ML20077B696
| ML20077B696 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 07/18/1983 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20077B688 | List: |
| References | |
| NUDOCS 8307250382 | |
| Download: ML20077B696 (86) | |
Text
UNITED STATES 0F AMERICA NUCLEAR REGULAT0RY C0MMISSION In the matter of )
General Electric Company.) Docket No. STN 50-447 Standard Plant )
AMENDMENT NO. 18 TO APPLICATION FOR REVIEW 0F 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
General Electric Company, applicant in the above captioned proceeding, hereby files Amendment No.18 to the 238 Nuclear Island General Electric Standard Safety Analysis Report (GESSAR II).
Amendment No. 18 further amends GESSAR II by:
- 1. Incorporation of the following General Electric submittals:
- a. April 28, 1983 letter pertaining to post-accident monitoring instrumentation (Appendix 1D).
- b. June 15, 1983 letter pertaining to Question 410.24 of the Commission's August 25, 1982 information request and a summary of the measures taken to avoid intergranular stress corrosion cracking.
- c. June 28, 1983 letter pertaining to a revised response to the CP/ML Rule (Appendix 1G).
- 2. Updating the GESSAR II/FSAR interface tables of Subsection 1.9.1.
- 3. Clarifying portions of the text where obvious discrepancies exist.
Respectfully submitted, General Electric Company
\ j By: iw i- Asu/A97
~
G1dffff G. 'Sherwood, Manager Nuclear Safety & Licensing Operation STATE OF CALIFORNIA COUNTY OF SANTA CLARA )
) ss On this 8th day of July in the year 1983, before me, Karen S. Vogelhuber, Notary Public, personally appeared Glenn G. Sherwood, personally proved to me on the basis of satisfaction evidence to be the person whose name is pggQgtgjQggggggg and acknowledged that he executed it.
g OFFICIAL SEAL 8 KAREN S. VOGELHUBER Mk/ [
NOTARY PUBUC.CAUFORNIA 2- -
SANTA CLARA COUNTY $ dMh( gb
$ My Commission Expires Dec. 21,1984 h NOTA M PUBLIC, STATE OF CALIFORNIA bcncncncwcucncwcwc>ocxc>:cxcuse 107072*-1 8307250392 830718 PDR ADOCK 05000447 K PDR
UNITED STATES 0F AMERICA NUCLEAR REGULAT0RY C0MMISSION O In the matter of General Electric Company )
Standard Plant
)
)
Docket No. STN 50-447 AMEN 0 MENT NO. 18 TO APPLICATION FOR REVIEW OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
General Electric Company, applicant in the above captioned proceeding, hereby files Amendment No. 18 to the 238 Nuclear Island General Electric Standard Safety Analysis Report (GESSAR II).
Amendment No. 18 further amends GESSAR II by:
- 1. Incorporation of the following General Electric submittals:
- a. April 28, 1983 letter pertaining to post-accident monitoring instrumentation (Appendix 1D).
- b. June 15, 1983 letter pertaining to Question 410.24 of the Commission's August 25, 1982 information request and a summary 1 of the measures taken to avoid intergranular stress corrosion l cracking.
- c. June 28, 1983 letter pertaining to a revised response to the CP/ML Rule (Appendix 1G).
- 2. Updating the GESSAR II/FSAR interface tables of Subsection 1.9.1.
3.
Clarifying portions of the text where obvious discrepancies exist.
Respectfully submitted, General Electric Company By: s/Glenn G. Sherwood Glenn G. Sherwood, Manager Nuclear Safety & Licensing Operation l STATE OF CALIFORNIA )
COUNTY OF SANTA CLARA ) ss On this 8th day of July in the year 1983, before me, Karen S. Vogelhuber, Notary Public, personally appeared Glenn G. Sherwood, personally proved to me on the basis of satisfaction evidence to be the person whose name is subscribed to this instrument, and acknowledged that he executed it.
By: s/ Karen S. Vogelhuber Notary Public - California Santa Clara County p
V My Commission Expires December 21, 1984 175 Curtner Avenue San Jose, CA 95125 107072*-2
,-ew r- e e a e---v , v-,...vac.- ---,-.--mrwa-----e.an-e--,, --------n----- - .- mww
I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 i INSTRUCTIONS FOR FILING AMENDMENT NO.18 Remove and insert the pages listed below. Dashes (----) in the remove or insert column indicate no action required.
Remove Insert Chapter 1 1.5-13.1.5-15/1.5-16, 1.5-13,1.5-15/1.5-16, 1.8.117-1,1.9-4.1-19, 1.5-17,1.5-18,1.8.117-1, 1.9-4.1-20,1.9-4.3-3, 1.9-4.1-19,1.9-4.1-20, 1.9-4.3-3a, and 1.9-4.3-4 1.9-4.3-3,1.9-4.3-4, and 1.9-4.3-4a Appendix 1D 10.1-1,10.2-5,1D.2-14, I D.1 - 1,1 D. 2- 5,10. 2 -5 a ,
10.2-15, and 1D.2-16 10.2-12a,10.2-14,10.2-15, 10.2-16,1D.4-0, and 1D.4-0a through 10.4-Oe Appendix 1G IG-v/1G-vi,1G.12-1/1G.12-2, ano 1G-v/1 G-vi ,1G.12-1/1G.12-2 ,
1G.21-2 IG.12-3/1G.12-4,1G.21-2, and 1G.21-2a Chapter 3 3.2-2,3.2-16,3.2-31,3.2-34, 3.2-2,3.2-16,3.2-31,3.2-34, 3.2-38,3.2-57/3.2-58,3.6-17, 3.2-38,3.2-57,3.2-58,3.6-17, 3.9-36e, and 3.9-38 3.9-37, and 3.9-38 Caapter 4 4.2-lb,4.2-Ic,4.2-3,4.2-5, 4.2-Ib,4.2-1c,4.2-3,4.2-5, and 4.2-7/4.2-8 and 4.2-7/4.2-8 4
Chapter 5 5.1-12,5.2-v/5.2-vi,5.2-ix/ 5.1-12,5.2-v/5.2-vi,5.2-ix/
1 5.2-x,5.2-7,5.2-25,5.2-103, 5.2-x,5.2-7,5.2-25,5.2-103, and 5.2-104 and 5.2-104
\
Amendment 18 July 8,1983
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 9
Remove Insert Chapter 19 19.3.9.24-2 19.3.9.24-2 and 19.3.9.24-3/
19.3.9.24-4 O
O Amendment 18 July 8, 1983
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 18 b Table 1.5-2 COMMITMENT ITEMS Reference Where Additional Technical Information Item Discussed
- 1. Preop Piping Vibration Test Program . . . . . . . . . . . . . . . Subsection 3.9.2
- 2. Reactor Internals Preop Vibration Test Program Results . . . . . Subsection 3.9.2
- 3. Dynamic Analysis of Peactor Internals and Piping . . . . . . . . Section 3.9
- 4. . Seismic Oualification of Class IE Electrical Equipment . . . . . Subsection 3.9.2.2 & 3.10
- 5. Environmental Qualification of Class IE Electrical Equipment . . Section 3.11
' Electrical Isolation Devices Test Program Results . . . . . .
6.
- 7. Fuel Experience Update - NEDO-10505 . . . . . . . . . . . . . .
. Chapter 7 See Note A ]
- 8. . Fuel Surveillance Program Results . . . . . . . . . . . . . . . . See Note B
- 9. Fuel Assembly Components Stress Report . . . . . . . . . . . . See Note C
- 10. Fuel Assembly Pressure and Temperature Capability . . . . . . . . See Note D
- 11. Fuel Assembly Dynamic Analysis . . . . . . . . . . . . . . . . . See Note C
- 12. Fuel Assembly Analysis Method for Creep-Rupture . . . . . . . . . See Note E
- 13. Fuel Assembly Design Limit for Instability . . . . . . . . . . . See Note D
- 14. Fuel Channel Defornation Analysis Methods . . . . . . . . . . See Note F
- 15. Fuel Assembly Stress Limits '
. . . . . . . . . . . . . . . . . . S ee No te D
- 16. Fuel Rod 0.060 Inch Deflection Justification . . . . . . . . . . See Note D
- 17. Gadolinia Rods Performance Experience . . . . . . . . . . . . . See Note G
- 18. Process Computer Performance Evaluation Accuracy Update . . . . . See Note H
- 19. Lattice Physics Methods Verification . . . . . . . . . . . . . . See Note I
- 20. Boiling Water Reactor Sinulator Verification . . . . . . . . See Note J
- 21. Void and Doppler Reactivity Coefficients . . . . . . . . . . . . See Note K
- 22. Full Power Scram Reactivity Function . . . . . . . . . . . . . . See Note L
- 23. Feedwater Flow Rate Uncertainty Justification . . . . . . See Note M
- 24. Resolution of Feedwater Nozzle Design and Verification.. .. .. . . See Note N
- 25. Description of MUJ1 Code and Loads on Internals During LOCA . . . See Note O 26.
27.
Safety / Relief Valve Surveillance Program Details Update PGCC LTR NEDO-10466 Jection lE.3 See Note P
)
- 28. Analytical Methods of Plant Transient Evaluation . . . . . . . Chapter 15 29.
\s_,/ 30.
ATWS . . . . . . . . . . . . . . . . . . . . . . . . . . . . Section 15.8 Test Program for Safety / Relief Valve Solenoids . . . . . . . . . See Note R
]
- 31. Fire Protection for PGCC . . . . . . . . . . . . . . . . . . . . See Note P Appendix 9A
- 32. Primary Coolant Pump Seals Leakage Characteristics . . . . . . . See Note S
- 33. Large Scale Mark III Test . . . . . . . . . . . . . . . . . . . . . Appendix 3B
- 34. Environmental Design of Isolation valves and Safety Related Equipment . . . . . . . . . . . . . . . . . . . . . . . . See Note T
-35. Post LOCA Panual Operator Actions . . . . . . . . . . . . . . . . See Note U 36.
