ML20034F578
| ML20034F578 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 02/26/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9303040020 | |
| Download: ML20034F578 (11) | |
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February 26,1993 Docket No. STN 52-001 i
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Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Review Schedule - Chapter 3 and 9 COL Action Items
Dear Chet:
Enclosed are SSAR markups addressing ABWR DFSER COL Action items 3.3.2-1, 3.5.1.2-1,3.5.2-1,3.10-1 and 9.3.5-1.
It should be noted that COL Action Item 9.3.5-1 was previously Confirmatory Item 9.3.5-1.
Sincerely, 9 %p Jack Fox Advanced Reictor Programs cc: Gary Ehlert (GE)
Norman Fletcher (DOE)
Cal Tang (GE) l
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t y z u 9 ;ll JI*11L-41 9303040020 930226 PDR ADOCK 05200001 A
co'- 3 3}. 2-}
ABW11 23raoore Standard Plant nev s are provided on all air intake and exhaust 3.
Deleted openings. These dampers are de.igned to
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withstand a negative 1.46 psi pressure.
33.23 Effed of Failure of Structures or Components Not Designed for Tornado [ mads 4 Bechtel Topical Report BC-TOP-3-A, Revision 3, Tornado and Extreme Hind Design Criteria All safety-related system and components are for Nuclear Power Plants.
protected within tornado-resistant structures.
See Subsection 3333 forg,o t i cu-g de ~M u. n c
aceteqmrement.
3.3.3 Interfaces 333.1 Site-Specific Design Basis Wind The site-specific design basis wind shall not exceed the design basis wind given in Table 2.0-1 (See Subsection 2.2.1).
3 33.2 Site Specific Design Basis Tornado The site-specific design basis tornado shall not exceed the design basis tornado given in Table 2.0-1 (See Subsection 2.2.1).
wow-b.Sec Co 443 cq I 3.3.3.3 Effect of[Rs.--............ Struc-(-
turesh and Components not Designed for Tornado Loads
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%e Qok a s c. of n o w f e. \\S' W C All remainder of plant structures, systems, c
and components not designed for tornado loads shall be analyzed for the site-speenfic loadings g D,9 7 4
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i to ensure that their mode of failure will not effect the ability of the Seismic Category I ABWR W 3 up of & ^6WU-S MAA Standard Plant structures, systems, and compo-f f Evk u, Il w a }.t>do.qare ked lkad' 54 t fw C.
nents to perform their intended safety functions.
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3.3.4 References ke co a.d a f"" ###
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1.
ANSI Standard A58.1, Minimum Design Loads
( g 54 for Buildings and Otiter Structures, Committee A. 58.1, American National Standards Institute.
2.
ASCE Paper No. 3269, Wind Forces on Structures, Transactions of the American Society of Civil Engineers, Vol.126, Part II.
J Amendment 25 B2
ABWR coti..s.d-s 2mmose Standard Plant nrv n missic-consequence mitigation by structural not considered credible. However, credible sec-
' walls and slabs. These walls and slabs are ondary missiles, e.g., concrete fragments, may be designed to withstand internal missile effects; formed followingimpact of primary missiles. See the applicable seismic category and quality group Subsection 3.5.4.4 for COL license information classification are listed in Section 3.2.
requirements.
Penetration of the structural walls by internally generated missiles is not considered credible.
3.5.1.2.4 Evaluation of Potential Gravitational Missiles inside Containment For local shields and barriers see the response to Question 410.9.
Gravitational missiles inside the containment have been considered as follows:
3.5.1.2 laternally Generated Missiles (Inside Containment)
Seismic Category I systems, components, and structures are not potential gravitational mis-Internal missiles are those resulting frorn site sources.
plant equipment failures within the contain-ment. Potential missile sources from both Non-Seismic Category I items and systems rciating equipment and pressurized components are inside containment are considered as Fo!!ows:
considered.
(1) CableTray 3.5.1.2.1 Rotating Equipment All cable trays for both Class IE and non-By an analysis similar to that in Subsection Class 1E circuits are scismically supported 3.5.1.1.1, it is concluded that no items of whether or not a hazard potential is evident.
rotating equipment inside the containment have the capability of becoming potential missiles. (2) Conduit and Non-Safety Pipe All reactor internal pumps are incapable of achieving an overspeed condition and the motors Non-Class 1E conduit is seismically sup-and impellers are incapable of escaping the ported if it is identified as a potential casing and the reactor vessel wall, respectively.
hazard to safety-related equipment. All Nuclear Island non safety related piping that 3.5.1.2.2 Purssurized Components is identified as a potential hazard is seismically analyzed per Subsection 3.7.3.13.
