequate core cooling, such as primary coolant saturation meters in PWR's, and a suitable combination of signals from indicators of coolant level in the reactor vessel and in-core thermocouples in PWR's and BWR's. (II.F.2)
Response
& M 6 L 4a,2 ;, lA.31, CrE A d & w r g 4 Q w 'd;vd + Jaar M ~cM gfaucig M' 4 S#
g4 man& x &s is- x m -#.
;., 5 238 k% W "T'
- Th Nec M6*1d E A* # I" #
W H 'y' q )I H_ q 4 Q f&+AwhEW$ p y -3 o ce). gu no. o m i. 64A _ gj& Q s- ) p.x p kmx y g g a h.6k .44 ^4*m-r
' dr.; wM wpss* E s V y k i r,&, I G .2.9 L
I .. ? l 1G.30-1/lG.30-2
- y. .- . - . ---
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 gjtsptubtg/VTMid FDAt fMMFMtw Ace!9tMr CON!/r/MS (Af6. Sulpf fo47) y 1G.31 4[ Item (2) (xix)] NRC Position
't Il
'! Provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage. (II.F.3)
\
l L Response 4
)
t . 5 { The GESSAR II design is assessed against Regulatory Guide 1.97 Revision 2 1 in Appendix 1D with supplementary justification for deviations provided z /' I on the docket (cover letter G. G. Sherwood to D. G. Eisenhut dated April 28, 1983). The assessment including the deviations and justifica-tions are still under review by the NRC. GE agrees that the GESSAR II design will meet designated portions of the i guide; and any deviations to the remainder of the guide will be justified i to the first satisfaction of the staff prior to referencing GESSAR II by the applicant. t t h L I e j -
=
8 N' i a 1G.31-1/lG.31-2 , i t H.. ___. __. m , , . , _ ._
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
'; . pagg $OWLitj FMl HlGUmtt2iR ggLIEF VALVES, f stecK VALVf$
rn"000nI;;a $ LEVEL LN9ICAfin{ I::OTnU:-::::T.*.TI'= [ Item (2) (xx)]
. 1G.32
'I j NRC Position
, Provide power supplies for pressurizer relief valves, block valves,
'j and level indicators such that~: (A) Level indicators are powered 4 from vital buses; (B). motive and control power connections to [ the emergency power sources are through devices qualified in
'accordance with requirements applicable to systems important to l
safety and (C) electric power is provided from emergency power sources. (Applicable to PWR's only). (II.G.1) l I
Response
}
r l This requirement is not applicable to the 238 Nuclear Island. It applies only to PWR-type reactors. t t 1' l? (' i. .i 4 k 3
' 1G.32-1/lG.32-2
- - . . _ . , . , _ _y,
- GESSAR II 22A7007
.s i 238 NUCLEAR ISLAND Rev. 17 QESCRl6E AUTOMAnc & MA8vnt. AcnoNSFoM fRetER FVMNdN/M6- of
\
Auxill ARY HEAT REMwAt. systems WHEN FW.SYJWM N*T* 0fERA$LE 1G.33 f.UXILIAnY :: EAT RZiK/iAL OYCTE". DECIG;; AI;0 rn0CEOUn:C
. [ Item (2) (xxi)]
NRC Position i j Design auxiliary heat removal systems such that necessary auto-j matic and manual actions can be taken to ensure proper functioning I when the main feedwater system is not operable. (Applicable to l BWR's only). (II.K.l.22)
Response
i 2Xe & or n @ y A~t ~d @ ch <- y a w A M s L J 6 /A ,37. & 4es u w a ruees- 0179, lact-, s-s 2, gg y y s --1 6 4 r-17. t
= . 1G.33-1/lG.33-2
_.L.;;:q:,1~_';r.=r. ..a=.......
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 17 l A$ALY3IS OF U9SRA9tMS L 1G.34 4MiliArAOF INTEGRATED CONTROL SYSTEM [ Item (2) (xxii)] NRC Position i I Perform a failure modes and effects analysis of the integrated
- l; control system (ICS) to include consideration of failures and
~.
6 effects of input and output signals to the ICS. (Applicable to j B&W-designed plants only). (II.K.2.9) I . I
! Response i It i This requirement is not applicable to the 238 Nuclear Island. ) applies only to PWR-type (B&W designed) reactors.
I
't l
i I s t t 4 f. s i I t If N. 1G.34-1/,1G.34-2 l l i
!.t.
r--- -- m ,y-,~,.- ,_
.~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
~
1 HAND-WMED SAFErf-GMAPE rg
' Item (2) (xxiii)] ,
1G.35jjANTICIPATORY REACTOR TRI NRC-Position 1 Provide, as part of the reactor protection system, an anticipatory i reactor trip that would be actuated on loss of main feedwater and on turbine trip. (Applicable to B&W-designed plants only). (II.K.2.10)
Response
This requirement is not applicable to the 238 Nuclear Island. It applies only to PWR-type (B&W designed) reactors. i i 1G.35-1/lG.35-2 e
=
GESSAR II 22A700.7 238 NUCLEAR ISLAND Rav. 17 I i; CEMntAL REco@W6- ' 1G.36 POST = Reef 9fftt? A WATER LEVEL g litSTfMGMENPA9fett [ Item (2) (xxiv) ] NRC Position j l Provide the capability to record reactor vessel water level in one j location on recorders that meet. normal post-accident recording
-l requirements. (Applicable to BWR's only) . (II.K.3.23) -
i
- } Response
& W' ~_-.-
, y g -n u.t % >> M d a M A i jp.1 3 y. d.% caul ~I & a ffn M A'W
! 2 sum;A/ i., Af c A e@ l 0-s
.1 L
e '\ ,s t i s I i -5 i 1G . 36-1/ lG . 36-2 l' .
,,3
.;m r- - , - r.. ,7 -- -- -- ;. ,- - s7--._... .
GESSAR II 22A7007 l R3v. 17 ~ 238 NUCLEAR ISLAND - I' Uf6hf6 LICENSE EMcR&EM)f S0!MRT )Dictiny &' 1G . 37, O!!CITZ OUTIGRT CL;I:nO 4 [ Item (2) (xxv)]
.j
- I NRC Position N Provide an onsite Technical Support Center, an onsite Operational Support Center, and, for construction permit applications only, a nearsite Emergency Operations Facility. (III.A.1.2)
I I i Response 6 -
, afL .
{ ty & s & f- 'I l i. a_ r e 4
}
e v 5 Y I I
?
a s I
! 1G.37-1/lG.37-2
....---- ----.- . . . a _ . 2_ ;p . __ _ ._ _ .,_ , . , . . . . . . , _ . ,
5 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 PRIMMY CoolENT SOURCES - THE p 1G. 38 LCAKAGE CC::"'nOL ""OCP'.". POn CCT"3E OUTSIDE4CONTAINMENT SMucTv4E [ Item (2) (xxvi)] NRC-Position Provide for leakage control and detection in the design of systems outside containment that contain (or might contain) TID 14844 source term radioactive materials following an accident. Appli-t
- cants shall. submit a leakage control program, including an initial I
. test program, a schedule for retesting these systems, and the i
j actions to be taken for minimizing leakage from such systems. The j goal is to minimize potential exposures to workers and public, and
; to provide reasonable assurance that excessive leakage will not i prevent the use of systems needed in an emergency. (III.D.l.1) filc Ja.bf dd hts d k W
Response
4-I A . 'l'l . NRc & l" )'v~1 is W"U WW) jedCOn 1' 3- 4 (fy M b - p TI^I 5 i l h s, 1 1G.38-1/lG.38-2
.f--
cr _ , .
... :e_- qs. , _. . . . , , . . . . . - .. .
GESSAR II 22A7007
- I, 238 NUCLEAR ISLAND Rnv. 17
=
5 l AqdNITDRWG-
- 1' 1G.39 INPLANT RADIATION '"O AIROOn"I: "J IO.*.C"TJITT i.
z_7 [ Item (2) (xxv11)]/r ' l i il NRC-Position i 1
; Provide for monitoring of inplant radiation and airborne radio-acitivity as appropriate for a broad range of routine and accident conditions. (III.D.3.3) 1 h.
I
; Response h f k h Nf $W i
l ImN6 W' t b [
}
3 i i i i k i i 'k k i i t k e t I ' r. I t I 5 1G.39-1/lG.39-2 l P t L . ---. , .; . -. - .. . ,~,. --~,---,.-----v .- - - . . - - - .
"- - GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 c
4 j - 1G.40 CONTROL ROOM HABITABILITY [ Item (2) (xxviii)] 2 .1 I<, NRC Position O
'.[l Evaluate potential pathways for radioactivity and radiation that y
- i may lead to control room habitability problems under accident
.k y conditions resulting in a TID 14844 source term release, and make -l necessary design provisions to preclude such problems. (III.D.3.4) s
Response
l p ; W /A 7 9 , W&} s '"' M ^'1*"*' j y
.; y 4 p , o r me>~<
l! l
%Li A -
ii I l i c
?
)
t I Ii 3 !i k l t i i L f 1 I t t i i t I k
>m _
l t l l l l ) 1G.40-1/lG.40-2 ( l __e-a
. . _ , , ...e _ , .