37.
Instrument and Control Systems . . . . . . . . . . . . . . . . .
HPCS Onsite Electrical Systens . . . . . . . . . . . . . . . . .
Chapter 7 Chapter 8 ]
- 38. Fire Protection for Nuclear' Island Conformance . . . . . . . . . Appendix 9A I
Developnent and Verification Test Programs
- 1. Fuel Surveillance Program . . . . . . . . . . . . . . . . . . . . See Note B
- 2. Subsection 5.2.2 i
3.
Safety Relief Valve Surveillance Progran . . . . . . . . . . . .
Core Spray Distribution . . . . . . . . . . . . . . . . . . . . See Note V
]
- 4. Fast Scran Design Verification . . . . . . . . . . . . . . . . . Se6 No te W
- 5. .Foedwater Nozzle Design Verification . . . . . . . . . . . . . . See Note N
- 6. Long Tern Pipe Replacement Progran . . . . . . . . . . . . . .
- See Note Z )
- 7. Instrunentation for Vibration and Loose Parts Detection . . . . .X
- 8. Pressure Suppression Design Verification . . . . . . . . . . . . Y
.9. Suppression Pool Dynamics . . . . . . . . . . . . . . . . . . . Appendix 3B
- 10. Evaluate Effects of Relief Valve Blow-Down . . . . . . . . . . . Appendix 3B and Chapter 15 r
G i 1.5-13 L
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.5-2 ll FDA COMMITMENT ITEMS (Continued)
NOTES A. (1) Information Report, NEDO-20922-P, Experience with BWR Fuel through September 1974, 7/17/75 (2) Information Report, NEDO-21660-P, Experience with BWR Fuel through December 1976, 12/14/77 B. (1) Letter R. Engel to D. Ross, dated 7/11/77 (2) Letter G. G. Sherwood to D. Ross, dated 4/7/78 C. Licensing Topical Report, NEDO-21175, BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake and Loss-of-Coolant Accident Loadings D. Licensing Topical Report, NEDO-20948, BWR/6 Fuel Design, 2/20/76 E. Licensing Topical Report, NEDE-20606-A, Creep Collapse Analysis at BWR Fuel Using Safe-Colaps Model, Approved by NRC 8/76.
F. Licensing Topical Report, NEDO-21354, BWR Fuel Channel Mechanical Design and Deflections G. Licensing Topical Report, NEDO-20943, Urania-Gadolinia Nuclear Fuel Physical and Irradiation Characteristics and Material Properties l H. Licensing Topical Report, NEDO-20340, Process Computer Performance Evaluation Accuracy, 7/17/74 I. Licensing Topical Report, NEDO-20939-A, Lattice Physics Methods Verification, Approved by NRC 9/22/76.
J. Licensing Topical Report, NEDO-20946-A, BWR Simulator Methods Verification, Approved by NRC 9/22/76 K. Licensing Topical Report, NEDO-20964, Generation of Void and
. Doppler Reactivity Feedback for Application to BWR Design, I
2/13/76 L. Appendix A of the Odyn Report M. Letter G. G. Sherwood to E. Case, dated 5/1/78 O
1.5-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 18
{}
Table 1.5-2 FDA COMMITMENT ITEMS (Continued)
NOTES (Continued)
N. Information Report NEDO-21821-A Boiling Water Reactor Feed-water Nozzle /Sparger Final Report.
O. Licensing Topical Report, NEDO-24048, Evaluation of Acoustic Pressure Loads on BWR/6 Internal Components, 12/1/77 P. Licensing Topical Report, NEDO-10466-A, Power Generation Control Complex Design Criteria and Safety Evaluation, Approved by NRC 7/31/78 Q. Not used. 3 R. (1) Letter G. G. Sherwood to E. Case, dated 2/18/78 (2) Letter S. Varga to G. G. Sherwood, MFN-183-78 (3) Information Report NEDO-23978, ADS Solenoid Valve Reliability Demonstration s,,) S. Licensing Topical Report, NEDO-24083, Recirculation Pump Shaft Seal Leakage Analysis, 12/12/78~
T. Letter G. G. Sherwood To N. Denton, dated, 10/11/78 U. Letter G. G. Sherwood to H. Denton dated 3/22/79 i
V. (1) NEDO-10846, Boiling Water Reactor Core Spray Distribution (2) NEDO-20566-3 Effect of Steam Environment on BWR Core i
Spray Distribution W. NEDO-24142, Fast Scram Control Rod Drive Qualification Program X. Letter, G. G. Sherwood to E. Case, dated 3/8/78
- Y. Licensing Topical Report, NEDO-20533, The GE Mark III Pressure Suppression Containment System Analytical Model, approved by i NRC 8/14/75 (Amended 6/30/78) i i
i OT i l
l l.5-15/1.5-16 i
- GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18
, ( Table 1.5-2 FDA COMMITMENT ITEMS (Continued)
NOTES (Continued) l Z. The GESSAR II design complies with Regulatory Guiue 1.44, Rev. O, and with the requirements of NUREG-0313, Rev. 1.
j Regulatory Guide 1.44 addresses 10CFR50, Appendix A, GDCs 1 and 4, and Appendix B, i.e., requirements for components i important to safety shall be' designed in accordance with appropriate codes and to accommodate the effects of and to
! be compatible with the environmental conditions associated with reactor operations. Experience with operating reactors has demonstrated that certain wrought austenitic stainless steels, when. welded, are sensitized to the degree that they are susceptible to stress corrosion cracking in reacter coolant water environments. With selection'of alloys and fg control of thermomechanical processing, intergranular stress
\s # corrosion cracking (IGSCC) of reactor coolant pressure boundary components can be avoided.
All custenitic stainless steel material that is. fabricated into components which see temperatures in reactor environ-ments greater than 200*F is purchased as low carbon grade or nuclear grade and in the solution annealed condition, in accordance with the applicable ASTM and ASME specifications.
Cooling rates from solution annealing heat treatment tem-peratures are required to be rapid enough to prevent sen-sitization. Resistance to IGSCC is verified using ASTM
'A262, Practice A methods.
Material changes have been made to minimize the possibility of IGSCC. All welded wrought austenitic stainless steel in ,
the reactor coolant pressure boundary is low carbon nuclear 1.5-17
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 Table 1.5-2 FDA COMMITMENT ITEMS (Continued)
NOTES (Continued)
Z. (Continued) grade 316LN with 0.02% maximum carbon content and nitrogen control for strength. There is no piping which is service sensitive or nonconforming as defined in NUREG-0313, Rev. 1.
All other applications of stainless steel (Types 304 and 316) are of the L grade (0.03% maximum carbon content).
Welding heat input controls are required for all stainless steel welds. For machine, automatic and manual welding, interpass temperatures are restricted to 3509F maximum for all stainless steel welds. High heat welding processes such as block welding and electroslag welding are not per-mitted. All weld filler metal consumable inserts and cast-ings are required by specification to have a minimum of 5%
ferrite.
The above practices avoid IGSCC and are reflected in Sub-sections 1.8.44, 4.5.1, 4.5.2, and 6.1.1.
O 1.5-18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18*
3 1.8.117 Regulatory Guide 1.117, Revision 1, Dated April 1978
-d
Title:
Tornado' Design Classification This guide describes a method for identifying those structures,
- systems, and components of light-water-cooled reactors that should be designed to withstand the effects of the design basis i
tornado (Regulatory Guide 1.76,-Design Basis Tornado for Nuclear Power Plants) including tornado missiles and remain functional.
Evaluation The BWR/6 Mark III Standard Plant Tornado Design Classification of structures, systems, and components including their foundations and supports that are designed or protected to withstand the 4
effects of a design basis tornado per Regulatory Guide 1.76 (including tornado missiles) without loss of capability to per-fg form their safety function) are those listed in the Appendix k_s) of Regulatory Guide 1.117 except for portions that are not within the Nuclear Island scope. "
. ~
, The analysis of structures, shields, and barriers indicates tornado winds could damage non-Category I structures. However, ability to shutdown the reactor, integrity of the containraent, and capability of the essential Heat Removal Systems are not impaired. All Seismic Category I systems are protected by being housed in tornado-resistant structures. Collapse of non-Category I towers or stacks will not endanger Category I structures since plant arrangement provides sufficient distance between them.
4 O
1.8.117-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 1.8.117 Regulatory Guide 1.117, Revision 1, Dated April 1978 (Continued)
The specific plant Applicant is responsible for the design and protection of GE out-of-scope-of-supply structures, systems, and components.
O O
1.8.117-2
. - . - - - - . - . . . - - . -- . ~ - - -- -- -.- - . - _.
i' 4
Table 1.9-1
^ .,
j CHAPTER 1
]_ GESSAR II/FSAR INTERFACES (CONTINUED)
+
ITEM L l NO. SUBJECT INTERFACE g DESCRIPTION PAGE SUBSECTION CATEGORY f
i 1.127 Alternative . Provide a costs and benefits com- 1G.12-1 1G.12 3 Hydrogen Con- !
{ parison of the. alternative !
- trol System systems considered for a Hydro- 3 l gen Control System. For the .
- l . selected system, provide design !
- -description, function layout, i l analyses and test data to verify "
] compliance with the requirements 00 s
of (f) (2) (ix) of 10CFR50.34. z. t CO '
l 1.128 Long-Term 1G.13-1 Training Establish a training program which addresses-the concerns related 1G.13 3 hE M U2
" t
, Upgrade to Item.I.A.4.2 of NUREG 0718. $ $. i 4 1.129 Long-Term Establish a program for integrating 1G.14-1 e Program of .and expanding current efforts to 1G.14 3 $U
, . Upgrading of improve plant procedures. -[5:
Procedures 0-hl
] 1.130 Hydrogen Provide a Hydrogen Control System 1G.21-2 1G.21 3
- Control capable of-handling equivalent of
- System a 100% active fuel-clad metal water' reaction. I i '
1.131 Purging Provide performance information 1G.27-1 1G.27 3
' 3
. of purge valves i
ww eM.