Identification of potential missiles and their consequences outside containment are specified in (3) Equipment for Maintenance Subsection 3.5.1.1.2. "The same conclusions are p,... b drawn for pressurized components inside of con-All other equipment, such as hoists, that is taiment. For example, the ADS accumulators are require during maintenance will either be moderate energy vessels and are therefore not removed femag operation, moved to a location considered a credible missile source. One where it is not a potential hazard to safety-additional item is fine motion control rod drives related equipment, or seismically restrained revent it from becoming a missile. See (FMCRD) under the reactor vessel. The FMCRD to 8 * **
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5 mechanisms are not credible missiles. The FMCRD housings are designed (Section 4.6) to prevent 3.5.1.3 Turtine Missiles any significant nuclear transient in the event of a drive housing break.
See Subsection 3.5.1.1.13.
3.5.1.23 Missile Barriers and Loadings Credit is taken in some cases of rotating and pressurized components generating missile for missile-consequence mitigation by structural walls and slabs. Penetration for the containment walls, floors and slabs by potential missiles is 354 Amendment 23
MM COL 3.3.2'I am Standard Plant prv n generated from other natural phenomena. The 3.5.1.6 Alteraft Hazards design basis tornado for the ABWR Standard Plant is the maximum tornado windspeed corresponding to Aircraft hazards are not a designpasis event a probability of 10E-7 per year (300 mph). The for the Nuclear Island (i.e.110 per year).
other characteristics of this tornado, summerized See Subsection 3.5.4.3 for COL license in Subsection 3.3.2.1. The design basis tornado information requirements.
missiles are per SRP 3.5.1.4, Spectrum I.
3.5.2 Structures, Systems,and Cornponents to be Protected from Externally Generated Missiles The sources of external missiles which could affect the safety of the plant are identified in Subsection 3.5.1. Certain iten;in the plant are required to safely shut down the reactor and maintain it in a safe condition assuming an Using the design basis tornado and missile additional single failure. These items, whether spectrum as defined above with the design of the they be structures, systems, or components, must Seismic Category I buildings, compliance with all therefore all be protected from externally of the positions of Regulatory Guide 1.117, generated missiles.
" Tornado Design Classification," Positions C.1 and C.2 is assured.
These items are the safety related items listed in Table 3.21. Appropriate safety The SGTS charcoal absorber beds are housed in classes and equipment locations are given in this the tornado resistant reactor building and table. All of the safety-related systems listed therefore are protected from the design basis are located in buildings which are designed as tornado missiles. The offgas system charcoal tornado resistant. Since the tornado missiles absorber beds are located deep within the turbine are the design basis missiles, the systems, building and it is considered very unlikely that structures, and components listed are considered these beds could be ruptured as a result of a to be adequately protected. Provisions are made design basis tornado missile. These teatures to protect the charcoal delay tanks against assure compliance with Position C.3 of Regulatory tornado missiles.
M*4 *E
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Guide 1.117.
3 See Subsectiong3.5.4.1[for COL license An evaluation of all non safety-related informatior. re,quirements.
structures, systems, and components (not housed in a tornado structure) whose failure due to a 3.5.3 Barrier Design Procedures design basis tornado missile that could adversely impact the safety function of safety related The procedures by which structures and systems anc' components will be provided to the barriers are designed to resist the missiles NRC by the applicant referencing the ABWR described in Subsection 3.5.1 are presented in design. See Subsection 3.5.4.2 for COL license this section. The following procedures are in information requirements.
accordance with Section 3.5.3 of NUREG-0800 (Standard Review Plan).
3.5.1.5 Site Proximity Missiles Except 3.5.3.1 14 cal Damage Prediction i
Aircraft External missiles other than those generated The prediction of local damage in the impact by tornadog are not considered as a design basis area depends on the basic material of construc-tion of the structure or barrier (i.e., concrete (i.e.1 10 per year).
or steel). The corresponding procedures are presented separately. Composite barriers are not utilized in the ABWR Standard Plant for missile j
protection.
1 l
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cot 3. S', t. 2 4 ABM 234simre
. Standard Plant g
3,3, q_ g nev u valent static load concentrated at the impact impact the safety function of a safety-related area is determined. The structural response to systems and components will be provided to the this load, in conjunction with other appropriate NRC by the applicant referencing the ABWR design loads, is evaluated using an analysis design. (See Subsection 3.5.1.4).
procedure similar to that in Reference 6 for rigid missiles, and the procedure in Reference 7 3.5A.6 Turbine System Maintenance Program for deformable missiles.
A turbine system maintenance program 3.5.4 Interfaces including probability calculations of turbine missile generation meeting the minimum 3.5A.1 Protection of Ultimate lleat Sink requirement for the probability of missile generation shall be provided to the NRC (See Compliance with Regulatory Guide 1.27 as Subsection 3.5.1.1.3).
related to the ultimate heat sink and connecting x
1 r4 S56T N S V T P ACE conduits being capable of withstanding the 3.5.5 References effects of externally generated missiles shall be demonstrated (See Subsection 3.5.2).
