ML20072H530
| ML20072H530 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 06/15/1983 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20072H519 | List: |
| References | |
| NUDOCS 8306290515 | |
| Download: ML20072H530 (174) | |
Text
. _
UNITED STATES 0F AMERICA g NUCLEAR REGULATORY COMMISS10N V
In the matter of )
General Electric Company) Docket No. STN 50-447 Standard Plant )
AMENDMENT NO. 17 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. 17 to the 238 Nuclear Island General Electric Standard Safety Analysis Report (GESSAR II).
Amendment No. 17 further amends GESSAR II by:
- 1. Furnishing a new Appendix 1G which provides the response to the CP/ML Rule 10CFR50.34(f).
- 2. Furnishing a revised Section 1.8 which now responds to the SRP Rule 10CFR50.34(g).
- 3. Clarifying portions of the text where obvious discrepancies exist.
() Respectfully submitted, General Electric Company By: s/ Joseph F. Quirk Joseph F. Quirk, Manager BWR Systems Licensing STATE OF CALIFORNIA
) ss:
l COUNTY OF SANTA CLARA )
On this 15th day of June in the year 1983, before me, Karen S. Vogelhuber, Notary Public, personally appeared Joseph F. Quirk, personally proved l, 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 My Commission Expires December 21, 1984 8306290515 830615 175 Curtner Avenue f ^ " U'UUU $ San Jose, CA 95125 JFQ: cal:hmm/K061314-2 l
6/15/83 l __. _ _ - _ _ _- . __ _ . _ _ - _- _ _ _ _ .._.._. . _ _ _ __.
GESSAR II 22A7007 238 NUCLEAR ISLAND ReV. 17 a
i INSTRUCTIONS FOR FILING AMENDMENT N0.17 A tab is also included for Appendix 1G.
Remove and insert the pages listed below. Dashes (----) in the remove or insert column indicate no action required.
Remove Insert
- Chapter 1 1.8-1,1.8.0-1/1.8.0-2, 1.8-i,1.8-xi/1.8-xti, 1.8.122-1/1.8.122-2, 1.8.0-1,1.8.0-2,1.8,0-2a 1.9-4.4-1/1.9-4.4-2, through 1.8.0-2e, 1.8.122-1/
and 1.9-4.15-2 1.8.122-2,1.9-4.1-3a, 1.9-4.1-13,1.9-4.1-19, 1.9-4.1-20,1.9-4.4-1/
1.9-4.4-2,1.9-4.11-2a, and 1.9-4.15-2 Appendix 1A Appendix 1A title page, Appendix 1A title page, p 1A.0-1,1AC-7/1AC-8, 1A.0-1,1AC-7/1AC-8, IAC-9, and 1AC-11 1AC-9, and 1AC-11 Appendix ID 10.2-3 10.2-3 and 1D.2-3a Appendix IG j ----
Insert new Appendix 1G Chapter 3 3.7-34 3.7-34 and 3.7-34a Chapter 4 4.2-1,4.2-4,4.4-10, 4.2-1,4.2-la,4.2-lb,4.2-Ic, and 4.6-29 4.2-3a ,4.2 -4,4.4-10, and 4.6-29 Chapter 5 5.2-37 and 5.2-43 5.2-37,5.2-37a,5.2-43, and 5.2-43a Chapter 6 b
V 6.4-2,6.8-5 and 6.8-6 6.4-2,6.7-3a,6.8-5, and 6.8-6 Amendment 17 June 15,1983
r GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 Remove Insert O
Chapter 8
8.3-14a Chapter 9 9.5-1 9.5-1,9.5-la,9.5-lb, and 9.5-Ic Chapter 11 11.5-23 11.5-23 and 11.5-23a Chapter 12 12.1-4a,12.2-1, and 12.1-4a,12.2-1,12.2-10, 12.2-10 and 12.2-10a Cha7ter 15 15.3-13,15.4-11,15.4-14, 15.3-13,15.3-13a,15.4-11, and 15.4-16 15.4-11a,15.4-14,15.4-14a, 15.4-16 and 15.6-15a Chapter 19 19.3.3.52-4,19.3.3.74-1/ 19.3.3.52-4,19.3.3.74-1/
19.3.3.74-2,19.3.6.28-2, 19.3.3.74-2,19.3.6.28-2, 19.3.6.28-3, and 19. 3. 6.28-3, and 19.3.15.1-1 19.3.15.1-1 0
Amendment 17 June 15, 1983
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 SECTION 1.8
^
{-s 1 CONTENTS Section Title Page i
l 1.8 CONFORMANCE WITH STANDARD REVIEW PLAN 1.8.0-1 3 i -
1 1.8.0 Purpose 1.8.0-1 1.8.0.1 Difference From Standard Review Plan 1.8.0-1 1.8.0.2 Consistency with NRC Regulatory Guides 1.8.0-2 _
l.8.1 Regulatory Guide 1.1, Revision 0,
' Dated November, 1970 1.8.1-1 1.8.2 Regulatory Guide 1.2, Revision 0, Dated November, 1970 1.8.2-1 l.8.3 Regulatory Guide 1.3, Revision 2, Dated June, 1974 1.8.3-1 l.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 1.8.5-1 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, l 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 r
1.8.13 Regulatory Guide 1.13, Revision 1, y
Dated December, 1975 1.8.13-1 1.8.14 Regulatory Guide 1.14, Revision 1, Dated August, 1975 1.8.14-1 1.8-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 1.8 CONTENTS (Continued)
Section Title Page 1.8.15 Regulatory Guide 1.15, Revision 1, (Withdrawn July, 1981) 1.8.15-1 1.8.16 Regulatory Guide 1.16, Revision 4, Dated August, 1975 1.8.16-1 1.8.17 Regulatory Guide 1.17, Revision 1, Dated June, 1973 1.8.17-1 1.8.18 Regulatory Guide 1.18, Revision 1, (Withdrawn July, 1981) 1.8.18-1 1.8.19 Regulatory Guide 1.19, Revision 1, (Withdrawn July, 1981) 1.8.19-1 1.8.20 Regulatory Guide 1.20, Revision 2, Dated May, 1976 1.8.20-1 1.8.21 Regulatory Guide 1.21, Revision 1, Dated June, 1974 1.8.21-1 1.8.22 Regulatory Guide 1.22, Revision 0, Dated February, 1972 1.8.22-1 1.8.23 Regulatory Guide 1.23, Revision 0, Dated February, 1972 1.8.23-1 1.8.24 Regulatory Guide 1.24, Revision 0, Dated March, 1972 1.8.24-1 1.8.25 Regulatory Guide 1.25, Revision 0, Dated March, 1972 1.8.25-1 1.8.26 Regulatory Guide 1.26, Revision 3, Dated February, 1976 1.8.26-1 1.8.27 Regulatory Guide 1.27, Revision 2, Dated January, 1976 1.8.27-1 1.8.28 Regulatory Guide 1.28, Revision 2, Dated February, 1979 1.8.28-1 1.8.29 Regulatory Guide 1.29, Revision 3, Dated September, 1978 1.8.29-1 O
1.8-ii
@ f f
GESSAR II 22A7007 1
238 NUCLEAR ISLAND Rev. 17 1
SECTION 1.8 TABLES f
Table Title Page !
I j 1.8.0-0 Summary of Differences from SRP 1.8.0-2 [
4
- 1.8.0-1 Regulatory Guide Cross Reference 1.8.0-3 l l
1 4
l r
t l
A i
i t
i i !
! l 1
lO i
i l
i r
t i
I i
l 1.8-xi/1.8-xii t I
-e p-s-mw w -=om m m . , , . . .-. ..an - . . . - - - - - - - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 h)
\.
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 10CFR50.34 (g) . Since the NRC regulatory guides are an integral part of NUREG-0800, this section also shows the consist-ency of the design with the regulatory guides.
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 responsi-i bility 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.
4
[A- Finally, GE has chosen for commercial reasons to delay the sub-mittal of certain 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 demon-strate that the GESSAR II design satisfies all of the NUREG-0800 requirements. However, GE has reviewed the GESSAR II design against the relevant 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 satisfied.
The Applicant will provide a summary of deviations from NUREG-0800 for those plant design features covered by 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 satisfied. ~
O 1.8.0-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 O
1.8.0.2 Consistency with NRC Regulatory Guides ,
The NRC (AEC) began in 1970 to issue regulatory guides (safety guides) which 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 1.1-1.7 5) , 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.
O 1.8.0-2 _
O O O L
Table 1.8.0-0
SUMMARY
OF DIFFERENCES FROM SRP GESSAR II j
Subsection Specific SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed 3.7.1 II.l.b - Design time history For higher damping values, 19.3.3.48 i (Rev. 1) and damping values criteria. the response spectra from synthetic time history are not in agreement with the enveloping values of the criteria 3.7.3 II.2.b - Determination of For equipment and components 3.7.3.2.2 other than piping, 10 rather *
(Rev. 1) number of OBE cycles.
than 50 peak OBE stress cycles g
. are used. om 4 m em
- Mm t
o 4.2 II . A. l . (b) - Sets limit on the NEDE-240ll sets a more con- 4.2.1 g>*
b (Rev. 2) number of strain fatigue servative limit than that in
]j
- cycles. the SRP. ((
i 4.2 II. A. l . (c) - Fretting wear of Wear limits are not stated. 4.2.1 $
D (Rev. 2) structural members should be j stated.
4.2 II . A. l . (g) - States that " Worst Design basis allows up to 4.2.1 (Rev. 2) case hydraulic loads" may not 0.52 inch " lift-off".
exceed the hold down capability of the fuel assembly.
4.2 II. A. 2. (e) - Prohibits any Design basis allows fuel 4.2.1 j (Rev. 2) fuel melting melting that is not 4 " excessive".
mw 4.2 II. A. 2. (g) - Specifies uniform Elastic strain not included 4.2.1 $y
. (Rev. 2) strain (elastic & plastic) in the 1% limit. -
g limit of 1%. ro aw
Table 1.8.0-0
SUMMARY
OF DIFFERENCES FROM SRP (Continued)
GESSAR II Subsection Specific SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed 4.2 II. A. 2. (i) - Limits applied Topical report is under 4.2.1 (Rev. 2) stress to < 90% of the review.
irr diated yield stress.
4.2 II . A. 3. (e) - Analytical pro- Topical report is under 4.2.1 (Rev. 2) cedures are prescribed. review, w
4.2 II.B - Lists parameters to be Fuel description does not 4.2.1 m
,- (Rev. 2) included in fuel description. include all parameters listed z w
in SRP. @@
Mm
? 4.2 II .C. 3. (a) - Lists models to Gadolinia fuel properties not 4.2.2 $$
WN
@ (Rev. 2) be included in thermal appropriate in model.
calculations. yH 4.2 II.C.3.(d) - Describes accept- Topical report is under 4.2.2 @
(Rev. 2) ance criteria for design review. O evaluation.
5.2.3 II. 3.b. (1) (a) - Welding proce- Minimum preheat and maximum 5.2.3.3.2.1 (Rev. 2) dure qualification. interpass temperature not specified.
5.2.3 II . 3.b. (3) - Regulatory Guide Alternate position employed. 5.2.3.4.2.3 (Rev. 2) 1.71, Welder Qualification for Areas of Limited Accessibility.
mw 6.2.1.1.C II.9 - Compliance with GESSAR II analysis takes 19.3.6.10 ow (Rev. 5) NUREG-0783. credit for weir wall annulus (Comparison to Id water. Section 5.7.1 ~8 of NUREG- 44 0783)
O O O
O O O Table 1.8.0-0
SUMMARY
OF DIFFERENCES FROM SRP (Continued)
GESSAR II Subsection Specific SRP Summary Description Where SRP Section Acceptance Criteria of Difference ,
Discussed 6.2.1.2 II.B.1 - Humidity for shield 1% relative humidity used 19.3.6.14 j (Rev. 2) wall annulus analysis. in analysis.
t 6.3 III.19 - Operator action GESSAR II requires operator 19.3.6.56
- (Rev. 1) following LOCA. action within 10 minutes for l some events, n
w 6.7 II.1 - MSIV leakage control Exception taken to Position 1.8.96 m w
(Rev. 2) meeting Regulatory Guide 1.96. C.9 of Regulatory Guide 1.96. z CO >
nm m
o 7.1 (Rev. 2)
II - Regulatory Guide 1.75 Alternates to portions of 7.1.2.10.18 yy i
(Table 7-1). Regulatory Guide 1.75 are utilized.
>g W
Q HH 7.2 II.1 and II.2 - IEEE 279-1971 Some RPS inputs come from Table $
(Rev. 2) and GDC 2. devices mounted on non- 19. 3. 7.14-1 (j ) @
seismically qualified equip- O ment and/or are located in non-seismically qualified enclosures.
7.3 II - TMI Item II.K.3.21: Core Spray and LPCI systems lA.63 (Rev. 2) Restart of Core Spray and Low- do not automatically restart Pressure Coolant Injection after being on low water Systems (Table 7-2) level if the initiation sig-
- nal is still present.
7.3 II - Paragraph 4.17 of IEEE HPCS, LPCS, LPCI, ADS, and 19.3.7.42 xw i (Rev. 2) 279 the containment spray mode Qy of RHR share common inter- w locks between the automatic ~$
WW and manual initiation modes.
Table 1.8.0-0
SUMMARY
OF DIFFERENCES FROM SRP (Continued)
GESSAR II Subsection Specific SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed 7.5 II - Regulatory Guide 1.97 Exception taken to some of Appendix 1D.
(Rev. 2) (Table 7-1) the requirements.
8.3.2 BTP PSB-1 Section 1. (c) . (3) - GESSAR II design based on (Rev. 2) Second level of undervoltage maximum fluctuation of 5% 19.3.8.5 protection for Class 1E on grid voltage. w equipment. g
." 9.5.1 II.2.a - Implementation of Lack of 3-hr-fire-rated dampers 9.5.1.1 $a
, (Rev. 3) fire protection program in in ventilation system. py o accordance with BTP CMEM 9.5-1. mm M
8 NN a 12.1 II.2 - Instructions to No specific instructions 12.1.2.2.1 gg (Rev. 2) designers and engineers provided. mH regarding ALARA.
12.2 II.6 - Contained source Size and shape of vessels 12.2.1.1 (Rev. 2) descriptions. with contained sources not provided.
12.2 II.6 - Buildup of activated Buildup of activated cor- 12.2.1.2.7.2 (Rev. 2) contaimment sources. rosion products provided only for recirculation piping.
15.3.3- II.8 - Use of non-safety Credit is taken for non- 15.3.3.2.2 15.3.4 grade equipment. safety grade equipment and (Rev. 2) failure of non-safety yM grade equipment is not <>
assumed. $
so QM O O O
- - - - - += - --- - - + - wee -w r-vly pr-
[)\ (D (J c LJ Table 1.8.0-0
SUMMARY
OF DIFFERENCES FROM SRP (Continued)
GESSAR II Subsection Specific SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed 15.3.3 - II.10 - Coincident loss of Not analyzed with coincident 15.3.3.2.2 j 15.3.4 offsite power. loss of offsite power.
(Rev. 2) 15.4.4 - II . 2. (b) - Fuel cladding MCPR not calculated. 15.4.4.3.2, 15.4.5.3.2.1. w 15.4.5 integrity.
(Rev. 2) & $
15.4.5.3.2.2 z
" co Radiological analysis for Part lb to h I$
h 15.6.5 II. (2) - Distribution of LOCA assumes 25% of iodine 19.3.5.1 Qy o Appendix B iodine inventory.
is in suppression pool. WN
, b (Rev. 1) HH O
Cl) H V
, Z O
(D M
. 4 O
i HO WM l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
N 1.8.122 Regulatory Guide 1.122, Revision 1, Dated February 1978 4
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 exception is provided in GESSAR II Subsection 3.7.2.9 which _
illustrates the conservative assumptions that have been included g in the calculation of the floor response spectra.
^
n- l 4
O 1.8.122-1/1.8.122-2
)
i'
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Table 1.9-1 CHAPTER 1 GESSAR II/FSAR INTERFACES (Continued)
ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 1.15 Power Provide design characteristics 1.3-13 Table 1.3-2 1 Conversion for the circulating water, and System condensate and feedwater systems 1.3-14 for comparison to other nuclear power plants per R.G. 1.70 Subsection 1.3.1.
l 1.16 Engineered Provide the design characteristics 1.3-16 Table 1.3-3 1 Safety Features for the essential service water h)
, system for comparison to other $l l nuclear power plants per R.G. 1.70 7' Suhtection 1.3.1. ggj c o un g 1.17 Radioactive Provide the gaseous radwaste 1.3-23 Figure 1.3-5 4
. Waste release point height above ground g lui[
F' Management for comparison to other nuclear d3 power plants per R.G. 1.70 g[j Subsection 1.3.1. un 1.18 Applicant Identify the Applicant per R.G. 1.4-1 1.4-1 3 Identification 1.70 Subsection 1.4. 6 1.19 Turbine Identify the turbine generator 1.4-2 1.4.4 3 Generator vendor and the appropriate division of responsibility per R.G. 1.70 Subsection 1.4.
1.20 Consultants Identify the prime agents or 1.4-2 1.4.5 3 contractors for the design, $gh3 construction, and operation of the nuclear power plant, as well <>
as division of responsibility, c!
per R.G. 1.70 Subsection 1.4. Ed o ik w
l
)
Table 1.9-1 CHAPTER 1 i
GESSAR II/PSAR INTERFACES (Continued)
ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATECORY 1.20.1 Differences Identify differences in plant 1.8.0-1 1.8.0-1 3 from standard design features covered by Review Plan Tables 1.9-1 through 1. 9-19 with evaluations that describe the basis to conclude that the underlying requirements are met M
W CD
. Z e CO i OM a MM
, M U2 i g >>
a WW j HH U3 H d
O 1
Ww (D w
<>-a O
HO 4Q O O O
O O O Table 1.9-1 CHAPTER 1 GESSAR II/FSAR INTERFACES (Continued)
ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 1.21 R.G. 1.8 Describe-personnel selection 1.8.8-1 1.8.8 3 and training program.