- -a i O HO
- 00 4 l r l
~
Table 1.9-1 CHAPTER 1 GESSAR II/FSAR INTERFACES (CONTINUED)
ITEM INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION CATEGORY 1.132 Upgrade Provide Technical Support Center, 1G.37-1 1G.37 3 License an Onsite Operational Support Emergency Center and a near site Emergency Support Operations Facility.
Facility w
1.133 In-Plant Provide monitoring of in-plant 1G.39-1 IG.39 3 $
Radiation radiation and airborne radioac-7 Monitoring tivity for routine and accident $o y conditions.
- c. Mm 1.134 Feedback of Provide administrative procedure 1G.41-1 1G.41 3 y%
i Operating, for evaluating operating, design y Design and and construction experience and y[
Construction ensure applicable important M Experience industry experience is provided to other plants. @
O 1.135 Expansion Ensure that the Quality Assurance 1G.42-1 1G.42 3 of QA List list required by Criterion II, App. B. 10CFR50, includes all structures, systems, and compon-ents important to safety.
1.136 Containment Provide details of containment 1G.44-1 1G.44 5 Penetration penetration arrangement.
1.137 Containment Provide containment vessel design 1G.45-3 IG.45 3 @kj Integrity capability of 45 psig for Service < >
Level C. $
eo 00 -4 I I O O O
i
< O O k
Table 1.9-3 CHAPTER 3 GESSAR II/FSAR INTERFACES (Continued) 2 ITEM NO. SUBJECT DESCRIPTION RELATED INTERFACE PAGE SUBSECTION QUESTION CATEGORY 3.15 Seismic Category Describe other seismic Category I I Structure 3.8-98 3.8.4.1.8 1 I
Descriptions structures Island.
not within the Nuclear 3.16 Codes, Standards Provide applicable codes, standards 3.8-102 and Specifica- 3.8.4.2.5 and specifications for other 1 4
tions seismic Category I structures not within Nuclear Island. .
N
- 3.17 Loads and Load u Combinations Provide loads and load combina- 3.8-114 3.8.4.3.8 *
! H tions appropriate to Seismic 1 j
- Category I structures not within e Nuclear Island g@O n
a tn l
f 3.18 Design and Provide design and analysis g%
w Analysis 3.8-126a 3.8.4.4.8 1 h*
O Procedures procedures for seismic Category I structures now within Nuclear g Island tn 3.19
, (Deleted)
?
%N m
<: >N
. 4 j O HO mW i
I i
l i
i
Table 1.9-3 CIU4PTER 3 GESSAR II/GSAR INTERFACES (Continued)
ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 3.20 Materials Provide structural acceptance 3.8-128 3.8.4.5.3 1 Criteria criteria for Seismic Category I structures not within Nuclear Island.
3.21 Materials, QA Describe materials, quality 3.8-130 3.8.4.6.7 1 and Special control, and special construction Construction techiques for Seismic Category I Techniques structures not within Nuclear M Island.
a3 f 3.22 Testing and Inservice Describe testing and inservice, ,
3.8-131 3.8.4.7.4 1 $g e inspection requirement 3 tor Selsmic n F1 Inspection Category I structures not within f Nuclear Island.
p@
w >. >.
us us 1 3.23 Verification of Provide verification, through 3.8-142 3.8.6.2 3 g, g, Foundation Soil Applicant's soils consulting U3 H engineer, that the Nuclear Island foundation soil meets the soil %
Z requirements of Appendix 3A, O subsection 3.8.6 and 3.7.
3.24 Containment Provide a summary of the contain- 3.8-146 Table 3.8-3 1 Vessel Analysis ment vessel analysis.
Summary 3.24.1 NUREG-0 8 0 3 Update computer programs and 3.9-la 3.9.1.2 3.151[MEB (DSER) 3 Compliance on indicate method of verification Item No. 39]
Computer and the version used.
Programs 3.24.2 Pump and Motor Provide a list of the pump and 3.9-lb 3.9.1.2 3.151[MEB (DSER) 2 Vendor Computer motor vendor computer programs Item No. 39) mu Programs along with statement of verifi-cation approach. Qy
. q O
3.24.3 Program Verification Provide method of verification and version of PDG1 used.
3.9-24 3.9.1.2.6.3.1.2 5 [Oy i
1 O O O
N. -
J4 Table 1.9-3 CHAPTER 3
- GESSAR II/FSAR INTERFACES (Continued) 1 i
ITEM NO. SUBJECT RELATED INTERFACE DESCRIPTION PAGE SUBSECTION _ QUESTION CATEGORY 3.24.4 Inservice Provide details of the pump and Testing of 3.9-125 3.9.6 3 valve inservice testing program, and Pumps and . including test schedules and 3.9-126 i Valves frequencies. Also, applications i
for written relief from Section XI
- l Addendum requirements, pursuant to 10CFR50, Section 55 a(g) (6) y w
' (i). "
E 3.25 Major Safety- Provide equipment loading condi- z j y Related tions, stress criteria, limiting 3.9-177 Table 3.9-11 2,4 $0 a- Mechanical stress types, allowable and calcu-(2) t< us g Components lated stresses on a component-by- 3.9-179 $! $
i Table 3.9-11 ms ss component basis.
W through (3) p, g, j 3.9-185 U2 H H
i 3.9-187 Table 3.9-11 $
through (4) t) i 3.9-190 3.9-191 Table 3.9-11 through (5) 3.9-192 i
4 i
4 i
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. 4 O
l8 0
' 00 4
. I i
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3 _ _ - _ _ . . ---.m m _;--u -__.a-, .u _ _; -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 18 LO I
APPENDIX 1D ASSESSMENT OF
! REGULATORY GUIDE 1.97, REVISION 2
{ AGAINST GESSAR II DESIGN i
1D.1
SUMMARY
1 This report provides an assessment of the 238 Nuclear Island design described by GESSAR II against the guidelines of Regulatory Guide.l.97, Revision 2, dated December 1980, " Instrumentation for Light Water Cooling Nuclear Power Plant to Assess Plant and Envir-onmental Conditions During and Following an Accident." This
- i assessment is summarized in Table 1D-0.
1D.l.1 Introduction Regulatory Guide 1.97 describes a method acceptable to the NRC staff for compliance with the Commission's regulation to provide instrumentation to monitor plant variables in systems during and i-following an accident at light water cooled nuclear power plants.
Although the intent of the Regulatory Guide has been met, several exceptions have been made to the Regulatory Guide as written.
These exceptions and their bases are discussed further in this report. These exceptions are viewed as acceptable means to imple-ment the guide's intent. '
) All variables identified by the regulatory guide are identified in this report. Table 1D-1 summarizes these variables. However, since some variables are outside of the 238 Nuclear Island scope,
, they are left for the Applicant to assess.
io i
i 1D .1 -, . _--- -, =..,..-,-..-,_-...,.,_._.,-_.__e....., - . - - - - . _ _ .,.-.r_---...-.. . . . . . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.l.1 Introduction (Continued)
Since there are other modifications being made to address post-TMI changes (Appendix 1A), the assessment makes use of the exist-ence of these changes and does not repeat a description of them here.
1D.l.2 Objective This assessment is conducted in three stages. The first stage develops an implementation position which defines the specific requirements against which each variable is assessed. The imple-mentation position is included as Attachment A and is discussed in Subsection 1D.2.1.
The second stage of the assessment defines the Type A variables.
These variables are identified from a review of design basis accidents in Chapter 15 and the Emergency Procedure Guidelines.
A more detailed discussion of the Type A variables is given in Subsection 1D.2.2.
The third stage of the assessment is a review of each variable shown on Table 1D-1. For this review, statements are provided to either justify the current design or to provide a conceptual definition of changes necessary to comply with the regulatory guide.
These discussions may be found in Subsection 1D.2.3.
1D.l.3 Conclusions Table 1D-2 summarizes the results of the assessment.. The degree of compliance is shown by the information in the table and the accompanying remarks. i l
Table lD-2 shows that the 238 Nuclear Island, as modified by the recommendations under review as a result of this assessment, fully !
complies with the intent of Regulatory Guide 1.97, Revision 2.
' 'lD .1-2
_ GESSAR II 22A7007 s
238 NUCLEAR ISLAND Rev. 18
/N 1D.2.3.3 Reactor Coolant Boron Concentration b '
Boron concentration in the reactor water is determined by analysis of a reactor water sample obtained from the post-accident sample station (refer to Subsection 1D.2.3.38). Recommended boron analy-sis procedures are included with the operation and maintenance manual supplied with the sample station. Actual procedures are left for the applicant to provide.
1D.'2.3.4 RPV Water Level Indication 1
The existing post-accident water level indication system consists of three wide-range instruments (calibrated for full pressure) covering the range -160" to +60" and two fuel zone instruments (chlibrated for atmospheric pressure) covering the range -326" to
-116". Narrow-range indication is not included in the assessment.
ji The RPV water level indication system is the primary variable
'- iddicating 't;he availability of adequate core cooling and is con-sidered acceptable, without diverse methods of indication, pro-vided adequate redundancy and unambiguity are provided over the entire range of interest.
The range for coolant level in the reactor specified on Table 1D-1 of the implementation position has been modified to require indi-cation only "above normal reactor water level" rather than the
" centerline of main steam line" as specified in the regulatory guide. This change was made because the high reactor water level transient has been shown to be not a safety concern (Subsec- ]
tion lA.23) especially considering the existence of improved high reactor water level trips on HPCS. Furthermore, since the presence of reactor water on the normal range assures adequate core cooling, higher ranges of indication are not needed for this purpose.
s Although Contingencies 5 and 6 of the Emergency Procedure Guide- -
)
()
\>
lines discusses > flooding of the main steam lines for alternate shutdown cooling and instrument reference legs refill, these ,
, , 1D.2-5 v.
