ML20035A326
| ML20035A326 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 03/17/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9303250149 | |
| Download: ML20035A326 (39) | |
Text
f GENuclear Energy
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,rn cs nu sv u w;s March 17,1993 Docket No. STN 52-001 Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submitta! Supporting Accelerated AllWR Review Schedule - Resolution of DFSER Outstanding Items for Chapters 3,4,6,10,14 and 20
Dear Chet:
Enclosed are SSAR markups addressing the following DFSER outstanding items:
O_yn items Confirmatory items COL Action Items 6.2.4.1-1 (1/22/93*)
6.5.1-1 (Amendment 24*)
3.93.2-1 6.5.1-2 (1/22/93')
7.7.1.5-1 3.933-1 203-2 7.1.4-1 203-3 10 3.1-1 203-4 14.133.7.2-1 14.133.4.2-1 14.133.2.1-1 14.133.9.2-1 14.133.6.4-2 14.133.6.4-1 14.133.43-1 14.133.5.10-1
- Amplification of a previous markup or amendment Jrs60 9303250149 9303 U FDR ADOCK 052 1
-A
s Chet Posiusny, Jr.
Page 2 March 17,1993
)
i In addition to the above, GE recommends that the first two sentences of the fifth paragraph of DFSER Page 14-18 be replaced with the following-i "The effect of pipe support stiffness on the piping response shall be considered in the analytical model. Supports shall be modeled in accordance with the -
l' SSAR. If supports are not modeled as stated in the SSAR, justification will be provided to validate the stiffness values used in the piping model."
t Finally, GE recommends that the following COL Action items be deleted:
Item Rationale
}
14.1.3.3.6.12-1 COL Action item 14.1.3.3.2.1-1 embraces the Staff-endorsed version of NF incorporating N-690.
14.1.3.3.5.1-1 SSAR will not include an option for the COL applicant to generate site-specific amplified building response spectra.
i Sincerely, I
-i Y
ek Fox l
Advanced Reactor Programs 6
cc: Norman Fletcher (DOE)
Bernie Genetti(GE) i Maryann Herzog (GE) j e
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ABWR Ico'- 3 9 3. 2_]
23smoosc Standard Plant mc 1A.2.8 Rule Making Proceeding or Degraded Core Accidents [II.B.8)
Response to this TMI action plan item is ad-
_/
dressed in Appendix 19A.
IA.2.9 Coolant System Valves-Testing Requirements [II.D.1]
NRC Position Pressurized water reactor and boiling water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for IA.2.10 Relief and Safety Valve Position design-basis transients and accidents.
Indication [II.D3]
Response
IGC Position The ABWR safety / relief valve (SRV) is postu-Reactor coolant system relief and safety valves lated to discharge steam only, not liquid or two phase shall be provided with a positive indication in the flow under expected operating conditions for design-control room derived from a reliable valve position basis transients and accidents.
detection device or a reliable indication of flow in the discharge pipe.
A generic test program was conducted through the BWR Outers Group (Reference 7) to satisfy the
Response
discharge of steam. These steam discharge test results wi'lbe used as the qualification basis for plant The ABWR Standard Plant safety relief valves specific SRV models and discharge piping that are are equipped with linear variable differential trans-sufficiently similar to those reported in Reference B.
formers (LVDT's) which are qualified as Class IE Plant specific SRV models and discharge piping that components. These LVDT's are mounted on the are not similar will be tested in accordance with valve operators and are highly reliable sensors for NUREG-0737 recuirements. Se e SOsccb a monitoringvalve position.
iA.3. s-c cot t e,4 J ~%.
The ABWR system logic for response to high In addition, the deutstream pipe from each valve water level conditions is described in Subsection line is equipped with the mocouples stich signal the 7.3.1.1.1.1(3) and is considered to be sufficiently annunciator and the plant process computer when redundant that the probability of steam line flooding the temperature in the tailpipe exceeds the predeter-by ECCS is extremely low. There is no high drywell mined serpoint.
pressure s:gnal that would inhibit this logic system.
These sensors are shown on Figure 5.1-3 In the ABWR design, each of three RHR shut.
(Nuc1 car Boiler System P&ID).
down cooling lines has its own separate containment penetration and its own separate source of suction 1A.2.11 Systems Reliability [II.EJ.2]
from the reactor vesset Alternate shutdown using the SRV is therefore not required for ABWR in This TMl action plan item superseded by USl order to meet single failure rules. Hence, the A-45. USI A-45 is aMressed in Appendix 19B.
ABWR does not require SRV testing with liquid under low pressure conditions associated with this 1A.2.12 Coordinated Study of Shutdown event as required in past BWRs.
Heat Removal Requirements [II.E33]
This TMl action plan item superseded by USl A-45. USI A-45 is addressed in Appendix 19B.
/
1AM Amenement 16
ABWR-usu=4c Standard Plant nev c 1AJ COLLICENSEINFORMATION changes, proposed or implemented, deemed appropriate, to improve the availability of the emer-TAJ.1 Emergency Procedures and gency core cooling equipment. (See Subsection i
Emergency Procedures Training Program 1A ? ? 9
)
i Emergency procedures, developed from the i A.3. 6 Te 3 of S EtN/ o.4 Dssc,( y j emergency procedures guidelines, shall be provided and implemented prior to fuelloading. (See Subsec-P. p3 tion IA.2.1).
% COL. aU e.o.d wdt co d e e h
i 1 A 3.2 R e n.ew and Modify Procedures for Removing Safety-Related Systems From 4bd o g S R.V5 ov-M sc hc j
4 p s p q g r sh U e-c) th od- - t,r Procedures shall be reviewed and modified (as n ok a m,\\ q < 4 o M % $.a. b a.k h
required) for removin'g safety-related systems from g
g,4g gp Q
(
service (and restormg to service) to assure operabil-ity status is known. (See Subsections 1A.2.18 and 4 c_3 k d t n A c c.o v-cd p c tv wd
[
1 5mbsa-ch.e i A. '2 9 1A33 In-Plant Radiation Monitoring I
q Equipment and training procedures shall be h
provided for accurately determining the airborne io-j dine concentration in areas within the facility where l plant personnel may be present during the accident.
l (See Subsection IA.235).
l 1A3.4 Reporting Failures of Reactor System Relief Valves l
Failures of reactor system relief valves shall be i
reported in the annual report to the NRC. (See Sub-section 1A.23.21.3).
- t 1A3.5 Report on ECCS Outages l
Starting from the date of commercial opera-l tions, an annual report should be submitted which in-cludes instance of emergency core cooling system un-availability because of component failure, mainte..
t nance outage (both forced or planned), or testing, i
the following information shall be collected:
f (1) Outage date (2) Duration of outage l
(3) Cause of outage 3
(4) Emergency core cooling system or i
component invohed (5) Corrective action taken The above information shall be assembled into a 3
report, which will also include a discussion of any l
t 1A3-1
[
Amendment D l
M MTR C
c.o' t o i 3 3 7 z -1
,mm Standard Plant Rrv. n (c)
The assemblies are subjected to a single As a result of piping re-analysis due to pressure test at a pressure not less differences between the design configuration than its design pressure.
and the as-built configuration, the highest stress or cumulative usage factor locations (d)
The assemblies do not prevent the access may be shifted; however, the initially required to conduct the inservice determined intermediate break locations need examination specified in item (7).
not be changed unless one of the following conditions exists:
(7) A 100% volumetric inservice examination of all pipe welds would be conducted during (i) The dynamic effects from the new each inspection interval as defined in (as-built) intermediate break locations IWA-2400, ASME Code,Section XI. See are not mitigated by the original pipe S A5'ch 3.6 5 3 b COVg_5 #%, whip restraints and jet shields.
3.6.2.1.4.3 ASME Code Section Class 1 Piping In Areas Other Than Containment (ii) A change is required in pipe parameters Penetration such as major differences in pipe size, wall thickness, and routing.
With the exception of those portions of piping identified in Subsection 3.6.2.1.4.2, breaks in 3.6.2.1.4.4 ASME Code Section III Class 2 and ASME Code,Section III, Class I piping are 3 Piping in Areas Other Than Containment postulated at the following locations in each Penetr ation piping and branch run:
With the exception of those portions of (a)
At terminal ends' piping indentified in Subsection 3.6.2.1.4.2, breaks in ASME Codes,Section III, Class 2 and 3 (b)
At intermediate locations where the piping are postulated at the following locations maximum stress range as calculated by in those portions of each piping and branch run:
Eq. (10) exceeds 2.4 Sm.
(a) At terminal ends (see Subsection if the calculated maximum stress range 3.6.2.1.4.3, Paragraph (a))
l of Eq.(10) exceeds 2.4 Sm, the stress rang: calculated by both Eq.(12) and (b) At intermediate locations selected by one of Eq.(13) in Paragraph NB-3653 should meet the following criteria:
the limit of 2.4 Sm.
(i) At each pipe fitting (e.g., elbow, tee, (c)
At intermediate locations where the cross, flange, and nonstandard cumulative usage factor exceeds 0.1.
fitting), welded attachment, and valve. Where the piping contains no fittings, welded attachments, or Extremities of piping runs that connect to valves, at one location at each extreme structures, components (e.g., vessels, pumps, of the piping run adjacent to the valves), or pipe anchors that act as rigid protective structure.
constraints to piping motion and thermal expansion. A branch connection to a main (ii) At each location where stresses calcu-piping run is a terminal end of the branch lated (see Subsection 3.6.2.1.4.2, run, except where the branch run is classified Paragraph (1)(d)) by the sum of Eqs.
as part of a main run in the stress analysis (9) and (10) in NC/ND-3653, ASME Code, and is shown to have a significant effect on Section III, exceed 0.8 times the sum the main run behavior. In piping runs which of the stress limits given in NC/ND-are maintained pressurized during normal plant 3653.
conditions for only a portion of the run (i.e., up to the first normally closed valve)
As a result of piping re-analysis due a terminal end of such runs is the piping to differences between the design connection to this closed valve.
configuration and the as-built configuration, the highest stress 3M Amendment 23 i
l
H.1 3. 3.h,2 - 1 ABWR c o t-Standard Plant m,n
=-
3.6 leak-Before-Btrak Analysis Report As required by Reference 1, and LBB analysis report shall be prepared for the piping systems proposed for exclusion from analysis for the dynamic effects due to failure of piping failure. The report shall be prepared in accrodance with the guidelines presented in Appendix 3E and Submitted by the COL applicant to l the NRC for approval. (See Subsection 3.63).