-.-.+++me
^ emme38-= * ,e e =a. -
~
GESSAR II 22A7007 238' NUCLEAR ISLAND Rav. 17
]
PROCEpoQES Fe4 FEEP64cX of ofEMA T7N(rj 1 pgst6 N AND CJNJ740CYled EXfERIENet aII'ISTraTI'." P".OCEDURE5 FOR EVEUA11HG EXPERIn::CE r t 1G.41 [ Item (3) (i) ] 1 NRC Position li Provide administrative procedures for evaluating operating. design and construction experience and for ensuring that applicable important industry experiences will be provided in a timely manner 1 2 to those designing and constructing the plant. (I.C.5)
.t .; Response
[h A N/ f.-
} fr . wf-5 l
5 t f 1
)
. .I
+_
6 e t 4
+
t j 1G.41-1/lG.41-2 i P i g a
! , :w-
----+ -
.=t;2..res_ - ,. , , - ,
GESSAR II 22A7007 i.! 238 NUCLEAR ISLAND Rev. 17 M
'] EXPAMP QA usr ; ;j 1G.42 EX"?l!OCO 0"?.LITI A55Url:CE ' 07. , I.IOT 3 [ Item (3) (ii) ] p 21 NRC Position
- p. .
,j Ensure that the quality assurance (QA) list required by criterion II, App. B. 10 CFR Part 50 includes all structures, systems, and components important to safety. (I.F.1) 1 I
Response
t r As d s s cuss tel s'd 5Ast Aow t % l . 2. y l M e identification of safety-related structures, systems, and /-
.j components (Q-list) to be controlled by the quality assurance program is the responsibility of the Applicant. The Applicant I
i will supp(lement N EC Q.au andhowclarify 7 6 0its 3 )0-list
. in accordance with Ques-j tion 17.3A The appropriate items will be added to Table 3.2-1.
l Therefiningitemswillbesubjecttothepertinentrequirements l of GE's and/or the Applicant's QA programs unless otherwise
! justified.
h,
- e.
i. t i li.tt i$ I 1 j 1G.42-1/lG.42-2 I I l A ,. . n m~~w- _ - - - - - _ -- _. ems m_am_,_ .
-~
;d GESSAR II 22A7007
{-j 238 NUCLEAR ISLAND Rsv. 17 1 ; PEVELef MORE j 1G.43 ADETAILED QA CRITERIA [ Item (3) (iii) ] 9'5 SI
;f d NRC Position i} Establish a quality assurance (QA) program based on consideration
}.- of: (A) Ensuring independence of the organization performing
~
checking functions from the organization responsible for perform-ing the functions; -(B) performing quality assurance / quality control functioning at construction sites to the maximum feasible extent; (C) including QA personnel in the documented review of and concurrence in quality related procedures associated with
.j a
design, construction and installation; (D) establishing criteria f for determining QA programmatic requirements; (E) establishing qualification requirements for QA and QC personnel; (F) sizing the QA staff commensurate with its duties and responsibilities;
;, (G) establishing procedures for maintenance of "as-built"
! documentation; and (H) providing a QA role in design and analysis
! activities. (I.F.2)
Response
r I. l .- % 1 L A TER i i f
? ~
t ( I ii l 'i i t i s ..I
. 1G.43-1 ...
A 'i i*..-.,-... - _ . - - - - -
,j7
? M A:@ M ~lG5 238 NUCLEAR ISLAND Rev. 17 j DEDICATED .CO'NTNINNEN'$ PENETRATIONS p ,3;. ... {0} (i;l }
/' ..'t
- 1G.44 ASN8W 3 - 8"- 0 0T' 0\ AM*1s(L opts.Js se C NRC Position Provide one or more dedicated containment penetrations, equivalent in size to a single 3-foot diameter opening, in order not to pre-
~
clude future installation of systems to prevent containment failure, such as a filtered vented containment system. (II.B.8)
Response
, i ,
i o M Nbk k & RVW l I
' l bmt ! . i Su Fmal DM u A, ,n6v d ,. ', GGh
, .c %, A- A Yo .
, l ! .
O M E M Op bP [_e_Y l h6 84. lI ! l . l3 D NA A\ w. 4 WAM a.m 6w; % 4
!; is i _
ui ,
! b v'% k o\\ s- g Aa G ok h o n tved AJ +M cu.L. .
eku kJ W T' G %i-b'w3
% 7J 3 W l :
! mdv J : 1 l '" l - ;
!a v a- A L o.L J cAo 6.1 AL hiv J.
i i ~ ! l. : J o r *jvv G o W t
&d 4have t
beeu dedic_afed s J
. i i. ! l _
i
~
&n. l- bc- -
/
' j i .
e
- , i -
a : t , WL M i
! C.Ok Aie 3 Q_ @ _d5__
', I i d5
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i J fle_J b\ s L I iA , Wm L._J LL -W i
) \ ! H-
~
i e- = * - *w* he GUSEMI - l+. 4L3mc,- Ap2a 1G.44-1/lG.44-2
22A7007 GESSAR II 238 NUCLEAR ISLAND Rav. 17 1G.45 CONTAINMENT INTEGRITY [ Item (3) (v) ] NRC Position a Provide preliminary design information at a level of detail con-sistent with that normally required at the construction permit stage of review sufficient to demonstrate that: (II.B.8) (A) (1) Containment integrity will be maintained (i.e., for steel containments by meeting the requirements of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsubarticle NE-3220, Service Level C Limits, except that evaluation of instability is not required, con-sidering pressure and dead load alone. For concrete containments by meeting.the requirements of the ASME Boiler Pressure Vessel Code, Section III, Division 2 Subsubarticle CC-3720, Factored Load Category, consider-ing pressure and dead load alone) during an accident that releases hydrogen generated from 100% fuel clad metal-water reaction accompanied by either hydrogen burning or the added pressure from post-accident inerting assuming carbon dioxide is the inerting agent. As a minimum, the specific code requirements set forth above appropriate for each type of containment will be met for a combination of dead load and an internal pres-sure of 45 psig. Modest deviations from these criteria i wiIl be considered by the staff, if good cause is shown
/ GM yAapplicant. System.s necessary to ensure containment i t'egrity shall also be demonstrated to perform their function under these conditions.
(2) Subarticle NE-3220, Division 1, and subarticle CC-3720, Division 2, of Section III of the July 1, 1980 ASME Boiler and Pressure Vessel Code, which are refer-enced in paragraphs (f) (3) (v) (A) (1) and (f) (3) (v) (B) (1) 1G.45-1
GESSAR II 22A7007
, 238 NUCLEAR ISLAND Rav. 17 1G.45 CONTAINMENT INTEGRITY [ Item (3) (v)] (Continued)
/0CFAlSC 34 of thir certier, were approved for incorporation by reference by the Director of the Office of the-Federal Register. A notice of any changes made to the material incorporated by reference will be published in the Federal Register. Copies of the ASME Boiler and Pres-sure Vessel Code may be purchased from the American Society of Mechanical Engineers, United Engineering Center, 345 East 47th St., New York, NY 10017. It is also available for inspection at the Nuclear Regulatory Commission's Public Document Room, 1717 H St., NW.,
Washington, D.C. (B) (1) Containment structure loadings produced by an inadvertent full actuation of a post-accident inerting hydrogen control system (assuming carbon dioxide), but not including seismic or design basis accident loadings will not produce stresses in steel containments in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsub-article NE-3220, Service Level A Limits, except that evaluation of instability is not required (for concrete containments the loadings specified above will not produce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Section III, Division 2, Subsubarticle CC-3720, Service Load Category, (2) The containment has the capability to safely withstand pressure tests at 1.10 and 1.15 times (for steel and concrete containments, respectively) the
- pressure calculated to result from carbon dioxide inerting.
l I - lG.45-2 i
' GESSAR 11 22A7007 -
- 238 NUCLEAR ISLAND Rev. 17 IG.45 CONTAINMENT INTEGRITY [ Item (3) (v) ] (Continued)
.: Response (A)Tbe ,
c.ow4=se 6 F da.3gw b a3if gy 4 15f.st3. A pp e dtv G o f S wb .s e. 5+to w. g5 0 3 I (.PRA) pvovtAno -4 k corre.3powd % f e e&4 a. m %i up=WL]hr %d I % e hows - Nm+.s.el\
! a~ .]I$ eo%v.