. ~.
GESSAR II 22A7007 239 NUCLEAR ISLAND Rev. 18 1D.2.3.4 RPV Water Level Indication (Continued) actions are not needed to achieve adequate core cooling. Other -
instrumentation is used to carry out high water level flooding procedure (see Attachment A).
O 1
e O
1D.2-Sa
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10
() kD.2.3.4 RPV Water Level Indication (Continued)
The requirement in Table 1D-1 of the regulatory guide for BWR core thermocouples has been replaced with a requirement that the reactor water level indication should extend from below the core plate to above the pressure vessel Level 1 as adequate indication of core cooling. Through work by General Electric and BWR Owners' Group (References 2 and 3), the NRC staff has recently retracted the requirement for BWR core thermocouples (Reference 4), but they have requested that an investigation continue to identify an acceptable diverse means of indicating adequate core cooling.
This investigation is still in progress. Work by General Electric and the BWR Owners' Group (Reference 3) has provided the NRC staff with extensive information on the relationship between the reactor water level and adequate cooling of the core. These studies have shown that as long as at least one of the water delivery systems is available and flowing to the reactor pressure vessel, that ade-I)
V quate core cooling exists. Indication of these flows is already required by the Regulatory Guide. Consequently, it is General Electric's view that no instrumentation other than RPV level indi-cation is required to assure indication of adequate core cooling.
Subsection C.1.3.1.b states that additional instrumentation should be provided to allow the operator to determine the actual condi-tions of the plant "when failure of one accident monitoring channel results in information ambiguity." Three independent channels of indication are considered acceptable to meet this provis o During a small-break accident which results in increasing drywell temperatures, the accuracy of the RPV water level indication varies significantly. This effect is described in detail in Reference 2.
A small drywell break could lead to ambiguity in all instrument
() ranges either because the redundant channels would not agree (if one failed) or because of increasing drywell temperatures and its 1D.2-6
-... . - _ _ . . _ = - _ - _ _ _ . - . - . . . . - . _ - _ .
. _ ~ _ _ - _ . . - - . . . - - . .
i GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 18 !
5 1D.2.3.12 Primary Containment Pressure i
i Also, 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> are available before reaching the top of the j indicating range (Table 18.2-3a) , which provides suf ficient time to open MSIVs or re-establish containment cooling.
i i
l !
! 1 I ,
i i
i
?
I .
1 i
i t LO l
l i
i F
l O
i 1D.2-12a
_ . - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - - _ _ . _ _ _ . - . ~ _ _ _ . - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 10 1D.2.3.12 Primary Containment Pressure (Continued)
Two channels of pressure instrumentation are adequate in case of a single failure because diverse indications to monitor containment integrity are provided.
~
No modifications are needed for this instrumentation. However, the instruments including displays should be included in Tables 3.10-1 and 3.11-9 to ensure qualification.
1D.2.3.13 Drywell/ Containment Hydrogen Concentration Drywell and containment hydrogen concentration may be determined 1
by analysis of samples obtained from the post-accident sample station (Subsection 1D.2.3.38). On-line instrumentation is the responsibility of the Applicant to provide, fs ID.2.3.14 Secondary Containment Area Radiation j
Area radiation' levels in the secondary containment are defined in Section 12.3. The specific location and range of area radiation monitors are the responsibility of the Applicant to provide.
1D.2.3.15 Secondary Containment Noble Gas Effluent i'
Applicant to provide.
1D.2.3.16 Containment Noble Gas Effluent Applicant to provide.
1D.2.3.17 Suppression Pool Temperature i
Two temperature sensors per quadrant of the suppression pool are provided in the 238 Nuclear Island design with control room (V~]
1D.2-13
- - - - , . - - - -..-r ,, ,-,-~<__.-,--r,,y-,, .,, - ,--, . . , - - - , _ , , . _ , , - - , , , - - , - - , - , , - . ---. -,---,,,y
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 18 1D.2.3.17 Suppression Pool Temperature (Continued) indication and recording. The instruments are discussed in Subsections 7.6.1.11 and 7.6.1.12.
The sensors are located in the upper third of the suppression poal and thus provide a conservative indication of suppression pool temperature for use on the Emergency Procedure Guidelines.
No modifications are needed to these instruments. The instruments including the displays should be included in Tables 3.10-1 and 3.11-9 to ensure qualification.
1D.2.3.18 Drywell Air Temperature Two Class lE channels of drywell temperature indication are pro-vided on the control room BOP benchboard (P870) as shown in Fig-ure 9.4-5.
The range of the display (up to 400*F) is adequate to carry out the functions prescribed by the Emergency Procedure Guidelines t (Reference 1). Highest post-LOCA drywell temperature is approx-imately 340'F for a main steam line break (Figure 6. 2-12) .
No modifications to the design are needed. The instruments l including displays should be included on Tables 3.10-1 and 3.11-9 1
to ensure qualification.
1D.2.3.19 Coolant Radiation No instrumentation is provided in the current design to monitor l radioactivity levels in the primary coolant and no changes to the l plant design are planned.
1 1
The specified range for the potential instrument (1/2 Technical Specification Limits (TSL) to 100 times TSL) suggests that the l
l 1D.2-14 l
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 18 1D.2.3.19 Coolant Radiation (Continued) purpose of this instrument is to assess coolant radiation level during routine plant operation. The current design provides sampling capability for reactor coolant as described in Subsec-tion 9.3.2 and provides offgas and mainstream line process radi-ation measurement as discussed in Section 11.5 for detection of fuel cladding breaches.
]
The value for the technical specification limit has not been established by the staff in standard technical specifications for BWR/6. Subsection 16.3/4.4.5, however, indicates that the TSL is 2 pCi/g Iodine-131 equivalent. On-line reactor coolant monitoring of this level of coolant activity may be impractical during normal operation because of the additional contributions to the detector from other isotopes such as circulating Nitrogen-16 or Cobalt-60 deposited on reactor coolant piping.
O_/
s Purthermore, should a significant breach of the fuel cladding occur the expected levels of iodine in the reactor coolant would far exceed the upper range specified by the regulatory guide for this instrument. The samples provided by the post-accident sam-ple system' (Subsection 1D.2.3.38) will provide quantification of the coolant radioactivity level. Post-Accident Sample Station "
identifies gross gamma indication (Figure 1AB.1-la). Under these .
conditions, an on-line monitor would serve no useful purpose.
1D.2.3.20 Coolant Gamma Sample The radioactivity content of the reactor water is determined by analysis of a reactor water sample obtained from the post-accident -
sample station (Subsection 1D.2.3.38). Recommended procedures to determine the gross activity in the coolant are included with the operation and maintenance manual supplied with the sample station. ..
N Actual procedures are the responsibility of the Applicant.
J 1D.2-15
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 18 1D.2.3.21 MSIV Leakage Control System Pressure The current design uses a Class lE positive leakage control system as described in Section 6.7. Proper system function is monitored and recorded by air system flows rather than system pressure as specified by the regulatory guide. System isolation automatically occurs on high flow or low differential pressure between the RPV and the pressurized lines.
The flow monitoring instrumentation is considered adequate to meet the intent of the regulatory guide to indicate proper system function. No changes are planned for this system.
1D.2.3.22 RHR System Flows The RHR system Loops A and B serve a variety of functions among them being low pressure coolant injection, containment spray, suppression pool cooling, and shutdown cooling. Loop C is only used for LPCI mode of operation. The valving arrangements (refer to Figure 5.4-12a) required to achieve these di.fferent functions of the RHR System occur downstream of the flow element and flow transmitter which is used to indicate RHR System flow. This instrument channel therefore provides the operator with flow indication during any of these operating modes for the RHR System.
From an operational point of view, proper functioning of the con-tainment spray mode of the RHR System is provided by the contain-ment temperature (Subsection 1D.2.3.ll) or containment pressure instrumentation (Subsection 1D.2.3.12). Should the containment spray mode be used, it is anticipated that the operators would only initiate flow long enough to decrease these containment parameters at which time flow in the containment spray mode would be termi-nated. Thus, the primary indicator of proper containment spray mode operation is the containment pressure or temperature indica-tion rather than the RHR System flow.
1D.2-16
O O O Table 1D-0
SUMMARY
OF ASSESSMENT OF REGULATORY GUIDE 1.97 AGAINST GESSAR II DESIGN GESSAR II Variable Deviation Justification Reference Remarks
- 1. Neutron Flux New system not yet e Will meet Reg. Guide 1.97 1D.2.3.1 -
approved by NRC
Pressure sidered as alter- (MSIV excess leakage) or Relief nate variable (FRS zero flow indicates R607, FRS 627) system malfunction. m
- 3. RHR Heat RHR Service Water e Meets all Category 2 1D.2.3.23 -
F' Exchanger Temperature flow (FI R602A, B) considered an alter-requirements $o OM nate variable e Positive indication of N$
i (Figure 9.2-la) heat removal capability $$
O
- 4. SLC Flow SLC pressure con- e Indicates proper system 1D.2.3.25 Qualification $U sidered an alter- function of indicator nate variable requ red.