AN 3.6.
References
[
- 1. Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Rupture, Federal Recister. Volume 52, No. 207, Rules and Regulations, Pages 41288 to 41295, October 27, 1987
- 2. RELAP 3, A Computer Program for Reactor Blowdown Analysis, IN-1321, issued June 1970, Reactor Technolcev TID-4500.
- 3. ANSI /ANS-58.2, Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture.
- 4. Standard Review Plan; Public Comments Solicited, Federal Recister. Volume 52, No.
167, Notices, Pages 32626 to 32633, August 28,1987.
- 5. NUREG-1061, Volume 3, Evaluation of Potential for Pipe Breaks, Report of the U.S. NRC Piping Review Committee, November 1984.
- 6. Mehta, H. S., Patel, N.T. and Ranganath, S.,
Application of the Leak-Before-Break Approach to BWR Piping, Report NP-4991, Electric Power Research Institute, Palo Alto, CA, December 1986.
t 3CI Amendment ll3
t t
f COL Action Item No. 14.1.3.3.7.2-1
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ATTACEMENT A
3.6.5.3 Inservice Inspection of Piping in Containment Penetration Areas The COL Applicant shall perform a 100 percent volumetric examination of circumferential and longitudinal pipe welds for those portions of piping within the break-exclusion region. The examination shall be performed in accordance with the requirements specified in Subsection 3.6.2.1.4.2(7).
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21A61mAE MN Co L M 1 - 3 3 A. 7 - 1 nry n Standard Plant excitation in each of the three axes is considered to act simultaneously. The excitations are combined by the SRSS method.
3.73.8.1.4 Rexible Subsystems if the piping subsystem has more than two supports, it cannot be considered a rigid body and must be modeled as a multi-degree-of. freedom subsystem.
The subsystem is modeled as discussed in Subsection 3.7.3.3.1 in sufficient detail (i.e.,
number of mass points) to ensure that the lowest natural frequency between mass points is greater than 33 Hz. The mathematical modelis analyzed using a time-history analysis technique or a response spectrum analysis approach. After the natural frequencies of the subsystem are obtained, a stress analysis is performed using the inertia forces and equivalent static loads obtained from the dynamic analysis for each mode.
1 In a response spectrum dynamic analysis, modal 3.73.8.1.5 Static Analysis responses are combined as described in Subsection 3.7.3.7. In a response spectrum or time-history A static analysis is performed in lieu of a dynamic analysis, responses due to the three dynamic analysis by applying the following orthogonal components of seismic excitation are forces at the concentrated muss locations combined as described in Subsection 3.7.3.6.
(nodes) of the analytical model of the piping
'I"*"
9 S e.e Sob s_n_ Ao r 3 7. C. t W' I"
-So v-C.0 L.
(1) horizontal static load, Fh" h
hcw ufovaksow, of the horizontal principal directions; (2) equal static load, F, in the other h
horizontal principal direction; and (3) verticalstaticload,F = C W; y
y where C,C
= multipliers of the gravity h y acceleration, g, determined from the horizontal and ver-tical floor response spectrum curves, respectively. (They are functions of the period and the appropriate damping of the piping system); and i
weight at node points of the W
=
analytical model.
J 3.7-20 Amendment 23
C ' 19-l-b eJ w
' MWR Standard Plant nty a 1
(1) ASME Code Case N-411-1 damping is not used.
excitation are combined as described in Subsection 3.7.3.6.
t (2) A support group is defined by supports which have the same time-history input. This 9 See SOM WS37.61 l
usually means all supports located on the Qev COL.It %
j same floor, or portions of a floor, of a J,.,,how, structure.
(3) The responses due to motions of supports in two or more different groups are combined by the SRSS procedure.
In lieu of the response spectrum analysis, the time history method of analysis subjected to distinct support motions may be used for multi-supported systems.
3.73.8.2 NSSS Piping Subsystems 3.73.8.2.1 Dynamic Analysis As described in Subsection 3.7.3.3.1, pipe line is idealized as a mathematical model consisting of lumped masses connected by clastic members. The stiffness matrix for the piping subsystem is determined using the clastic properties of the pipe. This includes the effects of torsional, bending, shear, and axial deformations as well as changes in stiffness due to curved members.
Next, the mode shapes and the undamped natural frequencies are obtained. The dynamic response of the subsystem is usually calculated by using the response spectrum method of analysis. When the connected equipment is supported at more than two points located at different elevations in the building, the response spectrum analysis is performed using the envelope response spectrum of all attachment points. Alternatively, the multiple excitation analysis methods may be used where acceleration time histories or response spectra are applied at all the equipment and piping attachment points.
In a response spectrum dynamic analysis, modal responses are combined as described in Subsection 3.7.3.7.
In a independent support motion response spectrum analysis, group responses are combined as described in Subsection 3.7.3.8.1.10.
In response spectrum or time-history dynamic analysis, responses due to the three irthogonal components of seismic 3.7-22.2 Amendment 23
ABWR cou 14.'3 3. 4 2_
i nw=u Standard Plant uv a the seismic switch location.
3.7.4.5 in-Service Surveillance The peak acceleration level experienced by :he Each of the seismic instruments will be reactor building basemat is available immediately demonstrated operable by the performance of the following the earthquake. This is obtained by channel check, channel calibration, and channel i
playing back the recorded THA data from the functional test operations at the intervals basemat location and reading the peak value from specified in Table 3.7 9.
a strip chart recorder.
3.7.5 Seismic Parameters Significant response spectra from the reactor building basemat are available immediately The design basis horizontal g value is 0.3g following an earthquake for comparison with the for SSE. This is the maximum free. field ground OBE and SSE response spectra.
acceleration at the site as measured at the existing grade level near the ABWR. The 3.7.4.4 Comparison of Measured and Predicted response spectra are presented in Subsection Responses 3.7.1. The range of site parameters used to establish the design basis seismic parameters is Initial determination of the carthquake level presented in Appendix 3A. _
comparing the measured response spectra from the 3.7.( References
/ gMgd I
is performed immediately after the earthquake by reactor building basemat with the OBE and SSE 7
response spectra for the corresponding location.
- 1. General Electric Company BWR/6-238 Standard If the measured spectra exceed the OBE response Safety Analysis Report (GESSAR), Docket Nu.
spectra, the plant is shut down and a detailed STN 50-447, November 7,1975.
analysis of the earthquake motion is undertaken.
- 2. E. H. Vanmarcke and C. A. Cornell, Scismic After any earthquake, the data frotu all Risk and Design Response Spectra, ASCE seismic recorders and recording instruments are Specialty Conference on Safety and retrieved. When the OBE has been exceeded, the Reliability of Metal Structures, Pittsburgh, data from these instruments are analyzed to Pennsylvania, November 1972.
obtain the seismic accelerations experienced at the location of major Seismic Category 1
- 3. NUREG-0800, Standard Review Plan, Section structures and equipment. The measured response 3.7.1.
from the time-history accelerographs, peak-recording accelerographs, and response spectrum
- 4. L. K. Liu, Seismic Analysis of the Boiling recorders are used to determine the response Water Reactor, symposium on seismic analysis spectra at the location of each Seismic Category of pressure vessel and piping components, I structure and system. These spectra are First National Congress on Pressure Vessel compared with those used in the design to and Piping, San Francisco, California, May determine whether the structure or system is 1971.
still adequate for future use. Peak recording accelerographs mounted on equipment are used to determine whether the design limitation of that specific equipment has been exceeded.
The theoretical structural response and mea-sured structural responses are compared to assess the degree of conservatism in the analytical pre-dictions. Seismic levels are established to de-2 termine whether the plant can be brought back on line. The criteria consider system design and dynamic analysis in establishing the acceptable levels for continued operation.
Amendment 23 3 %26
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C O L.
1 A.1. '3. 3 4. '2.
I 3,7. 6 c o t.
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l 3.7. G. 4 Piping Analysis, Modeling of Piping Supports The COL fpplicant shall provide justification for methods used other than those described in Subsection 3.7.3.3.1.6 for determining pipe support stiffnesses used in the piping analysis.
l The justification should include verification that the pipe support stiffness values are representative of the types of supports used in the piping system. The alternative approach used i
to determine pipe support stiffness values and its bases should be submitted to the NRC staff for review and approval before its use.
( 3 ee_ sA. ch. -
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. ABWR 23MtmE RfN B Standard Plani The results of the data analyses, vibration COL license information requirements.
amplitudes, natural frequencies, and mode shapes are then compared to those obtained from the Thermal stratification of fluids in a piping theoretical analysis.
system is one of the specific operating conditions that is included in the loads and Such comparisons provide the analysts with load combinations that are contained in the added insight into the dynamic behavior of the piping design specifications and design reactor internals. The additional knowledge reports. It is known stratification can occur gained from previous vibration tests has been in the feedwater piping during plant startup and utilized in the generation of the dynamic models when the plant is in hot standby conditions k seismic and loss of coolant accident (LOCA) following scram (see Subsection 3.9.2.1.3). If, i
analyses for this plant. The models used for during design or startup, evidence of thermal this plant are similar to those used for the stratification is detected in any other piping vibration analysis of earlier prototype BWR system, then stratification will be evaluated.
t If it cannot be shown that the stresses in the plants.
pipe are low and that movement due to bowing is 3.9.3 ASME Code Class 1,2, and 3 acceptable, then stratification will be treated Components, Component Supports, and as a design load. In general, if temperature Core Support Structures differences between the top and bottom of the pipe are less than 50 F, it may be assumed 3.93.1 Loading Combinations, Design design specification and stress reports need not l
Transients, and Stress Limits m u.ias,s.m.. }be revised to include stratification. he S M (::kh
/ '5. 3 7.1 o 4 cot _ 6m J.
L.