1.
C. V. Moore, The Design of Barricades for Ha:ardous Pressure Systems, Nuclear 3.5A.2 MissIIes Generated by Natural Phenomena Engineering and Design, Vol. 5,1967.
from Remainder of Plant Structures, Systems and Components 2.
F. J. Moody, Prediction of Blowdown Thrust and Ier Forces, ASME Publication 69-HT-31, The remainder of plant structures, systems, August 1%9.
and components shall be analytically checked to ensure that during a site-specific tornado they 3.
A. Amirikan, Design of Protective Struc-will not generate missiles exceeding the missiles tures, Bureau of Yards and Docks, Publica-considered under Subsection 3.5.1A.
tion No. NAVDOCKS P-51, Department of the Navy, Washington, D.C., August 1960.
3.5A.3 Site Proximity Missiles and Aircraft llazards.
4.
A. E. Stephenson, Full-Scale Tornado-Mis-site Impact Tests, EPRI NP-440, July 1977, Analyses shall be provided that demonstrate prepared for Electric Power Research that the probability of site proximity missiles Institute by Sandia Laboratories.
(including aircraft) impacting the ABWR Standard Plant and causing consequences greater than 10CFR 5.
W. B. Cottrell and A. W. Savolainen, U. S.
Part 100 exposure guidelines is s.10" per year Rescror Containment Technology, ORNL-(See Subsection 3.5.1.6).
NSIC 5, Vol.1, chapter 6, Oak Ridge Na-tional Laboratory.
3.5AA Secondary Missiles laside Containment 6.
R. A. Williamson and R. R. Alvy, Impact Protection against the secondary missiles Effect of Fragments Striking Structural inside containment described in Subsection Elements, Holmes and Narver, Inc., Revised 3.5.1.2.3 shall be demonstrated.
November 1973.
7.
J. D. Riera, On the Stress Analysis of 3.5A.5 Impact of Failurt of Non Safety Related Structures Subjected to Aircraft Impact Structures. Systems, and Components Due to a forces, Nuclear Engineering and Design, Design Basis Tornado North Holland Publishing Co., Vol. 8,1968.
An evaluation of all non safety-related 8.
Deleted structures, systems, and components (not housed in a tornado structure) whose failure dt'e to a design basis tornado missile that could adversely 3.5-8 Amendment 25
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i ABWR n-t Standard Plant uv e by dynamic analysis using appropriate exceeded when the tubing is subjected to the response spectra.
loads specified in Subsection 3.9.2 for i
Class 2 and 3 piping, j
(b) Floor Response Spectra 3.10.4 Operating License Review (Tests and (i) Floor response spectr3 used are Analyses Results) those generated for the supporting fl o o r.
In case supports are See Subsection 3.10.5.2 for COL bcense i
attached to the walls or to two information requirements.
j different locations, the upper i
bound envelope spectra obtained by 3.10.5 COL License Information superimposing are used.
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3.10.5.1' Equipment Qualification $tecordg cot (ii) 1o many cases, to facilitate the
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design, several floor response The equipment qualification records spectra are combined by an upper including the reports (see Subsections j
bound envelope obtained by 3.10.2.1 A and 3.10.2.2.3) shall be maintained M superimposing.
in a permanent file and shall be readily available for audit.
3.103.2.3 Local Instrument Supports
-j 3.103.2 Dynamic Qualification Report For field-mounted Seismic Category I instruments, the following is applicable:
A dynamic qualification report (DQR) shall be prepared identifying all Seismic Category I (1) The mounting structures for the instruments instrumentation and electrical parts and have a fundamental frequency above the equipment therein and their supports. The DOR excitation frequency of the RRS.
shall contain the following: (1) A table or file for each system that is identified in (2) The stress leve! in the mounting structure Table 3.21 to be safety related or having i
does not exceed the material allowable Seismic Category I equipment shall be included l
stress when the mounting structure is in the DOR containing the MPL item number and -
subjected to the maximum acceleration level name, the qualification method and the input j
for its location.
motion for all Seismic Category I equipment
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and the supporting structure in the system, l
3.10.3.2A Instrument Tubing Support and the corresponding qualification summary table or vendor's qualification report. (2)
The following bases are used in the seismic The mode of safety-related operation (i.e.,
and other RBV dynamic loads design and analysis active, manual active or passive) of the i
of Seismic Category I instrument tubing supports:
instrumentation and equipment along with the manufacturer identification and model numbers l
(1) The supports are qualified by the response shall also be tabulated in the DQR. The j
spectrum method; operational mode identifies the instrumentation or equipment (a) that performs (2) Dynamic land restraint measures and analysis the safety related functions automatically, 7
for the supports are based on combined (b) that is used by the operators to perform l
l limiting values for static load, span the safety related functions manually, or (c) length, and computed dynamic response; and whose failure can prevent the satisfactory
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accomplishment of one or more safety-related (3) The Seismic Category I instrument tubing functions. (See Subsection 3.10A).
systems are supported so that the allowable i
stress permitted by Section !!! of ASME i
Boiler and Pressure Vessel Code are not t
InsM r 37.10.E.]