- High pressure indicates (PI R600) flow blockage ghangeto Essential
- Erratic or low pressure MPL classi- !
indicates line break fication e Back-up information available
- Squib valve position
- Neutron flux
- SLC tank level mw aw O
i l
Table 1D-0
SUMMARY
OF ASSESSMENT OF REGULATORY GUIDE 1.97 AGAINST GESSAR II DESIGN (Continued)
GESSAR II Variable Deviation Justification Reference Remarks
Vent Flow an alternate (Figure 7A.3-9A)
Rate variable e Design flow rate may be used for release assess-ment if isokinetic flows are unavailable w w
y and B injection are qualified; indication $o valve position and RHR flow is alter-in control room py mm i nate variable e RPV water level is a $$
$ (Figure 5.4-12b) back-up variable. gg mH
- 7. Contaigment Containment spray e Valve position sensors 1D.2.3.22 -
Spray slow valve position and are qualified indications RHR flow is alter- U in control room nate variable e Containment pressure indication is back-up variable
- 8. RPV Level Alternate range for e Existing shutdown range 1D.2.3.4 -
Category I indica- is adequate to main steam tion is to top of lines (meets Category 3) wide range zone (Note: Deviation gy assumes enhanced Qg level instrument .w (ELI) to meet the 98 m4 intent of RG 1.97)
O O O
o O O Table 1D-0
SUMMARY
OF ASSESSMENT OF REGULATORY GUIDE 1.97 AGAINST GESSAR II DESIGN (Continued)
Variable Deviation GESSAR II Justification Reference Remarks
Temperature measurement satis- indication of approach fies requirements to and existence of ICC for core cooling i
indication e EPGs provide operator actions to prevent ICC e ELIs Third Division U eliminates need for "
diversity.
g gg o o to u
e ECCS flow indication E$
provides indication of a recovery from ICC gg tr (supported by NEDO-24708-A analysis) $[
- 10. Containment Existing range and e 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> available before
- Pressure method of indica-1D.2.3.12 Change design U i
reading top of indicating to 0-60 psig
' tion adequate range - time available to range. Change open MSIVs or re-establish pressure j
conCainment cooling indicator to (Table 8.2-3a) recorders.
i (RG 1.97, e Trend information is use- Rev. 2 re-ful, but not essential quires all to follow EPG. Category 1 channels to be recorded. $U
?D O HO CD 4 I
Table 1D-0
SUMMARY
OF ASSESSMENT OF REGULATORY GUIDE 1.97 AGAINST GESSAR II DESIGN (Continued)
GESSAR II Variable Deviation Justification Reference Remarks
- 11. Suppression The existing range e Low End of Scale 1D.2.3.9 -
Pool Water is adequate for Level .
expected situations - Upper pool dump ini-tiates at 18'5" (raises water level)
- High suppression pool level causes auto trans-fer to pumps to contain- [
ment to pool (20'5") m s
,U
- Lower tap risks failing following SRV action or
$a OM
^
vent clearing $$
$ High End of Scale NN
- Drywell sump level $ U.
indicates weir flow
- ADS causes 2 5' pool
{o rise (within scale of indication)
- 12. Drywell Air The lower range is Highest post-LOCA drywell 1D.2.3.18 -
Temperature acceptable temperature is =340*F for a main steam line break.
(Figure 6.2-12)
- 13. Coolant Post-Accident Sample e Interference by N-16 and 1D.2.3.19 -
Radiation System provides Co-60 may make monitor-adequate information ing of TSL impractical @U or ambiguous <g o
HO mW O O O
O O O i
Table 1D-0
SUMMARY
OF ASSESSMENT OF REGULATORY GUIDE 1.97 AGAINST GESSAR II DESIGN (Continued)
GESSAR II Variable Deviation Justification Reference Remarks
- 13. Coolant e Significant fuel release Radiation would exceed 100 TSL (100 (cont) TSL is not a public health risk) i e Post-accident sample station identifies gross gamma indication (Figure N 1AB.1-la) - indicates
- radiation of circulating z j sample prior to sample g
O $@
l 'u 14. Post-Accident Sump samples are e Sumps are isolated (not 1D.2.3.38 -
E$
f Samples not needed sources of release) **
a HH i
e RPV or suppression liquid $"
sample provides indica- E 4
tion of extent of core O damage e Process radiation monitors are used for release assessment e Analysis of sumps would be ambiguous Ww l @M v <>.J l
o l Ho mw 6
l i
Table 1D-0
SUMMARY
OF ASSESSMENT OF REGULATORY GUIDE 1.97 AGAINST GESSAR II DESIGN (Continued)
GESSAR II Variable Deviation Justification Reference Remarks
- 15. Drywell Sr_mp Category 3 (commer- e Sumps isolated by 1D.2.3.6 Add Category Level cial, single chan- accident 3 differen-nel) requirements tial pressure are acceptable e Back-up indication to wide range drywell pressure and instrumenta- ,
radiation level tion e Early indication only [j m
, cQ nm a e Air supply reliability M o is assured until 5ym isolated air supply ss bleeds off, mH E
O
- w O
HO WM O O O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 f
APPENDIX 1G LIST OF TABLES TABLE TITLE PAGE 1G.0-1 GESSAR II - CP/ML Rule Cross Reference 1G.0-3
, 1G.12-1 Comparison of Hydrogen Control Alternatives 1G.12-3 ]
' O LIST OF ILLUSTRATIONS FIGURE TITLE PAGE 1G.47-1 GESSAR II Organizational and Management Structure 1G.47-4 1G.47-2 Top Level Management Oversight and Technical Control 1G.47-5 l
l O
1G-v/lG-vi
. ~ . . . . . , . . . . . - . . , . , , . , . . , - - , _ _ , . , _ , . , . - - . - - . - , . . . , - . . , - , . . - - - - . _ . . . . -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 1G.12 EVALUATION OF ALTERNATIVE HYDROGEN CONTROL SYSTEMS O [ Item (1) (xii)]
NRC Position Perform an evaluation of alternative hydrogen control systems that would satisfy the requirements of paragraph (f) (2) (ix) of 10CFR50.34(f). As a minimum include consideration of a hydrogen ignition and post-accident inerting system. The evaluation shall include:
(A) A comparison of costs and benefits of the alternative systems considered.
(B) For the selected system, analyses and test data to verify compliance with the requirements of (f) (2) (ix) i of 10CFR50.34.
() (C) For the selected system, preliminary design descrip-tions of equipment, function, and layout.
Response
I
- (A) GE has conducted evaluations of the various hydrogen control concepts for the GESSAR II design. These con-cepts included distributed ignition, catalytic burners and post-accident inerting with carbon-dioxide or halon.
Of the concepts considered, only distributed ignition (igniters) and carbon-dioxide post-accident inerting appear to be viable alternatives. The costs and bene-fits of these two alternatives are summarized in Table 1G.12-1. -
(B) The Applicant will provide the analyses and test data to verify compliance with the requirements of 10CFR50. 34 (f) (2) (ix) ,
lO
! (C) The Applicant will provide the design descriptions of equipment, function, and layout.
1G.12-1/lG.12-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 73 Table 1G.12-1
\ ,)
COMPARISON OF HYDROGEN CONTROL ALTERNATIVES Item Igniters Post-Accident Inerting Description Distributed ignition systems Liquid CO2 discharged controlled burn at low H2 into containment air-concentration space (prevents combustion)
Cost (Order $1 Million* $10 Million*
of Magnitude)
R&D Concerns Flammability, mixing, pres- Possible partial sure response inerting flammability characteristics.
Mixing, effects on electronic equipment R&D Programs Underway EPRI/NRC None planned Pros e Minor impact of inadver- o No heat loads tent operation e No dependence on H2 e Low cost generation rate e Minimum design impact e Minor impact on e Lower containment existing equipment
\ pressures e AC power not required for inerting Cons e Potential for large equip- e Inadvertent actua-ment qualification program tion has potential e Assurance of combustion at adverse impact on low concentrations plant operation e Sensitize to hydrogen e High containment generation rate and con- pressure tainment entry point e High cost e Requires active heat e Some redesign of removal containment piping to accommodate o Potential adverse effects from low temperatures during injection
- These costs do not include the cost of corresponding equipment qualification programs. Inclusion of equipment qualification costs could result in nearly equal total costs for the first Applicant referencing GESSAR II.
C\
V 1G.12-3/lG.12-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.21
(x HYDROGEN CONTROL SYSTEM PRELIMINARY DESIGN [ Item (2) (ix)]
V NRC Position 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 preferred system option of those being evaluated in paragraph (1) (xii) of 10CFR50.34(f) is sufficient at the construction per-mit 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 metal-water reaction, or that the post-accident atmosphere will not r's. support hydrogen combustion.
tV)
(B) Combustible concentrations 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.
(C) Equipment necessary for achieving and maintaining safe 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 including the environmental conditions created by activation of the hydrogen control system.
/\ *
! 3
'J 1G.21-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 18 1G.21 HYDROGEN CONTROL SYSTEM PRELIMINARY DESIGN [ Item (2) (ix)]
(Continued)
(D) If the method chosen for hydrogen control is a post-accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation.
Response
The Applicant will provide a Hydrogen Control System capable of handling hydrogen generated by the equivalent of a 100% active fuel-clad metal water reaction. Detailed descriptions of the j selection of the hydrogen generation event, the progression of the event, and the methodology used in evaluating the _tydrogen generation rate during the hydrogen generation event and contained in the April 1982 BWR/6 Mark III Hydrogen Control Owners' Group report.*
The Hydrogen Control System shall provide with reasonable assur-ance that:
(1) Uniformly distributed hydrogen concentrations in the 1
l containment do not exceed 10% during and following an l accident that releases an equivalent amount of hydrogen as would be generated from a 100% fuel-clad metal l
l water reaction, or that the post-accident atmosphere will not support hydrogen combustion.
l l
l
- S. S. Dua, et al., "BWR/6 Mark III Hydrogen Control Owners' Group Report on Hydrogen Control: Accident Scenarios, Hydrogen Generation Rates and Equipment Requirements," General Electric Company Report, April 1982. _
1G.21-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 18 1G.21 O HYDROGEN CONTROL SYSTEM PRELIMINARY DESIGN [ Item (2) (ix)]
(Continued)
Response (Continued)
(2) Combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
(3) Equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integ-rity will perform its safety function during and after being exposed to the environmental conditions attendant with the release of hydrogen generated by the equivalent O
O 1G.21-2a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 0 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS Certain structures, components, and systems of the nuclear plant are considered important to safety because they perform safety actions required to avoid or mitigate the consequences of abnormal operational transients or accidents. The purpose of this section is to classify structures, components, and systems according to the importance of the safety function they perform. In addition, design requirements are placed upon such equipment to assure the proper performance of safety actions when required.