1.22 R.G. 1.16 l'rovide program for the reporting 1.8.16-1 1.8.16 '3 of operating information.
l.23 R.G. 1.17 Describe program for protection 1.8.17-1 1.8.17 3 against industrial sabotage.
1.24 R.G. 1.23 Describe onsite meteorological 1.8.23-1 1.8.23 3 w program.- co F 1.25 R.G. 1.27 Describe ultimate heat sink 1.8.27-1 1.8.27 3 $h!
n v1 i
e j, 1.26 R.G. 1.32 Describe use of IEEE 308-1974 1.8.32 1.8.32 3 E$
. during operating and training. >W s W **
I 4, 1.27 R.G. 1.33 Define Quality Assurance Program 1.8.33-1 1.8.33 3 4 84 U) requirements during operation.
1.28 R.G. 1.39 Demonstrate how housekeeping 1.8.39-1 1.8.39 3 requirements will be met.
1.29 R.G. 1.65 Describe in-service inspection 4 plant for RPV closure studs.
1.30 R.G. 1.68.1 Define preoperational and startup 1.8.68. 1.8.68.1 3 testing of feedwater and 1-1 4
condensate systems.
WN 1.31 R.G. 1.72 Meet requirements for spray pond piping made from fiber glass 1.8.72-1 1.8.72 3 yyq reinforced thermosetting resin. g i bQ l
I
O O O Table 1.9-1 CHAPTER 1 GESSAR II/FSAR INTERFACES (Continued)
ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 1.92 Administrative Establish administrative controls IC.2.3-5 1C.2.3 3 Controls minimize fire hazards per 10CFR50, App. R Section III.K. ,
- 4. 1.93 Initial Test Respond to the NRC guidance 1C.2.6-1 1C.2.6 3 Programs concerning, typically, those questions raised in the review of the Summer and San Onofre 2/3 g applications regarding the initial w test program. m 1.94 Inservice Inspections Provide an inservice inspection plan.
1C.2.8-1 1C.2.8 3 bb trJ >
H 1.95 Q-List Supplement and clarify Q-list in IC.2.12-1 1C.2.12 2 g 28 y
accordance with NRC guidelines wH s
th 1.96 RG 1.97 Assure that GESSAR II design 1D.2-3a 1D.2.3 5 variable meets designated portions of Assessments RG 1.97 and any deviations to and the remainder of the guide are Conformance justified 1.97 Boron Analysis 3 Provide boron analysis procedures. 1D.2-5 1D.2.3.3 ,
as N 1.98 Radiation Monitoring Determine the locations and ranges for area radiation monitors in the 1D.2-10 1D.2.3.8 3 $$4 (Primary containment. ga Containment) aa i I l
Table 1.9-1 CHAPTER 1 GESSAR II/FSAR INTERFACES (Continued)
ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 1.99 Radiation Determine the locations and ranges 1D.2-13 1D.2.3.14 3 Monitoring for area radiation monitors in the (Secondary secondary containment.
Containment) 1.100 Noble Gas Provide systems for the monitoring 1D.2-13 1D.2.3.15 1 Effects of primary and secondary containment noble gas effluents.
N) u 1.101 Coolant Gamma Provide procedures to determine ao H Sample the gross gamma activity in the 1D.2-15 1D.2.3.20 3 g coolant. h g) m
, o vi S$ 1.102 Condensate Provide for monitoring of the 1D.2-19 1D.2.3.29 1 E "'
g Storage Tank Level condensate storage tank level. gb 4
4, ss g
1.103 Particulate Provide for monitoring of 1D.2-22 1D.2.3.35 1 ui Halogen Release particulate / halogen releases.
1.104 Environs Provide for monitoring of the 1D.2-22 1D.2.3.36 1 Monitoring the environs radioactivity.
1.105 Meteorology Provide for monitoring of 1D.2-22 1D.2.3.37 1 meteorology.
1.106 Variables In addition to qualification, for 1D.4-5 Table 1D-2 1 variables outside of the Nuclear through Island scope, determine 1D.4-7 x N) redundancy, range, power supply, y gj and type of control room display, .y and also provide quality assurance. O Address R.G. 1.97 guidance. $
O O O
Table 1.9-2 ,
I CHAPTER 1 '
i GESSAR II/FSAR INTERFACES F I
ITEM INTERFACE
, NO. SUBJECT DESCRIPTION PAGE SUBSECTION CATEGORY
, (
i l 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 Eydro-gen Control System. For the selected system, provide design
- description, function layout, analyses and test data to verify E' compliance with the requirements ao
- l of (f) (2) (ix) of 10CFR50.34. 2 i g c: o 1.128 Long-Tenn Establish a training program which 1G.13-1 1G.13 3 OE! ,
i i Training addresses the concerns related b un ,
! A
, Upgrade to Item I.A.4.2 of NUREG 0718. $$
l.129 Long-Term Establish a program for integrating 1G.14-1 1G.14 3 u! U j u> Program of and expanding current efforts to C
. Upgrading of improve plant procedures. E Procedures U
~
l.130 Hydrogen Provide a Hydrogen Control System 1G.21-2 1G.21 3 i control capable of handling equivalent of
- System a 100% active fuel-clad metal i water reaction.
l 1.131 Purging Provide performance information 1G.27-1 1G.27 3 of purge valves :
MM l
mw I
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HO
. %3 m3 t
i j
Table 1.9-2 CHAPTER 1 GESSAR II/PSAR 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 k) 1.133 In-Plant Provide monitoring of in-plant 1G.39-1 1G.39 3 $
Radiation radiation and airborne radioac-7 Monitoring tivity for routine and accident ho j) conditions. p$
u 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 HH gg Construction ensure applicable important &
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 1G.45 3 $U Integrity capability of 45 psig for Service ,,
Level C. c3 wo 44 O O O
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Table 1.9-4 CilAPTER 4 GESSAR II/FSAR INTERFACES ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION 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 Provide reference core loading 4. 3-9a Figure 4.3-1 3 Pattern pattern figure.
4.3 Loose Parts Describe monitoring equipment and 4.4-10 4.4.6.1 1 procedures to be used to detect g
excessive vibration and the occurrence of loose parts per
[j a3 ua R.G. 1.70 Subsection 4.4.6 and 1
R.G. 1.133.
{Q O U3 A 4.4 Safety Provide information for safety 4.4-19 Table 4.4-6 3 t' U)
[ Injection Lines injection lines.
's Q l}
50 F" 4.5 Post- Provide a post-irradiation 4.2-4 4.2.4 3 **
Irradiation surveillance program y3 i Surveillance (( ""
em A Program b
g p
4.6 The rma l Provide stability analysis using 4.4-10 4.4.4 3 M liydraulic decay ratio agreed to by the Stability NRC staff.
4.7 Reactor Assure that the Process Monitoring 4.4-10 4.4.5 3 Coolant System is capable of 3% pressure Pressure drop in the coolant flow with a Drop flow test every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- %J O
>-* O 4 %J s
I s
. . . _ _ _ _ __ ___ _ _ . - _ _ _ . . . . . _ _ _ _ _ . . _ . . _ ~ _ . . _ . _ _ _ _
1
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l
! Table 1.9-11 t i
CilAPTER 11 GESSAR II/FSAR INTERFACES (Cont i nued )
! RELATED INTERFACE
! ITEM CATEGORY NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION j _ _ __ _ _ _ _ _ . _
f Provide in the plant technical 11.5-23 11.5.3.4 3 11.12 Isolation '
- Valve specification setpoints for
! Setpoints actuation of valves, dampers or I diversion valves.
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W CO 3
- OM D t* in i M (n i
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Table 1.9-15 CHAPTER 15 GESSAR II/FSAR INTERFACES
. ITEM RELATED INTERFACE l NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY i
i' 15.1 S/RV Provide the manufacturer for the 15.0-18 Table 15.0-1 2 Manufacturer safety / relief valves.
l 15.2- Transient Provide core-wide transient 15.0-24 Table 15.0-4 3 Analysis analysis results for listed Results transients for reference core j loading patterns.
1 l 15.3 Initial Core Provide initial core MCPR values 15.0-24 Table 15.0-5 3 MCPR Value for events listed for reference b) core loading patterns.
f'e 15.4 Loss of Feed-water Heating Reanalyze,the loss of feedwater 15.1-4 15.1.1.3.2 3
%Q event, using the plant specific n un I
core configuration per R.G. 1.70, & un f Chapter 15.
H %%
ns f 15.5 Feedwater Reanalyze the feedwater controller 15.1-7 15.1.2.3 3 "
H controller failure-maximum demand event, g un Failure using the plant specific core configuration per R.G. 1.70, Chapter 15.
15.6 Pressure Reanalyze the pressure regulation 15.2-5 15.2.1.3.2 3 Regulation downscale failure event, using the Downscale plant specific core configuration ;
Failure per R.G. 1.70, Chapter 15.
l 1
15.7 Generator Reanalyze the generator rejection 15.2-11 15.2.2.3.2.2 3 r Load with failure of bypass event, "
Rejection 3, 9 using the plant specific core os ha configuration per R.G. 1.70, Chapter 15. f3}o :
Ho A sJ 1
1 1
l l
1
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 Bundle accident using the plant specific Accident core configuration per R.G. 1.70, Chapter 15 15.9.1 Fuel Loading Include in the Plant Operating 15.4-16 15.4.7.1.1 3
[3 Errors Procedures / Technical Specifica- os g, tions, provisions for potential
. fuel loading errors. f5h) e o un j, 15.10 Dispersion Data Provide site boundary and low 15.4-37 Table 15.4-12 4 t' U)
. Control Rod population zone distances, using h)hy Fa Drop both design and realistic gs assumptions in the control rod H
[
ba drop accident. (l"*
15.11 Dispersion Data Provide site boundary and low 15.4-37 Table 15.4-12 4 Control Rod population zone distances, using Drop both design and realistic 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 assump-tions in the steamline break accident. gg trj to 15.13 Dispersion Data Provide site boundary and low 15.6-37 Table 15.6-7 4 < 3" LOCA population zone distances, using 2f both design and realistic assump- r4 c) tions in the loss-of-coolant -J 'J accident.
I I O . O O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 O
APPENDIX 1A l
~
RESPONSE TO TMI RELATED MATTERS OF NUREG-0737 1
1 1O
GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 17 APPENDIX 1A RESPONSE TO TMI RELATED MATTERS OF NUREG-0737 ,
~
(
1A.0 INTRODUCTION The investigations and studies associated with the j TMI-2 accident produced several documents specifying results and recommendations, which prompted the issuance by the NRC of various bulletins, letters, and NUREG's providing guidance and requiring specific actions by i 1
the nuclear power industry. In May 1980, the issuance of NUREG-0660 (Reference 1) provided a comprehensive and integrated plan and listing of requirements to correct or improve the regulation and operation of nuclear facilities based on the experience from the accident at TMI-2 and the studies and investigations of m the accident. NUREG-0737 (Reference 2), issued in November 1980, listed items from NUREG-0660 approved by the NRC for implementation, and included additional information concerning schedules, applicability, method of implementation review, submittal dates, and clarifi-cation of technical positions.
This Appendix 1A reports GE's responses for the 238 Nuclear Island to the NRC positions taken regarding the "TMI Action Plan Requirements for Applicants for an Operating License" as referenced in NUREG-0737, Enclosure 2.
These responses have developed as the NRC positions have evolved and been clarified by the issuance of subsequent documentation by the NRC.
In general, the responses demonstrate the methods of compliance to ensure that the NRC requirements are t
l 1A.0-1
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.0 INTRODUCTION (Cont'd)
O satisfactorily fulfilled for the 236 Nuclear Island.
For each item a summary 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 documentation that may be referenced for complete clarification of our position. Where a particular requirement is not applicable to the 238 Nuclear Island, a statement to that effect is provided in the response.
GE has been and continues to be a participant in the BWR Owners' Group program. Several responses to generic issues are based on the results and conclusions of test programs and studies sponsored by this organization to specifically address these respective generic items.
Generally the response references correspondence associated with the issue under consideration to the NRC from the BWR Dwners' Group, which is accompanied by a thorough explanation of the relevance or bearing of that analysis to the 238 Nuclear Island.
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 1A is complete in that nearly all of the "TMI Action Plan Requirements for Applicants for an Operating License" approved for implementation by the NRC as listed in NUREG-0737, Enclosure 2, have been addressed where they apply to the 238 Nuclear Island.
1A.0-2 O
16A37
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
() 1AC.3.2 RCIC Steam Supply / Lines (Cont'd)
This additional stage to the control switch prevents the isolation valves from opening simultaneously as a con-sequence of resetting the trip logic function. By adding another stage to the control switch with the same type of configuration (i.e., both stages currently close their contacts in the "OPEN" position switch mode), separation of the trip logic reset function from the deliberate operator action of opening the closed isolation valves is assured after an isolation has been cleared.
Since the control switch must be in the "OPEN" position mode for normal RCIC operation, the added stage of closed contacts provided by the proposed design change effectively blocks the RESET function until the isolation valves are closed.
To open the isolation valve (s), the operator must get the
() key for the control switch, reposition the switch to the "CLOSE" mode and depress the " RESET" pushbutton; this action resets the isolation valve trip logic so that the trip logic function is armed and ready to respond to another system isolation, if required. But only when the operator returns
( the switch again to the "OPEN" position mode, can the isolation valve be opened under operator manual control.
The design changes described above for the RHR Sample Line, l Reactor Water Sample Lines and RCIC Steam Supply Lines provide the deliberate and separate operator actions required by NUREG-0737, Item II.E.4.2 position Item 4.
l O'
1AC-7/1AC-8
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
(< ,)
v 1AC.4 DISCUSSION OF CONTAINMENT PURGE DESIGN This section provides justification of the Mark III Containment normal operation purge system design relative to the criteria of position (6). Position (6) states:
" Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the staff interim position of October 23, 1979, must be sealed closed as defined in SRP 6.2.4, Item II .3. f during operational conditions 1, 2, 3, and 4. Furthermore, these valves must be verified to be closed c' ',ast every thirty-one days."
Branch Technical Position CSB 6-4 provides guidance on the use of containment purge systems during normal operation.
The key points of that position are:
g-)
'd (1) The on-line purge system isolation design should be highly reliable (several specific criteria are mentioned).
Response: The design of the containment purge system (containment pressure control) is discussed in Section 9.4.5.6. This design features single exhaust and supply lines which receive redundant isolation signals j
on containment (purge exhaust line) high radiation at a level determined by ALARA considerations of 10CFR50.
i The normal operation flow is about 5000 CFM which I
penetrates containment through a 9-in. bypass valve.
7 l
Diversity in closure means (air pressure and spring) further accentuates the reliability of this design.
v 1AC-9 l
i GESSAR II 22A7007 l 238 NUCLEAR ISLAND REV. 4 1AC.4 DISCUSSION OF CONTAINMENT PURGE DESIGN (Cont'd)
O (2) "The purge system should not be relied upon for temperature or humidity control."
Response: The Mark III containment complies with this requirement. See Section 9.4.5.15 for a discussion of containment temperature and humidity control.
(3) " Provisions should be made to minimize the need for purging of the containment by providing containment atmosphere cleanup systems within the containment".
Response: Provisions for such systems are unnecessary in the Mark III containment because of the presence of negligible amounts of airborne radioactivity during normal operation. For abnormal events, the release of primary system fluid (and associated fission products) is to the suppression pool rather than the containment air space. Halogens and particulates are expected to be retained in the water by the scrubbing effect of the suppression pool (see Section 15D.2.2). The suppression pool cleanup system (Section 9.5.9) provides an effective means of removal of those fission products. The containment purge lines isolate due to low RPV water level or, as a backup, high radiation in the exhaust line.
(4) " Provisions should be made for testing the availability of the isolation function and the leakage rate of the isolation valves during reactor operation".
Response: Provisions are made as discussed in .
Subsection 6.2.6.
O 1AC-10 16I15
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
,/ 1AC.4 DISCUSSION OF CONTAINMENT PURGE DESIGN (Cont'd)
(5) Analysis
- a. "... radiologic consequences of a loss of coolant accident... the radiologic consequences should be within 10CFR Part 100 guideline values".
Response: The containment normal operation purge system is designed to isolate on low reactor water level, high drywell pressure, or high exhaust radiation level. The high exhaust radiation isolation level is set by ALARA considerations of 10CFR50. This setpoint is set in accordance with the generic technical speci-fication at a level to ensure offsite radiologic consequences are well within 10CFR100 values. The loss of coolant radiological analyses (Subsection 15.6.5)
(s,) assumes closure of the containment isolation valves instantaneously (due to low reactor water level or high drywell pressure). This isolation occurs prior to the release of significant activity to the drywell or containment atmospheres.
The exhaust line isolation valve closure times are within six seconds. Since the actuation of this closure occurs prior to the release of significant activity and any potential long-term leakage is con-trolled by a positive leakage control system (Sub-section 6. 5. 3. 3) , any differences between the calcu-l lated offsite exposure and that calculated in Sub-section 15.6.5 will be small. Consequently, no addi-tional analysis is necessary.
- b. .... acceptability of the provisions made to
! O)
( ,
protect structures and safety-related equipment...
located beyond the purge system isolation valves....".
I 1AC-ll
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AC.4 DISCUSSION OF CONTAINMENT PURGE DESIGN (Cont'd)
Response: There are no safety-related structures or O
equipment located outside of the Mark III containment which could be affected by exhausting air or steam.
- c. ... reduction in the containment pressure resulting from the partial loss of containment atmosphere during the accident for ECCS back pressure determination".
Response: No credit is taken on transient or accident analysis for containment pressure when determining adequate Net Positive Suction Head (NPSH) for ECCS equipment. Consequently, there is no need to conduct such an analysis.
- d. "The maximum allowable leak rate of the purge isolation valves....".