This section delineates the criteria for The design life for the ABWR Standard Plant selection and defiuition of design limits and is 60 years. A 60 year design life is a loading combination associated with normal requirement for all major plant components with 4
i operation, postv'ated accidents, and specified reasonable expectation of meeting this design seismic and other reactor building vibration life. However, all plant operational components j
(kBV) events for the design of safety.related and equipment except the reactor vessel are ASME Code components (except containment designed to be replaceable, design life not
]
components which are discussed ic Section 3.8).
withstanding. The design life requirement allows for refurbishment and repair, as i
This section discusses the ASME Class 1,2, appropriate, to assure the design life of the and 3 equipment and associated pressure retaining overall plant is achieved. In effect, parts and identifies the applicable loadings, essentially all piping systems, components and ll calculation methods, calculated stresses, and equipment are designed for a 60 year design j
allowable stresses. A discussion of major life. Many of these components are classified j
equipment is included on a component.by-component as ASME Class 2 or 3 or Quality Group D. In the basis to provide examples. Design transients and event any non-Class I components are subjected dynamic loading for / m Class 1, 2, and 3 to cyclic loadings, including operating equipment are covere
.obsection 3.9.1.1.
vibration loads and thermal transient effects, l
Seismi:-related loads,a > mamic analyses are of a magnitude and/or duration so severe that discussed in Section.$.7. The suppression the 60 year design life can be assured by pool-related RBV loads are described in Appendix required Code calculations. COL applicants will 3B. Table 3.9-2 presents the combinations of identify these components and either provide an dynamic events to be considered for the design appropriate analysis to demonstrate the required and analysis of all ABWR ASME Code Class 1,2, design life or provide designs to mitigate the and 3 components, component supports, core magnitude or duration of the cyclic loads.
support structures and equipment. Specific Components excladed from this requirement are loading combinations considered for evaluation of (1) tees where mixing of hot and cold fluids each specific equipment are derived from Table occurs and thern.al sleeves have been provided in 3.9-2 and are eontained in the design accordance with the P& ids, (2) cornponents, such specifications and/or design reports of the as the quencher, for which a fatigue analysis i
respective equipment. See Subsection 3.9.7.4 for has already been performed, providing the com-g 3
,19. L 5, 3.9. -r. c d
- 3. 9.L g 3.9-18 Amendment 23 c o L.1 3.1. b 3 2,. I t i
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udyr M Me dc grg 4WH of con @
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Standard Plant GeChor loon }O0d1 2 n C CCordance ugm j
,1 3.9.3.4 Component Supports [ EW" M NF-323f. The critical buckling loads for the f
N 5*^- E-Class 1 piping supports subjected to faulted The design of bolts for component supports loads that are more severe than normal, upset i
is specified in tbc ASME Code Section III, and emergency loads, are determined by using Subsection NF. Stress limits for bolts are given the methods discussed in Appendices F and XVII in NF-3225. The rules and stress limits which of the Code. To avoid buckling in the piping l
must be satisfied are those given in NF 3324.6 supports, the allowable loads are limited to multiplied by the appropriate stress limit factor two thirds of the determined critical buckling for the particular service loading level and loads.
stress category specified in Table NF-3225.2-1.
Maximum calculated static and dynamic Moreover, on equipment which is to be, or def!cetions at support locations are checked may be, mounted on a concrete support, sufficient to confirm that the support has not rotated holes for anchor bolts are provided to limit the beyond the vendor's recommended cone of action i
anchor bolt stress to less than 10,000 psi on the or the recommended arc of loading.
.3 nominal bolt area in shear or tension.
y-g Suppoprts for ASME Code Section III Concrete anchor bolts (including under-cut instrumentation lines are designed and type anchor bolts) which are used for pipe analyzed in accordance with ASME Codefection g
e
't support base plates will be designed to the III; Sa*osec. fica NE U
Aapplicable factors of safety which are defined in h{
IAE Bulletin 79-02," Pipe Support Base Plate The design of all supports for non-nuclear y
Designs Using Concrete Expansion Anchor Bolts,"
piping satisfies the requirements of ANSI \\ASME l Revision 2 dated Ncvember 8,1979.p B31.1 Power Piping Code, Paragraphs 120 and l l
4
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121.
1
'3.9.3.4.1 Piping j
j$
For the major active valves identified in j
j-Supports and their attachments for essential Subsection 3.9.3.2.4, the valve operators are
- W ASME Code Section III, Class 1,2, and 3 piping not used as attachment points for piping M*
are designed in accordance with Subsection NF* up supports.
o to the interface of the building structure, with jurisdictional boundaries as defined by The design criteria and dynamic testing re-Subsection NF. The loading combinations for the quirements for the ASME III piping supports p
various operating conditions correspond to those are as follows:
gg,gggefjen 97 used for design of the supported pipe. The component loading combinations are discussed in (1) Piping Supports - All piping supports are l
Subsection 3.9.3.1. The stress limits are per designed, fabricated, and ass (mbled so i
ASME III, Subsection NF and Appendix F.
that they cannot become disengaged by the Supports are generally designed either by load movement of the supported pipe or equip-l rating method per paragraph NF-3260 or by the ment after they have been installed. All stress limits for linear supports per paragraph piping supports are designed in accordance j
with the rules of Subsection NF of the i
ASME Code up to the building structure i
- Augmented by the following: (1) application of interface as defined by the jurisdictional Code Case N-476, Supplement 89.1 which governs boundaries in Subsection NF.
-l the design of single angle members of ASME Class 1,2,3 and MC linear component supports; and (2)
(2) Spring Hangers - The operating load on when eccentric loads or other torsionalloads are spring hangers is the load caused by dead l
not accommodated by designing the load to act weight. The hangers are calibrated to en-l i
through the shear center or meet " Standard for sure that they support the operating load Steel Support Design", analyses will be performed at both their hot and cold load settings.
in accordance with torsional analysis methods Spring hangers provide a specified down such as:" Torsional Analysis of Steel Members, travel and up travel in excess of the i
USS Steel Manual", Publication T114-2/83.
specified thermal movement. Deflections 1 Amendment 23 3D col-A.1.3 3 6,4-z.
.ABM IMMMAE REV B Standard Plant l
l 3.9.7 COL License Information Subsection 3.9.3.1.)
3.9.7.1 Reactor Internals Vibration Analysis, 3.9.7.3 Pump and Valve Inservice Testing Measurement and Inspection Program Program The first COL applicant will provid.:, at COL applicants will provide a plan for the the time of application, the results of the detailed pump and valve inservice testing and vibration assessment program for the ABWR inspection program. This plan will prototype internals. These results will include the following information specified in Regulatory (1) Include baseline pre-service testing to Guide 1.20.
support the periodic in-service testing of the components required by technical R. G.1 20 Subiect specifications. Provisions are included to disassemble and inspect the pump, check C.2.1 Vibration Analysis valves, and MOVs within the Code and Program safety-related classification as necessary, C.2.2 Vibration Measurement depcnding on test results. (See Program Subseetions 3.9.6, 3.9.6.1, 3.9.6.2.1 and C.2.3 Inspection Program 3.9.6.2.2)
C.2.4 Documentation of Results (2) Provide a study to determine the optimal frequency for valve stroking during NRC review and approval of the above inservice testing. (See Subsection information on the first COL applicant's docket 3.9.6.2.2) will complete the vibration assessment program requirements for prototype reactor internals.
(3) Address the concerns and issues identified in Generic Letter 89-10; specifically the in addition to the information tabulated method of assessment of the loads, the above, the first COL applicant will provide the method of sizing the actuators, and the information on the schedules in accorda,cc with setting of the torque and limit switches.
the applicable portions of position C.3 of (See Subsection 3.9.6.2.2)
Regulatory Guide 1.20 for non-prototype internals.
3.9.7A Audit of Design Specification and Design Reports Subsequent COL applicants need only provide the information on the schedules in accordance COL applicants will make available to the i
with the applicable portions of position C.3 of NRC staff design specification and design Regulatory Guide 1.20 for non-prototype reports required by ASME Code for vessels, l internals. (See Subsection 3.9.2.4).
pumps, valves and piping systems for the-l purpose of audit. _{5:: 59::% 0.0.3.1) 3.9.7.2 ASME Class 2 or 3 or Quality Group D e tNSGT A
Components with 60 Year Design Ufe 3.9.8 References I
COL applicants will identify ASME Class 2 1.
BWR fuel Channel Afechanical Design and ggg or 3 or Quality Group D components that are Deflection, NEDE-21354-P, September 1976.
g subjected to cyclic loadings, including operating vibration loads and thermri transients effects, 2.
BWR/6 Fuel Assembly Evaluadon of Combined of a magnitude and/or duration so severe the 60 Safe Shutdown Earthquake (SSE) and year design life can not be assured by required Loss-of-Coolant Accident (LOCA) Loadings, i
Code calculations and, if similar designs have NEDE-21175-P, November 1976.
not already been evaluated, either provide an appropriate analysis to demonstrate the required 3.
NEDE 24057-P (Class III) and NEDE-24057 design life or provide designs to mitigate the (Class I) Assessment of Reactor Internals.
magnitude or duration of the cyclic loads. (See Vibration in BWR/4 and BWR/5 Plants, 3NS Amendment 23 i
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14.1.3.3.2.1-1)
Q Action Item No. -
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TheCOL[pplicantshallensurethatthepipingsystemdesignis consistent with the construction practices, including inspection and examination methods, of the ASME Code edition and addenda as endorsed in 10 CFR 50.55a in effect at the time of application.
l TheCOLIpplicantshallidentifyASMECodeeditionsandaddenda other than those listed in Tables 1.8-21 and 3.2-3, that will be used to design ASME Code Class 1,2 and 3 pressure retaining 3
components and supports. The applicable portions of the ASME Code editions and addenda shall be identified to the NRC staff for review and approval with the COL application.