3.2.1 Seismic Classification Plant structures, systems, and components important to safety are designed to withstand the effects of a safe shutdown earthquake (SSE) and remain functional if they are necessary to assure:
/~' (1) the integrity of the reactor coolant pressure boundary; or (2) the capability to shut down the reactor and maintain it in a safe condition; or J
. (3) the capability to prevent or mitigate the consequences
- of accidents which could result in potential offsite
- exposures comparable to the guidelines exposures of 10CFR100.
Plant structures, systems, and components, including their founda-tions and supports, that must remain functional in the event of a safe shutdown earthquake are designated as Seismic Category I as.
indicated in Table 3.2-1, l
r Structures, components, equipment, and systems designated as Safety Class 1, Safety Class 2, or Safety Class 3 (see Subsec-tion 3.2.3 for a discussion of safety classes) are classified as 3.2-1 l
GESSAR II 238 NUCLEAR ISLAND 22A7007 pey, 18 3.2.1 Seismic Classification (Continued)
Seismic Category I except those portions of the radioactive waste treatment handling and disposal systems whose postulated simulta-neous failure would not result in conservatively. calculated poten-tial offsite exposures comparable to the guideline exposures of 10CFR100.
All Seismic Category I structures, systems, and components are designed to withstand, without loss of function, the appropriate seismic loads (as discussed in Section 3.7) in combination with other appropriate loads.
The seismic classifications indicated in Table 3.2-1 and shown in Figure 3.2-2 meet the requirements of NRC Regulatory Guide 1.29 except as otherwise noted in the table and as discussed in Subsection 1.8.29. 3 3.2.2 System Quality Group Classifications System quality group classifications as defined in NRC Regulatory Guide 1.26 have been determined for each water, steam, or radio-active waste containing component of those applicable fluid systems relied upon to:
(1) prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary; (2) permit shutdown of the reactor and maintain it in the safe shutdown conditions; and (3) contain radioactive material.
O 3.2-2
~
V Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality Safety Classi-Principal Component" Assurance Seismic Classb Location" fication Requirement
- Category f Conunents II (Continued)
- 8. Pipin - other beyond outer-most isolation valves
- a. RPV heat vent 3 D C B I
- b. Main steam drains 2 A B B I y ta
- 9. Piping - instrumentation beyond 2/Other D "
B/D B/N/A I/N/A (g) g outermost isolation valves
. @g
- 10. O to Y Safety / relief valves 1 D A B e
(n I N$
] 11. Valves - main steam and feed- 1 C,D A MM
- B I water isolation valves to H j 12. Valves, other - isolation valves N
} and within outermost isolation g
, valves j a. RPV heat vent valves 1 D A B I (g)
- b. Main steam drain valve 1 D,A A B I (g)
- c. Main steam - first valve 2 A A B I (g)
] downstream of isolating valves
=
- 13. Valves - instrunentation beyond 2/Other A B/D outermost isolation valves B/N/A I/N/A (g) 4 f
j 14. Mechanical modules - instru- 2 C N/A B I mentation with safety function m
<: >w l 15. Electrical modules with safety 2 C .w N/A B I (i)
{ function
_. AM 3.116-g ,
I l
Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality Safety Classi- Assurance Seismic Principal Component" Classb Location fication d Requirement
- Category f Comments II (Continued)
- 16. Cable with safety function 2 C,D,A,X N/A B I III Reactor Recirculation System
- 1. Piping 1/2 D A/B B I (g) [
m
- 2. Pipe suspension - recirculation 1 D A B I Z g line CO
. nm En 7
H
- 3. Pipe restraints - recirculation other D N/A N/A N/A >$
m line ##
HH
- 4. Pumps 1 D A B I
- 5. Valves Z 1/2 D A/B B I (g) U
- 6. Motor - pump other D N/A B I
- 7. Cable with safety function 2 X,C,A,X N/A B I
- 8. LFMG Set Other T N/A N/A N/A IV CRD Hydraulic System
- 1. Valves - scram discharge 2 C B B I (g) volume lines y<: >U
- 2. Valves insert and withdraw 2 C B B I (j)
- o lines yy 3.121 3.121 3.121 e e e
O O Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality Safety Assurance seismic g Principal Component Classb Iocation Classi d Requirement, Category fication Comunents XXIX (Continued)
- 4. Branch lines of MFL between the 2 A B B I (u) second isolation valve and the shutoff valve from the branch point at MFL to and including the first valve in the branch y line g
- 5. Main steamline piping between Other T D N/A N/A (u,v) @g g
O tG
- the MSL shutoff valve and the
" turbine main stop valve N$
w >>
H WW
- 6. Turbine bypass piping Other T D N/A N/A (u) tn H Branch lines of the MSL between Other T D N/A N/A (u,v) {
7.
Z the MSL shutoff valve and the C2 turbine main stop valve
- 8. Turbine stop valve, turbine Other T D N/A N/A bypass valves, and the main steam leads from the turbine control valve to the turbine
! casing i
! 9. Feedwater system components Other T D N/A N/A
- beyond the MFL shutoff valve I Nw Other T N/A N/A N/A (D M i.
- 10. Turbine generator <>-a
- 11. Condenser Other T N/A N/A N/A g@
7 co -J l t i I t I
i
Table 3.2-1 EQUIPMENT CLASSIFICA'iION (Continued)
Quality Group Quality Safety Classi-Principal Component" Assurance Seismic Classb Location fication Requirement Category Comments XXIX (Continued)
- 12. Air ejector equipment Other T N/A N/A N/A
- 13. Turbine gland sealing system Other T D N/A N/A components w
w XXX Offgas System co w
w 1. Tanks Other T D N/A (p) 5a l N/A O]
N tn N 2. Heat exchangers Other T D N/A N/A (p) $$
- 3. Piping Other T D N/A N/A (p, q, s) HH tn H Pumps h
- 4. Other T D N/A N/A (p, s) y
- 5. Valves - flow control Other T D N/A N/A (p, q, s)
- 6. Valves - other Other T D N/A N/A (p, q, s)
- 7. Mechanical modules with safety Other T,A D N/A N/A (p, q, s) function
- 8. Pressure vessels Other T,A D N/A N/A (p)
- 9. Recombiners Other T D N/A N/A (p) lC N (D w
- -a
.H < O 1
I e -
6 0
s PJ O Table 3.2-1
. EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality i Safety Classi- Assurance Seismic
- Principal Component" Classb Location fication Requirement" Category Comments
, XXXI Standby Gas Treatment System j 1. Filters 2 R N/A B I
- 2. valves - ductwork 2 A,R,C I
N/A B I
, 3. Cable with safety function 2 R,A,C,X N/A B I N w
co
- 4. Fans and motors 2 R N/A B I 2 g 1 w CO O to NI' Chilled Water Systems w
XXXII yy ,
" WW
- 1. Control Building 3 D,C,A C B I HH in H
- 2. Electrical switch gear 3 A C B I Z
- 3. Other buildings U Other A,R,W D N/A N/A XXXIII HPCS Service Water System This system is included under group /MPL XXXIV/P41.
1 XXXIV Essential Service Water System
- 1. Piping 2,3 O,A,C B/C B (g)
I gM
- 2. Pumps <>
3 P C B I
- y Ho j 3. Pump motors 3 P N/A B I l
l t i i
3.116
Tablo 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality Safety Classi- Assurance Seismic Principal Component" Classb Locationc fic tion Requirement Category Comments XXXIV (Continued) ' .
- 4. Valves - isolation 2 C,A B B I
- 5. All valves that are not, iso- 3 A,C,R,S C B I lation valves
- 6. Electrical modules with safety 3 A N/A B I j
function w-cn
- 7. Cable with safety function 3 A,0,P,X N/A B I g
[ 8. f:enessential portions Other A,C,R,W D N/A N/A bO w td tn I M th w >>
XXXV Closed Cooling Water System ,
HH th H
- 1. Piping and valves forming part 2 A,C,R B B I
{
of primary containment z boundary U
- 2. All piping and valves not Other A,C,D,R,W D N/A N/A
' forming part of primary con-tainment boundary NXXVI Condensate and Demineralized Water Storage and Transfer
- 1. Piping and valves forming part 2 C B B I of the containment boundary mN w
- 2. Condensate storage tank Other O D N/A N/A (w) o#
< q O
H .O CD 4 1 l 9 O O
3 a
) ..
(" 'j
. . .~ , '. f ,.
z
% a e "i Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Qua'lity '
)
Group Quality s-
- / Safety Principal C.amponenta c Assurance Classi d Requirement e Seismic f Class L
- i' Location fication Category Comments ;
i XLI Water Positive Seal; System , ' /
a p
< l .' Pipirdg '
2 A,C B B I i-i
- 2. Valves , 2 A,C B B I i
- 3. Tank 2 A B B I w w
CD t w XLII Air Positive Seal System N
'$on tn i 1. Accumulator 2 A,R M to W B B I M Cn
.a >>
- 2. Piping 2 MM A B B I gg tn H
- 3. Valves 2 A B B I h 2
- 4. Compressor 2 A t3 B B I
- 5. Compressor motors 2 A B B I
- 6. Electrical modules 2 A B B I
- 7. Cables 2 A N/A B I XLIII Breathing Air System M
- 1. Piping and valves gy Other C,A D N/A N/A !