Response: There is no need to calculate this value because leakage from the purge supply and exhaust lines are prevented by the air positive leakage control system. Refer to Section 6.5.3.3, The staff interim position of October 23, 1979 contains two key points:
(1) " ... emphasis should be placed... on limiting all purging and venting times to as low as achievable. To justify venting or purging, there must be an established need to improve working conditions..." ,
1AC-12 O
16Il7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 i
, N- 1D.2.3 Variable Assessments (Continued)
- environmental qualifications of safety-related electrical equip-ment. In addition, the Regulatory Guide requires that "the
- seismic portion of qualification should be in accordance with Regulatory Guide 1.100."
The equipment requiring qualification is summarized in Tables 3.10-1 [MPL] (for seismic qualification) and Tables 3.11-9 [MPL]
(for Environmental Qualification). The environmental qualification is to be in accordance with IEEE 323-1974 methods and the envelopes summarized in Tables 3.11-2 through 3.11-8. The seismic qualifica-tion is to be in accordance with IEEE 344-1975 methods. A summary of the qualification methods and results of the qualification evaluation is to be provided by the Applicant. Tables 3.10-1 and 3.11-9 have been reviewed in the course of this assessment to deter-("' mine whether additions to the equipment lists are needed because 1
of the requirements of Regulatory Guide 1.97. Where changes are needed, a statement to that effect is included in the individual paragraphs that follow. Qualification of these additional instru-ments is the responsibility of the Applicant.
I Where performance requirements are required to be specified (imple-mentation position paragraphs C.1.b and C.2.4) , the accuracy for i various components of each instrument channel have been assessed as they apply to normal operation. Assessment of any potential instru-ment accuracy variation due to post-accident environments and their relationship to the accuracy needs of the instruments during post-i accident periods has not been addressed as part of this appendix.
Such an evaluation should be made as part of the environmental qualification program for the post-accident instrumentation and is the Applicant's responsibility to address. As a minimum, Type A variables should be evaluated in this regard.
!O 1D.2-3
. - - . . , , ,n- --, ,- - - - , , , , . - , , , . - - - - - - - - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 The GESSAR II design will meet designated portions of Regulatory Guide 1.97; and any deviations to the remainder of the guide will
~
be justified to the satisfaction of the NRC staff before refer-encing of GESSAR II by the first Applicant.
O O
1D.2-3a
_- _.-- - -~ - .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10
() 1D.2.3.1 Neutron Flux The current neutron monitoring system consists of the source range monitor (SRM), intermediate range monitor (IRM), and the power range monitor (LPRM/APRM). While significant redundancy exists in the design, the current design does not meet the require-ment from Subsection C.1.3.1.b to be "... electrically independent and physically separated from each other and from equipment not classified important-to-safety in accordance with Regulatory Guide 1.75." In addition, because the SRM and IRM must be inserted into the core following a reactor scram and because the drive equipment is not likely to survive a post-accident environment for seismic disturbance, the SRM and IRM do not meet the requirements for avail-ability to monitor the variable during the time of interest.
General Electric is currently evaluating the design of a system which will comply with the criteria specified in Regulatory Guide C\
( j 1.97. The system, called the Wide Range Neutron Monitoring (WRNM) System, will be a replacement for the SRM and IRM indica-tions. Subsequent review and acceptance by the NRC will be required after the system design is finalized.
i 1D.2.3.2 Control Rod Position Indication Control Rod Position Indication is classified as Category 3 by the Regulatory Guide because it serves as a backup to neutron flux.
Its post-accident monitoring function is only to verify function of a reactor protection system and, consequently, this function is l only required for a brief period of time. Because of these reasons, I
the equipment function time and local environment are not specified for the control rod position indication as called for by the imple-l mentation position. No recommendations for change are made for Control Rod Position Indication.
l i O 1D.2-4
1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 l i
APPENDIX 1G RESPONSE TO CP/ML RULE 10CFR50. 34 (f)
O 1
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. _ - - e _ -
l l
l 1
O I
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 APPENDIX 1G CONTENTS i
4 TITLE PAGE SECTION 1G.0 Introduction 1G.0-1 1G.1 Probabilistic Risk Assessment - Item (1) (i) 1G.1-1 1G.2 Auxiliary Feedwater System Evaluation -
i Item (1) (ii) 1G.2-1 1G.3 Impact of.RCP Seal Damages Following Small-Break LOCA With Loss of Offsite Power - 1G.3-1 Item (1) (iii) 1G.4 Report on Overall Safety Effect of PORV Iso- 1G.4-1 lation System - Item (1) (iv) 1G.5 Separation of HPCS and RCIC System Initiation i Levels - Item (1) (v) 1G.5-1 1G.6 Reduction of Challenges and Failures of Safety Relief Valves - Feasibility Study and 1G.6-1 System Modification - Item (1) (vi) 1G.7 Modification of ADS Logic-Feasibility Study and Modification for Increased Diversity of l
' Some Event Sequences - Item (1) (vii) 1G.7-1 1G.8 Restart of Core Spray and LPCI Systems on Low Level - Design and Modification -
l 1G.8-1 i Item (1) (viii) i i 1G.9 Confirm Adequacy of Space Cooling Study for HPCS and RCIC - Item (1) (ix) lG.9-1
.l 1G.10 Verify Qualification of Accumulators on ADS Valves - Item (1) (x) 1G.10-1 f
i 1G.ll Evaluate Depressurization with Other Than 1G.ll-1 i Full ADS - Item (1) (xi) 4 1G.12 Evaluation of Alternative Hydrogen Control i
Systems - Item (1) (xii) 1G.12-1 IG.13 Long-Term Training Upgrade - Item (2) (i) 1G.13-1 i
1G.14 Long-Term Program of Upgrading of Procedures - 1G.14-1 l
Item (2) (ii) 1 1G-i
GESSAR II 22A7007 1 238 NUCLEAR ISLAND Rev. 17 APPENDIX 1G CONTENTS (Continued) O SECTION TITLE PAGE 1G.15 Control Room Design Reviews - Item (2) (iii) 1G.15-1 1G.16 Plant Safety Parameter Display Console (SPDS) - Item (2) (iv) 1G.16-1 1G.17 Safety System Status Monitoring - Item (2) (v) 1G.17-1 1G.18 Reactor Coolant System Vents - Item (2) (vi) 1G.18-1 1G.19 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation - Item (2) (vii) 1G.19-1 1G.20 Post-Accident Sampling - Item (2) (viii) lG.20-1 1G.21 Hydrogen Control System Preliminary Design - Item (2) (ix) 1G.21-1 1G.22 Testing Requirements - Item (2) (x) 1G.22-1 1G.23 Relief and Safety Valve Position Indication - Item (2) (xi) 1G.23-1 1G.24 Auxiliary Feedwater System Automatic Initiation and Flow Indication - Item (2) (xii)1G.24-1 1G.25 Reliability of Power Supplies for Natural Circulation - Item (2) (xiii) 1G.25-1 1G.26 Isolation Dependability - Item (2) (xiv) 1G.26-1 1G.27 Purging - Item (2) (xv) 1G.27-1 lG.28 Design Evaluator - Item (2) (xvi) 1G.28-1 1G.29 Additional Accident Monitoring Instrumentation - Item (2) (xvii) 1G.29-1 1G.30 Identification of and Recovery from Conditions Leading to Inadequate Core Cooling - Item (2) (xviii) 1G.30-1 1G.31 Instrumentation for Monitoring Accident Conditions (Regulatory Guide 1.97) -
Item (2) (xix) 1G.31-1 1G.32 Power Supplies for Pressurizer Relief Valves, Block Valves and Level Indication -
Item (2) (xx) 1G.32-1 1G-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 i
APPENDIX 1G L( )
CONTENTS (Continued)
SECTION TITLE PAGE 1G.33 Describe Automatic and Manual Actions for 2 Proper Functioning of Auxiliary Heat Removal Systems when FW System Not Operable -
Item (2) (xxi) 1G.33-1 1G.34 Analysis of Upgrading of Integrated Control System - Item (2) (xxii) 1G.34-1
- 1G.35 Hand-Wired Safety-Grade Anticipatory Reactor Trips - Item (2) (xxiii) 1G.35-1 1G.36 Central Water Level Recording - Item (2) i (xxiv) 1G.36-1 1G.37 Upgrade License Emergency Support Facility -
Item (2) (xxv) 1G.37-1 1G.38 Primary Coolant Sources Outside the Contain-ment Structure - Item (2) (xxvi) 1G.38-1 i
0 1G.39 Implant Radiation Monitoring - Item (2) (xxvii)lG.39-1 i
1G.40 Control Room Habitability - Item (2) (xxviii) 1G.40-1 l
1G.41 Procedures for Feedback of Operating, Design and Construction Experience - Item (3) (i) 1G.41-1
Item (3) (iii) 1G.43-1 l
l 1G.44 Dedicated Containment Penetrations Equivalent to a Single 3-Foot Diameter Opening -
Item (3) (iv) 1G.44-1 Containment Integrity - Item (3) (v) 1G.45-1 1G.45 1G.46 Dedicated Penetration - Item (3) (vi) 1G.46-1 1G.47 Organization and Staffing to Oversee Design and Construction - Item (3) (vii) 1G.47-1
, /~
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)
~
, 1G-iii/1G-iv
! GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1
! t i APPENDIX 1G
! LIST OF TABLES j TABLE TITLE PAGE 1G.0-1 GESSAR II - CP/ML Rule Cross Reference 1G.0-3 1
i i
i i
i I
i i
1 LIST OF ILLUSTRATIONS
- FIGURE TITLE PAGE ,
f 1G.47-1 GESSAR II Organizational and Management i Structure 1G.47-4 4
1G.47-2 Top Level Management Oversight and Technical ,.
Control 1G.47-5 !
1G-v/1G-vi l
_ - _ .. _ _ _ _ _ _ _ - _ _ _ .- _ __ .. _._._______ _ ~ . _ _ _
4 l
GESSAR II 22A7007
, 238 NUCLEAR ISLAND Rev. 17 APPENDIX 1G RESPONSE TO CP/ML RULE 10CFR50. 34 (f) 1G.0 INTRODUCTION l On January 15, 1982 (47 FR 2286) the NRC amended 10CFR50.34 to include paragraph (f), " Additional TMI-Related Requirements."
These additional requirements were directed to each applicant for a light-water-reactor Construction Permit or Manufacturing License (CP/ML) whose application was pending as of February 16, 1982.
In its " Proposed Coramission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation," on April 13,
- 1983 (48 FR 16014), the NRC proposed to extend its policy such that future CP applications or reactivations of CP applications l
previously docketed also comply with the CP/ML rule. This I
appendix reports General Electric's responses for the 238 Nuclear
! Island to the CP/ML rule.
l The responses demonstrate that the NRC requirements are satis-factorily fulfilled for the 238 Nuclear Island. For each item, j
a summary of the NRC position is given and followed by a response.
The response clarifies the issue as it pertains to the 238 f Nuclear Island and/or provides a listing of applicable GESSAR II l l
sections, relevant correspondence, or other necessary documenta-tion that may be referenced for complete clarification of our 1 position. Where a particular requirement is not applicable to l l the 238 Nuclear Island, a statement to that effect is provided in the response.
l l
l l
l lG.0-1 l
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 17 1G.0 INTRODUCTION (Continued)
O For items 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 is complete in that all of the
" Additional TMI-Related Requirements" approved for implementation by the NRC as listed in 10CFR50.34 (f) have been favorably addressed where they apply to the 238 Nuclear Island.
The bracketed item numbers at the end of each title correspond with the subsections in 10CFR50.34 (f) . Alphanumeric designations at the end of each "NRC Position" statement correspond to the related action plan items in NUREG-0718 and NUREG-0660 (provided in 10CFR50. 34 (f) for information only).
Table 1G.0-1 is provided as a convenient cross-reference which consolidates pertinent information associated with each of the 47 requirements. This includes the 10CFR50.34 (f) subsection, the action plan numbers, the GESSAR II Appendix G section number, the item title, and the GESSAR II reference detailing resolution.
O 1G.0-2
t .
Table 1G.0-1 3
GESSAR II - CP/ML RULE CROSS REFERENCE .
CP/ML Rule Item GESSAR II l Section Action Plan Section Title __
GESSAR II Reference (1) (i) II.B.8 1G.1 Probabilistic Risk Appendix 15D Assessment l
(ii) II.E.1.1 1G.2 Auxiliary Feedwater System Not Applicable (PWR Only)
Evaluation (iii) II.K.2.16 & 1G.3 Impact of RCP Seal Damages Sections lA.46 and 1A.66 Following Small-Break LOCA
- II.K.3.25
- with Loss of Offsite Power gg w nm
.O (iv) II.K.3.2 1G.4 Report on Overall Safety Not Applicable (PWR Only) yy o Effect on PORV Isolation >N l b '
System HH s
m (v) II.K.3.13 1G.5 Separation of HPCS and RCIC Section 1A.58 &
System Initiation Levels $
0 (vi) II.K.3.16 1G.6 Reduction of Challenges and Section lA.60 Failures of Safety Relief ,
Valves - Feasibility Study and System Modification I
(vii) II.K.3.18 1G.7 Modification of ADS Logic- Section lA.62 Feasibility Study and Modi-fication for Increased Diversity of Some Event
- Sequences xw (viii) II.K.3.21 1G.8 Festart of Core Spray and Section lA.63 @y ;
LPCI Systems on Low Level- - 4 Design and Modification r$
Q -J
Table 1G.0-1 GESSAR II - CP/ML RULE CROSS REFEPENCE (Continued)
CP/ML Rule Item GESSAR II Section Action Plan Section Title GESSAR II Reference (1) (ix) II.K.3.24 1G.9 Confirm Adequacy of Space Section lA.65 Cooling Study for HPCS and ;
RCIC j (x) II.K.3.28 1G.10 Verify Qualification of Section lA.68 Accumulators on ADS Valves II.K.3.45 1G.ll Evaluate Depressurization Section lA.72 . U (xi) with Other Than Full ADS {
"o Zm E (xii) --
1G.12 Evaluation of Alternative 1G.12 ! $$
a Hydrogen Control Systems !
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(2) (i) I.A.4.2 1G.13 Long-Term Training Upgrade Section 18.2 f "U H
l m
(ii) I.C.9 1G.14 Long-Term Program of Applicant Responbibility y I z Upgrading of Procedures i o (iii) 1.D.1 1G.15 Control Room Design Reviews Appendix 18 l (iv) 1.D.2 1G.16 Plant Safety Parameter Appendix 18B Display Console j (v) I.D.3 1G.17 Safety System Status 1G.17 [
Monitoring l t
(vi) II.B.1 1G.18 Reactor Coolant System Section lA.19 Vents gy (vii) II.B.2 1G.19 Plant Shielding to Provide Section lA.20 *j Access to Vital Areas and [o y
Protect Safety Equipment for Post-Accident Operation 9 e
p r} p
'\.Y 'w ) O Table 1G.0-1 GESSAR II - CP/ML RULE CROSS REFERENCE (Continued)
CP/ML Rule Item GESSAR II Section Action Plan Section Title GESSAR Il Reterence ___
(2) (viii) II.B.3 1G.20 Post-Accident Sampling Section 1A.21 (ix) II.B.O 1G.21 Hydrogen Control System 1G.21 Preliminary Design (x) II.D.1 1G.22 Testing Requirements Section lA.23 w
II.D.3 1G.23 Relief and Safety Valve Section lA.24 w (xi) "
Position Indication zo H Cm O (xii) II.E.1.2 1G.24 Auxiliary Feedwater System Not Applicable (PWR Only) om o Automatic Initiation and E$
& Flow Indication yW H
(xiii) I.E.3.1 1G.25 Reliability of Power Not Applicable (PWR Only) $"
Supplies for Natural %
Circulation g (xiv) II.E.4.2 1G.26 Isolation Dependability Section lA.29 ,
(xv) II.E.4.4 1G.27 Purging 1G.27 k (xvi) II.E.5.1 1G.28 Design Evaluator Not Applicable (B&W Only)
(xvii) II.F.1 1G.29 Additional Accident Moni- Appendix 1D toring Instrumentation ,
(xviii) II.F.2 1G.30 Identification of and Section lA.31 mg Recovery from Conditions oy Leading to Inadequate Core .w Cooling g ww
Table 1G.0-1 GESSAR II - CP/ML RULE CROSS REFERENCE (Continued)
CP/ML Rule Item GESSAR II Section Action Plan Section Title GESSAR II Reference (2) (xix) II.F.3 1G.31 Instrumentation for Monitor- Appendix lD ing Accident Conditions (Regulatory Guide 1.97)
(xx) II.G.1 1G.32 Power Supplies for Pres- Not Applicable (PWR Only) surizer Relief Valves, Block Valves and Level Indication (xxi) II.K.1.22 lG.33 Describe Automatic and Section lA.38 Manual Actions for Proper Zo a Functioning of Auxiliary $$
o Heat Removal Systems When yy 4 FW System Not Operable gx H
(xxii) II.K.2.9 1G.34 Analysis of Upgrading of Not Applicable (B&W Only) yH Integrated Control System g Z
(xxiii) II.K.2.10 1G.35 Hand-Wired Safety-Grade Not Applicable (B&W Only) O Anticipatory Reactor Trips (xxiv) II.K.3.23 1G.36 Central Water Level Sections lA.39 and
, Recording lD.2.3.4 (xxv) III.A.l.2 1G.37 Upgrade License Emergency Applicant Responsibility Support Facility (xxvi) III.D.l.1 1G.38 Primary Coolant Sources Section lA.77 Outside the Containment Structure aw (xxvii) III.D.3.3 1G.39 In-Plant Radiation Applicant Responsibility IDo Monitoring C O O O
O O O Table 1G.0-1 GESSAR II - CP/ML RULE CROSS REFERENCE (Continued)
CP/ML Rule Item GESSAR II Section Action Plan Section Title GESSAR II Reference (2) (xxviii)III.D.3.4 1G.40 Control Room Habitability Section lA.79 (3) (i) I.C.5 1G.41 Procedures for Feedback of Applicant Responsibility 4
Operating, Design and Con-struction Experience (ii) I.F.1 1G.42 Expand QA List 1G.42 w w
(iii) I.F.2 1G.43 Develop More Detailed QA 1G.43 Criteria $$
a cn b$
1 (iv) II.B.8 1G.44 Dedicated Containment Pene- 1G.44 l ,0 trations, Equivalent to $#H O a Single 3-foot Diameter g m Opening p (v) II.B.8 1G.45 Containment Integrity Section 1G.45 y I
(vi) II.E.4.1 1G.46 Dedicated Penetration Section lA.28 l (vii) II.J.3.1 1G.47 Organization and Staffing 1G.47 j
to Oversee Design and Construction NN (D M
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i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 i
() 1G.1 PROBABILISTIC RISK ASSESSMENT (Item (1) (i))
NRC Position I
Perform a plant / site specific probabilistic risk assessment, the aim of which is to seek such improvements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant. (II.B.8)
Response ,
A BWR/6 Mark III Probabilistic Risk Assessment (PRA) was submitted -
as Section 15D.3 on March 19, 1982. The PRA is currently under-
, going NRC staff review.