CTa Sub s cM\\ow 3,9.3, t) t u
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I mssgr 6 3.9.7.5 ASME Class 1,2 and 3 Piping System Clearance Requirements ASME Class 1,2 and 3 piping systems shall be designed to provide i
clearance from structures, systems, and components where necessary l
1 for the accomplishment of the structure, system, or component's safety function as specified i
designdescription.TheCOL1,1}gg(ge4respectivestructureorsystem 4
nsee shall verify that the maximum l
calculated piping system deflections under service conditions do 1
not exceed the minimum clearances between the piping system and nearby structures, systems, or components. The COL licensee shall i
document in the certified design stress report that the clearance requirements have been met.CScu S-b sa dson 1.73.1) e 1
i i
3.9.7.6 As-Built Reconciliation Analysis For j
ASME Class 1,2 and 3 Piping Systems l
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For ASME Class 1,2 and 3 piping systems, the COLAxicansca shall 3
reconcile the as-pg(lt, piping system with the as-designed piping i
system. The COL laccnscc will perform an as-built inspection of the pipe routing, location and orientation, the location, size, clearances and orientation of piping supports, and the location l
l and weight of pipe mounted equipment. This inspection will be performed by reviewing the as-built drawings containing l
Verification stamps, and by performing a visual inspection of the i
installed piping system. The piping configuration and component l
i location, size, and orientation shall be within the tolerances specified in the certified as-built piping Stress Report. The l
1 tolerances to be used for reconciliation of the as-built piping system with the as-designed piping system are provided in the EPRI report, " Guidelines for Piping System Reconciliation (NCIG-05, Revision 1)," NP-5639 dated May 1988. A reconciliation analysis i
using the as-built and as-designed information shall be performed.
The certified as-built Stress Report shall document the results of the as-built reconciliation analysis.( 3% mbs4 cb u,3,G j
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3.9.7.7 Pipe Support Baseplate and Anchor Bolt Design COL fpplicants shall provide justification for the use of safety factors for concrete anchor bolts other than those specified in
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Subsection 3.9.3.4. This justification shall be submitted to the l
NRC stuff for review and approval prior to the installation of the concrete anchor bolts.
(cot _ h. i. 2. s. s.4 -0 COL /pplicants shall account for pipe support base plate flexibility in the calculation of concrete anchor bolt loads in
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accordance with Subsection 3.9.3.4.
G m. i. s. 3.3.Y-i) 3.9.7.8 Pipe-Mounted Equipment Allowable Loads l
The COL gpplicant shall inspect the piping design reports and f
document that the pipe applied loads on attached equipment; such j
as valves, pumps, tanks and heat exchangers, are less than tbe equipment vendor's specified allowable loads. ( Su S wksu%
3.9. 3.1 )
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3.9.7.9 Benchmark Requirements for Computer Codes used to perform Piping Dynamic Analysis The COL,(pplicant shall benchmark their computer code used for i
piping system dynamic analysis against the NRC Benchmark Problems I
for ABWR, defined in Reference 5. The results of the COL applicant's piping dynamic analysis shall be compared with the results of the Benchmark Problems provided in Reference 5. The piping results to be compared and evaluated and the acceptance criteria or range of acceptable values are specified in Reference 5. Any deviations from these values as well as justification for such deviations shall be documented and submitted by the COL applicant to the NRC staff for review and approval before initiating the final certified piping analysis.
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10.3 Main Steam Supply System 103.2.1 cenersi Description i
The function of the main steam supply system The main steam supply system is illustrated in is to convey steam generated in the reactor to the Figure 103-1. The main steam piping consists of four i
turbine plant. This section discusses that portion of 2S-inch diameter lines from the outboard main steam the main steam system which ranges between, but line isolation valves to the main turbine stop vahrs.
does not include, the outermost containment The four main steam lines are connected to a header isolation valves and the turbine stop valves.
upstream of the turbine stop valves to permit testing i
of the main steam line isolation valves during plant The main steam line pressure relief system, operation with a minimum load reduction. This main steam line flow restrictors, main steam line header arrangement is also provided to ensure that isolation valves (MSIVs), and main steam piping the turbine bypass and other main steam supplies are from the reactor nahles through the outboard main connected to operating steam lines and not to idle steam isolation valve (MSIV) are described in Sub-lines. The main steam process downstream of the sections 5.2.2,5.4.4,5.4.5, and 5.4.9 respectively.
turbine stop valves is illustrated in Figure 103-2.
10.3.1 Design Bases The design pressure and temperature of the main steam piping is 1250 psig and 600 F,
10 3.1.1 Safety Design Bases respectively, the same values as the design parameters of the reactor. The main steam lines are classified as The main steam supply system is not required dWud in Section 3.2.
to effect or suppcrt safe shutdown of the reactor or to perform in the operation of reactor safety A drain line is connected to the low points of features, however, the main steam supply system is each main steam line, both inside and outside the i
designed:
containment. Both sets of drains are headered and connected, with isolation valves to allow drainage to 1
l (1) To comply with applicable codes and standards the main condenser. To permit intermittent draining in order to accommodate operational stresses of the steam line low points at low loads, orificed lines such as internal pressure and dynamic loads are provided around the final valve to the main r
without risk of failures and consequential condenser. The steam line drains maintain a continu-releases of radioactivity in excess of the ons douward slope from the steam system low established regulatory limits; points to the orifice located near the condenser. The drain line from the orifice to the condenser also l
(2) To accommodate normal and abnormal emi-slopes doutward. To permit emptying the drain lines j
ronmentalconditions; and for maintenance, drains are provided from the line low points, going to the radwaste system.
l (3) With suitable accesses to permit inservice testing and inspections.
The drains from the steam lines inside contain-ment are connected to the steam lines outside con-i 10 3.1.2 Ponr Generation Design Bases tainment to permit equalizing pressure across the main steam line isolation valves during startup and Power Generation Desien Basis One - The system following a steam line isolation.
is designed to deliver steam from the reactor to the y
turbine-generator system for a range of flows and (103.2.2 Component Description pressures varying from warmup to rated conditions.
It also provides steam to the reheaters, the steam jet The main steam system lines are made of air ejectors, the turbine gland scaling and the carbon steel and are sized for a normal steady state deaerating section of the main condenser and the velocity of 150 feet per second, or less. The lines are l
turbine bypass system.
designed to permit hydrotesting following construc-tion and major repairs without addition of temporary 10.3.2 Description pipe supports.
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welds in locations of restricted direct physical and visual accersibility.
I (a) The performance qualification should require i o.3 /7,2 M 5 LV Wh % e.
testing of the welds when conditions of accessibility to production welds are less than 30 to 35 cm (12-14 inches) in any direction
-Th. col cx p p ks cad w s l l from the joint.
grevscl4 O e. c% o u d o (b)
Requalification is required for Jifferent
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accessibility conditions or when other 4k N S C - ( 3 "
essential variables listed in the Code, WV\\.w Section IX, are changed.
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1 (c) Tbc qualification and requalification tests reouited by (a) and (b) above may be waived pr:rnded that the joint is to be 100% radio-graphed or ultrasonically examined after completion of the weldment. Examination procedures and acceptance standards should mee: the requirements of the ASME Code Section III. Records of the examination reports and radiographs should be retained and made p.-t of the Quality Assurance documentativa af the completed weld.
(2) Regulatory Guide 1.37, Quality Assura.cc Requirements for Cleaning of Fluid Systems and Associated Compon.:nts of Water-Cooled Nuclear Power Plants, describes acceptable procedures for cleaning and handling Class 2 components of the steam and feedwater systems. Vented tanks with deionized or demineralized water are an acceptable source of water for final cleaning or flushing of finished surfaces. The oxygen content r
of the water in these vented tanks need not be controlled.
(3) Acceptance criteria for nondestructive examina-tion of tubular products are given in the ASME Code,Section III. aragraphs NC 2550 chrough 2570.
10.3.7 COL License Information 103.7.1 Procedurrs to Avoid Stum Itammer and Discharge Loads The COL applicant will provide operating and maintenance procedures that include adequate precautions to avoid steam hammer and discharge loads (see Subsection 1033).
1013 Amendment 24
20,3-2.
ABM zmmma Rev c S_tandard Plant 6.7 HIGH PRESSURE NITROGEN GAS bottles. Normally, outlet valves from five of SUPPLY SYSTEM the ten bottles are kept open. Each division has a pressure control valve to depressurize the -
6.7.1 Functions nitrogen gas from the bottles.
/
The high pressure nitrogen gas supply system The bottles are mechanically restrained to is divided into two independent divisions, with preclude generation of high-pressure missiles each division containing a safety-related during an SSE. Tbc oottles are also covered by emergency stored nitrogen supply. The essential a heavy steel plate, which serves as a barrier stored nitrogen supply is Safety Class 3, Seismic to potential missiles.
Category 1, designed for operation of the main steam S/R valve ADS function accumulators.
Flow rate and capacity requirements are divided into an initial requirement and a '
continuous supply. An initial requirement for The function of the nonsafety-related, makeup each ADS SRV provides for actuatioas of the nitrogen gas supply system is:
valve against drywell pressure. Fifty gallon accumulators supplied for each main steam ADS (1) relief function accumulators of main steam SRV actuator fulfill the steam valve S/R valves, requirement. The continuous supply is divided into safety and nonsafety onrrions.
(2) pneumatically operated valves and instruments inside the PCV, Compressed nitrogen at a rate adequate to make up the nitrogen leakage of each serviced (3) leak detection system radiation monitor valve is provided by the safety portion. This calibration assumes an air leakage rate for each valve of I scfh for a period of at least seven days. The (4) ADS function accumulators to compensate for essential system with associated lines, valves the leakage from main steam S/R solenoid and fittings are classified as Safety Class 3, valves during normal operation Scismic Category I.