.+ ) b
- 4-Le L,.2 F4 s - u to-a+ o g 1 -
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22A7007 GESSAR II Rev. 17
; 238 NUCLEAR ISLAND 9EPICATEP fEME7MAridW 3 , 1G.46 EXTED N *'0".^"EN ""CO".S INE Sx [ Item (A) i)] g-NRC Position For plant designs with external hydrogen recombiners, provide rdundant dedicated containment penetrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere. (II.E.4.1)
Response
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l e-' GESSAR II 22A7007 i , 238 NUCLEAR ISLAND Rev. 17 4 1G.47 CR&M/2fi/6H AND 3rMFF/M6-
"'"ECS:~;;I FLRii141:!C 70 CVERSEE AI;D FRGCEDURE5 FOR FLA;;T- DESIGN AND CONSTRUCTION [ Item (3) (vii) ] Z- , NRC Position 4
1-l Provide a description of the management plant for design and I construction activities, to include: (A) the organizational and I management structure singularly responsible for direction of [ design and construction of the proposed plant; (B) technical i resources director by the applicant; (C) details of the inter-action of design and construction within the applicant's organiza-tion and the manner by which the applicant will ensure close integration of the architect engineer and the nuclear steam supply vendor; (D) proposed procedures for handling the transition to operation; (E) the degree of top level management oversight and technical control to be exercised by the applicant during design and construction, including the preparation and implementation of procedures necessary to guide the effort. (II.J.3.1)
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y - x e ATTACHMENT NO. 2
- i i
v l'. W i b l DRAFT OF GESSAR II AMENDMENT NO.17 RESPONSE TO SRP RULE i I
k ! GESSAR II 22A7007 { 238 NUCLEAR ISLAND Rsv. ,kW l'? 1 SECTION 1.8 i CONTENTS Section Title Page ~ l wtTH STmwDat) REMW PLAN 1.8- CONFORMANCE 1.8.0-1 - n f:': *., p k?"h.a e-U m. b * *^% I'..I' RYhI1YaYoryGu5$'l R sion 0, T 7 Dated November, 1970 1.8.1-1 , 1.8.2 Regulatory Guide 1.2, Revision 0, Dated November, 1970 1.8.2-1 1.8.3 Regulatory Guide 1.3, Revision 2, Dated June, 1974 1.8.3-1 1.8.4 Regulatory Guide 1.4, Revision 2, , Dated June, 1974 1.8.4-1 1.8.5 Regulatory Guide 1.5, Revision 0, Dated March, 1971 l . 9
- 5-1 1.8.6 Regulatory Guide 1.6, Revision 0, Dated March, 1971 1.8.6-1 1.8.7 Regulatory Guide 1.7, Revision 2, Issued in 1978 1.8.7-1 1.8.8 Regulatory Guide 1.8, Revision 1-R, Dated May, 1977 1.8.8-1 1.8.9 Regulatory Guide 1.9, Revision 2, Dated December, 1979 1.8.9-1 1.8.10 Regulatory Guide 1.10, Revision 1 (Withdrawn July,1981) 1.8.10-1 1.8.11 Regulatory Guide 1.11, Revision 0,
[ Dated March 1971 and Supplement, dated February, 1972 1.8.11-1 1.8.12 Regulatory Guide 1.12, Revision 2, Dated July 1981 1.8.12-1 1.8.13 Regulatory Guide 1.13, Revision 1, Dated December, 1975 1.8.13-1 1.8.14 Regulatory Guide 1.14, Revision 1, Dated August, 1975 1.8.14-1 1.8-i M** 't"* age .- m--,,-, ,-.m..
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V I 1.8 CONFORMANCE WITH THE STANDARD REVIEW PLAN [ 1.8.0 Purpose The purpose of this section is to provide an evaluation of the GESSAR II design against the Standard Review Plan (NUREG-0800) as required by 10 CFR 50.34(g). [ Since the NRC regulatory guides are an integral part of NUREG-0800, this section l also shows the consistency of the design with the regulatory guides. i , 1.8.0.1 Differences from Standard Review Plan Since the GESSAR II design scope is limited to the Nuclear Island, there are Balance of Plant (BOP) portions that are the responsibility of the Applicant. In addition, the Applicant is responsible for information within the scope of the ! Nuclear Island that will not be available until GESSAR II is utilized by an Applicant, j Finally, GE has chosen for commercial reasons to delay the submittal of certain f information until the first Applicant references GESSAR II. All of this information is presented in Tables 1.9-1 through 1.9-19. Hence, it is not possible at this time to demonstrate that the GESSAR II design satisifies all of the NUREG-0800 requirements. However, GE has reviewed the GESSAR II design against the revelent portions of NUREG-0800, and concludes that it meets the applicable acceptance criteria, except as noted in Table 1.8.0-0. The cited references include evaluations that describe the basis by which GE concludes that the underlying requirements are satisified. The Applicant will provide a summary of deviations from NUREG-0800 for those plant design features covered by the GESSAR II/FSAR interface Tables 1.9-1 through 1.9-19 with corresponding evaluations that describe the basis by which the Applicant ! concludes that the underlying requirements are satisified. l l l l l l l l
'2.
l 1.8.0-1 /l. 9.0 W. .= . : - . c: . =. . . = -
- . ~~ .
7 _ _ _ - - _ _ - _ _ _ . . _ ,_ _ _ _ _ . _ . . __ , . . - .. .. ,,u.___.._. . , _ , _ _ __ _ _ _ _ _ . , if [ Tabla 1.8.0-0 .-
~
SUPMARY OF DEVIATIONS FROM NUREG-0800 NUREG-0800 NUREG-0800 GESSAR II ( SECTION CRITERIA DIFFERENCE StBSECTION l 3.7.1 II.1.b - Design time history and damping For higher damping values, the response 19.3.3.48 (Rev. 1) values criteria. spectra from synthetic time history are i not in agreement with the enveloping values of the criteria.
} 3.7.3 II.2.b - Determination of number of For equipment and components other than 3.7.3.2.2 (Rev. 1) OBE cycles. piping. 10 rather than 50 peak OBE stress c cycles are used.
J
?
lO 1 N \ b
Tablo 1.8.0-0 (c:ntinu;d) SUMARY OF DEVIATIONS FROM NUREG-0800 NUREG-0800 NUREG-0800 GESSAR II SECTION CRITERIA DIFFERENCE S WSECTION 4.2 II.A.1.(b) - Sets Limit on the number of NEDE-24011 sets a more conservative 4,2,1 (Rev. 2) strain fatigue cycles. limit than that in the SRP, 1 l 4.2 II.A.1.(c) - Fretting wear of structural Wear limits are not stated 4,2,1 f, (Rev. 2) members shoQld be stated. 4.2 II.A.I.(g) - States that " Worst case Design basis allows up to 0,52 4.2.1 (Rev. 2) hydraulic loads" may not exceed the "l i f t-o f f" . i hold down capability of the fuel ass'y. I 4.2 II. A.2.(e) - Prohibits any fuel melting Design basis allows fuel melting that 4.2.1 (Rev.2) is not " excessive", 4.2 II. A.2.(g) - Specifies uniform strain Elastic strain not included in the 1% 4.2,1 (Rev. 2) (elastic & plastic) limit of 1%. limit. , 4.2 II. A.2.(1) - Limits applied stress to S Topical Report is under.feview. 4.2.1
- (Rev. 2) than 90% of the irradiated yield stress.
4.2 II. A.3.(e) - Analytical procedure are Topical Report is under review. 4.2.1 (Rev. 2) prescribed. 4.2 II.B-Lists parameters to be included Fuel description does not include all 4,2.1 (Rev. 2) in fuel description. parameters listed in SRP, 4.2 II.C.3.(a) - Lists models to be included Gadolinia fuel properties not appropriate 4.2.2 (Rev. 2) in thermal calculations. in model. 4.2 II.C.3.(d) - Describes acceptance Topical Report under review 4.2.2 (Rev. 2) criteria for design evaluation.
Table 1.8.0-0 (c:ntinued) 4
SUMMARY
OF DEVIATIONS FROM NUREG-0800 -} NUREG-0800 NUREG-0800 GESSAR II SECTION CRITERIA DIFFERENCE StBSECTION a 5.2.3 II.3.b.(1)(a) - Welding procedure Minimum preheat and maximum interpass 5,2,3,3,2,1 q!* (Rev. 2) qualification. temperature not specified, d d 5.2.3 II.3.b.(3) - Regulatory Guide 1.71, Alternate position employed, 5.2,3,4,2,3 (Rev. 2) Welder Qualification for Areas of Limited Accessibility. i s' ' 6.2.1.1.C II.9 - Compliance with NUREG-0783. GESSAR II analysis takes credit for 19,3,6,10 (Rev. 5) ' weir wall annulus water, (. Comparison to Section 5,7,1 of N.UREG-0783) 6.2.1.2 II.B.1 - Humidity for shield wall 1% relative humidity used in analysts, 19,3,6,14 j (Rev. 2) annulus analysis.
?
6.3 III.19 - Operator action following GESSAR II require operator action 19,3.6,56 (Rev. 1) LOCA. within 10 minutes for some events, 9 6.7 II.1 - MSIV leakage control meeting Exception taken to Fosition C,9 of 1,8,96 i (Rev. 2) Regulatory Guide 1.96. Regulatory Guide 1,96,
) 7.1 II - Regulatory Guide 1.75 (Table 7-1). Alternates to portions of R.G,1,75 are 7,1,2,10,18 (Rev. 2) utilized.
7.2 II.1 and II.2 - IEEE 279-1971 and Some RPS inputs come from devices mounted Table (Rev. 2) GDC 2. on non-seismically qualified equipment 19,3,7.14 1(,j) and/or are located in non-seismically qualified enclosures.
Ji d Table 1.8.0-0 (ctntinuid) Suf#tARY OF DEVIATIONS FROM NUREG-0800 l ' ' ' NUREG-0800 NUREG-0800 GESSAR II
] SECTION CRITERIA DIFFERENCE S WSECTION 1 7.3 II - TMI Item II.K.3.21: Core Spray and LPCI systems do not 1A.63 (Rev. 2) Restart of Core Spray and Low-Pressure automatically restart after on low -)
Coolant Injection Systems (Table 7-2) water level if the initiation signal is a syll. 7.3 II - Paragraph 4.17 of IEEE 279 HPCS, LPCS, LPCI, ADS, and the 19.3.7.42 (Rev. 2) containment spray mode of RHR share common
. .i interlocks between the automatic and manual initiation modes.
i
; 7.5 II - Regulatory Guide 1.97 Exception taken to some of the Appendix ID.