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! GESSAR II 22A7007 2
238 NUCLEAR ISLAND Rev. 17 1G.2 AUXILIARY FEEDWATER SYSTEM EVALUATION
[ Item (1) (ii) ]
NRC Position j Perform an evaluation of the proposed auxiliary feedwater system (AFWS), to include (applicable to PWR's only) (II.E.1.1): -
l (A) A simplified AFWS reliability analysis using event-tree and fault-tree logic techniques.
l (B) A design review of AFWS.
(C) An evaluation of AFWS flow design bases and criteria.
Response t 4
This requirement is not applicable to the 238 Nuclear Island.
It applies only to PWR-type reactors.
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u,- _ _ . _ . _ . . _ , _ . . _ _ . . _ _ _ _ _ . . . . . _ . _ _ _ , - . . _ , . _ .
GESSAR II 22A7007 l
238 NUCLEAR ISLAND Rev. 17 II ) 1G.3 IMPACT OF RCP SEAL DAMAGES FOLLOWING SMALL-BREAK LOCA WITH LOSS OF OFFSITE POWER [ ITEM (1) (iii)]
1 NRC Position Perform an evaluation of the potential for and impact of reactor i coolant pump seal damage following small-break LOCA with loss of offsite power. If damage cannot be precluded, provide an analy-sis of the limiting small-break loss-of-coolant accident with
- subsequent reactor coolant pump seal damage. (II.K.2.16 and II.K.3.25) i
Response
The consequences of a loss of cooling water to the reactor recirculation pump seal coolers (II . K. 2.16 and II . K. 3. 2 5) have j
been analyzed with favorable results as indicated in Section 1A.66.
The NRC approval of the test is found in NUREG-0979, page 15-7
{( (Section 15. 2) .
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_ ~ .. . .- _. _ . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 2
() 1G.4 REPORT ON OVERALL SAFETY EFFECT OF PORV ISOLATION SYSTEM
[ Item (1) (iv)]
i NRC Position i
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-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 i
operate when the reactor coolant system pressure falls after the PORV has opened. (Applicable to PWR's only). (II.K.3.2)
Response
This requirement is not applicable to the 238 Nuclear Island. It applies only to PWR-type reactors.
}
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 bl O
1G.5 SEPARATION OF HPCS AND RCIC SYSTEM INITIATION LEVELS
[ Item (1) (v)]
NRC Position Perform an evaluation of the safety effectiveness of providing for separation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) system initiation levels so that the RCIC system initiates at a higher water level than the HPCI system, and of providing that both systems restart on low water level.
(For plants with high pressure core spray systems in lieu of high pressure coolant injection systems, substitute the words, "high pressure core spray" for "high pressure coolant injection" and
("HPCS" for "HPCI") (Applicable to BWR's only). (II.K.3.13) .
Response
4
() The evaluation was performed and is described in Section lA.58.
Separation of initiation levels was found to be unnecessary.
However, the RCIC will be modified to allow automatic restart (following high level trip). NRC staff approval of these conclu-sions is documented in NUREG-0979, page 5-22.
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- .- -- . .-~ _.- - .. - - - , .-- --. -. . . - ..
GESSAR II 22A7007 4 238 NUCT: EAR ISLAND Rev. 17
() 1G.6 REDUCTION OF CHALLENGES AND FAILURES OF SAFETY RELIEF VALVES - FEASIBILITY STUDY AND SYSTEM MODIFICATION
[ Item (1) (vi)]
NRC Position Perform a study to identify practicable system modifications that 1 would reduce challenges and failures of relief valves, without compromising the performance of the valves or other systems.
(Applicable to BWR's only.) (II.K.3.16) i
Response
A study was performed and is described in Section 1A.60. As a result of the study, a modification was proposed and will be
) implemented following NRC staff approval. The Staff acknowledged the modification proposal in NUREG-0979, page 5-7.
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1G.6-1/lG.6-2
GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. 17 I \ 1G.7 MODIFICATION OF ADS LOGIC-FEASIBILITY STUDY AND
- MODIFICATION FOR INCREASED DIVERSITY OF SOME EVENT SEQUENCES [ Item (1) (vii)]
i 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
- adequate core cooling. (Applicable to BWR's only.) (II.K.3.18)
Response
l The study was performed by the BWR Owners' Group and is described in Section lA.62. The favorable options and NRC concurrence is presented in NUREG-0979, page 6-41.
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4 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 l
1G.8 RESTART OF CORE SPRAY AND LPCI SYSTEMS ON LOW LEVEL - DESIGN
{ AND MODIFICATION [ Item (1) (viii)]
NRC Position 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) 1
Response
The study was performed by GE in conjunction with the BWR Owners' Group. As explained in Section lA.63, the study concluded that ,
no changes are necessary or appropriate. The NRC staff has approved GE's position in NUREG-0979, Subsection 7.3.2.4 (page 7-30).
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1G.8-1/lG.8-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.9 CONFIRM ADEQUACY OF SPACE COOLING STUDY FOR HPCS AND RCIC O. [ 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) systems, following a complete loss of offsite power to the plant for at least two (2) hours. (For plants with high pressure core spray systems in lieu of high pressure coolant injection systems, substitute the words, "high pressure core spray" for "high pressure coolant injection" and "HPCS" for "HPCI") (Applicable to BWR's only). (II.K.3.24)
Response
fw The completed study is described in Section lA.65. Results
(__) indicated that the cooling systems for the HPCS and RCIC Systems are adequate and no plant modifications are required.
1G.9-1/lG.9-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
.1G.10 VERIFY QUALIFICATION OF ACCUMULATORS ON ADS VALVES
[ Item (1) (x)]
NRC Position Perform a study to ensure that the Automatic Depressurization System, valves, accumulators, and associated equipment and instrumentation will be capable of performing their intended functions during and following an accident situation, taking no credit for non-safety related equipment or instrumentation, and accounting for nora al expected air (or nitrogen) leakage through valves. (Applicable to BWR's only). (II.K.3.28)
Response
The study and its favorable results are described in Section 1A.68.
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GESSAR II 22A7007 230 NUCLEAR ISLAND Rev. 17 1
1G.ll EVALUATE DEPRESSURIZATION WITH OTHER THAN FULL ADS O lItem (1) (xi)]
- l NRC Position 4
Provide an evaluation of depressurization methods, other than by full actuation of the automatic depressurization system, that would reduce the possibility of exceeding vessel integrity limits
- during rapid cooldown. (Applicable to BWR's only) (II.K.3.45)
Response
i t The completed evaluation is described in Section lA.72. It was concluded that no change in depressurization rate is required or appropriate. NRC concurrence is indicated in NUREG-0979, page 6-42.
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) . _ _ _ _ . _ _ _ , _ _ _ _ - - _ - _ - .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 g)
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1G.12 EVALUATION OF ALTERNATIVE HYDROGEN CONTROL SYSTEMS
[ 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 comparisen 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) of 10CFR50.34.
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(C) For the selected system, preliminary design descrip-tions of equipment, function, and layout.
l
Response
(A) Comparison of costs and benefits of the alternative systems considered will be provided by the Applicant.
l (B) The Applicant will provide the analyses and test data to verify compliance with the requirements of 10CFR50. 34 ( f ) (2 ) (ix) .
1 (C) The Applicant will provide the design descriptions of
( equipment, function, and layout.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 i
1G.13 LONG-TERM TRAINING UPGRADE [ Item (2) (i)]
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1 NRC Position 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)
Response
The GE training simulator, near Tulsa, Oklahoma, was used as part
! of the control room human factors system / task analysis discussed in Section 18.2. This simulator was selected for use because it closely resembles the control room design for the 238 Nuclear Island except as noted in Chapter 18. Small-break LOCAs were among the transients selected for review. No limitations in the simulator's capability to simulate small breaks were noted.
C\
i Q It shall be the Applicant's responsibility to establish a training program which addresses the concerns related to Item I.A.4.2 of NUREG-0718.
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t GESSAR II 22A7007 238 NUCLEAR ISLAND' Rev. 17 i'
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1G.14 LONG-TERM PROGRAM OF UPGRADING OF PROCEDURES [ Item (2) (ii)]
NRC Position j
, Establish a program, to begin during construction and follow into i
operation, for integrating and expanding current efforts to im-prove plant procedures. The scope of the program shall include emergency procedures, reliability analyses, human factors engi-neering, crisis management, operator training, and coordination i with INPO and other industry efforts. (Applicable to construction permit applicants only) (I.C.9)
Response
i Plant procedures are the Applicant's responsibility to provide.
The Emergency Procedure Guidelines (EPGs) are established as part l of the 238 Nuclear Island design as recommendations for operator 1
action. The human factors system / task analysis (Sections 18.2 and 18.3) is based on actions consistant with the EPGs. A representative sample of emergency procedures for the 238 Nuclear Island is provided in Appendix 18A.
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i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.15 CONTROL ROOM DESIGN REVIEWS (Item (2) (iii)]
I NRC Position i
Provide, for Commission review, a control room design that reflects state-of-the-art human factor principles prior to committing to fabrication or revision of fabricated control room panels and layouts. (I.D.1)
- Response Chapter 18 provides a control room review, consistent with the Standard Review Plan (SRP), NUREG-0800, Section 18. It provides an EPG-based system / task analysis (Sections 18.2 and 18.3) as well as a control room panel and arrangement review (Section 18.4) of selected panels. These were based on state-of-the-art human j factors principles as established by GE and the BWR Owners'
() Group.
The Human Factors Engineering Branch (HFEB) of the NRC reviewed
~
the text of the document and performed an audit of the finished control room modeled by the Black Fox Simulator near Tulsa, Oklahoma. The audit was conducted utilizing the human factors principles contained in NUREG-0700. The NRC review and results are contained in the GESSAR II Safety Evaluation Report (NUREG-0979). Also indicated in that report is GE's 5-point plan f for resolution of human engineering discrepancies (HEDs) identi-
, fied as a result of the control room audit. The plan was judged acceptable by the Staff as indicated in NUREG-0979, page 18-4 l (Section 18.6).
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i-. - -.- - .-- .- ._
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
() 1G.16 PLANT SAFETY PARAMETER DISPLAY CONSOLE (SPDS) [ Item (2) (iv)]
i NRC Position
- Provide a plant safety parameter display console that will display
, to operators a minimum set of parameters defining the safety status of the plant, capable of displaying a full range of important plant 1
i parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded. (I.D.2)
Response
The 238 Nuclear Island design includes an Emergency Response Information System (ERIS) which satisfies the requirements of NUREG-0718, Item I.D.2. The ERIS is described in Appendix 18B.
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1G.16-l/lG.16-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
() 1G.17 SAFETY SYSTEM STATUS MONITORING [ Item (2) (v) ]
NRC Position i
Provide for automatic indication of the bypassed and inoperable
- l status of safety systems. (I.D.3)
I Response The 238 Nuclear Island design fully complies with Regulatory Guide 1.47 (see Subsection 19.3.7.16). The' automatic indication of bypassed and inoperable status of safety systems is therefore inherent in the design.
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- Assumed " operable" incorrect; changed to " inoperable."
1G.17-1/lG.17-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.18 REACTOR COOLANT SYSTEM VENTS [ Item (2) (vi)]
NRC Position Provide the capability of high point venting of noncondensible gases from the reactor coolant system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an vn.,ccept-able increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment inLugrity. (II.B.1)
Response
The vent provisions are part of the plant's original design and are covered by the original design bases. A complete description and comparison with requirements is given in Section lA.19. NRC
(_- Staff approval is given in NUREG-0979, page 5-7.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.19 PLANT SHIELDING TO PROVIDE ACCESS TO VITAL AREAS AND O- PROTECT SAFETY EQUIPMENT FOR POST-ACCIDENT OPERATION IItem (2) (vii)]
NRC Position Perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain TID 14844 source term radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipment from the radiation environment. (II.B.2)
Response
The details of the completed design review are described in Attachment A to Appendix 1A. As indicated in the attachment and in the response in Section 1A.20, no corrective action is required.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
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/ 1G.20 POST-ACCIDENT SAMPLING [ Item (2) (viii)]
ks)'
NRC Position Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID 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 certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chl oride , and boron concentrations. (II.B.3)
Response
A post-accident sample system has been added as indicated in the
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xs response in Section lA.21. Technical descriptions are found in Attachment B to Appendix 1A and Subsection 19.3.9.17. NUREG-0979, Section 9.3.2.2 (page 9-21), has approved those portions of the system within GE's scope of supply. The remainder shall be supplied by the Applicant.
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1G.20-1/lG.20-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
()
v 1G.21 HYDROGEN CONTROL SYSTEM PRELIMINARY DESIGN [ Item (2) (ix)}
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 I (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
() support hydrogen combustion.
(B) Combustiole 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.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.21 HYDROGEN CONTROL SYSTEM PRELIMINARY DESIGN (Item (2) (ix)}
(Continued) h (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.
The Hydrogen Control System shall provide with reasonable assur-ance that:
(1) 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 support hydrogen combustion.
(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 1G.21-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 I i 1G.21 HYDROGEN CONTROL SYSTEM PRELIMINARY DESIGN [ Item (2) (ix)]
\- # (Continued) of a 100% fuel-clad metal water reaction, including the environmental conditions created by activation of the hydrogen control system.
The following criteria will be used to design the Hydrogen Control 1 System:
(1) The system will be single active failure proof.
(2) Operation of the Hydrogen Control System will not adversely affect the safe shutdown of the plant.
(3) The system will be protected from tornado and external missile hazards.
(4) The system will not compromise the containment design.
(5) If the method chosen for hydrogen control is a post-accident inerting system, inadvertent actuation of the system must be safely accommodated during plant operation.
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GESSAR II 22A7007
- 238 NUCLEAR ISLAND Rev. 17 i
1G.22 TESTING REQUIREMENTS (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 3
expected under operating conditions, transients and accidents.
l Consideration of anticipated transient without scram (ATWS)
I conditions shall be included in the test program. Actual testing under ATWS conditions need not be carried out until subsequent 5 phases of the test program are developed. (II.D.1) l
Response
4 j, A generic test program has been conducted through the BWR Owners' Group and is described in the response in Section lA.23. A plant
() modification was recommended as a result of the program, and has been approved by the NRC Staff (see NUREG-0979, Section 7. 3. 2.1,
. ~Page 7-29). Therefore, the deletion of the high drywell pressure j interlock will be implemented in the GESSAR II design before its reference by the first Applicant.
s Consideration of ATWS conditions shall be included in the test i program following ATWS rulemaking, as indicated in Section 15.8.
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GESSAR II 22A7007 238 NUCLEAF ISLAND Rev. 17
() 1G.23 RELIEF AND SAFETY VALVE POSITION INDICATION [ Item (2) (xi)}
NRC Position Provide direct indication of relief and safety valve position (open or closed) in the control room. (II.D.3)
Response
A system providing positive position indication is described in the response in Section 1A.24. Subsequent approval by the NRC
{ staff was indicated in NUREG-0979, Section 7.3.2.3. Therefore, the design modification will be implemented before GESSAR II's reference by its first Applicant.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.24 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW
}
INDICATION [ Item (2) (xii))
i NRC-Position ;
i Provide automatic and manual auxiliary feedwater (AFW) system initiation, and provide auxiliary feedwater system flow indication B
in the control room. (Applicable to PWR's only) (II.E.1.2) [
I Response i
This requirement is not applicable to the 238 Nuclear Island.
It applies only to PWR-type reactors.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. ,17
() 1G.25 RELIABILITY OF POWER SUPPLIES FOR NATURAL CIRCULATION
[ Item (2) (xiii) ]
NRC Position Provide 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 power available. (Applicable to PWR's only) (II.E.3.1)
Response
This requirement is not applicable to the 238 Nuclear Island. It applies only to PWR-type reactors.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
[dD 1G.26 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, (B) For each non-essential penetration (except instrument lines) have two isolation barriers in series, (C) Do not result in reopening of the containment isolation valves on resetting of the isolation signal, (D) Utilize a containment set p7 int pressure for initiating containment isolation as low as is compatible with
[V } normal operation, (E) Include automatic closing on a high radiation signal for all systems that provide a path to the environs.
Response
The containment isolation system has been reviewed in accordance with NUREG-0737, Item II.E.4.2. See Section lA.29 and Attachment C to Appendix 1A.
~
Crite rio'n' [C") above required design modifications as described in Subsection LAC.3. NRC staff approval was given in NUREG-0979, Section 6.2.7 (Page 6-33).
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GESSAR II 22A7007
]
238 NUCLEAR ISLAND Rev. 17
() 1G.27 PIIRGING [ Item (2) (xv)]
NRC Position Provide a capability for containment purging / venting designed to i minimize the purging time consistent with ALARA principles for occupational exposure. Provide and demonstrate high assurance that the purge system will reliably isolate under accident conditions. (II.E.4.4) i
! Response The general safety concern over containment purging stems from the presumption that the purge line provides a path for accident releases prior to isolation, and further, that the dynamic effects of the accident may interfere with effective isolation of the purge line.