COL applaanh wil mvide. /%tspen'AL saisofE ml dywn't i
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3.9,7 COL License Inforrnation Subsecsion 3.93.1.)
3.9.7.1 Reactor laternals Vibration Analysis, 3.9.73 Pump and Valve laservice Testing i
Measurement and laspection Program Program The first COL applicant wjll provide, 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 fer 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 SuMeet 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 Measurernent depending on test results. (See Program Subsections 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 accordance with setting of the torque and limit switches.
the applicable portions of position C.3 of (See Subsection 3.9.6.2.2) c4 Regulatory Guide 1.20 for non-prototype
/psppy 3.4,7.3 o y, j p /
interna 1s.
3.9.7.4 Audit of Design Specification and g 7,yjg.g Design Reports Subsequent COL applicants need only provide the information on the schedules in accordance COL applicants will make available to the 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, j internals. (See Subsection 3.9.2.4),
pumps, valves and piping systems for the purpose of audit. (See Subsection 3.9.3.1) 3.9.7.2 ASME Class 2 or 3 or Quality Group D Components with 60 Year Design IJre 3.9.8 References i
COL applietats will identify ASME Class 2 L BWR Fuel Channel Mechanical Design and or 3 or Quality Group D components that are Deflection. NEDE-21354-P, September 1976.
subjected to cyclic loadings, including operating vibration loads and thermal transients effects, 2.
BRR/6 Tucl Assembly EmAanion 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.
Code calculations and, if similar designs have NEDE-21175-P, November 1976.
not already been evaluated, either provide an i
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.
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Barriers have been considered to assure SLCS check valves. Position indicator lights in the protection from pipe break (Section 3.6).
control room indicate that the local valve is closed for test or open and ready for It should be noted that the SLCS is not operation. Leakage from the reactor through the required to provide a safety function during any first check valve can be detected by opening the postulated pipe break events. This system is same test connection in the line between the only required under an extremely low probability check valves when the reactor is pressurized.
event, where all of the control rods are assumed to be inoperable while the reactor is at normal
'The test tank contains condensate water for full power operation. Therefore, the protection approximately 3 minutes of pump operation.
provided is considered over and above that Condensate water from the makeup system or the required to meet the intent of ASB 3-1 and MEB condensate storage system is available for 3-1.
refilling or flushing the system.
This system is used in special plant Should the boron solution ever be injected capability demonstration events cited in Appendix into the reactor, either intentionally or in.
A of Chapter 15; specifically, Events 54 and 56, advertently, then after making certain that the which are extremely low probability nondesign normal reactivity controls will keep the reactor basis postulated incidents. The analyses given subcritical, the boron is removed from the reac-there are to demonstrate additional plant safety tor coolant system by flushing for gross dilu-considerations far beyond reasonable and tion followed by operating the reactor cleanup conservative assumptions, system. There is practically no effect on reac-tor operations when the boron concentration has 9.3.5.4 Testing and Inspection Requirements been reduced below approximately 50 ppm.
Operational testing of the SLCS is performed The concentration of the sodium pentaborate in at least two parts to avoid inadvertently in the solution tank is determined periodically injecting boron into the reactor, by chemical analysis.
With the valves to the reactor and from the Electrical supplies and relief valves are storage tank closed, and the valves to and from also subjected to periodic testing.
the test tank opened, condensate water in the test tank can be recirculated by locally starting The SLCS preoperational test is described in either pump.
Subsection 14.2.12.
During a refueling or maintenance outage, the
.3.5.5 Instrumentation Requirements injection portion of the system can be functionally tested by valving the suction line The instrumentation and control system for to the test tank and actuating the system from the SLCS is designed to allow the injection of the control room. System operation is indicated liquid poison into the reactor and the in the control room.
maintenance of the liquid poison solution well above the saturation temperature. A further After functional tests, all the valves must be discussion of the SLCS instrumentation may be returned to their normal positions as indicated found in Section 7.4.
in Figure 9.3-1.
l After closing a local locked-open valve to the reactor, leakage through the injection valves can be detected by opening valves at a test connection in the line between the drywell (
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