(Rev. 2) (Table 7-1) requirements. q 8.3.2 BTP PSB-1 Section 1.(c).(3) - Second GESSAR II design based on maximum 19.3.8.5 l (Rev. 2) level of undervoltage protection for fluctuation of + s% on grid voltage. Class IE equipment. l 9.5.1 II.2.a- Implementation of fire protection Lack of 3-hour-fire-rated dampers in 9.5.1.1
)< (Rev. 3) program in accordance with BTP CMEM 9.5-1. ventilation system.
1
)
12.1 II.2 - Instructions to designers and No specific instructions provided. 12.1.2.2.1
, (Rev.2) engineers regarding ALARA.
- i. 12.2 II.6 - Contained source descriptions. Size and shape of vessels with 12.2.1.1 (Rev. 2) contained sources not provided.
12.2 II.6 - Buildup of dctivated containment Buildup of activated corrosion products 12.2.1.2.7.2 (Rev. 2) sources. provided only for recirculation piping. l
a
- 1
[] Table 1.8.0-0 (c:ntinu::d) j SLMiARY OF DEVIATIONS FROM NUREG-0800 1
! NUREG-0800 NUREG-0800 GESSAR II SECTION '
CRITERIA DIFFERENCE SLSSECTION 15.3.3 - II.8 - Use of non-safety grade equipment. Credit is taken for non-safety grade 15.3.3.2.2 15.3.4 equipment and failure of non-safety (Rev. 2) grade equipment is not assumed. 15.3.3 - II.10 - Coincident loss of offsite Not analyzed with coincident loss of 15.3.3.2.2 15.3.4 power. offsite power. (Rev. 2) f j 15.4.4 - II.2.(b) - Fuel cladding integrity. MCPR not calculated. 15.4.4.3.2, a 15.4.5 15.4.5.3.2.1 & J (Rev. 2) 15.4.5.3.2.2 15.6.5 II.(2) - Distribution of Iodine Radiological analysis for LOCA assumes Part Ib to 3 Appendix B Inventory. 25% of Iodine is in suppression pool. 19.3.5.1 j ( Rev. 1) 4 e l e 1 e
GESSAR II 22A70 238 NUCLEAR ISLAND Rev. [07lg 1.8 CONFORMANCE TO NRC REGULATORY GUIDES __
The purpose of this section is to show that the design of the 238 Nuclear Island is consistent with the requirements of the regulatory guides issued by the NRC.
~
f N 1,9 O.2.Cowsssk*wC-h # b NEC D "(^4D Ano w The NRC (AEC) began in 1970 to issue regulatory guides (safety guides) whi6h state, in detail, methods acceptable to the NRC staff of meeting applicable. Federal Regulations. Since that time, new and revised regulatory guides have been issued on an on going basis. During the construction permit stage, GE agreed in GESSAR PDA to comply with the appropriate regulatory guides issued through March 1, 1974 (Regulatory Guides 1.1-1. 75) , plus Regulatory Guides ( 1.76, 1.89, and 1.96. For the FDA, however, GE elected to base GESSAR II on compliance with all regulatory guides in effect as
, of the date of docketing. Therefore, Regulatory Guides 1.1 through 1.150, 8.8 and 8.19 with the revisions in effect as of February 22, 1982 are applicable to GESSAR II.
Table 1.8.0-1 lists these regulatory guides (and revisions) used as design bases and defines the GESSAR II subsection that describes the manner in which the design complies with the applicable regu-latory guide. l l - sa
@ 0-1/l.8,o_,
GESSAR II 22A700 238 NUCLEAR ISLAND Rnv. 37 1.8.122 Regulatory Guide 1.122, Revision 1, Dated February 1978
Title:
, Development of Floor Design Spectra for Seismic Design of Floor Supported Equipment in Components This guide describes methods for developing design response spectra at various floors or other equipment support locations of interest from the time-history motions resulting from the dynamic analysis of the supporting structures. Evaluation GE complies with all guidelines of the regulatory guide except Position C.2 where, instead of using 15 percent in frequency for
~
spectrum broadening, GE uses 10 percent. Justification of this 3.7 2.9 exception is provided in GESSAR II Subsection 3_'_?_2 which ~ illustrates the conservative assumptions that have been included in the calculation of the floor response spectra. O
.l.8.122-1/1.8.122-2
() D D. . 0 s,- 7 g(% - i Table 1.9-1 i CHAPTER 1
- - GESSAR II/FSAR INTERFACES (Continued)
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l} r ^ lf q , , . , Table 1.9-4 . CHAPTER 4 GESSAR II/FSAR INTERFACES ITEM RELATED NO. INTERFACE SUBJECT DESCRIPTION PAGE SUBSECTION QUES'rION CATEGORY 4.1 Core Loading Provide fuel designations and 4.3-6 Table 4.3-1 3 Pattern number loaded for the reference core loading pattern. ] 4.2 Core Loading Pattern Provide reference core loading 4.3-9a Figure 4.3-1 3 pattern figure. l 4.3 Loose Parts Describe monitoring equipment and 4.4-10 4.4.6.1 1 g procedures to be used to detect ta excessive vibration and the m g occurrence of loose parts per go e R.G. 1 70 Subsection 4.4.6 % od d . C, . CM
- 6. t33 Om 1 4.4 Safety Provide information for safety 4.4-19 Table 4.4-6 3 p1 >
Injection Lines injection lines. gW
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J Table 1.9-15 CHAPTER 15 GESSAR II/FSAR INTERFACES (Continued) ITEM ' RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 15.8 Rod Withdrawal Provide results of either the 15.4-7 15.4.2.3.2 3 Error generic or plant specific rod withdrawal error event per R.G. 1.70, Chapter 15. 15.9 Misplaced Analyze the misplaced bundle 15.4-17 15.4.7.3 3 i Bundle accident using the plant specific Accident core configuration per R.G. 1.70, 4 Chapter 15. N 15.10 Dispersion Data Provide site boundary and low 15.4-37 Table 15.4-12 4 $ g Control Rod population zone distances, using e Drop both design and realistic assumptions in the control rod Egp3
} n en ; drop accident.
[ 15.11 Dispersion Data Provide site boundary and low, 15.4-37 Table 15.4-12 4 W g N a control Rod population zone distances, using ww Drop both design and realistic in assumptions in the control rod drop accident. 15.12 Dispersion Data Provide site boundary and low 15.6-31 Table 15.6-2 4 Steamline Break population zone distances, using both design and realistic assumptions in the steamline break accident. 15.13 Dispersion Data Provide site boundary and low 15.6-37 Table 15.6-7 4 l LOCA population zone distances, using w
; botn design and realistic M
assumptions in the loss-of-coolant . D accident. o
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GESSAR II 22A7007 238 NUCLEAR ISLAND R v. l} 3.7.3.2 Determination of Number of Earthquake Cycles
\
(} 3.7.3.2.1 Piping ' Fifty (50) peak OBE cycles are postulated for fatigue evaluation 4 N 3.7.3.2.2 Other Equipment and Components _ y /W/427 ':: - 3. 'T. 3. 2.t. To evaluate the number of cycles engendered by a given earthquake, a typical Boiling Water Reactor Building reactor dynamic model was excited by three different recorded time histories: May 18, 1940, El Centro NS component, 29.4 sec; 1952, Taft N69' W component, 30 sec; and March 1957, Golden Gates 89'E component, 13.2 sec. The modal response was truncated so that the response of three different frequency bandwidths could be studied, 0+-to-10 Hz, 10-to-20 Hz, and 20-to-50 Hz. This was done to give a good approx-imation to the cyclic behavior expected from structures with dif-ferent frequency content. t
.) Enveloping the results from the three earthquakes and averaging the results from several different points of the dynamic model, the cyclic behavior given in Table 3.7-51 was formed.
Independent of earthquake or component frequency, 99.5% of the i stress reversals occur below 75% of the maximum stress level, and l 95% of the reversals lie below 50% of the maximum stress level. ,o l l In summary, the cyclic behavior number of fatigue cycles of a component during an earthquake is found in the following manner: (1) the fundamental frequency and peak seismic loads are l found by a standard seismic analysis (i.e., from eigen extraction and forced response analysis) ; (2) t-3.7-34
- m m
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. - .w,-- -
' INSERT 3.7.3.2.2 The SRP 3.7.3 criteria II.2.b recommends that at least one safe shutdown earthquake (SSE) and five operating basis earthquake (OBE) should be assumed during the plant li fe. It also recommends that a minumum of 10 maximum stress cycles per earthquake should be assumed (i.e.10 cycles for SSE and 50 cycles for OBE.) For equipment and components other than piping,10 peak OBE stress cycles are postulated for fatigue evaluation based on the following justification.
l l
.- - rx , - - _ _ _ _ . _
,y.
GESSAR II 22A7007
, 238 NUCLEAR ISLAND Rev.
4.2 FUEL SYSTEM DESIGN See Appendix A, Section A.4.2 of Reference 1. 4.2.1 Design Bases See Appendix A, Subsection A.4.2.1 of Reference 1. 4 TNSE1t.T 4. 2.1,1 - 4.2.2 Description and Design Drawings See Appendix A, Subsection A.4.2.2 of Reference 1.