6.7.2 System Description The nonsafety portion provides compressed nitrogen at a rate adequate to recharge the ADS Normally, nitrogen gas for both the essential SRV accumulators. The nonessential system has and nonessential makeup systems 's supplied from two pressure control valves to depressurize the the nitrogen gas evaporator via the makeup line nitrogen gas from the AC system. One is to to the atmospheric control (AC) system. The depressurize to 200 psi for the SRV accumulators nitrogen supply system shall supply nitrogen and the other is to depressurize to 100 psi for which is oil-free with a moisture content of less other pneumatic uses.
l than 2.5 ppm. This nitrogen is filtered in the HPIN system to remove particles larger than 5 The continuous supply portion of the microns. All equipment using this nitrogen shall pneumatic system, extending from the AC system be capable of operating with nitrogen of the to the isolation velve prior to the essential quality listed above. If nitrogen is not system is not safety related.
available from the AC system to supply the
, essential system, nitrogen is supplied from high Nonsafety piping and valves of the system are pressure nitrogen gas storage bottles. The designed to ANSI B31.1, Power Piping Code, and essential system is separat.:d into two the requirements of Quality Group D of divisions. There are fielines between the R egulatory Guide 1.26. Pressure vessels and nonessential and each division of the essential heat exchangers are designed to ASME Section system. Each ticline has a motor operated VIII, Division I.
shutoff valve. For details, see Figure 6.7-1 and
^-+he Table 6.7-1.
System design pressure e-29 sy ~ M r m " : = _ &
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sb ce 6 7' Q 4-Each division of the essential system has ten w
m C-1 Amendment 24
ABM oT 20,3
'2 ussiooxa REV C Standard Plant Category I, ASME Code III, Class 3, Quality Group isolation valves in order to verify their C and Quality Assurance B requirements, except leaktightness. Operation of valves and for the piping and valves for the containment and associated equipment used to switch from the drywell penetrations which are designed to nonsafety to safety nitrogen supply can be Seismic Category I, ASME Code III, Class 2, tested to assure operational integrity by manual Quality Group B and Quality Assurance B actuation of a switch located in the control requirements.
room and by observation of associated position indication lights. Periodic tests of the check The essential high pressure nitrogen gas valves and accumulators shall b-conducted to supply is separated into two independent assure valve operability. p divisions, with each division capable of supplying 100% of the requirements of the 6.7.5 Instrumentation Requirements division being serviced. Each division is mechanically and electrically separated from the A pressure sensor is provided for the safety other. /The system satisfies the componems' nitrogen supply, and an alarm signals low aitrogen demands during all plant operation nitrogen pressure.
condi: ions (normal through faulted).
A remote manual switch and open closed l
Safety grade portions of the high pressure position lights are provided in the control room l nitrogen gas supply system are capable of being for valve operation and position indication.
1 italated from the nonsafety parts and retaining their function during LOCA and/or seismic events under which any nonsafety parts may be demaged.
O Pipe routing of Division I and Division 2 nitrogen gas is kept separated by enough space so gg/
fjfggJ g that a single fire, equipment dropping accident, strike from a single high energy whipping pipe, gf.g g gg grg jet force from a single broken pipe, internally generated missile or wetting equipment with gg g gpg[ gg g spraying water cannot prevent the other division
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from accomplishing its safety function.
Separation is accomplished by spatial separation j
or by a reinforced concrete barrier, to ensure g g[g f/g[ Mg gg[g separation of each pneumatic air division from any systems and components which belong to the gggg [3 gMg Mg g
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other pneumatic air division.
d Vahte of 27 /&rs[c< A our 6.7.4 Inspect. ion and Testing Requ.irements CI SC hh Y~ 20Ch $3 C Per. dic inservice inspection of components, io..
in accordance with ASME Section XI, to ensure the q //g[,ja/q,
capability and integrity of the system is mandatory. Nitrogen quality shall be tested periodically to assure compliance with ANSI M C11.1.
The nitrogen 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 prmided at the containment 6.7-2 Amendment 13
g GVR 23A6tC0As Standard Plant nev 4 ation of the ADS. The results of this program were
Response
submitted to the NRC in a letter report from D. B.
Waters, Chairman of BWR Owners' Group, to D. G.
The ABWR primary containment is inerted and is, Eisenhut, Director (NRC), dated December 29, therefore, protected from hydrogen combustion re-1980. A summary of this evaluation follows.
gardless of the amount or rate of hydrogen genera-tion. In fact, n. creasing amounts of hydrogen moves The cases analyzed in the letter report above show the primary containment oxygen concentration fur-that, based on core cooling considerations, no signifi-ther from the flammable regime. The ABWR is also cant improvement can be achieved by a slower de-provided with permanently-installed recombiners pressurization rate. A significantly slower depressur-which prevent the buildup of oxygen, due to radiolysis, ization will result in increased core uncovery times from creating a potentially flammable mixture.
before ECCS injection. Furthermore, a moderate Radiolysis is the only potential source of oxygen in the decrease in the depressurization rate necessitates an ABWR primary containment.
earlier action time to initiate ADS. Such an earlier [
actuation time has the negative impact of providing f 19A.2.13 Long-Term Training Upgrade less time for the operator to st et high pressure
[ Item (2) (i)]
ECCS without obtaining a significant benefit to vessel fatigue usage. This earlier actuation time nc-NRC Position cessitates a higher initiation level which would result in an increased frequency of ADS actuation.
Provide simulator capability that correctly models the control room and includes the capability to simu-It should be noted that the ADS is not a normal late small-break LOCA's. (Applicable to construction core cooling system, but is a backup for the high permit applicants only.) [LA.4.2]
pressure core cooling systems such as feedwater, RCIC or HPCF. If ADS operation is required,it is
Response
because normal and/or emergency core cooling is threatened. As a full ADS blowdown is well within COL license information, see Subsection 19A3.1.
l the design basis of the RPV and the system is prop-erly designed to minimize th: threat to core cooling, 19 A.2.14 Long-Term Program of no change in depressurization rate is required or ap-Upgrading of Procedures [ Item (2) (ii) prop 6 ate.
NRC Position
~
19 A.2.12 Evaluation of Alternative Hydrogen Control Systems [ Item (1) (xii)]
Establish a program, to begin during construction and follow into operation, for integrating and expand-NRC Fosition ing current efforts to improve plant procedures. The scope of the program shallinclude emergency proce-dures, reliability analyses, human factors engineering, Perform an evaluation of alternative hydrogen n-trol systems that would satisfy the requirement of crisis management, operator training, and coordina-paragraph (f) (2) (ix) of 10CFR5034(f). As a ini-tion with INPO and other industry efforts. (Applica-mum include consideration of a hydrogen ig tion ble to construction permit applicants only.) [I.C.9]
and post-accident inerting system. The eva ation shallinclude:
Response
(A) A comparison of costs and bencfits the alter-COL license information, see Subsection 19A3.2.
l native systems considered.
19A.2.15 Control Room Design Reviews (B) For the selected system, analys and test data (Item (2) (iii)]
to verify compliance with th equirements of (f) (2)(ix) of 10CFR5034 NRC Position (C) For the selected syste preliminary design de-Provide, for Commission review, a control room scriptions of eq nt, function, and layout.
design that reflects state-of-the-art human factor prin-q) See S d seelhon G. 2. 7 1 or COL ts W H Amendment 23
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CA/)~m 9,7,j,y l ABWR Design occument 2.2.11 Process Computer System Design Description f
The Process Computer System (PCS) is classified as a non-safety-related system and has no functional safety design basis; however, it is designed so that the l
functional capabilities of safety-related systems are not affected by it. Input data i
for the PCS are derived from both Class 1E and nonClass 1 E sources. Divisio division and safety you non-safety PCS interfaces are made up of fiber optic l
l cables, which act as optical isolators for electrical separation.
The purpose of the PCS is to:
perform the functions and calculaticrx for the evaluation of nuclear (1) power plant operation; provide the capability for supenisory control of the plant by supplying (2) setpoir.t commands to independent automatic control systems as changing load demands and plant condit;ons dictate, j
^
(3) provide a permanent record and historical perspective for plant operating activities and abnormal events; t
provide analysis, evaluation and recommendation capabilities for start-(4) up, normal operation, safe plant shutdown and abnormal operating and emergency conditions; provide capability to monitor plant performance through presentation (5) of video displays in the main control room and elsewhere throughout the plant; provide the ability to directly control certain non-safety-related plant (6) equipment through onsreen technology and provide a plant simulator for training and for development and ana (7) of operation >.1 techniques.
tions performed by the PCS include process validation and conMon, nuclear system supply performance calculations and balanceef-plant performance calculations.
c#fN Failures of process input signals are isolated and identified by the process 7<74f-/,
computer.
.lme er A,-+The power for the PCS is supplied from a redundant, constant vo frequency uninterruptible, non-Class 1E power supply.
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,A ABWR D:sion 0: cum:nt The PCS consists of two subsystems, Performance Monitoring and Control I
(PMC) and Power Generation Control (PGC).
]
i Performance Monitoring and Control System j
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- i i
PMC)(provides nuclear steam supply (NSS) performance and prediction calculations, video display control, point log and alarm processing and balance i
of plant (110P) performance calculations.
l The NSS Performance module of PMC takes reactor and in-core data from a plant data acquisition system and procedures current state and predicted core performance information. NSS performance calculations are done to provide three-dimensional simulationg{Li r.fer-w6b ytsfn w a.
Power Generation Control System i
PGC monitors the overall plant conditions, exercises the algorithms for the automated control sequences associated with plant power range operation and j
issues reactor command signals to the automatic power regulator (APR) system, j
which implements them.
In the event that conditions which are not expected during normal plant operations occur in the plant or in the PGC, PGC reverts to the manual mode.
j Inspections, Tests, Analyses and Acceptance Criteria
+
Table 2.2.11 provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the PCS.
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Table 2.2.1[ Process Computer System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Test, Analysis Acceptance Criteria 1.
The Process Computer System identifies 1.
Tests will be performed by simulating input
- 1. The Process Computer System output and Isolates failure of nput signal failures to the Process Computer signal is based upon the remaining valid signals.
System.
input signals and failed inputs will be o
identified.
1 he Process Comptte(3 ils powered.
2.
Atost shall be ormed by simulating 2.
Th e is nolo s of pr so puter ctioptiuri
'the. loss of fall Jre oi each nt/cpstedl f
n power y re undantt' cope ant atge opet -
volt age.onstant s py.
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- 4. In the event that conditions which are not
- k. The system shall be tested by:
is Following the simulated failure, the PGC expected during plant operations develop L decouples and plant operation reverts to L
a.
Simulating communications failure manual mode.
in the plant or trouble occurs in the Process Computer Syrtem, PGC automatically with one low level controller.
decouples from the plant control circuits and plant operation reverts to the manual b.
