ML20198F928
| ML20198F928 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 08/07/1974 |
| From: | Maccary R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-0988, CON-WNP-988 NUDOCS 8605290145 | |
| Download: ML20198F928 (19) | |
Text
I AUG 7 1974 V. A. Moore, Assistant Director for Light Water Reactors, Group 2 Directorate of Licensing St.TETY EVALUATION, im?-1 Plant Mwe: UPPSS liuclear Project No. 1
~
Licensing Etss;e: CP Docket ren. : SD-460 Responsible RP Branch and Project Manager:
LL'R 2-3, T. Cox Responsihie TR 1 ranch ftEB Requested Coupletion Date: August 5, 1974 Review Status: Complete but with one item not fully resolved. An SER nupplement may be requfred when resolution his been achieved.
The information submitted by the applicant in the PSAR. through Amendment 8, has been reviewed and evaluated by the Mcchanical Tngineering Branch, Directorate of Licensing. The appropriate s'ections of the Safety Evaluation are enclosed.
In the enclosure, one of the findings will be found tra contain a qualification requiring resolution by codification of a stress criterion provided in the PSAR. The finding in question is 3.9.2.1, Plant Conditions an.1 Desitja Loading Continations for AS'tr. Code Class 2 and 3 Components. The stress criterion in question is the primary raembrann plus pritary bendin:; design stress linit under the emergency condition load combinations in Class 2 and 3 pressure veascla designed by the rules of Section VIII, Division 1.
An acceptable value of thin stress limit is 1.8 S whereas the value selected by the applicant is 2.25 C.
(Mginal sigied by R.R. Maccary 8605290145 740007 PDR ADOCK 05000460 R. R. Maccary, Assistant Director A
PDR for eer W Directorate of Licensing cc w/ enc 1:
ec v/o enel S.11. liannuer, DRTA A. Giambusso, L T. Schroeder, L U. C. licDonald, L A. Schwencer L Docket File 5 J. P. ::uicht, L L. Reading L MEB File v~
.,,,.h':bou,L LIMEB L:MEi,/
I f.
/
A. N 'fillef,'"E b
k y
ABMillfrijm JPKnight....
accary-8/.6 /74
.8/I'/74 8/t/74 r..
cac.n. < n,.. nn u os os.o
MECHANICAL ENGINEERING BRANCH DIRECTORATE OF LICENSING WPPSS NUCLEAR PROJECT NO. 1 DOCKET No. 50-460 3.0 Design of Structures, Components, Equipment and Systems 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping The applicant has submitted criteria for postulating design basis pipe breaks and design basis leakage cracks s's well as for protecting against the resulting dynamic effects in piping systems containing high energy fluids and located outside the containment.
The design of piping restraints as applied to the reactor coolant pressure boundary and related high energy fluid systems within containment will provide adequate protection of the containment structure, reactor coolant system components, and other systems important to saf ety which are either interconnected with the reactor coolant system, or in close proximity to other high energy fluid lines in which postulated pipe failures are assumed to occur as a design basis. The systems which are considered and the locations and types of pipe break to be provided for are consistent with Regulatory Guide 1.46 (Protection Against Pipe Whip Inside containment, 5/73). The method of analysis used will adequately account for the dynamic loadings that are associated with the pipe rupture postulated and will provide l
l l
adequate assurance that the containment structure, unaffected system components, and those systems important to safety which are in close proximity to the systems in which postulated pipe failures are assumed to occur, will be protected.
These provisions for protection against the dynamic effects associated with pipe ruptures and the resulting discharging coolant provide adequate assurance that, in the event of the occurrence of the combined loadings imposed by an earthquake of the magnitude specified for the SSE and a concurrent single pipe break of the largest pipe at any one of the design basis break locations, the fol'1owing conditions and safety functions will be accommodated and assdred:
1.
The design basis loss-of-coolant accident will not lead to multiple failures of piping, that could aggravate the con-sequences of a pipe rupture.
2.
The reactor emergency core cooling systems can be expected to perforn their intended function.
3.
Structures, systems and components important to safety will be appropriately protected.