--+ t u sser 4. 2, 2 ~]
J 4.2.2.1 Control Rods The control rods perform the dual function of power shaping and reactivity control. A design drawing of the control blade is seen in Figure 4.2-1 and 2. Power distribution in the core is controlled during operation of the reactor by manipulating selected pat. terns of control rods. Control rod displacement tends to counterbalance steam void effects at the top of the core and results in significant power flattening. The control rod consists of a sheathed cruciform array of stainless steel tubes filled with boron-carbide powder. The control rods are 9.868 in. in total span and are separated uniformly throughout the ff* core on a 12-in. pitch. Each control rod is surrounded by four fuel assemblies. The main structural member of a control rod is made of Type-304 stainless steel and consists of a top handle, a bottom casting with a velocity limiter and' control rod drive coupling, a vertical cruciform center post, and four U-shaped absorber tube sheaths. The top handle, bottom casting, and center post are welded into a single skeletal' structure. 4.2-1
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l INSERT 4.2.i Acceptance Criterion II. A.I.(b) of SRP Section 4.2 requires that the cumulative number 6f strain fatigue cycles on the structural members of the fuel system should be significantly less than the design fatigue lifetime, which is based on appropriate data and includes a safety factor of 2 on stress amplitude or a safety factor of 20 on the number of cycles. The design limit for fatigue cycling in Ref. I has the following limiting condition: Actual time at stress , Actual number of cycles at stress 4 Allowable time at stress Allowable number of cycles at stress ' 1.0 Since the Ref. I limit is more conservative than that of the SRP, the deviation is acceptable. Acceptance Criterion II. A.1.(c) requires that the allowable fretting wear on major structural members of the fuel assembly be stated. The GESSAR II fretting wear design basis design for fuel system components (letter, Charnley to Staff, Jan. 25,1983) is: the fuel assembly components. This statement plus the discussion on fretting wear in Section 2.6.3 of Ref.1 show that fretting wear is considered in the design analysis and the intent of the SRP is met. Acceptance Criteria II. A.I.(g) of SRP Section 4.2 requires that the fuel assembly hold down capability (gravity and sprimp) exceed the worst-case hydraulic loads for normal operation, which includes anticipated operation occurences. The GESSAR II design limit for fuel assembly lift off is 0.52" as documented in NUREG-0979. This limit was calculated to be the largest which would not permit sufficient lateral l displacement of the fuel assembly to result in control blade interference. Since c6ntrol blade interference is prevented, this design limit is acceptable. 1 l Acceptance Criterion II.A.2.(e) states that for normal operation and anticipated cperational occurrences centerline melting of the fuel is not permitted. The GESSAR II d2 sign basis for fuel pellet overheating (Letter, J.S. Charnley to NRC Staff, January 25,1983)is: the fuel rod is evaluated to ensure that fuel rod failure due to excessive fuel melting will not occur during steady - state operation. This design limitation clearly shown that the GE design objective is to avoid fuel failures due to fuel melting and thus meets the intent of the SRP criterion.
1 t INSERT 4.2.1 (Continued) Acceptance Criterion II.A.2.(g) states that the uniform fuel cladding strain (plastic & elastic) should not exceed 1.0% (steady-state creepdown and irradiation growth are excluded). The Reference 1 model for evaluation of the 1% strain limit does not include elastic strain. The basis for the model is contained in Appendix A. Subsection A.4.2.1 of Reference 1. Acceptance Criterion II.A.2.(i) limits the applied stress on the cladding to 90% of the irradiated yield stress at the appropriate temperature. The mechanical fracture analysis for the GESSAR II fuel design is given in a topical report on the LOCA and SSE loads evaluation, NEDE-21175-3, which is currently under evaluation by the NRC staff. In NEDE-21175-3, the maximum externally applied load on the fuel cladding is determined to be less than 60% of the irradiated ultimate tensile strength at the appropriate temperature. The cladding design is thus concluded to be adequate in terms of resistance to mechanical fracturing. Acceptance Criterion II. A.3.(e) describes analytical procedures for the determination of fuel assembly structural deformation. The GESSAR II feel assembly structural analysis is described in Topical Report NEDE-21175-3-P. In this report, each major fuel assembly component part is shown to be functionally adequate to withstand the separate and combined peak loadings from the dynamic and LOCA blowdown events without experiencing structural failure. _ . _ -y . .,,; _ s_ .
, _ _ _ , , m ._. . .._
INSERT 4.2.2 Acceptance Criterion II.B. lists design parameters and drainage to be included in the fuel system description. The GESSAR II fuel system description, given in Ref.1, does not include all of the design parameters listed in Acceptance Criterion II.B. However, sufficient information is given to provide a reasonably accurate representation of the GESSAR II fuel system, satisfying the intent of the SRP. I i l t i
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'22A7007 s.
238 NUCLEAR ISLAND Rdr. 6
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4.2.2.2 Velocity Limiter (Continued) The velocity limiter is in the form of two nearly mated, conical elements that act as a large clearance piston inside the control rod guide tube. The lower con'ical element is separated from the upper conical element by four radial spaciis 90 degrees apart and is at a 15-degree angle w relative tos the uppers conical
's element, with the peripheral separ'ation e
less tha'n;th'e central separation. -
~, -
) -
The hydraulic drag forces on a control rod are proportional to y approximately the square of the rod velocity andEare negliaiblei , at normal rod withdrawal or red insertion speeds. Ho'weUe'r , I n
\
during the scram stroke,-the rod reaches'high veloci~ty, sand'the? %- drag forces must be overcome by the drive mechanism. '
.% ~.
To limit control rod velocity during dropout, out not during scram, the velocity limiter is provided with a' streamlined profile in the scram (upward) direction. 8 Thus, when the control rod is scrammed, water flows over the smooth surface of the upper conical element into the annulus between the guide tube and the limiter. In the dropout direction, however, water is trapped by the lower conical element and dis-charged through the annulus between the two conical sections. Because this water is jetted in a partially reversed direction into water flowing upward in the annulus, a severe turbulence is created, thereby slowing the descent of the control rod assembly to less than 3.11 ft/sec. 4.2.3 Design Evaluation See Appendix A, Subsection A.4.2.3 of Reference 1. e 4.2-3
. . . - ~ . _ . , _ , . .
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( , Acceptance Criterion II.C.3(a) lists phenomonological models to be included in fuel -
. 1
~ system thermal calculations. The GESSAR II fuel thermal model does not include the s s use of approved gadolinia fuel properties. H'owever, as discussed with the NRC staff,
; w the General Electric Cogany does not license' material properties for design analyses but, rather, maintains tiese' analyses up-to-date. To fulfill our quality control obligations under 10CFR50, Appendix B, the latest property values are incorporated into i
w, design applications only after' they are qualified in the design code. An improved
, Q fuel rod thermai-mechanical design code has recently been developed and qualified
~
which includes the revised gadolinia fuel theml conductivity relations. The results of the_ fuel centerml' ting analysis using this improved fuel rod design code verifies that gadolinia fuel melting is not expected to occur during normal steady-state operat{o'n fr during the largest whole core anticipated operational transient. s s
\
The GESTAR II (NEDE-24011-P-A) amendment inco.7porating the application of the above thermal-mechanical design code is currently under review by the Core Performance I. Branch. Since GESSAR II references the " latest approved revision" of GESTAR II, this N issue,will be resolved when the GESTAR II' amendment is approved. Acceptance Criterion II.C.3.(d) describe's acceptance criteria for evaluation of fuel 3 assembly structural response to externally applied forces.
\
An analysis has been performed (NEDE-21175-3) to show that the GESSAR II fuel meets structural requirements (including 1iftoff) similar to those of Appendix A of Section 4.2 of the SRP (NUREG-0800). That analysis is currently under review by the NRC staff. Decause previous generic analytical methods presented in earlier versions of NEDE-21175 have been approved by the NRC staff (letter from 0, P. Parr (NRC), May 17, 1979) and because favorable sample results were also presented in Amendment 2 of NEDE-21175, the new GE analysis is expected to be approved.
\
4.2-3a
GESSAR II 22A7007 238 NUCLEAR ISLAND R^.v . (}
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( 4.2.4 Testing, Inspection and Surveillance Plans ' See Appendix A, Subsection A.4.2.4 of Reference 1. 4.2.5 References %
- 1. " General Electric Standard Applicaition for Reactor Fuel,"
NEDE-240ll-P-A, latest approved revision. Tl Th 4. ApMs ch WO PV8W SW T t h .s p q. c m o n vo v-a. m d o {vew b\ 0^ ow
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A typical program involves visual examination of selected assemblies (commonly 5 to 10% of the discharged fuel), concentrating on the lead bundles. Visual
' examinations normally include, but are not necessarily limited to, crud buildup, rod bowing, and missing components. Additional inspections should be performed )
depending and on inspections. the visual the results of operational monitoring including coolant activity 4.2-4 t
- '-*'___ , 7- --
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GESSAR II 238 NUCLEAR ISLAND 22A7007[7 Rav. / 4.4.3.6 Thermal and Hydraulic Characteristics Summary Table The thermal-hydraulic characteristics are provided in Table 4.4-1 for the core and tables of Section 5.4 for other portions of the reactor coolant system. 4.4.4 Evaluation See Appendix A, Subsection A.4.4.4 of Reference 1. - A 4.4.5 Testing and Verification See Appendix A, Subsection A.4.4.5 of Reference 1. , I 4.4.6 Instrumentation Requirements See Appendix A, Subsection A.4.4.6 of Reference 1. 4.4.6.1 Loose Parts 4 To be supplied by Applicant. 4.4.7 References I (
- l. " General Electric Ste.ndard Application for Reactor Fuel,"
(NEDE-240ll, latest approved revision). .