Simulating failure within PCS resulting mode.
In a loss of a PCS function.
c.
Simulating a plant transient.
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.A*c?pe nen critoria_,
e f 3 '4, The Process Computer sheE log both the
. The system shall be tested by:
7 A The process computer logs the tested tripped and reset conditions of the RPS-
- a. Simulating trips and resets of the RPS-conditions and the logs contain the specific t
related sensor instrument channels and the related sensor instrumerA channels.
trip variable, the divisional channel identity RPS automatic or manua; conditions of the and the specitic automatic or manual trip i
RPS-related sensor instrument channels
- b. Simulating RPS automatic or manual system for all conditions simulated.
and the RPS automatic or manualtrip conditions of the RPS-related sensor systems. The computer shallidentify the instrument channels and the RPS I
specific trip variable, divisional channel automatic or manualtrip systems.
t l
identity and specific automatic or manual I
trip system for all conditions that cause a reactor trip.
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s COL. 10. 3. l - t)
Standard Plant net 4 10.3 Main Steam Supply System 102.2.1 Generai Description The function of the main steam supply system The main steam supply system is illustrated in is to convey steam generated in the reactor to the Figure 103-1. The main steam piping consins of four turbine plant. This section discusses that portion of 28-inch diameter lines from the outboard main steam the main steam system which ranges between, but line isolation valves to the main turbine stop valves.
does not include, the outermost containment The four main steam lines are connected to a header isolation valves and the turbine stop valves.
upstream of the turbine stop valves to permit testing of the main steam line isolation valves during plant The main steam line pressure relief system, operation with a minimum load reduction. This main steam line flow restrictors, main steam line header arrangement is also provided to ensure that i
isolation valves (MSIVs), and main steam piping the turbine bypass and other main steam supplies are from the reactor nozzles through the outboard main connected to operating steam lines and not to idle steam isolation valve (MSIV) are described in Sub-lines. The main steam process downstream of the sections 5.2.2,5.4.4,5.4.5, and 5.4.9 respectively.
turbine stop valves is illustrated in Figure 103-2.
10 3.1 Design Bases The design pressure and temperature of the main s: cam piping is 1250 psig and 600 F,
103.1.1 Safety Design Bases respectzvely, the same values as the design parameters of the reactor. The main steam lines are classified as The main steam supply system is not required discussed in Secion 3.2.
to effect or support safe shutdown of the reactor or to perform in the operation of reactor safety A drain line is connected to the low points of features, however, the main steam supply system is each main steam line, both inside and outside the designed:
containment. Both sets of drains are headered and connected, with isolation valves to allow drainage to l (1) To comply with applicable codes and standards the main condenser. To permit intermittent draining in order to accommodate operational stresses of the steam line low points at low loads, orificed lines such as internal pressure and dynamic loads are provided around the final valve to the main without risk of failures and consequential condenser. The steam line drains maintain a continu-releases of radioactivity in excess of the ous downward slope from the steam system loiv established regulatorylimits; points to the orifice located near the condenser. The drain line from the orifice to the condenser also l (2) To accommodate normal and abnormal envi-slopes downward. To permit emptying the drain lines ronmentalconditions; and for maintenance, drains are provided from the line low points, going to the radwaste system.
l (3) With suitable accesses to permit inservice testing and inspections.
The drains from the steam lines inside contain-ment are connected to the steam lines outside con-103.1.2 Power Generation Design Bases tainment to permit equalizing pressure across the main steam line isolation valves during startup and Power Generation Desien Basis One - The sptem following a steam line isolation.
is designed to deliver steam frcm the reactor to the turbine-generator system for a range of flows and 103.2.2 Component Description pressures varying from warmup to rated conditions.
It also provides steam to the reheaters, the steam jet The main steam system lines are made of air ejectors, the turbine gland sealing and the carbon steel and are sized for a normal steady state deaerating secticn of the main condenser and the velocity of 150 feet per second, or less. The lines are turbine bypass system.
designed to permit hydrotesting following construc-tion and major repairs without addition of temporary 103.2 Description pipe supports.
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- C o t-to. 3.1 - i Standard Plant RmA welds in locations of restricted direct phpical and visual accessibility.
(a) The performance qualification should require i o,3,7,2 ra 3 iy L,. b g e, J
testing of the welds when conditions of accessibility to production welds are less than 30 to 35 cm (12-14 inches) in any direction nt COL a fI h c.c wil {
e from the jomt.
Pv ovs h O e. o. m o u d o I (b)
Requalification is required for different
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accessibility conditions or when other g k k N E-C - ( 3 4
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(c) The qualification and requalification tests required by (a) and (b) above may be waived provided that the joint is to be 100% radio-graphed or ultrasonically cramined after completion of the weldment. Examination procedures and acceptance standards should meet the requirements of the ASME Code Section III. Records of the examination reports and radiographs should be retained and made part of the Quality Assurance documentation of the completed weld.
(2) Regulatory Guide 1.37, Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants, describes acceptable procedures for cleaning and handling Class 2 components of the steam and feedwater systems. Vented tanks with deionized or demineralized water are an acceptable source of water for final cleaning or flushing of finished surfaces. The oxygen content of the water in these vented tanks need not be controlled.
(3) Acceptance criteria for nondestructive examina-tion of tubular products are given in the ASME Code,Section III, Paragraphs NC 2550 through 2570.
10.3.7 COL License Information 103.7.1 Procedures to Avo!J Steam llammer and Discharge leads The COL applicant will provide operating and maintenance procedures that include adequate precautions to avoid steam hammer and discharge loads (see Subsection 1033).
1013 Amendment 24
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ABM usuooxo Standard Plan?
uv c 6.7 HIGH PRESSURE NITRO GAS bottles. Normally, outlet valves from five of SUPPLY SYSTEM the ten bottles are kept open. Each division has a pressure control valve to depressurize the 6.7.1 Functions nitrogen gas from the bottles.
The high pressure nitrogen gas supply system The bottles are mechanically restrained to is divided into two independent divisions, with preclude generation of high pressure missiles each division containing a safety-related during an SSE. The bottles are also covered by emergency stored nitrogen supply. The essential a heavy steel plate, which serves as a barrier stored nitrogen supply is Safety Class 3, Seismic to potential missiles.
Category I, designed for operation of the main steam S/R valve ADS function accumulators.
Flow rate and capacity requirements are divided into an initial require nent and a continuous supply. An initial requirement for The function of the nonsafety-related, makeup each ADS SRV provides for actuations of the nitrogen gas supply system is:
valve against drywell pressure. Fifty gallon accumulators supplied for each main steam ADS (1) relief function accumulators of main steam SRV actuator fulfill the steam valve S/R valves, requirement. The continuous supply is divided into safety and nonsafety portions.
(2) pneumatically operated valves and instruments inside the PCV, Compressed nitrogen at a rate ah nate to make up the nitrogen leakage of eac:
c r viced (3) leak detection system radiation monitor valve is provided by the safety portion. This calibration assumes an air leakage rate for each valve of I scfh for a period of at least seven days. The (4) ADS function accumulators to compensate for essendal system with associated lines, valves the leakage from main steam S/R solenoid and fittings are classified as Safety Cisss 3, valves during normal operation Seismic Category 1.
6.7.2 Systern Description The nonsafety portion provides compressed nitrogen at a rate adequate to recharge the ADS Normally, nitrogen gas for both the essential SRV accumulators. The nonessential system has and nonessential makeup systems is supplied from two pressure control valves to depressurize the the nitrogen gas evaporator via the makeup line nitrogen gas from the AC system. One is to to the atmospheric control (AC) system. The depressurize to 200 psi for the SRV accumulators nitrogen supply system shall supply nitrogen and the other is to depressurize to 100 psi for which is oil-free with a moisture content of less other pneumatic uses.
than 2.5 ppm. This nitrogen is filtered in the HPIN system to remove particles larger than 5 The continuous supply portion of the microns. All equipment using this nitrogen shall pneumatic system, extending from the AC system be capable of operating with nitrogen of the to the isolation valve prior to the essential quality listed above. If nitrogen is not system is not safety related.
available from the AC system to supply the essential system, nitrogen is supplied from high Nonsafety piping and valves of the system are pressure nitrogen gas storage bottles. The designed to ANSI B31.1, Power Piping Code, and essential system is separated into two the requirements of Ouality Group D of divisions. There are ticlines between the Regulatory Guide 1.26. Pressure vessels and nonessential and each division of the essential heat exchangers are designed to ASME Section system. Each ticline has a motor operated Vill, Division 1.
shutoff valve. For details, see Figure 6.7-1 and Table 6.7-1.
System design pressure a 200 m wh the sy ' ~ Nr - W =
.m L'3"I'. a ~ 1 b gb4 F 7* 6 7-I-Each division of the essential system has ten m sb o-a 67-1 Amendment 24
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ABWR oT 20.3-2 2mweio Rrv c Standard Plant I
Category I, ASME Code III, Class 3, Quality Group isolation valves in order to verify their C and Quality Assurance B requirements, except leaktightness. Operation of valves and for the piping and valves for the containment and associated equipment used to switch from the drywell penetrations which are designed to nonsafety to safety nitrogen supply can be Seismic Category I, ASME Code III, Class 2, tested to assure operational integrity by manual Quality Group B and Quality Assurance B actuation of a switch located in the control requirements.
room and by observation of associated position indication lights. Periodic tests of the check The essential high pressure nitrogen grs valves and accumulators shall be conducted to supply is separated into two independent assure valve operability. j divisions, with each division capable of supplying 100% of the requirements of the 6.7.5 Instrumentation Requirements division being serviced. Each division is mechanically and electrically separated from the A pressure sensor is provided for the,afety other. The system satisfies the components' nitrogen supply, and an alarm signals fow nitrogen demands during all plant operation nitrogen pressure.
conditions (normal through faulted).
A remote manual switch and open-closed Safety grade portions of the high pressure position lights are provided in the control room l nitrogen gas supply system are capable of being for valve operation and position indication.
isolated from the nonsafety parts and retaining their funcdon during LOCA and/or seismic events under which any nonsafety parts may be damaged.