,O j These presumptions are not directly applicable to the Mark III l containment of the GESSAR II design. The reactor coolant system t
j piping is enclosed in the drywell which communicates with the con-tainment only through the suppression pool. Releases from the primary system are subjected to the quenching and scrubbing action of the suppression pool before entering the containment, so the purge system does not provide a path for primary system releases in the same sense as other containment designs. Even so, special care is being taken in the purge system design, specifically for valve operability assurance. Finally, the isolation valves are provided with positive leakage control.
l The specific points of this item are addressed below:
The basis for the purge system design is justified in the response i
to Question 6.28 (480.23). The present design provides for con-()
I tinuous purging of the containment during power operation at l
1G.27-1 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 Response (Continued) 5000 cfm through a 9-in. line to reduce airborne radionuclide concentrations to a level which permits continuous access. This is in keeping with occupational ALARA considerations, because extensive containment access for routine maintenance is required.
The performance of prototype 6-in. purge isolation valves has been evaluated and meets the requirements of BTP CSB 6-4 for isolation and dependability under accident pressures. The Appli-cant will provide performance information for the specific 9-in.
purge isolation valves. These purge valves are conservatively designed to close against the containment design pressure of 15 psig (the drywell pressure would only be a few psi at start of closure). The performance of the 42-in. purge valves will be demonstrated during refueling where the pressure differential is negligible.
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1G.27-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 Ch g%)
1G.28 DESIGN EVALUATOR [ 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.
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1G.28-1/lG.28-2
GESSAR II' 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.29 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION
[\ ' [ Item (2) (xvii) ]
NRC Position Provide 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
Discussion of each of these variables is contained in Appendix 1D.
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GESSAR II 22A7007 3 238 NUCLEAR ISLAND Rev. 17 1G.30 IDENTIFICATION OF AND RECOVERY FROM CONDITIONS LEADING TO INADEQUATE CORE COOLING [ Item (2) (xviii) ]
NRC Position i Provide instruments that provide in the control room an unambigu-j 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
As indicated in Section 1A.31, GE believes the existing highly
~
redundant direct water level instrumentation already provides an unambiguous indication of inadequate core cooling and does not plan to include core-exit thermocouples in the 238 Nuclear Island design.
The NRC staff agreed to broaden the issue from the specific l
requirements for in-core thermocouples to that of monitoring f inadequate core cooling (ICC). (See NUREG-0979, Section 4.4.7, page 4-35.) This issue is contained within Regulatory Guide 1.97 and will be resolved in conjunction with GESSAR II's conformance with the guide as indicated in the response in Section 1G.29.
O 1G.30-1/lG.30-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.31 INSTRUMENTATION FOR MONITORING ACCIDENT CONDITIONS (REGULATORY GUIDE 1.97) (Item (2) (xix)]
NRC Position i
- Provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage. (II.F.3)
Response
The GESSAR II design is assessed against Regulatory Guide 1.97 Revision 2 in Appendix 1D with supplementary justification for deviations provided on the docket (letter, G. G. Sherwood to _
- D. G. Eisenhut, dated April 28, 1983). The assessment, including the deviations and justifications, is still under review by the NRC.
, GE agrees that the GESSAR II design will meet designated portions v of the guide; and any deviations to the remainder of the guide will be justified to the satisfaction of the Staff before refer-encing of GESSAR II by the first Applicant.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.32 POWER SUPPLIES FOR PRESSURIZER RELIEF VALVES, BLOCK VALVES AND LEVEL INDICATION [ Item (2) (xx)]
NRC Position Provide power supplies for pressurizer relief valves, block valves, and level indicators such that: (A) Level indicators are powered 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 safety and (C) electric power is provided from emergency power sources. (Applicable to PWR's only). (II.G.1)
Response
This requirement is not applicable to the 238 Nuclear Island. It applies only to PWR-type reactors.
O 1G.32-1/lG.32-2
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GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. 17 1
C 1G.33 DESCRIBE AUTOMATIC & MANUAL ACTIONS FOR PROPER FUNCTION-ING OF AUXILIARY HEAT REMOVAL SYSTEMS WHEN FW SYSTEM NOT OPERABLE [ Item (2) (xxi) ]
NRC Position Design auxiliary heat removal systems such that necessary auto-matic and manual actions can be taken to ensure proper functioning when the main feedwater system is not operable. (Applicable to BWR's only). (II.K.l.22)
Response
The actions of the auxiliary heat removal system for these con-ditions is described in Section lA.38. The NRC Staff approval is documented in NUREG-0979, Section 5.4.2, pages 5-26 and 5-27.
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i GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 17
! 1G.34 ANALYSIS OF UPGRADING OF INTEGRATED CONTROL SYSTEM
. [ Item (2) (xxii)]
i L
f NRC Position
! Perform a failure modes and effects analysis of the integrated i control system (ICS) to include consideration of failures and
- - effects of input and output signals to the ICS. (Applicable to B&W-designed plants only). (II.K.2.9) l Response This requirement is not applicable to the 238 Nuclear Island. It 4 applies only to PWR-type (B&W designed) reactors, t
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
() 1G.35 HAND-WIRED SAFETY-GRADE ANTICIPATORY REACTOR TRIPS
[ Item (2) (xxiii)]
NRC-Position Provide, as part of the reactor protection system, an anticipatory 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.
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1G.35-1/lG.35-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
() 1G.36 CENTRAL WATER LEVEL RECORDING [ Item (2) (xxiv) ]
NRC Position Provide the capability to record reactor vessel water level in one location on recorders that meet normal post-accident recording requirements. (Applicable to BWR's only). (II .K . 3. 2 3 )
Response
j The reactor vessel water level instrumentation used for post-
! accident recording is described in Subsection 1D.2.3.4. A j technical description for enhanced level recording is described in Attachment C to Appendix 1D.
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GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. 17 1G.37 UPGRADE LICENSE EMERGENCY SUPPORT FACILITY (JteF; ( 2) (xxv) ]
l NRC Position [
d Provide an onsite Technical Support Center, an onsite Operational l
Support Center, and, for construction permit applications only, a I nearsite Emergency Operations Facility. (III.A.l.2) 4
Response
The response to this requirement will be provided by the Applicant.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
(' ' 1G.38 PRIMARY COOLANT SOURCES OUTSIDE THE CONTAINMENT STRUCTURE
\ [ Item (2) (xxvi))
NRC-Position I '
! 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-
! cants shall submit a leakage control program, including an initial test program, a schedule for retesting these systems, and the
- actions to be taken for minimizing leakage from such systems. The
- goal is to minimize potential exposures to workers and public, and
, to provide reasonable assurance that excessive leakage will not l prevent the use of systems needed in an emergency. (III.D.l.1)
Response
i The leakage control program is detailed in Section lA.77. NRC acceptance of the TMI action plan is found in NUREG-0979, Section 9.3.4 (Page 9-24).
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4 GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. 17 ,
F 1G.39 IMPLANT RADIATION MONITORING .-
s
[ Item (2) (xxvii)] i i i 1- ,
1
! NRC-Position i
Provide for monitoring of inplant radiation and airborne radio-
{'
activity as appropriate for a broad range of routine and accident '
- conditions. (III.D.3.3)
{
. Response
}s l The response to this requirement will be supplied by the Applicant.
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i GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 17 i
1G.40 CONTROL ROOM HABITABILITY [ Item (2) (xxviii) ]
(}
NRC Position l
j Evaluate potential pathways for radioactivity and radiation that i may lead to control room habitability problems under accident !
conditions resulting in a TID 14844 source term release, and make necessary design provisions to preclude such problems. (III.D.3.4)
- Response
) As indicated in Section lA.79, analysis demonstrated no changes to the 238 Nuclear Island are required to satisfy this item.
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1 GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. 17
- 1G.41 PROCEDURES FOR FEEDBACK OF OPERATING, DESIGN AND CONSTRUC-TION EXPERIENCE [ Item (3) (i)]
f l
i NRC Position
} Provide administrative precedures for evaluating operating, design j and construction experience and for ensuring that applicable
- important industry experiences will be provided in a timely manner to those designing and constructing the plant. (I.C.5) l Response The response to this requirement will be supplied by the Applicant.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
/~ 1G.42 EXPAND QA LIST [ Item (3) (ii)]
V)
NRC Position i
Ensure that the quality assurance (QA) list required by Criterion i II, App. B. 10 CFR Part 50 includes all structures, systems, and components important to safety. (I.F.1) i
Response
As discussed in Subsection 17.1.2, the identification of safety-related structures, systems, and components (0-list) to be con-trolled by the quality assurance program is the responsibility of the Applicant. The Applicant will supplement and clarify its Q-list in accordance with Question 17.3 (NRC Question 260.3). The appropriate items will be added to Table 3.2-1. The remaining items will be subject to the pertinent requirements of GE's and/or
() the Applicant's QA programs, unless otherwise justified.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 m
( 1G.43 DEVELOP MORE DETAILED QA CRITERIA (Item (3) (iii)]
)
NRC Position 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 design, construction and installation; (D) establishing criteria 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)
O Response (A) NEDO-ll209-04A, " Nuclear Energy Business Operations (NEBO) BWR Quality Assurance Program Description", con-4 forms to this requirement. See Paragraph 1.1 on page 1-1.
(B) NEBO services performed at the construction site are under the Owners' QA program. NEBO provides QA program support to the Owner as described in NEDO-ll209-04A, pages 1-3, 1-7, 11-1, and 11-2.
(C) The NEBO Quality Assurance and Reliability Operation (QA&RO) is responsible for preparing the top level NEBO i quality policy and instructions for issue by the Vice
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 Response (Continued)
President and General Manager, NEBO. QA&RO is also responsible for preparing and issuing several NEBO quality procedures. These documents are identified on pages 2-2 and 2-3 of NEDO-11209-04A.
In addition, QA&RO is responsible for developing, issu-ing, and controlling the BWR Engineering Operating Procedures (NEDE-21109), the ASME Quality Assurance Program Manual (NEDE-20387), and NEDO-ll209.
The NEBO line QA organizations are responsible for developing and documenting a quality system in com-pliance with NEBO policies, instructions and procedures, and applicable codes, standards, and regulatory require-ments. See NEDO-ll209-04A, Section 1.3, "QA Functional Responsibilities".
(D) NEDO-ll209-04A responds to each of the QA programmatic requirements of 10CFR50, Appendix B, and the require-ments of the regulatory guides and industry standards identified in Table 2-1. In addition, the NEBO QA pro-gram conforms to the requirements of the ASME Code.
l (E) NEDO-ll209-04A, Section 2-1, fourth paragraph, describes the qualification of training of NEBO personnel who per-form activities affecting quality. See also Subsection 1.4, "QA Personnel Responsibilities and Qualifications."
(F) The NRC has evaluated the NEBO QA Program implementation for several years and has found that the program, includ-ing sizing of the QA staff, is being implemented satis-factorily. See NRC letter from J. T. Collins to W. H. Bruggeman, Docket No. 99900403, dated May 24, 1983.
f 1G.43-2
GESSAR II 22'A7007 238 NUCLEAR ISLAND Rev. 17 f^\
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Response (Continued)
(G) NEDO-ll209-04A, Section 17, describes the NEBO commit-ments related to "as-built" documentation. The NEBO commitments are further detailed on pages 2-10, 2-11, and 2-13 thru 2-15.
f (H) QA&RO has the following responsibilities that are documented in NEDO-ll209-04A, Subsection 1.3:
i 1. Develop, issue, and control the Engineering Operat-1 ing Procedures for NEBO design and analysis activities i
- 2. Conduct or participate in independent design reviews
- 3. Conduct independent audits of the NEBO design con-trol program.
t i'
Based'on the foregoing evaluation, it is demonstrated that the NEBO QA program as described in NEDO-ll209-04A, and as currently l
approved by the NRC, includes full consideration of the matters identified in this item.
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1G.43-3/lG.43-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.44 DEDICATED CONTAINMENT PENETRATIONS EQUIVALENT TO A SINGLE 3-FOOT DIAMETER OPENING [ Item (3) (iv)]
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) i Response As a result of the NRC staff review for Final Design Approval, GE has agreed to provide separate penetrations for purging during operation and during refueling. The isolation valves for the 42-in. diameter refueling purge penetrations are now locked closed during normal operation and have been dedicated as containment
() penetrations as required by this item. The details of the pene-tration arrangement will be provided before the first Applicant references GESSAR II.
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1G.44-1/lG.44-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 f
O' 1G.45 CONTAINMENT INTEGRITY [ Item (3) (v)]
NRC Position 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) i (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 i
Boiler Pressure Vessel Code,Section III, Division 2 g 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 will be considered by the staff, if good cause is shown by an applicant. Systems necessary to ensure contain-ment integrity 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-
'O enced in paragraphs (f) (3) (v) (A) (1) and ( f) (3) (v) (B) (1) 1G.45-1
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 17 1G.45 CONTAINMENT INTEGRITY [ Item (3) (v)] (Continued) of 10CFR50.34, 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 Se 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.
O lG.45-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
() .
1G.45 CONTAINMENT INTEGRITY [ Item (3) (v)] (Continued)
Response (Continued)
(A) The containment design basis is 15 psig. Appendix G
- 'T of Subsection 15D.3 (PRA) provides the corresponding y containment capability analyses. These analyses demon-
', N, strate that all areas of the containment exceed 45-psig i ', Service Level C Limits except for the knuckle region of
- the 2:1 torispherical head. Preliminary analysis indi-cates that the knuckle region can also meet 45-psig
. Service Level C Limits by modifying the curvature of the s, head using a three-center design (no other modifications are necessary).
s 4
The containment vessel design capability of 45-psig Service Level C is included as an interface requirement.
(B) Containment structure loadings produced by an inadver-tent full actuation of a post-accident inerting Hydrogen Control System shall not exceed ASME Section III, NE3220 Service Level A Limits as calculated according to 10CFR50. 34 (f) (3) (v) (B) .
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 O
V 1G.46 DEDICATED PENETRATION [ Item (3) (vi)]
NRC Position For plant designs with external hydrogen recombiners, provide redundant dedicated containment penetrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere. (II.E.4.1)
Response
As indicated in Section lA.28, this requirement is not applicable to the 238 Nuclear Island (GESSAR II) design. NRC concurrence is given in NUREG-0979, Section 6.2.7 (page 6-33).
L.]
1G.46-1/lG.46-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1
() 1G.47 ORGANIZATION AND STAFFING TO OVERSEE DESIGN AND CONSTBUCTION [ Item (3) (vii) ]
NRC Position Provide a description of the management plant for design and construction activities, to include: (A) the organizational and
, management structure singularly responsible for direction of design and construction of the proposed plant; (B) technical 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 l and construction, including the preparation and implementation of l() procedures necessary to guide the effort. (II.J.3.1) j Response (A) The organization and management structure is depicted in I
Figure 1G.47-1. The Utility evaluates the type of LWR desired. Studies are conducted to select an appropriate site. A designer is selected and licensing of the proposed site is secured by the Applicant. General Electric supplies the pre-approved GESSAR II design and furnishes technical direction during procurement and construction. Changes, non-conformances and as-builts are reviewed and approved by General Electric as the engineer of record for the Nuclear Island. The con-structor is selected by the Applicant and is under his direct control.
O 1G.47-1
.. GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 1G.47 ORGANIZATION AND STAFFING TO OVERSEE DESIGN AND CONSTRUCTION (Item (3 ) (vii) ] (Continued)
(B) The Utility that applies GE3SAR II obtains a design that is basically complete. Therefore, an architect-engineer (A/E) is generally retained to furnish the closure engineering effort that may be required. The A/E is a subcontractor to the Utility and the Utility can mobilize the full resources of the A/E organization.
Likewise, General Electric has the capacity to mobilize a very large amount of qualified technical personnel if so requested or deemed necessary by the Utility.
(C) Utilization of the GESSAR II design greatly simplifies interfaces. The coordination between the NSSS and the balance of the Nuclear Island is performed internally by General Electric using existing procedures. Since the GESSAR II design is essentially complete, the amount of interfaces required is further simplified from custom designs.
The Utility has to verify that the preselected and pre-approved site meets the envelope requirements of the GESSAR II. Therefore, the services of an A/E organiza-tion are generally retained by the Utility. The Utility forms a project group under the leadership of an experienced executive who oversees all the principal organizations: General Electric; Contractor and " house" A/E.
This project group within the Utility organization is responsible for coordinating among contractors, verifying contractual performance, monitoring schedules, licensing and support construction activities.
O 1G.47-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 O
V 1G.47 ORGANIZATION AND STAFFING TO OVERSEE DESIGN AND CONSTRUCTION [ Item (3) (vii) ] (Continued)
(D) Transition of operations from construction to operation of the plant is Utility unique and should be supplied by each Applicant.
(E) Generally, each Project is managed by a senior, executive within the Utility who reports to the Vice President of Engineering. Timely reports are prepared on status of all major contracts. Discrete reviews are systematic-ally performed and problems are tracked to successful resolution.
The Vice President of Engineering is intimately familiar with the design and construction of each Nuclear plant I
('"g being added to the Company's grid, since it represents
\2 a very large percentage of the Utility's investments.
Delays and changes are extremely costly and are con-stantly monitored. Procedure and guideline implementa-tion is normally reported by the QA/QC organization and they impact directly the job's progress which is thoroughly reviewed by top management. Reviews are l periodically made by means of memos; letters; standard forms; meetings and telephone calls.
General Electric's top management is responsible for the implementation of all procedures and they must sign off on pertinent sheets to indicate agreement / disagreement with the contents. Periodic, systematic reviews are made by top management in order to responsibly control l
the work.
l 1G.47-3
- o
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 O
APPLICANT AND PLANT OWNER 1I NUCLEAR ISLAND GENER AL ELECTRIC _
1 I
Y U BUILDER CONSTRUCTOR O
Figure 1G.47-1. GESSAR II Organizational and Management Structure 1
l O
1G.47-4 L L------__
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
(
CEO SENIOR VICE PRESIDENT SERVICES I I I VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT OPERATIONS QA/O FINANCE VICE PRESIDENT VICE PRESIDENT ENGINEERING CONSTRUCTOR O
PROJECT CHIEF PROJECT MANAGERS ENGINEERS SUPERVISOR STAFF ---- STAFF STAFF GE A/E LICENSING SU8 CONTRACTORS Figure 1G.47-2. Top Level Management Oversight and Technical Control 1G.47-5/lG.47-6
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1 i' GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
,3 4
>(
Oj 3.7.3.1 Seismic Analysis Methods (Continued) are superimposed and the resulting spectrum is the upper bound envelope of all the individual spectrum curves considered.