The methods used for formulating the hydro-dynamic forcing functions induced by pipe rupture and the dynamic analysis for the pipe whip l
motion will provide an acceptable basis for restraint des'ign. The i
l i
l 1
l
3 t
criteria used for the identification, design, and analysis of piping i
systems where postulated breaks may occur constitute acceptable design bases for meeting the applicable requirements of GDC 1, 2, 4, 14, 15, 31 and 32.
t 1
k i
t J
l I
l i
?
I i
r l
l
_4-0 3.9 Mechanical Systems and Components 3.9.1 Dynamic System Analysis and Testing 3.9.1.1 Piping Vibration Operational Test Program The applicant will conduct a piping vibration operational test program in ar.cordance with the ASME Code,Section III, par.
NE-3622.3 and NC-3622, which requires that the designer be responsible by observation under startup or initial operating conditions, for ensuring that the vibration of piping systems is within acceptable levels. A preoperational vibration dynamic effects test program will be conducted on all ASME Class 1 and Class 2 piping systems and piping restraints during startup and the initial operating conditions testing.
The test will provide adequate assurance that the piping and piping restraints of the system have been designed to withstand vibrational dynamic effects due to valve closures, pump trips, and operating modes associated with the design operational transients.
The tests, as planned, will develop loads similar to those experienced during reactor operation and are consistent with recent Regulatory staff positions concerning preoperational piping dynamics effects test programs. Compliance with this test program constitutes an acceptable basis for satisfying of the applicable requirements of General Design Criterion 2.
i
. 1 3.9.1.2 Dynamic Oualification Procedures for Mechanical Equipment 1
Dynamic testing and analysis procedures will be implemented to confirm that all Category I mechanical equipment will function during and af ter an earthquake of magnitude-up to and including the SSE, and that all equipment support structures are adequately designed to withstand such seismic disturbances.
3.9.1.3 Preoperarional Vibrarion Assurance Program for Reactor Internals The combination of tests, predictive analyses, and inspections will provide adequate assurance that the reactor internals may b'e expected, during their service lifetime, to withstand the flow-induced vibrations of reactor ope.ations without gross loss of structural integrity. The integrity of the reactor internals in service iIs essential to assure the retention of all reactor fuel assemblies in place as well as to permit unimpaired operation of the control rod assemblies for safe reactor operation and shutdowns.
Implementation of these dynamic testing and analysis procedures constitutes an acceptable basis for satisfying the applicable requirements of General Design Criteria 2 and 14.
With regard to flow-induced vibrational testing of reactor internals of this facility, the applicant has stated that if the WNP-1 reactor is the first of the B & W 205 fuel assembly reactors to be ready for hot functional testing then it will be tested as a prototype reactor in accordance with Regulatory Guide 1.20 (Vibration Measurements on Reactor Internals, 12/71).
i l
l
If the staff has accepted another 205 fuel assembly reactor as a satisfactory prototype before this facility is ready for hot functional testing, the applicant will perform additional confirmatory vibration testing and subsequent visual inspection on this unit as part of the preoperational tests to provide added confirmation of the capability of the structural elements of the reactor internals to sustain flow-induced vibrations.
The proposed program is consistent with Regulatory Guide 1.20.
We will review at the FSAR stage the preoperational vibration test program proposed by the applicant to verify the design adequacy of the reactor internals under loading conditions comparable to those experienced during operation. The combination of tests, predictive analysis, and post-test inspection will provide adequate assurance that the reactor internals can be expected to withstand flow-induced vibrations without loss of structural integrity during their service lifetime. The preoperational vibration test program will be performed in accordance with Regulatory Guide 1.20 and as such sosstitutes an acceptable basis for demonstrating the design adequa.;y of the reactor interns 1s in satisfying the applicable requirements of General Design Criteria 2 and 14.
. 3.9.1.5 Analysis Methods Under LOCA Loadings To assure the structural design adequacy of the reactor internals, the applicant will perform a dynamic analysis of the reactor internals and of broken and unbroken piping loops. The dynamic system analysis will be performed under the combined effects of
' he postulated occurrence of a (LOCA) and safe shutdown earthquake t
(SSE).