< -J
- 2. Air purifier and other equipment Other C,A D N/A N/A o
o4 ;
l
N. Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued) s s
Quality Group Quality
' Safety Classi- Assurance Seismic Principal Component" '
Classb Location fication Requirement Category Comments XLIV Plant Electrical Systens (Applicrat to Supply)
XLV Auxiliary AC Power System
- 1. u All components with safety 2/3 A,C,X N/A (g) function -
B I $
f $g N a t:1 XLVI Diesel Generator Systems
- 1. Fuel oil storage and transfer 3 S,0 C B I (y) ys system MH
- 2. Cooling water system h 3 S C B (y)
I h
- 3. Starting air tanks receivers, 3 S C B I (y) piping from and including
- check valve and downstream piping and valves
- 4. Starting air compressor and Other S N/A N/A N/A motors
- 5. Lubrication system 3 S C B I (y)
- 1 M
- 6. Combustion air intake and exhaust system 3 S,0 C B I k$
o wo l I 3.116 I
i i 3.118 G G G
GESSAR II 22A700i 238 NUCLEAR ISLAND Rev. 18 Table 3.2-3
\_)
SUMMARY
OF MINIMUM SAFETY CLASS DESIGN REQUIREMENTS Safety Class Design Requirements 1 2 3 Other Quality Group Classification" A B C D Quality Assurance Requirement B B B N/A Seismic Category I I I/N/A N/A d -
Electrical Classification lE lE lE N/A .
NOTES
- a. Equipment containing radioactive material, water, or steam ~
shall be constructed in accordance with the indicated quality group listed in Table 3.2-1, and the code class defined in Table 3.2-2. A quality group is not applicable (N/A) for equipment not'containing radioactive material, water, or steam. '"
- b. B - The equipment shall be constructed in accordance with the quality assurance requirements of 10CFR50 Appendix B as delin-eated in Chapter 17.
l N/A - The equipment shall be constructed in accordance with the quality assurance requirements consistent with accepted practice for steam power generating stations.
- c. I - The equipment of these safety classes shall be constructed in accordance with the seismic requirements for the safe shutdown earthquake as described in Section 3.7.
N/A - The seismic requirements for the safe shutdown earth-quake are not applicable to the equipment of this classification.
km 3.2-57
,e - - - - w -y ,y--, , -, =- g -
--e - e -
m =+
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 Table 3.2-3
SUMMARY
OF MINIMUM SAFETY CLASS DESIGN REQUIREMENTS (Continued)
NOTES (Continued)
- d. All electrical equipment with Safety Class 1, 2 or 3 shall have an electrical classification of lE. An electrical classification is not applicable (N/A) to equipment of Safety Class Other. _
O O
3.2-58
GESSAR II 22A7007 -
238 NUCLEAR ISLAND Rnv. 18 3.6.2.1.4.2 Piping in Containment Penetration Areas (Continued)
)
(b) The cumulative usage factor must be less than 0.1
}
)
b i N
- (c) If the calculated maximum stress range of Equation m
10 exceeds 2.4Sm, then the stress ranges calculated by both Equation 12 and 13 of NB-3653 do not exceed 3
2.4Sm. _
1 (d) The maximum stress as calculated by Equation 9 of l NB-3652 under the loadings resulting from a postu-lated piping failure beyond the required restraints
- does not exceed 2.25Sm. Higher stresses between the isolation valves and restraints were permitted provided a plastic hinge was not formed and oper-O ability of the valves with such stresses was assured.
(2) For ASME Code Section III Class 2 piping, the following stress and fatigue limits are not exceeded.
i (a) The maximum stress ranges calculated by the sum of Equations 9 and 10 of NC-3652 for normal and upset plant conditions (including an operating basis j ' earthquake) does not exceed 0.8 (1.2Sh + Sa).
- (b) The maximum stress as calculated by Equation 9 of NC-3652 under the loadings resulting from a postulated piping failure beyond the required restraints does not exceed 1.8Sh. Higher stresses _
between the isolation valves and restraints were i permitted provided a plastic hinge was not formed i
- and operability of the valves with such stresses 3.6-17
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 3.6.2.1.4.2 Piping in Containment Penetration Areas (Continued) was assured. When the piping beyond the isolation valve is constructed in accordance with ANSI B31.1, this exception may be applied provided the pipe is either of seamless construction with full radi-ography of all circumferential welds or all longi-tudinal and circumferential welds are fully radiographed.
(3) The piping runs are straight.
(4) Welded attachments for pipe supports or other purposes were avoided unless the detailed stress analyses or tests were performed to demonstrate compliance with the
}
y stress limits given in items (1) and (2). _
(5) The number of circumferential and longitudinal piping welds and branch connections are minimized.
l Where guard pipes are used, the enclosed portions of piping are of seamless construction and have no circumferential welds unless specific provisions for access is made to permit 100% inservice volumetric examination of all welds.
(6) The length of these portions of piping are reduced to the minimum length practical.
(7) The design of pipe anchors or restraints (e.g., connec-tions to containment penetrations and pipe whip restraints) do not require welding directly to the outer surface of the piping (e.g., flued integrally-forged pipe fittings may be used) except where such welds are 100% volumetrically examinable in service and a O
3.6-18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 i
determining and modifying resonant sections, or changing operating conditions. After corrective action is taken, additional testing shall be performed to determine if the vibrations have been suffi-o
- ciently reduced to satisfy the acceptance criteria. $.
i m
l 3.9.2.1.2.2 Preoperational Thermal Expansion and Dynamic Testing I Preoperational thermal expansion and dynamic testing is provided i in Subsection 14.2.12.1.75. -
i l
O 3.9-37
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-v # ---.--,--,.._,--.r._,-_--.--my ,__e,., ,,r__e -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 3.9.2.1.3 Thermal Expansion Testing of Main Steam and Recirculation Piping ]
A thermal expansion pre-operational and startup testing program performed through the use of potentiometer sensors has been established to verify that normal thermal movement occurs in the piping systems. The main purpose of this program is to ensure the following:
(1) the piping system during system heatup and cool-down is free to expand and move without unplanned obstruction or restraint in the xy, y, and z directions; (2) the piping system does shakedown after a few thermal expansion cycles; (3) the piping system is working in a manner consistent with the assumption of the NSSS stress analysis; (4) there is adequate agreement between calculated values of O
displacements and measured value of displacement; and (5) there is consistency and repeatability in thermal dis-placements during heatup and cooldown of the NSSS systems.
Limits of thermal expansion displacements are established prior to start of piping testing to which the actual measured displacements are compared to determine acceptability of the actual motion.
If the measured displacement does not vary from the acceptance limits values by more than the specified tolerance, the piping system is responding in a manner consistent with the predictions and is therefore acceptable. Two levels of limits of displace-ments would be established to check the systems as explained in Subsection 3.9.2.1.5.
O 3.9-38
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18
- 4.2.1 Design Bases (Continued) b Acceptance Criterion II. A. 3 (e) of SRP Section 4.2 describes analytical procedures for the determination of fuel assembly structural deformation. The GESSAR II fuel 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 exper-iencing structural failure.
4.2.2 Description and Design Drawings See Appendix A, Subsection A.4.2.2 of Reference 1.
Acceptance Criterion II.B of SRP Section 4.2 lists design parameters and drainage to be included in the fuel system description. The GESSAR II fuel system description, given in Reference 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.
4.2.2.1 Control Rods The control rods perform the duel 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 patterns 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 maximum total span and are separated uniformly "
[~') fthroughoutthecoreonanominal12-in. pitch. Each control rod fis surrounded by four fuel assemblies.
4.2-lb
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 4.2.2.1 Control Rods (Continued)
The main structural member of a control is made of Type-304 and/
or 316L stainless steel and consists of a top handle, a bottom cacting 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.
O 1
l 9 .
4.2-1c
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 rw 4.2.2.2
[Ji 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 conical element is separated from the upper conical element by four radial spacers 90 degrees apart and is at a 15-degree angle relative to the upper conical element, with the peripheral separation less than the central separation.
The hydraulic drag forces on a control rod are proportional to approximately the square of the rod velocity and are negligible at normal rod withdrawal or rod insertion speeds. However, during the scram stroke, the rod reaches high velocity, and the drag forces must be overcome by the drive mechanism.
To limit control rod drag during scram, the velocity limiter is f'
\
provided with a streamlined profile in the scram (upward) ~
direction.
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 t'he dropout direction, however, water is trapped by the conical elements and discharged
~
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.
O b
4.2-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 Acceptance Criterion II.C.3 (a) of SRP Section 4.2 lists phenomeno-logical models to be included in fuel system thermal calculations.
The GESSAR II fuel thermal model does not include the use of approved gadolinia fuel properties. However, as discussed with the NRC staff, the General Electric Company does not license material properties for design analyses but, rather, maintains these analyses up-to-date. To fulfill the quality control obliga-tions under 10CFR50, Appendix B, the latest property values are incorporated into design applications only after they are qualified in the design code. An improved fuel rod thermal-mechanical design code has recently been developed and qualified which includes the revised gadolinia fuel thermal conductivity relations. The results of the fuel centerline melting analysis using this improved fuel rod design code verify that gadolinia fuel melting is not expected to occur during normal steady-state operation or during the largest whole core anticipated operational transient.
The Reference 1 amendment incorporating the application of the above thermal-mechanical design code is currently under review by the Core Performance Branch. Since GESSAR II references the
" latest approved revision" of Reference 1, this issue will be resolved when the Reference 1 mmendment is approved.
Acceptance Criterion II.C.3.(d) of SRP Section 4.2 describes acceptance criteria for evaluation of fuel 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 lift-off) 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.