-For vibrating systems and their supports, multi-degree-of-freedom models are used in accordance with the lumped-parameter modeling techniques and normal mode theory described in Subsection 3.7.2.1.1 and the references listed in Subsection 3.7.6. Piping analysis is described in Subsection 3.7.3.3.1.
When testing is used to qualify Seismic Category I subsystems and components, all the loads ncrmally acting on the equipment are simulated during the test. The actual mounting of the equipment is also simulated or duplicated. Tests are performed by supplying input accelerations to the shake table to such an extent that generated test response spectra (TRS) envelope the required
() response spectra.
For certain Seismic Category I equipment and components where I
dynamic testing is necessary to ensure functional integrity, test performance data and results reflect the following:
(1) performance data of equipment which has been subjected to dynamic loads equal to or greater than those experi-enced under the specified seismic conditions; (2) test data from previously tested comparable equipment which has been subjected under similar conditions to dynamic loads equal to or greater than those specified; and (3) actual testing of equipment in accordance with one of the methods described in Subsection 3.9.2.2 and
() Section 3.10.
3.7-33
~ .._ __.
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 e
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 3.7.3.2.2 Other Equipment and Components _
Criterion II.2.b of SRP Section 3.7.3 recommends that at least one safe shutdown earthquake (SSE) and five operating basis earth-quakes (OBEs) should be assumed during the plant life. It also recommends that a minimum of 10 maximur stress cycles per earth-quake 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. _
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.
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 stress reversals occur below 75% of the maximum stress level, and 95% of the reversals lie below 50% of the maximum stress level. o 3.7-34
_. .~ . . _ . . . - _ . . .- . . - _ _ _ _ _ - _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev.-17 3.7.3.2.2 Other Equipment and Components (Continued)
In summary, the cyclic behavior number of fatigue cycles of a cor.' e c. ant during an earthquake is found in the following manner:
(1) the fundamental frequency and peak seismic loads are found by a standard seismic analysis (i . e . , from eigen extraction and forced response analysis);
I i
?
1 I
3.7-34a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 f-w, 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.
~
Acceptance Criterion II. A. l. (b) of SRP Section 4.2 requires that the cumulative number of 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 Reference 1 has the following limiting condition:
Actual time at stress < 1*0 Allowable time at stress Actual number of cycles at stressAllowable number of cyc x
\ Since the Reference 1 limit is more conservative than that of the SRP, the deviation is acceptable.
Acceptance Criterion II. A.l. (c) of SRP Section 4.2 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, J.S. Charnley to NRC staff, January 25, 1983) is that the fuel assembly is evalu-ated to ensure that the fuel will not fail due to fretting wear of the fuel assembly components. This statement plus the discus-sion on fretting wear in Section 2.6.3 of Reference 1 show that fretting wear is considered in the design analysis and the intent of the SRP is met.
Acceptance Criterion II . A. l . (g) of SRP Section 4.2 requires that the fuel assembly hold-down capability exceed the worst-case hydraulic loads for normal operation, which includes anticipated O. operation occurrences. The GESSAR II design limit for fuel -
assembly lift-off is 0.52 inch as documented in NUREG-0979. This 4.2-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 4.2.1 Design Bases (Continued) limit was calculated to be the largest which would not permit sufficient lateral displacement of the fuel assembly to result in control blade interference. Since control blade interference is prevented, this design limit is acceptable.
Acceptance Criterion II. A.2. (e) of SRP Section 4.2 states that for '
normal operation and anticipated operational occurrences, center-line melting of the fuel is not permitted. The GESSAR II design basis for fuel pellet overheating (Letter, J.S. Charnley to NRC staff, January 25, 1983) is that 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 shows that the GE design objective is to avoid fuel fail-ures due to fuel melting and thus meets the intent of the SRP criterion.
Acceptance Criterion II. A. 2. (g) of SRP Section 4.2 states that the uniform fuel cladding strain (plastic and 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 Reference 1, Subsection A.4.2.1 of Appendix A.
Acceptance Criterion II. A.2. (i) of SRP Sebtion 4.2 limits the applied stress on the cladding to 1 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 cur-rently 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.
4.2-la
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
'N 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 l 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 total span and are separated uniformly throughout
(_) the core on a 12-in. pitch. Each control rod is surrounded by four fuel assemblies.
4.2-lb
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 4.2.2.1 Control Rods (Continued)
O 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.
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4.2-1c
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6
- O
, x_/ 4.2.2.1 Control Rods (Continued) i i
The U-shaped sheaths are resistance welded to the center post, handle and castings to form a rigid housing to contain the boron-carbide-filled absorber rods. Rollers at the top and bottom of the control rod guide the control rod as it is inserted and withdrawn from the core. The control rods are cooled by the core bypass flow. The U-shaped sheaths are perforated to allow the coolant to circulate freely about the abcorber tubes. Oper-ating experience has shown that control rods constructed as
, described above are not susceptible to dimensional distortions.
i The boron-carbide (B C) powder in the absorber tubes is compacted 4
, to about 70% of its theoretical density. The boron-carbide
. contains a minimum of 76.5% by weight natural boron. The boron-10 (B-10) minimum content of the boron is 18% by weight.
j g-w Absorber tubes are made of Type-304 stainless steel. Each
(/ absorber tube is 0.220 in, in outside diameter and has a 0.027 in, wall thickness. Absorber tubes are sealed by a plug welded into each end. The boron-carbide is longitudinally separated into individual compartments by stainless steel balls at approximately 17-in. intervals. The steel balls are held in place by a slight crimp of the tube. Should boron-carbide tend l to compact further in service, the steel balls will distribute the resulting voids over the length of the absorber tube.
4.2.2.2 Velocity Limiter i
The control rod velocity limiter (Figure 4.2-3) is an integral i
part of the bottom assembly of each control rod. This engi-l neered safeguard protects against high reactivity insertion i
i rate by limiting the control rod velocity in the event of a
- control rod drop accident. It is a one-way device in that the
. control rod scram velocity is not significantly affected, but
\_/ the control rod dropout velocity is reduced to a permissible limit.
4.2-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 O)
N, 4.2.2.2 Velocity Limiter (Continued)
The velocity limiter is in the form of two nearly mated, conical t
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.
f, To limit control rod velocity during dropout, but not during
( ,) scram, the velocity limiter is 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 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.
v 4.2-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 g 4.2.4 Testing, Inspection and Surveillance Plans s
See Appendix A, Subsection A.4.2.4 of Reference 1.
The Applicant will provide a routine fuel inspection program to provide information on irradiated and discharged fuel as indicated in SRP Section 4.2.II.D.3. A typical program involves visual examination of selected assemblies (commonly 5 to 10% of the dis-charged fuel) , concentrating on the lead bundles. Visual examin-2 ations normally include, but are not necessarily limited to, crud buildup, rod bowing, and missing components. Additional inspections should be performed depending on the results of operational monitoring, including coolant activity and the visual inspections.
4.2.5 References
" General Electric Standard Application for Reactor Fuel,"
l.
() NEDE-240ll-P-A, latest approved revision.
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4.2-4
4 GESSAR II '22A7007 238 NUCLEAR ISLAND Rev. 6
( 4.4.3.3.1 Flow Control (Continued) i speed, where the LFMG set will power the pump and motor. The flow control valve is then opened to the maximum position, at which point reactor heatup and pressurization can commence. When opera-ting pressure has been established, reactor power can be increased.
This power-flow increase will follow a line within Region I of the flow control map shown in Figure 4.4-1.
When reactor power is greater than approximately 20-28% of rated, 4 the low feedwater flow interlock is cleared and the main recir-culation pumps can be switched to the 100% speed power source. The flow control valve is closed to the minimum position before the speed change to prevent large increases in core power and potential flux scram. This operation occurs within Region II of the operating map. The system is then brought to the desired power-flow level within the normal operating area of the map (Region IV) by opening
) the flow control valves and by withdrawing control rods.
Control rod withdrawal with constant flow control valve position l will result in power / flow changes along lines of constant c sub (v) (constant position) . Flow control valve movement with constant control rod position will result in power / flow changes along, or nearly parallel to, the rated flow control line.
4.4.3.4 Temperature-Power Operating Map (PWR)
Not aprlicable.
l 4.4.3.5 Load-Following Characteristics See Appendix A, Subsection A.4.4.3.5 of Reference 1.
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4.4-9
.,-e -- , 4.- -# . , , . . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 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.
Results of a stability analysis will be provided before the first Applicant references GESSAR II. _
4.4.5 Testing and Verification See Appendix A, Subsection A.4.4.5 of Reference 1.
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The Applicant will include in the plant Technical Specifications the requirement that core flow will be checked at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to detect flow reduction. _
4.4.6 Instrumentation Requirements See Appendix A, Subsection A.4.4.6 of Reference 1.
4.4.6.1 Loose Parts To be supplied by Applicant.
4.4.7 References
- 1. " General Electric Standard Application for Reactor Fuel,"
(NEDE-240ll, latest approved revision).
O 4.4-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 A
t j 4.6.2.3.1.8 Mechanical Damage (Continued) i guide tube (see Subsections 4.2.3.3.7 through 4.2.3.3.8 for these analyses).
4.6.2.3.1.9 Evaluation of Control Rod Velocity Limiter The control rod velocity limiter limits the free-fall velocity of the control rod to a value that cannot result in nuclear system process barrier damage. This velocity is evaluated by the rod drop accident analysis in Chapter 15 (Accident Analysis).
4.6.2.3.2 Control Rod Drives 4.6.2.3.2.1 Evaluation of Scram Time The rod scram function of the CRD system provides the negative
() reactivity insertion required by safety design basis 4 . 6 .1.1.1.1. l (1) . The scram time shown in the description is adequate as shown by the transient analyses of Chapter 15.
4.6.2.3.2.2 Analysis of Malfunction Relating to Rod Withdrawal There are no known single malfunctions that cause the unplanned l
withdrawal of even a single control rod. However, if multiple l
malfunctions are postulated, studies show that an unplanned rod withdrawal can occur at withdrawal speeds that vary with the combination of malfunctions postulated.
l-i 4.6-29
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.1 Drive Housing Fails at Attachment Weld O
The bottom head of the reactor vessel has a penetration for each CRD location. A drive housing is raised into position inside each penetration and fastened by welding. The drive is raised into the drive housing and bolted to a flange at the bottom of the housing.
The CRD housing material at the vessel penetration is seamless, Alloy 600 tubing with a minimum tensile strength of 80,000 psi, and Type-304 stainless steel pipe below the vessel with a minimum strength of 75,000 psi. The basic failure considered here is a complete circumferential crack through the housing wall at an elevation just below the J-weld.
Static loads on the housing wall include the weight of the drive and the control rod, the weight of the housing below the J-weld, and the reactor pressure acting on the 6-in, diameter cross-sectional area of the housing and the drive. Dynamic loading results from the reaction force during drive operation.
If the housing were to fail as described, the following sequence of events is foreseen. The housing would separate from the vessel. The CRD and housing would be blown downward against the support structure, by reactor pressure acting on the cross-sectional area of the housing and the drive. The downward motion of the drive and associated parts would be determined by the gap between the bottom of the drive and the support structure and by the deflection of the support structure under load. In the cur-rent design, maximum deflection is approximately 3 in. If the collet were to remain latched, no further control rod ejection would occur 1; the housing would not drop far enough to clear the vessel penetration; reactor water would leak at a rate of approxi-mately 180 gpm through the 0.03-in. diametral clearance between the housing and the vessel penetration.
O 4.6-30
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
(g 5.2.3.3.2 Control of Welding V
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, }
I Subsection NA. Components are either held for an extended time i I
at preheat temperature to assure removal of hydrogen, or preheat
()'#
is maintained until post-weld heat treatment. The minimum pre-
\- heat and maximum interpass temperatures are specified and monitored.
Acceptance Criterion II.3.b. (1) (a) of SRP Section 5.2.3 for con-trol of preheat temperature requires that minimum and maximum l While the GESSAR II control interpass temperatures be specified.
of low-hydrogen electrodes to prevent hydrogen cracking (provided in Subsection 5.2.3.3.4) does not explicitly meet this requirement, the GESSAR II control will assure that cracking of components made from low-alloy steels does not occur during fabrication. Further, the GESSAR II control minimizes the possibility of subsequent cracking resulting from hydrogen being retained in the weldment. _
All welds were nondestructively examined by radiographic methods.
In addition, a supplemental ultrasonic examination was performed.
v 5.2-37
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 5.2.3.3.2.2 Regulatory Guide 1.34: Control of Electroslag Weld Properties No electroslag welding was perfcrmed on BWR components 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.
e l
e 5.2-37a 1
GESSAR II 22A7007
, 238 NUCLEAR ISLAND Rev. 0
'"'N 5.2.3.3.3 Regulatory Guide 1.66: Nondestructive Examination
& of Tubular Products Regulatory Guide 1.66 describes a method of implementing requirements acceptable to NRC regarding non-destructive examination requirements of tubular products used in RCPB.
Wrought tubular products were supplied in accordance with applicable ASTM /ASME material specifications. Additionally, the i specification for the tubular product used for CRD housings specified ultrasonic examination to paragraph NB-2550 of ASME Code Section III.
These RCPB components met the requirements of ASME Codes existing i at time of placement of order which predated Regulatory Guide 1.66. At the time of the placement of the orders 10CFR50 Appendix B, requirements and the ASME Code requirements assured
() adequate control of quality for the products.
This Regulatory Guide was withdrawn on September 28, 1977, by l the NRC because the additional requirements imposed by the guide were satisfied by the ASME Code.
7 l
For commitment and revision number, see Section 1.8.
5.2.3.3.4 Moisture Control for Low Hydrogen, Covered Arc Welding Electrodes All low-hydrogen-covered welding electrodes are stored in controlled storage areas and only authorized persons are permitted to release and distribute electrodes. Electrodes are received in hermetically sealed cannisters. After removal from l
the sealed containers, electrodes which are not immediately used are placed in storage ovens which are maintained at about 2500F (generally 2000F minimum).
(}
5.2-38
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
() 5.2.3.4.2.3 Regulatory Guide 1.71:
Areas of Limited Accessibility (Continued)
Welder Qualification for All ASME Section III welds are 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 is accomplished by mockup welding. Mock-up is examined with radiography or sectioning.
The Acceptance Criterion II.3.b.(3) of SRP Section 5.2.3 is based on Regulatory Guide 1.71. GESSAR II meets the intent of this regulatory guide by utilizing the alternate approach given in Subsection 1.8.71. _
5.2.3.4.3 Regulatory Guide 1.66: Nondestructive Examination of Tubular Products
. O 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 l
i This section discusses the inservice inspection and testing l
! program for the NRC Quality Group A components; i.e., ASME Boiler and Pressure Vessel Code Section III, Class 1, components. It f
will show how the program meets requirements of Section XI of i the ASME Code.
r j 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) is designed i
5.2-43
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 5.2.4.1 System Boundary Subject to Inspection (Continued) and fabricated to permit full compliance with ASME Code Section XI. Access is provided for volumetric examination of pressure l
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O 5.2-43a
d i GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. 0
/% 5.2.4.1 System Boundary Subject to Inspection (Continued)
V i
retaining welds from the external surface. The examination l procedures have been considered in the design of components, weld
, joint configurations, and system arrangements to assure inspec-tability. Periodic design reviews and onsite audits are made throughout the design and erection phases to assure that these objectives are being met.
1 The ASME Code Class 1 components (including supports and pressure retaining bolting) subject to inspection according to the method specified in Table IWB2600 of ASME Code Section XI include the reactor pressure vessel and piping, pumps, and valves within the following systems.
1 (1) Reactor pressure vessel (2) Main steamlines
) (3) Reactor feedwater lines (4) Reactor recirculation lines (5) Residual heat removal system lines
! (6) Core spray lines j (7) Reactor core isolation cooling system lines i (8) Standby liquid control / core AP line (9) Reactor water cleanup lines (10) Reactor drain line i
Where the system penetrates primary containment, the areas of examination on Class 1 components as defined in Table IWB2500 will be extended to include the first isolation valve outside containment.
O .
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5.2<44
_ _ _ - , . . _ . . _ . _ _ _ . - _ _ _ _ _ _ . _ _ _ _ _ _ _ - . _ _ . . - - - . ~ .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O b)
\_ 6.4 HABITABILITY SYSTEMS The control room habitability system is provided to ensure that the control room operators can remain in the control room and take actions to operate the plant safely under normal conditions and to maintain it in a safe condition under accident conditions.
The Habitability Systems include missile protection, radiation shielding, radiation monitoring, air filtration and ventilation systems, lighting, personne1'and administrative support, and fire protection.
The detailed descriptions of the various habitability systems and
, provisions are discussed in the following sections:
Conformance with NRC General Design Criterion 19 Section 3.1 Wind and Tornado Protection Section 3.3 Flood Design Section 3.4 Missile Protection Section 3.5 Protection against dynamic effects Section 3.6 associated with the postulated rupture of piping Seismic Design of Electrical Section 3.10 Components 4
Environmental Design of Section 3.11 Mechanical and Electrical
-- Equipment
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6.4-1
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 17 6.4 HABITABILITY SYSTEMS (Continued)
, Radiation Protection Section 12.3 and Chapter 15.0 Heating, Ventilating and Air Subsection 9.4.1 Conditioning (HVAC)
Fire Protection Subsection 9.5.1 Lighting Systems Subsection 9.5.3
, Fower 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.