We have reviewed the analytical methods described in B & W Topical Report BAW-10008 (Part 1, Revision 1 - Reactor Internals Stress &
Deflection Due to LOCA and Maximum Hypothetical Earchquake, 6/70).
We' find that an analysis using thes'e methods will provide adequate assurance that the combined stresses and strains in the components of the reactor coolant system and reactor internals will not exceed the allowable design stress and strain limits for the materials of construction as specified in Appendix F to the ASME Boiler and Pressure Vessel Code,Section III.
In addition, the resulting deflections or displacements of any structural elements of the reactor internals will not distort the reactor internals geometry to the extent that core cooling would be impaired.
The assurance of structural integrity of the reactor internals under the postulated SSE and the most severe LOCA conditions pr5vides added confidence that the design can be expected to withstand a spectrum of lesser pipe breaks and seismic loading combinations.
We have concluded that the use of the proposed analytical techniques will result in an acceptable structural design for the WNP-1 reactor internals, and constitutes an acceptable bases for satisfying the applicable requirements of GDC 10.
. 1
?
i 3.9.2 ASME Code Class 2 and 3 Components.
3.9.2.1 Plant Conditions and Design Loading Combinations All Category I systems, components and equipment outside of the reactor coolant pressure boundary will be designed to custain normal loads, anticipated transients, the Operating Basis Earth-quake, and the Safe Shutdown Earthquake within design limits which
~
are consistent with those outlined in Regulatory Guide 1.48 (Design Limits and Loading Conditions, 5/73). The specified design basis combinations of loading as applied to the design of the safety-related ASMI Code Class 2 and 3 pressure-retaining components in seismic Category I systems provide reasonable assurance that in the event (a) of an earthquake affecting the site, or (b) of an upset, emergency, or faulted facility transient occurring during facility operation, the resulting combined stresses imposed on the pressure-retaining components are not expected to exceed the allowable design stress and strain limits for the materials of construction.
Limiting the stresses under such loading combinations provides a conservative basis for the design of the system components to withstand the most adverse combination of loading. events without i
gross loss of structurh1 integrity. The design stress limit specified by the applicant for ASKE Class 2 and 3 pressure vessels, primary membrane plus primary bending stress category, under the l
emergency loading condition (2.25 S where S is the code allowable stress limit for each material at the service temperature) is higher than the corresponding limit provided in the ASME Boiler and Pressure Vessel Code (1.8 S).
The staff will not accept a limit for the t
. conditions described above that is greater than 1.8 S.
With an acceptable resolution of this matter the design load combinations and associated stress and deformation limits specified for ASME Code Class 2 and 3 components will constitute an acceptable basis for design in satisfying the applicable requirements for GDC 1, 2, and 4 and are consistent with those specified in Regulatory Guide 1.48 (Design Limits and Loading Combinations for Seismic Category I Fluid System Components, 5/73).
3.9.2.4 ASME Code Class 2 and 3 Active Component Operability Assurance Program The applicant has agreed to utilize an operability assurance program, in addition to the limits on stress and deformation, to qualify active ASME Class 2 and 3 seismic Category I pumps and valves.
The program will include component testing, or a combination of tests and predictive enalysis supplemented by seismic qualification testing, of motors, operators and component appendages to provide assurance that such components can' withstand postulated seismic l
l loads in combination with other significant loads without loss of l
structural integrity and can perform the " active" function (i.e.,
valve closure or opening or pump operation), when a safe plant i
shutdown is to be effected, or the consequences of an accident are l
l to be mitigated. This commitment to develop and utilize a component operability assurance program satisfactory to the staff constitutes an acceptable basis for implementing the requirements of GDC 1 as related to operability of ASME Code Class 2 and 3 active pumps l
l and valves.
l i
i i
3.9.2.5 Design and Installation Criteria, Pressure Relieving Devices (Class 2)
Design criteria and streso analyses for ASME Code Class 2 safety and relief valves stations will be in accordance with ASME B & PV Code Section III rules as applicable. Dynamic load factors will be calculated taking into account spring-mass characteristics and time history of the fluid forces resulting from the discharge of fluid.