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- 238 NUCLEAR ISLAND Rsv.
22A70[07 l 5.2.3.3.2 Control of Welding
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5.2.3.3.2.1 Regulatory Guide 1.50: Control of Preheat Temperature Employed for Welding of Low-Alloy Steel Regulatory Guide 1.50 delineates preheat temperature control requirements and welding procedure qualifications supplementing those in ASME Sections III and IX. The use of low-alloy steel is restricted to the reactor pressure vessel. Other ferritic components in the reactor coolant-pressure boundary are fabricated from carbon steel materials. Preheat temperatures employed for welding of low alloy steel meet or exceed the recommendations of ASME Code Section III, Subsection NA. Components were either held for an extended time at preheat temperature to assure removal of hydrogen, or preheat was maintained until post-weld heat treatment. The minimum preheat and maximum interpass temperatures were specified and monitored.
> rw ss pT 5 '2.1. 3 '2 . \
All welds were nondestructively examined by radiographic methods.
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In addition, a supplemental ultrasonic examination was performed. fvi mcr 4 t ent and rr"irir.- ...uct, mmm C;ction 1. t
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5.2.3.3.2.2 Regulatory Guide 1.34: Control of Electroslag Weld Properties No electroslag welding was performed on BWR componentr. 5.2.3.3.2.3 Regulatory Guide 1.71: Welder Qualification for
? Areas of Limited Accessibility Qualification for areas of limited accessibility is discussed in Subsection 5.2.3.4.2.3.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. [] 5.2.3.4.2.3 Regulatory Guide 1.71:
.T' Welder Qualification for Areas of Limited Accessibility (Continued)
All ASME Section III welds were fabricated in accordance with the requirements of Sections III and IX of the ASME Boiler and Pressure Vessel Code. There are few restrictive welds involved in the fabrication of BWR components. Welder qualification for welds with the most restrictive access was accomplished by mockup welding. Mock-ups were examined with radiography or sectioning. n x - < & --; ...; _~a-inn mmw , --- e--tin,l. _ { 5.2.3.4.3 w l Regulatory Guide 1.66: Nondestructive Examination of Tubular Products For discussion of compliance with Regulatory Guide 1.66, see Subsection 5.2.3.3.3. 5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary This section discusses the inservice inspection and testing program for the NRC Quality Group A components; i.e., ASME Boiler and Pressure Vessel Code Section III, Class 1, components. It will show how the program meets requirements of Section XI of the ASME Code. 5.2.4.1 System Boundary Subject to Inspection ' The reactor pressure vessel, system piping, pumps, valves, and components within the reactor coolant pressure boundary defined as quality Group A (ASME Code Section III, Class I) were designed and fabricated to permit full compliance with ASME Code Section XI. (Applicant will provide applicable code and addenda dates. ) Access is provided for volumetric examination of pressure [ %e oLc cepheM4 gi En 4e n.w 2.L As .ts) of' S R P s.2.3 is loased j %Aho G m o(n. t.71.G6ssA9~4 hae.hs r y{ ^ T wMbg
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev.[( 7 L 6.4 HABITABILITY SYSTEMS (Continued) L o
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L [ Radiation Protection Section 12.3 and Chapter 15.0 1 Heating, Ventilating and Air Subsection 9.4.1 f Conditioning (HVAC) Fire Protection Subsection 9.5.1 Lighting Systems Subsection 9.5.3 Power Systems Chapter 8 Radiation Instrumentation Subsections 7.6.1.2 and Monitoring and 12.3.4, and Section 11.5 Control Room Isolation Subsection 7.3.1.1.17
- Instrumentation and Controls Equipment and systems are discussed in this section only as necessary to describe their connection with control room habit-ability. References to other sections are made where appropriate.
S - The term " control building" @picall3 includes the main control [ room, areas adjacent to the main control room containing plant , information and equipment necessary to normal and emergency l j operations, andkitchengandsanitaryfacilities.AItisalso l the entire zone service / by the control room vent ilation system. - 1 ! " Emergency conditions" include such postulated ri tleases as radio-active materials. toxic gases, smoke and steam. l i 43* S Ydt A I tw c w dr k v oow s h c~e S e_ e l on
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p5- GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O V
} 6.7.1.1 Safety Criteria (Continued)
(8) The MSPLCS, including instrumentation and circuits necessary for the functioning of the system, is designed
~
to standards applicable to an engineered safety feature. l - (9) The MSPLCS controls include interlocks to prevent inadvertent operation of the system. In particular, i interlocks are provided to prevent damage to the MSPLCS, or to the main steam system, due to accidental t opening of any system isolation valves when the pressure in the connecting main steam piping exceed MSPLCS operating pressure. All such controls and interlocks are activated from appropriately designed safety systems or circuits. L (10) The MSPLCS is designed to permit testing of the oper-ability of controls and actuating devices during power operation to the extent practical, and complete testing h of system function during plant shutdowns. , (11) The MSPLCS is designed so that: (a) thermal stresses i and pressures associated with flashing and thermal
- deformations, under the loading conditions associated with the activated system shall not affect the structural integrity or operability of the main steam system or main steam isolation valves; and (b) any deformation of l
isolation valve internals shall not induce leakage of 1 the main steamline isolation valve beyond the capacity . or capability of the MSPLCS. (12) Equipment is provided (as part of the MSPLCS) to prevent l the release of valve stem packing leakage to the environ-ment from main steam system isolation valves outside the containment. 6.7-3
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i GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. $' g 7 I J
- 9.5 OTHER AUXILIARY SYSTEMS 9.5.1 Fire Protection System 9.5.1.1 Design Bases The bases for the design of the fire protection program are pre-sented in detail in Appendix 9A (Fire Hazard Analysis). The pro-gram's intent is to provide a " defense-in-depth" design resulting in an adequate balance in:
(1) preventing fires from starting; (2) quickly detecting and extinguishing fires that occur, thus limiting fire damage; and (3) designing safety-related systems so that a fire that starts in spite of the fire prevention program and burns out of control for a considerable length of time will not prevent safe shutdown. In addition, fire protection systems are designed so that their inadvertent operation or occurrence of single failure in any of these systems will not prevent plant safe shutdown. Possible fires that could affect safety-related systems and significant combustible loadings are presented in Appendix 9A on a room-by-room basis. Fire barriers and fire protection systems are discussed for each safety and nonsafety-related area. Each room is also analyzed for its potential radioactive release due to a postulated fire. Noncombustible or fire-resistive materials having a flame-spread, smoke-evolved and fuel-contri'buted index of 25 or 1 css are used wherever practicable. '
% T W SESLT 9. 5 l . l -
Containment isolation valves and included piping of che Fire Protection System are classified as ASME Section III, Class 2 and l 9.5-1 i _-y. : _ - - r w- r . 1 _. ______ _ _
l l INSERT 9.5.1.1 f. l SRP Acceptance Criterion II.2.a requires adherence to BTP CMEB 9.5-1. Three-ht . fire rated dampers (required by paragraph C.S.f of BTP CMEB 9.5-1) have not been provided in HVAC ducts in the smoke removal systems which have 3 hr fire rated I barriers in the Control bldg., the auxiliary b1dg., and in the HVAC ductwork that penetrates the reactor b1dg. wall from the auxiliary bldg. and fuel bldg. Some of these ventilation ducts are shared systems in that they also provide normal ventialation. Other ducts are for smoke venting only. Based on the discussion below the present GESSAR II design should be adequate and should be acceptable to the NRC. l The auxiliary building smoke removal system is shown on Figure 9.4-4 and described in Section 9.4.3.2.1.11. Each set of duct work serves and traverses only fire areas i of one safety division. There is a smoke vent intake in each fire area with a remote ?~ manually operated fire damper which is normally closed. There is a fusible link from l the air operator to the vanes so that the damper will close on high temperature. The fire rating of the dampers is ils hours. The duct is heavy gaga, welded construction which exceeds the requirements for 3 hour fire rated construction. Hence, the design is ( { considered completely adequate for the service. i l l One of the design objectivet of GESSAR II is to avoid fire dampers in smoke vents, as i s their automatic closure would render the smoke vent inoperative at the very time it was f needed. With two exceptions, suoke vents pass through safety areas only of the same ( division as the vented area. The two exceptions are the Division 2 cable tunnel vent and the primary containment vent. l l e The Division 2 cable tunnel located in the corridor of (-)6' 10 elevation of the l auxiliary building has a dedicated smoke removal system, which passes through the i division 1 area. The duct opening is 2.5 sq. ft. and is designed to with stand a l 3-hr. fire. I There is a containment vent and a containment supply. The supply takes air from the auxiliary building roof top intake. The fans are located in a room on the top floor I ef the auxiliary building. The boundaries of the room have a 3 hour fire rating. The supply duct goes directly into the reactor building from the room. A fire in the room cannot prevent safe or alternate shutdown. There is an inboard and a outboard isolation
- valve for the duct.