The term " control building" includes the main control room, areas .
adjacent to the main control room containing plant information and equipment necessary to normal and emergency operations, and kitchen and sanitary facilities. These areas include the rooms which com-ply with the requirements of SRP Section 6.4 for " Control Room Emergency Zone." It is also the entire zone serviced by the con-
~
trol room ventilation system. " Emergency conditions" include .
such postulated releases as radioactive materials, toxic gases, smoke and steam.
t O
6.4-2 t
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
() 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. t i
1 (9) The MSPLCS controls include interlocks to prevent inadvertent operation of the system. In particular, interlocks are provided to prevent damage to the MSPLCS, or to the main steam system, due to accidental 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
]
j activated from appropriately designed safety systems or circuits.
(10) The MSPLCS is designed to permit testing of the oper-O)
( ability of controls and actuating devices during power operation to the extent practical,'and complete testing i
of system function during plant shutdowns.
(11) The MSPLCS is designed so that: (a) thermal stresses and pressures associated with flashing and thermal deformations, under the loading conditions associated
! with the activated system shall not affect the structural i
integrity or operability of the main steam system or main steam isolation valves; and (b) any deformation of isolation valve internals shall not induce leakage of l
the main steamline isolation valve beyond the capacity or capability of the MSPLCS.
l (12) Equinment is provided (as part of the MSPLCS) to prevent the +-elease of valve stem packing leakage to the environ-ment from main steam system isolation valves outside the
() containment.
6.7-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 I
O (13) The design of the MSPLCS complies with the requirements of SRP Section 6.7 and Regulatory Guide 1.96 with the exceptions discussed in Subsection 1.8.96.
O l
1 1
0 6.7-3a l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
( System Description
}6.7.2 6.7.2.1 General Description Two independent systems (outboard and inboard) are provided to accomplish the leakage control function. The leakage control barrier is established by pressurizing an isolated volume in the main steamline system, thus isolating the containment with respect to the environment. The pressurized volume eliminates out-leakage through the closed MSIVs and main steam drain lines such that any leakage which does occur is inward from the pressurized volume into the reactor pressure vessel or containment. Both systems are connected to the offsite as well as onsite emergency power.
Figures 6.7-1, 6.7-2, and 7.3-6 are, the P&ID, process diagram and functional control diagram, respectively. The outboard system is connected to each of the main steam' system shutoff valves, drain lines (inboard and outboard MSIV), and outboard MSIV stem packing
(} leak-offlines. The inboard system is connected to the outboard MSIV body (inlet side), and to the inboard MSIV drain lines located outside the primary containment.
The MSPLCS is categorized in three sections: (1) an injection section, consisting of a system connection to the air positive seal system, maintenance valve pressure control valve (PCV), bypass valve including the restricting orifice, flow measuring device, pressure relief valve, swing-free disc check valve, and the air injection valve; (2) an isolation section, consisting of the two isolation valves, maintenance valves and including the process l line connected to the main steam system; and (3) a header section, which is a portion of the system between the isolation and injec-4 tion section up to and including the drain valve and a line connection to the radwaste system. '
1 p
l U 6.7-4 L.
GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 17 6.8.3 System Evaluation (Continued)
[L.))
which are designed to Seismic Category I, ASME Code III, Class 2, Quality Group B and Quality Assurance B requirements.
The pneumatic supply is separated into two independent divisions, with each division capable of supplying 100% of the requirements of the division being serviced. Each division is mechanically and electrically separated from the other. The system satisfies the components' air demands during all plant operation conditions (normal through f aulted) .
Safety grade portions of the Pneumatic Supply System are capable of being isolated from the nonsafety parts and retaining their function during LOCA and/or seismic events under which any nonsafety parts may be damaged.
/- m
('"') Pipe routing of Division 1 and Division 2 pneumatic air is kept separated by enough space so that a single fire, equipment dropping accident, strike from a single high energy whipping pipe, jet force from a single broken pipe, internally generated missile or wetting equipment with spraying water cannot prevent the other division from accomplishing its safety function. Separation is accomplished by spatial separation or by a reinforced concrete barrier, to ensure separation of each pneumatic air division from any systems and components which belong to the other pneumatic air division.
6.8.4 Inspection and Testing Requirements Periodic in-service inspection of components, in accordance with ASME Section XI, to ensure the capability and integrity of the system is mandatory. Air quality shall be tested periodically to m m
assure compliance with ANSI MCll.1-1976. -
O 6.8-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 6.8.4 Inspection and Testing Requirements (Continued)
The air-operated isolation valves are capable of being tested to assure their operational integrity by manual actuation of a switch located in the control room and by observation of associated position indication lights. Test and vent connections are provided at the containment isolation valves in order to verify their leaktightness. Operation of valves and associated equipment used to switch from the nonsafety to safety air supply can be tested to assure operational integrity by manual actuation of a switch located in the control room and by observation of associated position indication lights. Periodic tests of the check valves m m
and accumulators shall be conducted to assure valve operability. _
6.8.5 Instrumentation Requirements Supply-air-header-pressure and moisture-content sensors are located at the supply header of each division and monitored in the control room.
A pressure sensor is provided for the safety air supply, and an alarm signals low air pressure.
Automatic isolation valves are provided for the safety and non-safety air supply. Upon a LOCA and/or detection of nonsafety air low pressure, the system automatically isolates the nonsafety air supply and opens the isolation valve to the safety air supply.
A remote manual switch and open/ closed position lights are pro-vided in the control room for verification of proper valve operation. The compressor is shut down by a high backpressure signal or by manual override.
Air-operated isolation valves in series are provided for the pneumatic air pipe penetrations through the containment and drywell. These valves are not automatically closed by a LOCA isolation signal. Air supply for the isolation valve air The operators is from the Pneumatic Supply System itself.
6.8-6
GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 17 8.3.1.1.6.3 Bus Protection (Continued)
(3) 6.9-kV feeders for ESW substations have inverse time overload and ground fault protection.
(4) 6.9-kV feeders used for motor starters have instantaneous, inverse time overload, ground fault and motor protection.
(5) 480V bus incoming line and feeder circuits have inverse time overload and ground fault protection.
See single-line diagrams in Figure 8.3-2, 8.3-3 and 8.3-14 through 8.3-16.
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- . , _ _ _ _ __ _.__ __- _ , ~ _ _ , _ . . _ _ . ~ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ - _ _ _ _ . _ _ _ _ . _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 O)
(_, ' 9.5 OTHER AUXILIARY SYSTEMS i 1
9.5.1 Fire Protection System i
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:
i (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-contributed index of 25 or less are used wherever practicable.
O 9.5-1
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 17 9.5.1.1 Design Bases (Continued)
SRP Acceptance Criterion II.2.a of SRP Section 9.5.1 requires adherence to BTP CMEB 9.5-1.
The 3-hr 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 barriers in the Control Building, the Auxiliary Building, and in the HVAC ductwork that penetrates the Reactor Building wall from the Auxiliary Building and Fuel Building.
Some of these ventilation ducts are shared systems in that they also provide normal ventilation. Other ducts are for smoke vent-ing only. Based on the discussion below the GESSAR II design meets the intent of the SRP.
The Auxiliary Building smoke removal system is shown in Figure 9.4-4 and described in Subsection 9.4.3.2.1.11. Each set of duct work serves and traverses only fire areas 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 the air operator to the vanes so that the damper will close on high temperature. The fire rating of the dampers is 1-1/2 hours. The duct is heavy gage, welded construc-tion which exceeds the requirements for 3-hr fire rated construc-tion. Hence, the design is considered completely adequate for the service.
One of the design objectives of the GESSAR II design is to avoid fire dampers in smoke vents, since their automatic closure would render the smoke vent inoperative at the very time it was needed.
With two exceptions, smoke 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.
O 9.5-la
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
(~N 9.5.1.1 Design Bases (Continued) k The Division 2 cable tunnel located in the corridor of (-)6'-10" elevation of the Auxiliary Building has a dedicated smoke removal system, which passes through the Division 1 area. The duct open-ing is 2.5 ft 2 and is designed to withstand a 3-hr fire.
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 of the Auxiliary Building.
The boundaries of the room have a 3-hr 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 an outboard isolation valve for the duct.
The containment exhaust has two inboard (one manual) isolation valves and one outboard 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-hr rated walls. The room is directly accessible from the Fuel Building or the stair tower between the Fuel and Auxiliary Buildings. All return registers except for the pool sweep are located high in the containment so that bulk mixing, aided by the dome mixing system, would occur l before any combustion gases could enter the ventilation duct.
The containment is more sensitive to bulk air temperature than the ventilation duct. If a fire raised the bulk temperature exces-l sively, containment spray would be initiated to protect the con-tainment at a temperature well below the threshold of damage to l the ventilation duct. For these reasons, the 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-hr fire rating.
i l r~%
( 9.5-lb l
GESSAR II 221.7007 238 NUCLEAR ISLAND Rnv. 17 9.5.1.1 Design Bases (Continued)
The remaining smoke vents which do not have fire dampers are the two in the Control Building. Each one of these smoke vents serves and traverses one division. Since it is impossible for these smoke vents to allow the fire in the area of one division to spread to another division, again the GESSAR II design is considered to be adequate and proper.
Containment isolation valves and included piping of the Fire Protection System are classified as ASME Section III, Class 2 and 9
O 9.5-lc
I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
\
()
+
-9.5.1.1 Design Bases (Continued)
Seismic Category I. Refer to Subsection 6.2.4 for details of these penetrations.
Seismic design requirements are imposed on fire protection systems on an individual basis. Each Fire Protection System component which could damage safety-related systems or components as a result of its collapse due to an earthquake is designed with I Seismic Category I supports to prevent such an occurrence.
4 T'e wet standpipe portion of the Fire Protection System, up to and I including the isolation valves to the sprinkler system and other
}:
systems, as well as the hose racks, is designed to Seismic Cate-i gory I requirements. The remainder of the system outside the buildings is nonseismic.
() The quality assurance (QA) program, in accordance with APCSB 9.5-1 for the design of fire protection systems, is presented in Sub-section 9.5.1.5.3 (Applicant is responsible for QA on construction and operation).
The consequences of inadvertent operation of a suppression system i
and of moderate energy line cracks are discussed in Appendix 9A.
Except for fuel and lubricating oil located in the diesel-generator rooms, there are no storage areas in the Reactor Island buildings l
for flammable liquids, oxidizing agents, flammable compressed gases, corrosive materials or explosive or highly flammable materials. Nonflammable compressed gases (e.g . , air, nitrogen)
I do not represent a fire hazard.
, Small quantities of chemicals may be stored in listed or approved cabinets and containers for immediate use. The CRD maintenance area is an example where such storage is permitted. Control of
(
9.5-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 O
(m,/ 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.
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.
,fq t 6
(_/ 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
~
The Applicant will provide isolation valves, dampers or diversion valves with an automatic control feature that fails in the closed or safe position. Setpoints for actuation of automatic control features initiating actuation of isolation valves, dampers, or diversion valves should be specified in the plant Technical Specifications.
Setpoints are listed in Table 11.5-1.
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11.5-23
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 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 e
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ll.5-23a
_ =_ - __
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O 11.5.4.1 Implementation of General Design Criteria 60 (Continued)
()
discharge on indication of a high radiation level. These include:
(1) offgas posttreatment; (2) containment HVAC; and (3) liquid radwaste effluent.
The effluent isolation functions for each-monitor are given in Table 11.5-1.
11.5.4.2 Implementation of General Design Criteria 64 Radiation levels in radioactive and potentially radioactive process streams are monitored by the following process monitors:
(1) main steamline; l {~'))
x.
(2) offgas pretreatment and posttreatment; (3) carbon bed vault; (4) closed cooling water; and (5) essential service water.
t 11.5.4.3 Basis for Monitor Location Selection Monitor locations are selected to assure compliance with Regula-tory Guide 1.21 in that sample points are located where there is a minimum of disturbance due to fittings and other physical characteristics of the equipment and components. Sample nozzles are inserted into the flow or liquid volume to ensure sampling
()
V the bulk volume of pipes and tanks. In the case of both liquid
. 11.5-24
- GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 i
() 12.1.2.2 Equipment Design Considerations For ALARA Exposures 12.1.2.2.1 General Design Criteria No specific instructions have been given to component designers and engineers regarding ALARA design as provided by specific Acceptance Criterion II.2 of SRP Section 12.1. However, the ,
engineering design procedures require that the component design 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 design considerations made to implement ALARA.
12.1.2.2.2 Equipment Design Considerations to Limit Time Spent in Radiation Areas b
1-V (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
\
12.1-4a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17
) 12.2 RADIATION SOURCES 12.2.1 Contained Sources
- 12.2.1.1 Source Terms 1
With the exception of the vessel and drywell shields, shielding i designs are based on fission product and activation product sources consistent with Section 11.1. For shielding, it is con- I servative to design for fission product sources at peak values.
rather 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-(s ance is made for transient decay while at the same time providing i
1 for transient increase of the noble gas source, daughter product l formation and energy level of emission. Areas where fission i
products are significant relative to Nitrogen-16 include: (1) the condenser off-gas system downstream of the jet air ejector; j (2) liquid and solid radwaste equipment; (3) portions of the RWCS; i
l and (4) portions of the feedwater system downstream of the hotwell l 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).
The following paragraphs represent the source data in various pieces of equipment throughout the p] ant. General locations of t equipment are shovn in the general plant arrangement drawings of
~
Section 1.2. Specific Acceptance Criterion II.6 of Section 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 has been pro- '
.vided in Table 12.2-1 for the reactor and in Chapter 5 for the main steam and recirculation piping.
12.2-1
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 4 12.2.1.1 Source Terms (Continued) -
~
An alternate to the ANSI source term guidance has been used in the GESSAR II design. The equivalent list of isotopes is provided in Chapter 11. The fission product data and the reactor water cor-osion product data provided in Chapter 11 are based on Reference 6.
Reference 6 provides a summary of measured experience for noble gases, halogens, long term fission products and corrosion 1 products. In general, the noble gases, halogens, and cesiums O concentrations using the data of Chapter 11 are in excess of the concentrations in ANSI N237. The remaining Chapter 11 data are sometimes more or sometimes less than the ANSI N237 data. Chapter 11 contains specific details on how the Reference 6 data was developed and used.
In Chapter 12 the reactor water concentrations were used to develop sources in equipment containing reactor water or steam. .
O 4
O 12.2-la ]
.- - - -. .- . - _ _ _ . . - _- -- . . _ _ . = _ . . . _ - - , . . - - .
t GESSAR II 22A7007 238 NUCLEAR ISLAND. Rev. O i.
12.2.1.2.6.3 Radioactive Sources in the Gaseous Radwaste System (Continued) been evaluated for several possible operating modes. In all
- cases, a 1-yr operating time has been used to accumulate the decay activities. This is sufficient time for most isotopes to reach equilibrium.
1 12.2.1.2.6.4 Radioactive Sources in the Solid Radwaste System The solid radwaste system provides the capability for solidifying i
and packaging waste from the other radwaste systems (Sub-section 11.4.2). The wastes are not solidified separately by type or source. The final waste is placed in a steel container.
i The expected average radioactivity content of the solid waste per container is given in Table 12.2-15.
() 12.2.1.2.6.5 Radioactive Sources in the Fuel Pool Cleanup System The radiation source data used in the shield design of the Fuel i-j Pool Cleanup (FPCCU) System filter demineralizer system is given in Table 12.2-16.
12.2.1.2.6.6 Radioactive Sources in the Suppression Pool Cleanup System l
The radiation source data used in the shield of the Suppression Pool Cleanup (SPCU) System is given in Table 12.2-17.
12.2.1.2.7 Radioactive Sources in Piping and Main Steam Systems 12.2.1.2.7.1 Radioactive Sources in Main Steam System All radioactive materials in the Main c+eam System result from radioactive sources carried over from the reactor during plant
[
- operation. In most of the components carrying live steam, the i
12.2-9
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 17 12.2.1.2.7.1 Radioactive Sources in Main Steam System (Continued)
O source 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 is 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- ~
9 lation in Recirculation System Piping. Criterion II.6 of SRP N Section 12.2 provides that the buildup of activated corrosion products in various components and systems should be addressed and allowances made in design source terms should be explained.
Based on current data and analysis, activated corrosion products are most significant in the recirculation piping. Crud levels in steam piping are estimated to be about 1% of those in the recirculation piping.
12.2.1.2.8 Radioactive Sources in the Spent Fuel The radiation source for spent fuel is given in Sub-section 12.2.1.2.1.1.4 (Table 12.2-3) in terms of MeV/sec/W. The design calculation is carried out for a mean element for an appropriate decay time.
O 12.2-10
_ . = - - . - _ - . - . -
i GESSAR II 22A7007
! 238 NUCLEAR ISLAND Rev. 17 -
1 !
t l 12.2.1.2.9 other Radioactive Sources j 12.2.1.2.9.1 Reactor Startup Source j The reactor startup source is shipped to the site in a special j cask designed for shielding. The source is transferred under water !
l while in the cask and loaded into beryllium containers. This is then loaded into the reactor while remaining under water. The 4 !
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GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 17 f i k3l s
15.3.3.2 Sequence of Events and Systems Operations j 15.3.3.2.1 Sequence of Events 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,
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the analysis of this event assumes normal functioning of plant instrumentation and controls, plant protection, and reactor
~
protection systems. Acceptance Criterion II.8 of SRP Section 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 simula-tion 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. Justifica-tion for this difference is given in Section IE.ll. Acceptance criterion II.10 of SRP Section 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 off site power transient described in Subsection 15.2.6; however, this. event
/~'g would be less nevere because of the faster reactor flow coastdown
-/ and the earlier feedwater pump trips.
15.3-13
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 15.3.3.2.2 Systems Operation (Continued)
Operation of safe shutdown features, though not included in this simula tion, 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 vessel level (L8) trip are similar to the considerations in Subsection 15.3.1.2.3.2 (see Appendix 15A for further details) .
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GESSAR II 22A7007 238 NUCLSAR ISLAND Rev. 6 l' 'i
(_ / 15.3.3.3 Core and System Performance 15.3.3.3.1 Mathematical Model The nonlinear dynamic model described in Subsection S.2.2 of Reference 1 is used to simulate this event.
15.3.3.3.2 Input Parameters and Initial Conditions For the purpose of evaluating consequences to the fuel thermal limits, this transient event is assumed to occur as a consequence of an unspecified, instantaneous stoppage of one recirculation pump shaf t while the reactor is operating at 105% NBR steamflow.