Where more than one valve is mounted on a header, the additiva affect of multiple discharges on total stress will be taken into account.
The criteria used in developing the design and mounting of the safety and relief valves of ASME Class 2 systems provide adequate assurance that, under maximum discharging conditions, the resulting 1
stresses are expected not to exceed the allowable design stress and strain limits for the materials of construction. Limiting the allowable stresses under the loading combinations associated with the actuation of these pressure relief devices will provide a conservativu 'sasis for the design of the system components to with-stand these loads without loss of structural integrity and without impairment of the overpressure protection function.
The criteria used for the design and installation of overpressure relief devices in ASME Class 2 Systems constitute an acceptable j
design basis in meeting the applicable requirements of GDC 1, 2, I
4, 14 and 15.
l
. 3.9.3 Components Not Covered by the ASME Code - Mechanical Design of Control Rod Assemblies and Control Rod Drives The design procedures applied to the doutrol rod assemblies and control rod drives are based upon acceptable methods of analysis, specification of stress allowables, and definition of applied loads.
These design procedures are supported by specialized tests incliding life tests of a prototype of a control rod drive mechanism. Tha use of these methods and technology will provide reasonable asssurance that the control rod assemblies and control' rod drives will be expected to withstand the imposed. loads associated with normal reactor operation, anticipated operational transients, postulated accidents, and seismic evcats without gross loss of their structural integrity or impairment of function. Compliance with these design criteria fulfills the requirements of GDC 2 and 14 as these criteria relate to control rod assemblies, and control rod drives.
D
3.10 Seismic Qualification of Category I Instrumentation and Electrical Equipment Operability of the instrumentation and electrical equipment is essential to assure protective actions in the event of a safe shutdown earthquake (SSE) as nacessary for the operation of engineered safety features and standby power systems. The applicant has referenced IEEE Standard 344(IEEE Guide for Seismic Qualification of Class IE Electric Equipment for Nuclear Power Generating Stations, 1971) for seismic qualification of Category I electrical equipment.
This standard, however, is undergoing a major revision and in accordance with our request, the applicant has committed to our am'plification of that standard to assure that proper consideration will be given to cross-coupling in different directions and multi-frequency excitation. The general program as specified constitutes an acceptable basis for satisfying staff requirements and GDC 2.
A detailed presentation concerning the results of test and analysis will be provided in the FSAR for evaluation during our review of the application for an operating license.
l l
t O
I
0 5.0 Reactor Coolant System and Connected Systems 5.2.1 Design of Reactor Coolant Pressure Boundary Components 5.2.1.5 Design Transients (Plant Conditions, Loading Combinations and Stress Limits)
The design loading combinations specified for ASME Code Class 1 RCPB components have been appropriately categorized with respect to the plant condition identified as Normal, Upset, Emergency or Faulted. The design limits proposed by the applicant for these plant conditions are consistent with the criteria recommended in Regulatory Guide 1.48 (Design Limits and Loading Combinations for Seismic Category I Fluid System Components, May 1973). Use of the criteria recommended in Regulatory Guide 1.48 for the design of the RCPB components will provide reasonable assurance that, (1) in the event an earthquake should occur at the site, or (2) other system upset, emergency or faulted transient should develop the resulting combined stressee imposed on the system components will not exceed the allowable design stresses and strain limits for the materials of construction. Limiting the stresses and strains under such loading combinations will provide an acceptable basis for the design of the system components for the most adverse loading events which have been postulated to occur during the service lifetime without loss of the system's structural integrity. The design load combinations and associated stress and deformation limits specified for ASME Code Class 1 components constitute an acceptable basis for design in satisfying the related requirements of GDC 1, 2 and 4.
. 5.2.1.7 ASME Code Class 1 Active Component Operability Assurance Program The applicant has identified the active valves within the RCPB whose operation is relied upon to safely shutdown the plant and maintain it in a safe condition in the unlikely event of a safe shutdown earthquake or a design basis accident. The applicant has also stated that component operability test programs supplemented by analytical methods, will be developed to provide additional assurance that the capability of these active components will, (1) withstand the imposed loads associated with Normal, Upset, Emergency and Faulted plant conditions without loss of structural integrity, and (2) perform the " active" function (i.e., valve closure or opening) under conditions compar'able to those expected when safe plant shutdown is to be effected or the consequences of an accident are to be mitigated.