Because previous generic analytical methods presented in earlier versions of NEDE-21175 have been approved by the NRC staff (Letter j from O. 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.
l 4.2-3a l l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 O -
HANDLE O
\<
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h i VELOCITY LIMITER l
I f ,
l COUPLING SOCKET Figure 4.2-1. Control Rod Assembly 4.2-5
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cn w Figure 4.2-2. Control Rod Information Diagram e O O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 o i o g ROLLERS FUEL SUPPORT CASTING CORE SUPPORT PLATE
, I ORIFICE
& P CONTROL ROD
. l r--
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O Figure 4.2-3. Control Rod Velocity Limiter
, 4.2-7/4.2-8
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 1
CONTENTS (Continued)
Section Title Page I
5.2.5.8 Safety Interfaces 5.2-75 i 5.2.5.9 Testing and Calibration 5.2-75 5.2.5.10 Regulatory Guide 1.45: Compliance 5.2-75 .
l 5.2.6 References 5.2-77 I
e 5.2-v/5.2-vi
GESSAR II 22A7007 238 NUCLEAR ISLAND
~
Rsv. 18 SECTION 5.2 ILLUSTRATIONS Figure Title Page 5.2-1 Deleted 5.2-99 5.2-2 Deleted 5.2-100 5.2-3 MSIV Closure with Flux Scram and Installed Safety / Relief Valve Capacity 5.2-101 5.2-4 Deleted 5.2-102 5.2-5 Deleted 5.2-103 5.2-6 Deleted 5.2-104 _
5.2-7 Power-Actuated and Safety-Action Valve Lift Characteristics 5.2-105 I 'i 5.2-8 Reactor Vessel Pressure Following Transient
\m / Isolation Point 5.2-106 5~.2-9 Safety / Relief Valve Schematic Elevation 5.2-107 5.2-10 Safety / Relief Valve and Steamline Schematic 5.2-108 5.2-11 Nuclear Boiler System - P&ID Data 5.2-109 5.2-12 Schematic of Dual-Function Type Spring-Loaded Direct-Acting Safety / Relief Valve 5.2-111 5.2-13 Typical BWR Flow Diagram 5.2-112 5.2-14 Conductance Versus pH as a Function of Chloride Concentration of Aqueous Solution at 25*C 5.2-113 5.2-15 Leak Detection System - P&I Flow Diagram 5.2-115 5.2-16 Calculated Leak Rate Versus Crack Length as a Function of Applied Hoop Stress 5.2-119 O
V 5.2-ix/5.2-x
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 5.2.2.2.2.1 Operating Conditions O
(1) operating power = 3729 MWt (104.2% of nuclear boiler rated power);
(2) vessel dome pressure < 1045 psig; and 6
(3) steamflow = 16.71 x 10 lb/hr (105% of nuclear boiler rated steamflow).
l These conditions are the most severe because maximum stored energy exists at these conditions. At lower power conditions the transients would be less severe.
5.2.2.2.2.2 Transients See Appendix A, Subsection A.S.2.2.2.2.2 of Reference 3. -
I l
O 3.2-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 O
5.2.2.2.2.3 Safety / Relief Valve Transient Analysis Specification (1) Simulated valve groups:
power-actuated relief mode - 4 groups spring-action safety mode - 5 groups (2) opening pressure setpoint (maximum safety limit):
power-actuated relief mode - group 1 1125 psig group 2 1135 psig group 3 1145 psig group 4 1155 psig g
spring-action safety mode - group 1 1175 psig group 2 1185 psig group 3 1195 psig group 4 1205 psig group 5 1215 psig (3) reclosure pressure setpoint (% of opening setpoint) both modes:
maximum safety limit (used in analysis) 98 minimum operational limit 89 The opening and reclosure setpoints are assumed at a conserva-tively high level above the nominal setpoints. This is to account for initial setpoint errors and any instrument setpoint drift that might occur during operation. Typically the assumed 5.2-8
GESSAR II 007 j' 238 NUCLEAR ISLAND Rav. 18 i
5.2.3.2.2 BWR Chemistry of Reactor Coolant (Continued)
{ Several investigations have shown that in neutral solutions some f oxygen is required to cause stress / corrosion cracking of stainless steel. In the absence of oxygen no cracking occurs. One investi-f l
gation of the chloride oxygen relationship (Reference 2) ]
! showed that, . at high chloride concentration, little oxygen is required to cause stress / corrosion _ cracking of stainless steel, j .and, at high oxygen concentration, little chloride is required to
, cause cracking. These measurements were determined in a wetting
.and drying situation'using alkaline-phosphate-treated boiler water i
and, therefore, are of limited significance to BWR conditions.
l They are, however, a qualitative indication of trends.
a I: The water quality requirements are further supported by General f
I Electric stress corrosion test data summarized as follows.
- (1) Type 304 stainless steel specimens were exposed in a j flowing loop operating at 5370F. The water contained 1.5 ppm chloride and 1.2 ppm oxygen at pH 7. Test specimens were bent-beam strips stressed over their I yield strength. After 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br /> exposure, no -
cracking or failures occurred.
(2) Welded Type-304 stainless steel specimens were exposed in a refreshed autoclave operating at 5500F. The water contained 0.5 ppm chloride and 1.5 ppm oxygen at pH 7. Uniaxial tensile test specimens were stressed at 125% of their 550 F yield strength. No i
cracking or failures occurred at 15,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> exposure.
When conductivity is in its normal range, pH, chloride, and other impurities affecting conductivity will also be within their normal range. When conductivity becomes abnormal, chloride 5.2-25
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 5.2.3.2.2 BWR Chemistry of Reactor Coolant (Continued) measurements are made to determine whether or not they are also out of their normal operating values. This would not necessarily be the case. Conductivity could be high due to the presence of a neutral salt which would not have an effect on pH or chloride.
In such a case, high conductivity alone is not a cause for shutdown. In some types of water-cooled reactors, conductivities are high because of the purposeful use of additives. In BWRs, however, where no additives are used and where near neutral pH is maintained, conductivity provides a good and prompt measure of the quality of the reactor water. Significant changes in conductivity provide the operator with a warning to investigate and remedy the condition before reactor water limits are reached.
Methods available to the operator for correcting the off-standard condition include operation of the reactor water cleanup system, reducing the input of impurities, and placing the reactor in the cold shutdown condition. The major benefit of cold shutdown is to reduce the temperature-dependent corrosion rates and provide time for the cleanup system to re-establish the purity of the reactor coolant.
.The following is a summary and description of BWR water chemistry for various plant conditions.
(1) Normal Plant Operation The BWR system water chemistry is described by following the system cycle (Figure 5.2-13 and Table 5.2-6).
For normal operation starting with the condenser /
hotwell, condensate water is processed through a condensate treatment system. This process consists of filtration and demineralization resulting in effluent water quality represented in Table 5.2-6.
5.2-26
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 i
s i
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I
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5.2-103
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 18 O
Figure 5.2-6 Deleted O
O 5.2-104
)
GESSAR II 22A7007 238 Nuclear Island Rev. 14 19.3.9.24 O QUESTION / RESPONSE 9.24 (410.24)
QUESTION 9.24 Verify that the information provided in Section 9.1.3 of your FSAR is based on the new high density spent fuel pool storage capacity. Provide additional information regarding the spent fuel decay heat load for the maximum, normal and abnormal heat loads as discussed in Items 1.d and 1.h of the review procedures in Section 9.1.3 of the SRP. ( 9.1. 3 )
RESPONSE 9.24 Paragraph 9.1.3.1.2(4) for the Power Generation Design Basis states that the heat load is the sum of (1) the 37 percent core batch just removed at the last 18-month equilibrium fuel cycle, with 4-year exposure, and (2) the 37 percent core batch from the previous refueling outage. Therefore, the
()
heat load is a function of two 37 percent batches, which means that the entire heat capacity of the fuel storage pool does not enter the design. The fresh core supplies about 90 percent of the heat load and the aged core fraction supplies the other 10 percent of the design load. The density of the fuel racks would change the heat load calculation only if all of the potential batches stored within the pool were used toward the total design value. Even under these conditions, the design value would be only slightly affected.
Paragraph 9.1.3.2 describes that the above design core load for heat capacity is based upon maintaining 125 F in the pool. This is the system design maximum load and temperature l
combination. However, if conditions exist as described in Paragraph 9.1.3.3, wherein up to a full reactor core is
. O 19.3.9.24-1
v GESSAR II 22A7007 238 Nuclear Island Rev. 18 19.3.9.24 QUESTION / RESPONSE 9.24 (410.24) (Continued) -
placed into the pool, instead of the 37 percent batch, the pool may go to 150 F. .But adding the RHR cooling capacity will keep the temperature at a maximum of 125 F.
,. Item 1.h(ii) states that the normal maximum spent fuel heat load is set as one refueling load at equilibrium conditions
'after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decay, with one refueling load after 1 year of decay, and 140*F pool temperature. ~The GESSAR II design basis is more' conservative in that the refueling load is assumed at 112 hours0.0013 days <br />0.0311 hours <br />1.851852e-4 weeks <br />4.2616e-5 months <br /> of dacay and the maximum pool temperature is set at 125*F. The shorter fresh batch decay time adds to the total heat load sum 'of -the two batches.
Item 1.h(iii) states that the Spent Fuel Pool Cooling System
=
will have capacity for a full core at equilibrium and one refueling load, at 36 days, for a total of 1-1/3 core fraction. Item 1.h(iv) further adds 1/3 core for pool capacity over 1-1/3 batches. If RHR cooling capacity is included in the Spent Fuel Pool Cooling System, then the cooling capacity is more than adequate to meet these criteria. The two' fuel pool heat exchangers cover only the normal maximum, reflecting-ltem 1.h(ii), and the RHR covers any additional load while. the 5eactor is open.
The failure of one of the two active pumps or heat exchangers will reduce the capacity of the Spent Fuel Pool Cooling System.
The amount of cooling required by the pool is a function both -of the amount of core placed in the pool at the last refueling and of the time for decay of that core fraction. If the decay heat exceeds the removal capacity, the RHR System shall be employed to maintain fuel pool temperature. During this time of RHR use, the ,
l reactor will not be restarted following the refueling shutdown .
that placed the sp'ent~ fuel in the pool. When the decay heat is i
~
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l 1 9 rA
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- [k. GESSAR II 22A7007
-l 238 NUCLEAR ISLAND Rev. 18
, A s ,
s, 19.3.9.24 -s
~ QUESTION / RESPONSE 9 24 (410.24) (Continued) t less 'than 'the capacity of the Fuel Pool Cooling System, either ~
. by reduction due to time or by increased system cooling capacity, q the use of the RHR ~will be stopped and the reactor startup will p be allowed.
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