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t-l INSERT 9.5.1.1 (centinued). l ! . The containment exhause has two inboard (1 manual) isolation valves and one outboard f isolation valve. If a fire occurs, either the inboard valves or the outboard valve would be located out of the fire area and could be closed. The valve within the fuel building is located in a room with 2 hour rated walls. The room is directly l accessible from the fuel building or the stair tower between the fuel and auxiliary ! building. All return registers except for the pool sweep are located high in the containmen containment so that bulk mixing, aided by the dome mixing system, would occur before any } j combustion gases enter the ventilation duct. The containmer,t is more sensitive to bulk f air temperature than the ventilation duct. If a fire raised the bulk temperature i excessively, containment spray would be initiated to protect the containment at a f temperature well below the threshold of damage to the ventilation duct. For these i reasons, the current GESSAR II design for the containment ventilation is considered proper and adequate. The exhaust ducting which is schedule 20 welded pipe will be designed with a 3 hour fire rating. The remaining smoke vents which do not have fire dampers are the two in the control l building. Each one of these smoke vents serves and traverses one division. Since l it is impossible for these smoke vents to allow the fire in the area of one division h to spread to another division, the current GESSAR II design is considered to be f adequate and proper. l \ e l I
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GESSAR II 82A7007 l 238 NUCLEAR ISLAND R v. f 17 l
) 11.5.3.1 Basis for Monitor Location Selection Monitor locations are selected to assure that all effluent materials comply with regulatory requirements as covered in Regulatory Guide 1.21, Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Release of Radioactive Effluent from Light Water-Cooled Nuclear Power Plants. !
l 11.5.3.2 Expected Radiation Levels Expected radiation levels are in the ranges listed in Tables 11.5-2 and 11.5-3. 11.5.3.3 Instrumentation Radiation monitors used are listed in Table 11.5-1. Grab samples are analyzed to identify and quantify the specific radionuclides in effluents and wastes. The results from the sample analysis are used to establish relationships between the gross gamma monitor readings and concentrations or release rates of radionuclides in continuous effluent releases. 11.5.3.4 Setpoints
% Setpoints are listed in Table 11.5-1.
11.5.4 Process Monitoring and Sampling 11.5.4.1 Implementation of General Design criterion 60 All potentially significant radioactive discharge paths are equipped with a control system to automatically isolate the W. A g u M- will htCMM kek Vedrto , clamhW2, oc chwrsets vch/to M aeo s es M W M -ped an M cc 4 &- seJpA 4 cn te d t. tM . actud& or bb adu& v&chhw u, , oc avacs a sua u{anh s cn,ky%*a M M m*:M.$c fai&*All.5-23 - r- -- m,-, _ . , _ _ . , , m ., . _ ;. ,;
GESSAR II 22A7007 238' NUCLEAR ISLAND Rav. g 17 y 12.1.2.2 Equipment Design Considerations For ALARA Exposures 12.1.2.2.1 General Design Criteria q engineering design procedures require that the component esign engineer consider the applicable regulatory guides as a part of the design criteria. This includes Regulatory Guide 8.8. In this way, the radiation problems of a component or system are considered. A summary survey of the components designs was made to determine the factors considered. The following paragraphs cite some examples of desitn considerations made to implement ALARA. 12.1.2.2.2 Equipment Design Considerations to Limit Time Spent in Radiation Areas k (1) Equipment is designed to be operated and have its instrumentation and controls in accessible areas'both during normal and abnormal operating conditions. Equipment such as the Reactor Water Cleanup (RWCS) System and the Fuel Pool Cleanup (FPCCU) System are remotely operated, including the backwashing and precoat operations. Other equipment has been redesigned in order to lengthen service life. For example, seal water is applied to the recirculation pump seals to keep them clean. This increased the maintenance interval from
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No specific instructions have been given to component designers and engineers regarding ALARA design as provided by specific acceptance criterion II.2 of SRP 12.1. Howear f.
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# GESSAR II 238 NUCLEAR ISLAND 22A7007/
RLv.f G 12.2 RADIATION SOURCES 12.2.1 Contained Sources 12.2.1.1 Source Terms i a With the exception of the vessel and drywell shields, shielding designs are based on fission product and activation product sources consistent with Section 11.1. For shielding, it is con-servative to design for fission product sources at peak values rattier than an annual average, even though experience supports a lower annual average than the design average (Reference 1). It should be noted. that activation products, principally Nitrogen-16, control shielding calculations in most of the primary system. In areas where fission products are significant, conservative allow-ance is made for transient decay while at the same time providing for transient increase of the noble gas source, daughter product formation and energy level of emission. Areas where fission products are significant relative to Nitrogen-16 include: (1) the ( condenser off-gas system downstream of the jet air ejector; (2) liquid and solid radwaste equipment; (3) portions of the RWCS; and (4) portions of the feedwater system downstream of the hotwell including condensate treatment equipment. For application, the design sources are grouped first by location and then by equipment type (e.g., reactor building, core sources). l The following paragraphs represent the source data in various pieces of equipment throughout the plant. General locations of equipment are shorn in the general plant arrangement drawings of Section 1.24 Opecific acceptance criterion II.6 of SRP 12.2 provides that in addition to the location of contained sources, their approximate size and shape be shown. - Though this has not always been included, the source strength or concentration has been provided in Chapter 12 tables and detailed geometry is provided in Table 12.2-1 for the reactor, and in Chapter 5 for the main steam and a
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^ GESSAR II 238 NUCLEAR ISLAND 22A7007 o
Rsv. JF l] 12.2.1.2.7.1 Radioactive, Sources in Main Steam System (Continued) h cource is dominated by Nitrogen-16. In components where N-16 has decayed, the other activities carried by the steam become sig-nificant. During plant shutdown, there is a residual activity resulting from prior plant operations. These data will be-pro-vided by the Applicant. ~ 12.2.1.2.7.2 Radioactive Crud in Piping and Steam Systems The inside surfaces of the piping and all reactor and power sys-tems components become coated with activated corrosion products, commonly called crud. The quantity of crud on the components in dependent on a number of factors, including power history, water quality and fuel experience. The piping and components carrying reactor water are coated with higher levels of crud than piping and components carrying steam. Figure 12.2-2 shows the data used in the design of this plant to characterize crud accumu- ' " m lation in Recirculation System Piping.jgCrud levels in steam piping -4 cro estimated to be about 1% of those i n the recirculation piping.
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12.2.1.2.8 Radioactive Sources in the Spent Fuel The radiation source for spent fuel is given in Sub-cection 12.2.1.2.1.1.4 (Table 12.2-3) in terms of MeV/sec/W. The decign calculation is carried out for a mean element for an appropriate decay time. 12.2.1.2.9 Other Radioactive Sources 12.2.1.2.9.1 Reactor Startup Source Tho reactor startup source is shipped to the site in a special cack designed for shielding. The source is transferred under water while in the cask and loaded into beryllium containers. This is then loaded into the reactor while remaining under water. The 12.2-10 n e-- , ~ - , _ -m-
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INSERT m 1 2. 9.l.2 . 7 . 1 Criterion II.6 of SRP 12.2 provides that the buildup of activated corrosion products in various components and systems should be addressed and allowances made in design source terns should be explained. Based on current data and analysis, activated corrosion products are most significant. in the recirculation piping. 6 9 _. 1 -
s i GESSAR II 22A7007 I 238 NUCLEAR ISLAND Rsv. { l i : 15.3.3.2 Sequence of Events and Systems Operations 15.3.3.2.1 Sequence of Events l Table 15.3-5 lists the sequence of events for Figure 15.3-5. 15.3.3.2.1.1 Identification of Operator Actions The operator should ascertain that the reactor scrams from reactor water level swell. The operator should regain control of reactor water level through RCIC operation or by restart of a feedwater pump, and he should monitor reactor water level and pressure control after shutdown. 15.3.3.2.2 Systems . Operation In order to properly simulate the expected sequence of events, the analysis of this event assumes normal functioning of plant instrumentation and controls, plant protection, and reactor __, protection systems.4F
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'-- TV S54T l5.G 3.2.2 Operation of safe shutdown features, though not included in this simulation, is expected to be utilized in order to maintain adequate water level.
15.3.3.2.3 The Effect of Single Failures and Operator Errors Single failures in the scram logic originating via the high l vessel level (L8) trip are similar to the considerations in 1 t Subsection 15.3.1.2.3.2 (see Appendix 15A for further details). u 15.3-13 & ~_ : n - _ ,
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INSERT DEEEit 15 3.3,2,1 Acceptance criterion II.8 of SRP 15.3.3 provides that only safety grade equipment should be used to mitigate the consequences of this event. It also provides that safety functions be accomplished assuming the worst single
-failure of a safety system active component. The actual simulation used for this event provides a more conservative basis for evaluating system performance for this transient than would result from direct application of this SRP criterion.
Justification for this difference is given in Section IE.11. Acceptance criterion II.10 of SRP 15.3-3 also provides that the analysis assume turbine trip and coincident loss of offsite power. Should a coincident. loss of offsite power occur, the consequences would be similar to the consequences of the loss of offsite Power Transient described in Subsection 15.2.6; however, this event would be less severe due to the faster reactor flow coastdown and the earlier feedwater pump trips. t l l l
GESSAR II ' 238 NUCLEAR ISLAND 22A7007[{7 Rav. ) 15.4.4.3.2 Results (Continued) S f before decreasing after the cold water washed out of the loop at about 18 sec. No damage occurs to the fuel barrier and MCPR ' remains significantly above the safety limit as the reactor settles out at its new steady-state condition. Therefore, this y event does not have to reanalyzed for specific core '~ configurations. .I *t i.