The applicant has committed to utilizing a component operability assurance program satisfactory to the staff. We find that the program proposed by the applicant is acceptable and will provide reasonable assurance of valve operability and will meet the requirements of GDC 1 applicable to ASME Code Class 1 active valves.
6
5.2.2.2 Mounting of Pressure-Relieving Devices (Class 1)
The design and installation criteria for pressure relief devices on the RCPB will be in accordance with the acceptable rules of Subsection NB-3500 of the ASME Boiler and Pressure Vessel Code,Section III. The maximum full discharge loads resulting from the opening of ASME Code Class 1 safety and relief valves will be calculated by a time response dynamic analysis of the system.
The criteria used in developing the design and mounting of the safety and relief valves of ASME Code Class 1 systems provides adequate assurance that, under discharging conditions, the resulting str' esses will not exceed the allowable design stress and strain limits for the materials of construction. Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices provides a conservative basis for the design of the system components to withstand these loads without loss of structural integrity and impairment of the
, overpressure protection function. The criteria used for the design and installation of overpressure relief devices in ASME Code Cla,ss 1 Systems constitute an acceptable design basis and meet the requirements of GDC 1, 2, 4, 14 and 15.
6
MECHANICAL ENGINEERING BRANCH SAR REVIEW BIBLIOG,RAPHY The following references are employed b.y MEB personnel in their evaluation of Safety Analysis Reports.
This list of references is presented for each section of the Standard Format the MEB is
~
concerned with.
3.6 Pro tec tion Against Dynamic Ef f ects Associated with the Postulated Rupture of Picing
References:
1 and 10 3.9 Mechanical Systems and Components
References:
1, 2,
3, 4, 5, 6, 7, 8 and 9 3.10 Seismic Design of Category I Instrumentation
Reference:
11 4.2 Reactor Mechanical Design
References:
1 and 2 5.2 Integrity of Reactor Coolant Pressure Boundarv
References:
1, 2,
3, 4, 6 and 8
References:
1.
"ASME Boiler and Pressure Vessel Code" 1971 Edition,Section III and Addenda, The American Society of Mechanical l
Engineers, New York, 1971.
2.
" Criteria of the ASME Bbiler and Pressure Vessel Code for Design by Analysis in Sections III and VIII, Division 2, j
"The American Society of Mechanical Engineers, New York, i
l 1969.
O l
i._
~
\\
~s s
3.
U S A Standard B 16.5-1968, " Steel Pipe Flanges and Flanged Fittings," The American Socie'ty of Mechanical Engineers, New York.
1968.
4.
U S A Standard B 14.9-1964, " Wrought Steel Buttwelding Fitting,"
The American Society of Mechanical Engin'eers, New York, 1964.
5.
U S~A Standard B 31.l.0-1967, " Power Piping," The American Society of Mechanical Engineers, New York, 1967.
~
6.
U S A Standard B 31.7-1969, " Nuclear Power Piping". The American Society of Mechanical Engineers, New York, 1969.
7.
U S A Standard B 36.10-1959, " Wrought-Steel and Wrought-Iron Pipe," The American Society of Mechanical Engineers, New York, 1959.
8.
MSS Standard Practice SP-58, " Pipe Hangers and Supports-Materials, and Design,," Manufacturers Standardization Society, Arlington, Virginia, 1959.
9.
MS'S Standard Practice SP-66, " Pressure-Temperature Ratings for Steel Butt-Welding End Valves," Manufacturers Standardization Society, New York, 1959.
- 10. ANSI.N176 Draft, " Design Basis for Protection of Nuclear Power Plants Against the Effect of Postulated Pipe Rupture" April 5, 1974.
eS e
3 11.
IEEE draft standard 344, "IEEE Guide for Seismic Qualification i
of Class IE Electric Equipment for Nuclear Power Generating Stations," Revision 3 February 15, 1974, American National Standard Institute N41.7.
4 i
1 e
e I
e f
l?
l l
l e
e 6
. -