Pipe routing of Division 1 and Division 2
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g g[ gpg[ g g > g nitrogen gas is kept separated by enough space so that a single fire, equipment dropping accident, strike from a single high energy whipping pipe,
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jet force from a single broken pipe, internally generated missile or wetting equipment with gg g g /,'g gg [.[
spraying water cannot prevent the other division [ggg gg// g ggh from accomplishing its safety function.
Separation is accomplished by spatial separation j
or by a reinforced concrete barrier, to ensure g g.p g f/ g J-f/1 6 h O M separation of each pneumatic air division from any systems and components which belong to the gMqg f,5 g /J yg M g g g y other pneumatic air division.
d valve of 27 /dvs[e h our 6.7.4 Inspection and Testing Requirements I
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Period.ic inserv ce snspect. ion of components, in accordance with ASME Section XI, to ensure the g hd ya/pg,
capability and,ntegrity of the system is i
mandatory. Nitrogen quality shall be tested periodically to assure compliance with ANSI M C11.1.
The nitrogen isolation valves are capable of being tested to assure their operational integrity by manual actuation of a switch located in the control roc,m and by observation of associated position indication lights. Test and vent connections are provided at the containment l
6.7-2 Amendment 13 i
ABWR 23A6100AS Standard Plant w4 ation of the ADS. The results of this program were
Response
submitted to the NRC in a letter report from D. B.
Waters, Chairman of BWR Owners' Group, to D. G.
The ABWR primary containment is inerted and is, Eisenhut, Director (NRC), dated December 29, therefore, protected from hydrogen combustion re-1980. A summary of this evaluation follows.
gardless of the amount or rate of hydrogen genera-tion. In fact, increasing amounts of hydrogen moves The cases analyzed in the letter report: above show the primary containment oxygen concentration fur-that, based on core cooling considerations, no signifi-ther from the flammable regime. The ABWR is also i
cant improvement can be achieved by a slower de-provided with permanently-installed recombiners pressurization rate. A significantly slower depressur-which prevent the buildup of oxygen, due to radiolysis, ization will result in increased core uncovery times from creating a potentially flammable mixture.
before ECCS injection. Furthermore, a moderate Radiolysis is the only potential source of oxygen in the decrease in the depressurization rate necessitates an ABWR primary containment.
carlier action time to initiate ADS. Such an earlier actuation time has the negative impact of providing 19A.2.13 Long-Term Training Upgrade less time for the operator to start high pressure
[ Item (2) (i)]
ECCS without obtaining a significant benefit to vessel fatigue usage. This earlier actuation time ac-NRC Position cessitates a higher in;tiation level which would result in an increased frequency of ADS actuation.
Provide simulator capability that correctly models the control room and includes the capability to simu-It should be rioted that the ADS is not a normal late small-break LOCA's. (Applicable to construction core cooling s3 stem, but is a backup for the high permit applicants only.) [LA.4.2}
pressure core cooling systems such as feedwater, RCIC or HPCF. If ADS operation is required,it is
Response
because normal and/or emergency core cooling is threatened. As a full ADS blowdown is well within COL license information, see Subsection 19A3.1.
the design basis of the RPV and the system is prop-erly designed to minimize the threat to core cooling, 19 A.2.14 Long-Term Program of no change in depressurization rate is required or ap-Upgrading of Procedures [ Item (2) (ii) propriate.
NRC Position 19 A.2.12 Evaluation of Alternative Hydmgen Control Systems [ Item (1) (xii)]
Establish a program, to begin during construction and follow into operation, for integrating and expand-NRC Position ing current efforts to improve plant procedures. The scope of the program shallinclude emergency proce-dures, reliability analyses, human factors engineering, Perform an evaluation of alternative hydrogen n-trol systems that would satisfy the requirement of crisis management, operator training, and coordina-paragraph (f) (2) (ix) of 10CFR5034(f). As a ini-tion with INPO and other industry efforts. (Applica-mum include consideration of a hydrogen ig ition ble to construction permit applicants only.) [I.C.9]
and post-accident inerting system. The eva ation shallinclude:
Response
(A) A comparison of costs and benefits - the aher-COL license information, see Subsection 19A3.2.
l native systems considered.
19A.2.15 Control Room Design Reviews (B) For the selected system, analys and test data
[ Item (2) (iii)]
to verify compliance with th requirements of (f) (2) (ix) of 10CFR5034.
NRC Position (C) For the selected syste. preliminary design de-Provide, for Commission review, a control room scriptions of equi nt, function, and layout.
design that reflects state-of-the-art human factor prin-AlSee s a sec b G 2. 7. i for cot h est k b # ^ Os 19A.2 3
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Amendment 23 p
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u, OT 20 3-A MN 23A6100AB Standard Plant pry c these lines result in no significant safety con-6.2.43.2.2.2.1 RHR, RCIC and HPCF Unes sideration. All of the lines terminate below the minimum drawdown level in the suppression pool.
The RHR, RCIC, and HPCF suction lines contain motor-operated, remote. manually actuated The test return lines are also used for gate valves which provide assurance of isolating suppression pool return flow during other modes these lines in the event of a break. These of operation. In this manner, the number of valves also provide long. term leakage control.
penetrations are reduced, thus minimizing the in addition, the suction piping from the potential pathways for radioactive material suppression pool must be available for long-term release. Typically, pump minimum flow bypass usage following a design basis LOCA, and, as lines join the respective test return lines such, is designed to the quality standards downstream of the test return isolation valve. commensurate with its importance to safety. The The bypass lines are isolated by motor-operated RHR discharge line fill system suction lines valves in series with a restricting orifice.
have manual valves for operational purposes.
These systems are isolated from the containment 6.2.43.2.2.1.2 RCIC Turbine Exhaust and Pump by the respective RHR pump suction valves from Minimum Flow Bgass unes suppression pool.
The RCIC turbine exhaust line which 6.2.43.2.2.2.2 SPCU Suetion une penetrates the containment and discharges to the suppression pool is equipped with a normally The SPCU system suction line has two open, motor-operated, remote-manually actuated isolation valves. However, because the gate valve located as close to the containment n penetra' ion is under water, both the isolation possible. In addition, there is a simple check valves are located outside containment. The valve upstream of the gate valve which provides first valve is located as close as possible to positive actuation for immediate isolation in the the containment, and the second is located to event of a break upstream of this valve. The provide adequate separation from the first.
gate valve in the RCIC turbine exhaust is designed to be locked open in the control room 6.2.43.2.23 ACS Unes To Containment i
and is interlocked to preclude opening of the g
influent and effluent lines whic,h(ACS) has both The Atmospheric Control Syste inlet steam valve to the turbine until the penetrate the turbine exhaust valve is in its full open position. The RCIC pump minimum flow bypass line containment. Both isolation valves on these is isolated by a normally closed, remote manually lines are outside of the containment vessel to [
actuated valve outside containment, provide, accessibility to the valves. The 5 1
_ GFboarDvalvef1}icicated as close as practical
,Y 6.2.43.2.2.13 SPCU Discharge une to the containment vessel. The piping to both valves is an extension of the containment The suppression pool cleanup (SPCU) system boundary. nv5 F F T f, s. K J, ua. 2 discharge line to :he suppression pool (i.e.,
containment penetration, piping and isolation 6.2.43.2.2.4 Conclusion on Criterion 56 valves) is designed to Seismic Category 1, ASME Section IH, Class 2 requirements, in order to assure protection against the consequences of accidents involving release of 6.2.4J.2.2.2 Utleent unes trom Suppression significant amounts of radioactive materials, pipes that penetrate the containment have been Pool demonstrated to provide isoldion capabilities
- 3N Figure 6.2 38 identifies the isolation on a case-by-case basis in accordance with Criterion 56.
M.M% provisions in the effluent lines from the jo,3 4/
suppression pool.Each containment penetration and piping extension is single ronmental qualification of each safety h,3,;73 failure proof.
related valve assures common mode failure potential is acceptably low.
Similar piping and isolation is used in Susquehanna and Limerick BWR atmospheric control systems.
6230 Amendment
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23A61oors REV C Standard Plant Influent and effluent lines of this group are 62.4.4 Test and inspections isolated by automatic or remote-manualisolation valves located as close as possible to the The containment isolation system is containment boundary.
scheduled to undergo periodic testing during 7
i reactor operation. The functional capabilities l
l 62.43.2.4 Evaluation Against Regulatory of power-operated isolation valves are tested Guide l.11 remote-manually from the control room. By observing position indicators and changes in the lastrument lines that connect to the RCPB and affected system operation, the closing ability penetrated the containment have 1/4-inch orifices of a particular isolation valve is demonstrated.
and manual isolation valves, in compliance with l
Regulatory Guide 1.11 requirements.
Air-testable check valves are provided on influent emergency core cooling lines of the 6.2.433 Evaluation of Single Failure HPCF and RHR systems whose operability is relied upon to perform a safety function.
A single failure can be defined as a failure of a component (e.g., a pump, valve, or a utility A discussion of testing and inspection of l
such as offsite powei) to perform its intended isolation valves is provided in Subsection safety functions as a part of a safety system.