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15.4.4.4 Barrier Performance ' No evaluation of barrier performance is required for this event since no significant pressure increases are incurred during this transient (Figure 15.4-1). Radiological Consequences 15.4.4.5 An evaluation of the radiological consequences is not required for this event, s{ncenoradioactivematerialisreleasedfromthe fuel. 15.4.5 Recirculation Flow Control Failure with Increasing Flow 15.4.5.1 Identification of Causes and Frequency Classification 15.4.5.1.1 Identification of Causes Failure of the master controller of neutron flux controller can cause an increase in the core coolant flow rate. Failure within a loop's flow controller can also cause an increase in core coolant flow rate. 15.4.5.1.2 Frequency Classification This transient disturbance is classified as an incident of moderate frequency. 15.4-11
INSERT M l S . 4 4 .3. 2., l Acceptance criterion II.2.(b) of SRP 15.4.4 provides that fuel clad integrity , shalI be maintained by ensuring that the CPR remains above the MCPR safety limit. Since this event does not result in a significant increase in pressure and it is initiated from a low power condition, no MCPR calculation was performed. & , - , + + . e .
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. GESSAR II 22A7007j ;
238 NUCLEAP. ISLAND Rnv. A [s' 15.4.5.3.1 Input Parameters and Initial Conditions (Continued) ) Maximum stroking rate of a single recirculation loop value for a loop controller failure is limited by hydraulics to 30%/sec.
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15.4.5.3.2 Results "
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15.4.5.3.2.1 Fast Opening of One Recirculation Valve Figure 15.4-2 shows the analysis of a failure where one recircula-tion loop main valve is opened at its maximum stroking rate of 30%/sec. Table 15.4-4 provides the sequence of events of this failure. The rapid increase in core flow causes a sharp rise in neutron flux, initiating a reactor scram at approximately 1.3 sec. The peak neutron flux reached was 235% of NBR value, while the accompanying average fuel surface heat flux reaches 73% of NBR I at approximately 2.2 sec. MCPR remains considerably above the safety limit and average fuel temperature increases only 108*F. -
-Reactor pressure is discussed in Subsection 15.4.5.4.
hMSG4 15.4.5.3.2.2 Fast Opening of Two Recirculation Valves ! I 4-3+ 3.2.I
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Figure 15.4-2 illustrates the failure where both recirculation loop main valves are opened at a maximum stroking rate of 11%/sec. Table 15.4-5 shows the sequence of events for this failure. _It is very sim gar to ge ove g ansient. Flux scram occurs at approx-imately 1.6 sec, peaking at 162% of NB rated, while the average surface heat flux reaches 67% of NB rated at approximately 2.3 sec. MCPR remains considerably above the safety limit and average fuel temperature increases 80 F. Therefore, this event does not have l to be reanalyzed for specific core configurations. ) (S.4-14 (z : 77 n~~ 7 w y n 3 m ' ~w*q
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INSERT lGat t t5 ,4 ,f,3,'4.j i Acceptance criterion II.2.(b) of SRP 15.4.4 provides that fuel clad integrity shall be maintained by ensuring that the CPR remains above the MCPR safety limit. Since this event does not result in a significant increase in pressure i and it is initiated from a low power condition, no MCPR calculation was performed.
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. .- 238 NUCLEAR ISLAND 22A700[7 Rev. (7 15.4.7.1.1 Identification of Causes (Continued) (
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incorrect location or discharged. Third, the misplaced bundles would have to be overlooked during the core verification process performed following core loading. a ,, 15.4.7.1.2 Frequency Classification This unlikely event occurs when a fuel bundle is loaded into the wrong location in the core. It is assumed the bundle is misplaced to the worst possible location, and the plant is operated with the mislocated bundle. This event is categorized as an infrequency incident based on the following data: Expected Frequency: 0.002 events / operating cycle The above number is based upon past experience. 15.4.7.2 Sequence of Events and Systems Operation 15.4.7.2.1 Sequence of Events I The postulated sequence of events for the misplaced bundle accident (MBA) is presented in Table 15.4-6. 15.4.7.2.2 Systems Operation A fuel loading error, undetected by in-core instrumentation follow-ing fueling operations, may result in an undetected reduction in thermal margin during power operations. For the analysis reported herein, no credit for detection is taken and, therefore, no corrective operator action or automatic protection system functioning is assumed to occur.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 's . / 15.6.5.5 Radiological Consequences (Continued) 10CFR100 guidelines. This analysis is referred to as the " design basis analysis". (2) The second is based on assumptions considered to provide a realistic estimate of radiological consequences. This analysis is referred to as the " realistic analysis". A schematic of the transport pathway is shown in Figure 15.6-2. Additional parameters and information for specific design basis - g accidents are provided in Subsection 19.3.15.1. d 15.6.5.5.1 Design Basis Analysis The methods, assumptions and conditions used to evaluate this accident are in accordance with those guidelines set forth in Regulatory Guide's 1.3 and 1.7. The specific models, assumptions and computer code used to evaluate this event based on the above criteria are presented in Reference 2. Specific values of param-eters used in this evaluation are presented in Table 15.6-7. 15.6.5.5.1.1 Fission Product Release from Fuel It is assumed that 100% of the noble gases and 50% of the iodine are released from an equilibrium core operating at a power level of 3651 MWt for 1000 days prior to the accident. While not specifically stated in Regulatory Guide 1.3, the assumed release of 100% of the core noble gas activity and 50% of the iodine ac.ivity implies fuel damage approaching melt conditions. Even though this condition is inconsistent with operation of the ECCS system (Section 6.3), it is assumed applicable for the evaluation of this accident. Of this release, 100% of the noble gases and 50% of the iodine become airborne. The remaining 50% of the iodine is removed by plate-out and condensation; therefore, it is not available for airborne release to the environment. The activity airborne in the containment is presented in Table 15.6-8. 15.6-15
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l' o I i. I For determining equipment leakage contribution to the LOCA dose, it is assumed p that the 50% "plateout and condensation" fraction of the released iodine [ finds its way into the suppression pool water. This is consistfnt with R.G.1.3 though not:.with acceptance criterion II (2) of SRP 15.6.5 the latter document i] provides that 50% of the core iodine activity should be assumed to be missed in the sump water being circulated through the containment external piping. The Q ) assumptions used in this calculations are the more conservative with respect to [e BWR post-LOCA total dose calculations. See item 16 of Subsection 19.3.15.1 for a detailed description of equipment leakage contribution to off-site dose. 3 h I i 9 i l i e i i l t 6. 6 15 a _. g;,.,, ; -- 3 :,r, ~ :~ ~7 .. , .
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t . 238 NUCLEAR ISLAND Rcv. [ []
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.19.3.3.74 QUESTION / RESPONSE 3.74 (220.33) ,
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In Section 3.8.3.3.6.3.2 of your FSAR, you indicate that you satisfy three out of the four load combinations presented in ti - Item 1I.3.c (ii)(a) of Section 3.8.3 of the SRP for the
'N factored load conditions for steel structures using the elastic working stress design method. State why you omitted Equation (4) of Item II.3.c(ii)(a) and verify that you satisfy our position on the load combination represented by Equation (4). (3.8.3)
RESPONSE 3.74 Subsection .8.3.3.6.3.2 was revised to include the missing
- Equation 3 of SRP 3.8.3 II.3.c(ii)(a). Equation 4 is more l severe than and bounds Equation 3. Equation 4, not Equation 3, is applied to the current design of GESSAR II.
i b l l l E l l I l i l' I t i 19.3.3.74-1/19.3.3.74-2 1 -- ., - - . . , , , . . . - - . . ___f . , ,
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6 1 t' o GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 19.3.15 Chapter 15 - Responses 19.3.15.1 QUESTION / RESPONSE 15.1 (440.3) hUESTION15.1 Address each item identified in Item 1 of Table 15-4 of Regulatory Guide 1.70, Revision 3, or indicate an interface to (15.6.5) provide the information. RESPONSE 15.1 la. Hydrogen Purge Analysis As noted in Subsection 6.2.5, redundant Class lE hydrogen i recombiners are provided. Even assuming the arbitrary fail-ure of one recombiner, the remaining recombiner is capable of maintaining hydrogen concentrations below the ignitable or - l( detonable level; therefore, there will be no need to purge [ the containment and there will be no additional dose f contribution from this source. lb. Equipment Leakage Contribution to LOCA Dose The potential dose contribution from this source is deter-mined in a manner consistent with RG 1.3 and SRP 15.6.5 unless otherwise noted. - 0 f* j (1) Fission Product Source Term.Appenf s*MSRP1'i.6.5su]ggests5 the iodine contained in the core at shutdown is released {
;i to and contained within the suppression pool. RG 1.3 suggests that 50% of the iodine in the core is released to the containment where 50% remains airborne and 50%
is lost due to washout /plateout phenomena. RG 1.7 sug-
, gests that 50% of the iodine remains in the core. Since lk there is no question that the core cannot initially 19.3.15.1-1
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