Also, the reactor is assumed to be operating at thermally limited (S
\,_.) conditions.
The void coefficient is adjusted to the most conservative value (i.e., the least negative value in Table 15.0-2).
15.3.3.3.3 Results Results for this event are documented in Subsection S.2.5.5 of Reference 1. Based on these results, this event does not have to be reanalyzed for specific core configurations.
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15.3-14
i GESSAR II 22A7007 238 NUCLEAR ISLAND pey, 17 "x 15.4.4.3.2 Results (Continued)
^
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 event does not have to reanalyzed for specific core configurations. Acceptance Criterion II.2.(b) of SRP Section 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 and it is initiated from a low power condition, no MCPR calcula-i tion was performed. .
15.4.4.4 Barrier Performance 4
No evaluation of barrier performance is required for this event since no significant pressure increases are incurred during this f-w
(/ transient (Figure 15.4-1).
15.4.4.5 Radiological Consequences An evaluation of the radiological consequences is not required for this event, since no radioactive material is released from the fuel, i
. 15.4.5 Recirculation Flow Control Failure with Increasing Flow 4
I 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 i 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.
15.4-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 15.4.5.1.2 Frequency Classification This transient disturbance is classified as an incident of moderate frequency.
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GESSAR II 22A7007 1
238 NUCLEAR ISLAND Rev. 7
() 15.4.5.2 Sequence of Events and Systems Operation 15.4.5.2.1 Sequence of Events 15.4.5.2.1.1 Fast Opening of One Recirculation Valve Table 15.4-4 lists the sequence of events for Figure 15.4-2.
15.4.5.2.1.2 Fast Opening of Two Recirculation Valves Table 15.4-5 lists the sequence of events for Figure 15.4-3.
15.4.5.2.1.3 Identification of Operator Actions Initial action by the operator should include:
(1) transfer flow control to manual and reduce flow to
() minimum, and (2) identify cause of failure.
Reactor pressure will be controlled as required, depending on whether a restart or cooldown is planned. In general, the corrective action would be to hold reactor pressure and condenser vacuum for restart after the malfunctioning flow controller has been repaired. The following is the sequence of operator actions expected during the course of the event, assuming restart. The operator should:
(1) observe that all rods are in; (2) check the reactor water level and maintain abcVe low level (L2) trip to prevent MSLIVs from isolating;
() (3) switch the reactor mode switch to the STARTUP position; 15.4-12
L GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6
) 15.4.5.2.1.3 Identification of Operator Actions (Continued) '
i (4) continue to maintain vacuum and turbine seals; 1
(5) transfer the recirculation flow controller to the manual position and reduce setpoint to zero; (6) Survey maintenance requirements and complete the scram report; monitor the turbine coastdown and auxiliary systems; and (7)
(8) establish a restart of the reactor per the normal L procedure f NOTE: Time required from first trouble alarm to restart would be approximately 1 hr.
15.4.5.2.2 Systems Operation i
The analysis of this transient assumes and takes credit for normal functioning of plant instrumentation and controls and the reactor protection system. Operation of engineered safeguards is not expected.
l 15.4.5.3 Core and System Performance 15.4.5.3.1 Input Parameters and Initial Conditions ]
In each of these transient events, the most severe transient results when initial conditions are established for operation at the low end of the rated flow control rod line. Specifically, this is 54% NBR power and 33% core flow. The maximum stroking
() rate of the recirculation loop valves for a master controller failure driving two loops is limited by individual loop controls to ll%/sec 15.4-13 i
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 17 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.
15.-4.5.3.2 Results 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 sur ace heat flux reaches 73% of NBR at approximately 2.2 sec. MCPR remains considerably above the safety limit and average fuel temperature increases only 108 F. -
Acceptance Criterion II.2(b) of SRP Section 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 and it is initiated from a low power condition, no MCPR calculation was performed.
Reactor pressure is 4 3.'3dscd in Subsection 15.4.5.4. _
15.4.5.3.2.2 F a s +. Open.fA of Two Recirculation Valves Figure 15.4-2 illustrates the failure where both recirculation loop main valves are opened at a maximum stroking rate of ll%/sec.
Table 15.4-5 shows the sequence of ev0ats for this failure. It is very similar to the above transient. Flux scram occurs at approx-imately 1.6 sec, peaking at 162% of NB rated, while the average O
15.4-14 s
[ GESSAR II 22A7007 1 238 NUCLEAR ISLAND Rev. 17 i
15.4.5.3.2.2 Fast Opening of Two Recirculation Valves (Continued) j surface heat flux reaches 67% of NB rated at approximately 2.3 sec.
i MCPR remains considerably above the safety limit and average fuel temperature increases 80*F. Therefore, this event does not have to be reanalyzed for specific core configurations.
i
.l t
i lO r
O f 15.4-14a l
i.___._ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ , . _ _ _ _ . _ _ _ , - _ - . _ _ _ _ _ _ . _ _ . _ . . _ _ _ _ . , . , _ _ - _ _ , - . - ~ _ - - - . _ _ - - - - - - -
a GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 15.4.5.4 Barrier Performance 15.4.5.4.1 Fast Opening of One Recirculation Valve This transient results in a very slight increase in reactor vessel pressure (Figure 15.4-2) and therefore represents no threat to the i
RCPB.
15.4.5.4.2 Fast Opening of Two Recirculation Valves This transient results in a very slight increase in reactor vessel pressure (Figure 15.4-3) and therefore represents no threat to the RCPB.
15.4.5.5 Radiological Consequences 4
An evaluation of the radiological consequences is not required for
) this event, since no radioactive material is released from the fuel.
15.4.6 Chemical and Volume Control System Malfunctions i
Not applicable to BWRs. This is a PWR event.
15.4.7 Misplaced Bundle Accident 15.4.7.1 Identification of Causes and Frequency Classification 15.4.7.1.1 Identification of Causes The event discussed in this section is the improper loading of a fuel bundle and subsequent noern* inn of the core. Three errors must occur for this event to take place in the equilibrium core loading. First, a bundle must be misloaded into a wrong location
( in the core. Second, the bundle which was supposed to be loaded where the mislocation occurred would have to also be put in an 1
15.4-15
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 15.4.7.1.1 Identificati,a of Causes (Continued) incorrect location or discharged. Third, the misplaced bundles would have to be overlooked during the core verification process performed following core loading.
The Applicant will include in the plant Operating Procedures /
Technical Specification provisions for potential fuel loading errors.
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 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.
15.4-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 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. m'
_H 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 Guides 1.3 and 1.7. The specific models, assumptions
("w)g
\__ 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 activity 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
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 O
For determining equipment leakage contribution to the LOCA dose, it is assumed that the 50% "plateout and condensation" fraction of the released iodine finds its way into the suppression pool water. This is consistent with Regulatory Guide 1.3 though not with Acceptance Criterion II. (2) of SRP Section 15.6.5. The latter document provides that 50% of the core iodine activity should be assumed to be missed in the sung water being circulated through the containment external piping. The assumptions used in this calculation are the more conservative with respect to BWR post-LOCA total dose calculations. See Item 16 of Subsection 19.3.15.1 for a detailed description of equipment leakage con-tribution to offsite dose.
O O
15.6-15a
i GESSAR II 22A7007
, 238 NUCLEAR ISLAND Rev 0-() 15.6.5.5.1.2 Fission Product Transport to the Environment The transport pathway consists of leakage from the containment to the secondary containment-like structures by several different mechanisms and discharge to the environment through the Standby Gas Treatment System (SGTS):
(1) Containment leakage - The design basis leak rate of the primary containment and its penetrations (excluding the main steamlines) is 1.0%/ day for the duration of the accident. All of this leakage is to the secondary con-j tainment and from there to the environment via a 99%
SGTS. Credit is taken for mixing and holdup within the
]
secondary containment. The Shield Building exhaust rate, leakage rate, and mixing ratio are given on Tables 15.6-9 and 15.6-10.
( (2) Leakage from engineered safety feature (ESP) components outside primary containment.
i, ,
(3) Hydrogen purge - In the event of failure of the Hydrogen Recombiner System, purging of the containment may be necessary to control hydrogen concentration inside the primary containment. The earliest this purge may be i utilized is one hour after the accident rate of 100 scfm minimum. The purge would be processed by SGTS prior to release to the environment.
1 Fission product release to the environment based on the above
, assumption is given in Table 15.6-11.
i
!, 15.6.5.5.1.3 Results The calculated exposures for the design basis analysis are pre-y ,) sented in Table 15.6-12 are well within the guidelines of i 10CFR100.
15.6-16
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 15
( Conclusions A review of the seismic forces indicated that in a few locations they are slightly higher than the envelope (in the range of 3%).
However, the resulting stresses are within the allowables. The seismic response spectra for the various structures resulting from this analysis were generally within the envelopes previously generated. Since the containment structure was stiffened by the additional concrete and the Shield Building participation increases in the load distribution, some minor shifting in frequency was observed (in the range of 5 to 10%).
The revised response spectra are listed below:
Figure Title e 3.10-1 RPV Floor Response Spectra Horizontal Acceleration i OBE El 54.00, 2 Percent Damping 3.10-3 RPV Floor Response Spectra Horizontal Acceleration OBE El 17.83, 2 Percent Damping 3.10-7 Shield Wall Floor Response Spectra Horizontal Acceleration OBE, El 43.00, 2 Percent Damping 3.10-19 Containment Response Spectra Horizontal Acceleration OBE, El 149.00, 2 Percent Damping 3.10-20 Containment Response Spectra Vertical Acceleration OBE, El 149.00, 2 Percent Damping 3.10-21 Containment Response Spectra Horizontal Acceleration OBE, El 110.83, 2 Percent Damping 3.10-23 Containment Response Spectra Horizontal Acceleration OBE, El 93.20, 2 Percent Damping E 3.10-29 Shield Building Response Spectra Horizontal Acceleration OBE, El 115.90, 2 Percent Damping 19.3.3.52-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 i
Conclusions A review of the seismic forces indicated that in a few locations they are slightly higher than the envelope (in the range of 3%).
However, the resulting stresses are within the allowables. The seismic response spectra for the various structures resulting from this analysis were generally within the envelopes previously generated. Since the containment structure was stiffened by the additional concrete and the Shield Building participation increases in the load distribution, some minor shifting in frequency was observed (in the range of 5 to 10%).
The revised response spectra are listed below:
Figure Title 3.10-1 RPV Floor Response Spectra Horizontal Acceleration OBE El 54.00, 2 Percent Damping 3.10-3 RPV Floor Response Spectra Horizontal Acceleration OBE El 17.83, 2 Percent Damping 3.10-7 Shield Wall Floor Response Spectra Horizontal Acceleration OBE, El 43.00, 2 Percent Damping 3.10-19 Containment Response Spectra Horizontal Acceleration OBE,
! El 149.00, 2 Percent Damping I
! 3.10-20 Containment Response Spectra Vertical Acceleration OBE, El 149.00, 2 Percent Damping
(
3.10-21 Containment Response Spectra Horizontal Acceleration OBE, El 110.83, 2 Percent Damping 3.10-23 Containment Response Spectra Horizontal Acceleration OBE, El 93.20, 2 Percent Damping l
('
3.10-29 Shield Building Response Spectra Horizontal Acceleration OBE, El 115.90, 2 Percent Damping 19.3.3.52-3
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GESSAR II 22A7007 !
238 NUCLEAR ISLAND Rev. 17 l l
1 fg 19.3.3.74 QUESTION / RESPONSE 3.74 (220.33)
U QUESTION 3.74 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 Item II.3.c (ii)(a) of Section 3.8.3 of the SRP for the 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 3.8.3.3.6.3.2 was revised to include the missing _
Equation 4 of SRP 3.8.3 II.3.c(ii)(a). Equation 4 is more severe than and bounds Equation 3. Equation 4, not Equation 3, -
is applied to the current design of GESSAR II.
i Os- 19.3.3.74-1/19.3.3.74-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 f~s 19.3.6.28 QUESTION / RESPONSE 6.28 (480.23)
( I N/
QUESTION 6.28 You state on pages 1.8-171 and 9.4-62 of your FSAR that the containment purge valves are open during normal operation and that containment pressure is controlled through these valves.
However, we state our position on this matter in branch technical position CSB6-4 that the use of large purge and vent lines should be restricted to cold shutdown conditions and refueling outages. Provide your basis for purging the ]
containment continuously in light of our position on this matter. (6.2.4)
RESPONSE 6.28 Continuous purging of the primary containment outside of the
,,c) drywell during reactor operation is required for access,
( ,) inspection, and maintenance associated with the control rod drive - hydraulic control units (one per CRD); safety related instrument calibrations, water sampling of reactor water, suppression pool, and upper containment pools, RWCU system and feedwater.
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 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 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
- line.
V) 19.3.6.28-1 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 19.3.6.28 QUESTION / RESPONSE 6.28 (480.23) (Continued)
O During the NRC staf f review for Preliminary Design Approval, the containment purge penetrations were modified to reduce their size for normal operating continuous purge from a 42-in. diameter to an 18-in. diameter. A further reduction in size from an 18-in.
diameter to a 9-in. diameter for normal operating continuous purge was made as a result of the NRC staff review for Final Design Approval of the GESSAR II design. In addition, these 9-in, diameter penetrations for continuous purge during operation are now separate. Thus, the larger 42-in. penetrations are open 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.
Fast closing isolation valves are provided to close on LOCA sig- ]
nal or whenever the exhaust radiation sensors detected radiation levels high enough to exceed plant operating limits. During normal reactor operations with the 5,000 cfm 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 during reactor operation. ]
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 the upper containment pool surface so early detection of potential radioactivity can be detected during refueling when the reactor is open.
O 19.3.6.28-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 r^s
()
19.3.6.28 QUESTION / RESPONSE 6.28 (480.23) (Continued)
The suppression pool cleanup system ensures that radioactivity in the pool water can be kept low and reduces the amount of airborne radioactivity during abnormal plant operating events.'
~
This same topic is also an issue contained in NUREG-0737, Item III.E.4.2, which is addressed in Section 1A.29 " Containment Isolation Dependability." Studies of suppression pool scrubbing action have been found to retain considerable amounts of iodine in the event of a major pipe break and failure of fuel cladding. This same capability also applies to normal and abnormal reactor operations when potential
~
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 discussed with the staff during the various reviews and e-~ _
( )s licensing of GESSAR II.
Periodic isolation valve testing is required during reactor 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.
i 19.3.6.28-3
)
GESEAR II 22A7007 238 NUCLEAR ISLAND R v. 15 19.3.6.28 QUESTION / RESPONSE 6.28 (480.23) (Continued)
Studies have been made to evaluate possible containment ventilation isolation during each day. The maintenance functions were planned such that each of two shifts would schedule their work back-to-back. At the end of the second shift work, the containment could be isolated for a very short period because the containment would then require purging at a 25,000 cpm flow rate through the large 42-in.
valves to ensure that the environment would be habitable for the following two work shifts. The small amount of time that containment isolation could be achieved was not cost beneficial for good plant operation and maintenance.
~
In order to more fully comply with the intent of BTB CSB 6-4, separate penetrations for purging during normal operation will be added. In addition, 9-inch penetrations and valves will be used for these penetrations unless GE can demonstrate to the satisification of the NRC staff that a larger diameter is justified.* Details of the purge system will be provided before the first Applicant references GESSAR II 4
According to BTP CSB 6-4, the size of the purge lines should be about 8-inches in diameter for PWR plants and this line size may be overly conservative from a radiological viewpoint for the Mark III BWR plant; therefore, larger lines may be justified.
19.3.6.28-4 Y
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 17 I i 19.3.15 Chapter 15 - Responses V
19.3.15.1 QUESTION / RESPONSE 15.1 (440.3)
QUESTION 15.1 Address each item identified in Item 1 of Table 15-4 of Regulatory Guide 1.70, Revision 3, or indicate an interface to provide the information. (15.6.5)
RESPONSE 15.1 la. Hydrogen Purge Analysis As noted in Subsection 6.2.5, redundant Class lE hydrogen 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 ,
detonable level; therefore, there will be no need to purge the containment and there will be no additional dose contribution from this source.
lb. Equipment Leakage Contribution to LOCA Dose l
The potential dose contribution from this source is deter-i mined in a manner consistent with RG 1.3 and SRP 15.6.5 unless otherwise noted.
(1) Fission Product Source Term. Appendix B to SRP Section l 15.6.5 suggests 50% of the iodine contained in the core at shutdown is released 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 suggests that 50% of the iodine remains in the core. Since there is no question that the core cannot initially 19.3.15.1-1
/
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 4 19.3.15.1 QUESTION / RESPONSE 15.1 (440.3) (Continued) contain greater than 100% of its generated iodine activity and since there is no argument that TO% of the iodine remains in the core for the type of LOCA event evaluated, there can only be at most 50% of the core iodine activity released to the containment. Since the radiological analysis has assumed 25% of the core iodine inventory is airborne, there only remains, at most, 25% of the core inventory to be transported to the suppression pool. Therefore, the radiological analysis presented in this response assumes that 25% of the core iodine inventory is contained within the sup-pression pool. The iodine activity in the suppression pool as a function of time is shown in Table 19.3.15.1-1.
(2) Maximum Operational Leak Rate. The maximum operational leak rate through the HPCS, LPCS or RHR pump seals is 20 cc/Hr (%10-" gpm).
(3) Maximum Leakage Assuming Seal Failure. The ECCS is capable of withstanding the passive failure of valve stem packings and pump seals following a LOCA. The max-imum leakage due to a failure of this nature would be 23 gpm from seal failure of a HPCS, LPCS or RHR pump seal. Valve stem leakage would be significantly less than this amount. Based upon a leak of 23 gpm the leak-age would be detected within 35 minutes by the leakage detection system. During this time period a maximum of 805 gallons of suppression pool fluid would be discharged to secondary containment. Assuming an additional 10 min- ,
utes for operator action, the maximum amount of fluid discharged through the assumed seal failure is 1035 gallons.
i 19.3.15.1-2