6.2.1.6. Instrumts are periodically tested The purpose of the evaluation is to demonstrate and inspected. Test and/or calibration points that the safety function of the system will be are supplied with each instrument. Leakage l
completed even with that single failure. integrity tests shall be performed on the Appendix A to 10CFR50 requires that electrical containment isolation valves with resilient l
systems be desigaed specifically against a single material seals at least once every 3 months.
passive or active failure. Section 3.1 describes l
6 the implementation of these standards as well as 6.2.5 Combustible Gas Controlin General Design Criteria 17, 21, 35, 38, 41, 44, Containment The atmospheric control system (ACS-T31) is l l
54,55 and 56.
l Electrical as well as mechanical systems are provided to establish and maintain an inert designed to meet the single-failure criterion, atmosphere within the primary containment during regardless of whether the component is required all plant operating modes except during shutdown to perform a safety action. Even though a com-for refueling or equipment maintenance and ponent, such as an electrically operated valve, during limited periods of time to permit access is not designed to receive a signal to change for inspection at low reactor power. The state (open or closed) in a safety scheme,it is flammability control system (FCS-T49) is provided to control the potential buildup of assumed as a single failure if the system compon-ent changes state or fails. Electrically-oper-oxygen from design basis radiolysis of water. MU/
ated valves include valves that are electric-The objective of these systems is to preclude,29.3-3 ally piloted but air operated, as well as valves combustion of hydrogen and damage to essential that are directly operated by an electrical de-equipment and structures. /M&r d',p,r
[h.g 1
vice. In addition, all electrically-operated j
valves that are automatically actuated can also 6.2 5.1 Design Bases Therefore, a single failure in any electrical Following are criteria that serve as the
't be manually acguated from the main control room.
system is analyzed, regardless of whether the bases for design:
loss of a safety function is caused by a component failing to perform a requisite (1) Since there is no design requirement for mechanical motion or a component performing an the ACS or FCS in the absence of a LOCA and unnecessary mechanical motion.
there is no design-basis accident in the ABWR that results in core uncovery or fuel i
failures, the following requirements j
mechanistically assume that a LOCA f
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Included in the leak rate test summary report
'I will be, a report detailing the containment in-spection, a report detailing any repairs neces-sary to pas! ac tests, and the leak rate test l
results.
6.2.6.5 Special Testing Requirements -
[
The maximum allowable leakage rate into the secondary containment and the means to verify i
that the inleakage rate has not been exceeded, as j
well as the containment leakage rate to the environment, are discussed in Subsections 6.23
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l 6.2fReferences I
t 1.
W.J. Bilanin, The G.E. Mark 111 Pressure Suppressio_n Containment Analytical Model, l
June 1974, (NEDO-20533).
?
2.
F.3. Moody, Maximum Discharge Rate of -
l Liquid Vapor Mixtures from Vessels, Genernl Electric Company, Report No. NEDO-21052, September,1975.
3.
W.J. Bilanin, The G.E. Mark III Pressure Ssppression Containment Analytical Model, Supplement 1, September 1975 (NEDO-20533-1).
j i
g;4.
J.P. Dougherty, SCAM-Subcompartment.
Analysis Method, January 1977, (NEDE-e,
y 21526).
/
i t
l l
t
.i I
t 4
t i
i
'M Amendmenty I
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1 ABWR 21Aa10MB Standard Plant
,c refuelieg floor ventilation exhaust, both SGTS trains efficiencies are outlined in Table 6.5-1. Dose are automatically operated. When the operation of analyses of events requiring SGTS operation both the trains is assured, one train is placed in described in Subsections 15.6.5 and 15.7I standby mode. la the event a malfunction disables indicate that offsite doses are within the limits an operating train, the standbytrain is automatically established by 10 CFR 100.
initiated.
j, p p g e r e Im /4 (3) The SGTS is designated as an engineered 6.5.1.23.
Manual ggpfyg safety feature since it mitigates the CPM
- It consequences of a postulated accident by 45./4 The SGTS fon standby during normal plant controlling and reducing the release of operation.and may be manually initiatgd b'e e e radioactivity to the environment. The SGTS.
fy hog primary containmentfurging Ode inertingh except for the deluge, is designed and built to when required to limit the darge of contaminants the requirements for Safety Class 3 equipment
( to the environment within 10CFR20 limits. k Ipas defined in Section 3.2, and 10 CFR 50, p be manually initiated for 'a g a d :::=: b ree Appendix B.
rb
- d h a. M or
=ks.au:.:: Ad :s..dd =::" 3:4. fang The SGTS has independent, redundant active
[g j
%Aw;/lomte.
tratas. Should any active train fail, SGTS 6.5.1.233 Decay Heat Removal functi ns can be performed by the redundant train. Tb :Emke! 4 ;,a; d hm::M Cooling of the SGTS filters may be required to r ; :::::ypowered from separate C1 prevent the gradual accumulation of decay heint in L LE electrical busses. doid, p t the charcoal. This heat is generated by the decay of Wa. Jo radioactive iodine adsorbed on the SGTS charcoal.
(4) The SGTS is designed ismic Category 1 The charcoal is typically cooled by the air from the requirements as specified in Section 3.2. The process fan.
SGTS is housed in a Category I structure. All surrounding equipment, components, and A water deluge capability is also provided, but supports are designed to appropriate safety primarily for fire protection since redundant process class and seismic requirementa.
fans are provided for air cooling. Since the deluge is available, it may also be used to remove decay heat (5) A secondary containment draw-down analysis (
for sequences outside the normal design basis.
will be performed to demonstrate the capability (
[)
Temperature instrumentation is provided for control of the SGTS to maintain the design negative of the SGTS process and space electric heaters. This pressure followmg a LOCA 'mc!uding inleakage instrumentation may also be used by the operator to from the open, non-isolated penetration lines b
[re-] establish a cooling air flow post accident, if identified during construction engineering and j required.
the vent of the worst single failure of a secondary isolation valve to close. (See Water is supplied from the fire protection system Subsection 6.5.5.1 for COL license information and is connected to the SGTS via a spont piece.
requirements).
6.5.13 Design Evalenties 6.5.13.2 Slzing Basta 6.5.13.1 Genersi Figure 6.5 2 provides an assessment of the [
Q secondary containment pressure after the [
(1) A slight abative pressure is normally design-basp LOCA assuming an SGTS fan capacity i maintained in the secondary containment by of 6800 m hr (21 C, I atmosphere) per fan. Credit '
the reactor building fWAC system (Subsection for secondary containment as a fission product 9A.5). On SGTS initiation per Subsection control system is only taken if the secondary 6.5.1.23.1, the secondary containment HVAC containment is actually at a negative pressure by is automatically isolated.
considering the potential effect of wind on the ambient pressure in the vicinity of the reactor (2) The SGTS filter particulate and charcoal building. For the ABWR dose analysis, direct transport of containment leakage to the environment mc4
+x 7% ~ m ba.a;& k%ymt~ue(Selwaf'IbyN %}
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TABLE 6.5-1 i
STANDBY GAS TREATMENT SYSTEM I
COMPONENT DESCRIPTION i
?
i Filter Train Consists of a moisture separator,an electric (
i process heater, prefilter, pre HEPA filter, i charcoal adsorber, post HEPA filter and '
7 space clearic heaters.
?
Ouantity 2
h 3
Capacuy 6800 m /hr
/
k -
Motstore Separater General Woven wire, arminlm steel mesh pads
['
{
Ouantity 1 bank of Mandard size moisture separators ('
per filter train l.
i' t
Effiocacy per ASME N509, Section 5.4 L
/
Electric Process Hester General Electric, finned tubular ctkn Onnarity 1 pc aan f
I i
Rating 53 kW minimum,26.2 kW====um r
( CPrW.
/ (4l~l Relative humaidity
.g.
Inlet 100% @ 66,C 7
Outlet 70% @ 75"C f
Air AT 9C f
/
PruSher
[
f General Cartridge type M:
(,
f..
Omastity 1 bank of standard size filters per filter train ;
I
)
l Media Glass fiber j
Effidency Per ASME N509, Section 53
{
l f Ca y cLSy q -L. Luka is A u ysci d to 4.safu G h i
s lo t. &q h-f usu LaasP c
b l
w s a e..
I a
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evn reactor coolant pressure boundary and the release of radioactive materials from either the Instrumentation and control is provided to reactor coolant pressure boundary or from the automatically maintain an acceptable thermal l k
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fuel and equipment storage pools.
environment for safety equipment and operating personnel.
7.1.133 Wetwell and Drywell Spray Mode of RIIR 7.1.13.9 HVAC Emergency Cooling Water System Instrumentation and control provides ma ual initiation of werwell spray and manual initiation Automatic instrumentation and control isl n&
of drywell spray (when high drywell pressure provided to assure that adequate cooling is signal is present) to condenses steam in the con-provided for the main control room, the control tainment and remove heat from the containment, building essential electrical equipment rooms, The drywell spray has an interlock such that dry-and the diesel generator cooling coils.
well spray is possible only in the presence of a 7.1.13.10 high drywell pressure condition.
liigh Pressure Nitrogen Gas Supply Sptem 7.1.13A Suppression Pool CooIIng Mode of RitR (SPC RIIR)
Automatic instrumentation and control is l n R
provided to assure adequate instrument high Instrumentation and control is provided to pressure nitrogen is available for ESF equipment manually initiate portions of the RHR system to operational support.
effect cooling of the suppression pool water.
7.1.1.4 Safe Shutdown Systems 7.1.13.5 Standby Gas Treatment System 7.1.1A.1 Alternate Rod lasertion function (ARI) lustrumentation and Controlis provided to maintain negative pressure in the secondary
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containment and for automatically limiting Though not required for safety, instrumenta-airborne radioactivity release from containment tion and controls for the ARI provide a function
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for mitigatior, of the consequences of antici-pated transient without scram (ATWS) events.
7.1.13.6 Emergency Diesel Generator Support Upon receipt of an initiation signal (high reac-Systems tot dome pressure or low reactor water level),
the fine motion control rod drive (FMCRD) motor Instrumentation and control is provided to as-shall automatically drive all rods full-in.
A sure availability of electric control and motive This provides a method, diverse from the $
hydraulic control units (HCUs) for scramming the power under all design basis conditions. The reactor.
function of the diesel generator is to provide h j automatic emergency AC power supply for the safety related loads (required for the safe 7.1.1A.2 Standby Uguid ControI System (SIES) shutdown of the reactor) when the offsite source j
of power is not available.
Instrumentation and controls are provided for the manual initiation of an independent backup i
l 7.1.13.7 Reactor Building Cooling Water System system which can shut the reactor down from
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rated power to the cold condition in the event
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lastrumentation and control is provided to that all withdrawn control rods cannot be assure availability of cooling water for heat inserted to achieve reactor shutdown.4 removal from the nuclear system as required.
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Safety-related portions of this system start 7.1.1A3 Residual lient Removal (RHR) System /
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automatically on receipt of a LOCA and/or LOPP Shutdown Cooling Mode signal.
Instrumentation and controls provide manual 7.1.13.8 Essential HVAC Systems initiation of cooling systems to remove the decay and sensible heat from the reactor vessel.
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