ML19312C872
| ML19312C872 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/31/1972 |
| From: | Maccary R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML19312C873 | List: |
| References | |
| NUDOCS 8001130041 | |
| Download: ML19312C872 (15) | |
Text
_
e MAY 31 1372 Richard C. De"oung Annistant Director for LT.'s, Directorate of Licensing OCO:iEE NUCLEAR STATION, IEi1TS 2 AND 3, DOCEET NOS. 50-270 AND 50-287 Adequate responses to the enclosed requent for additional inforention are required before ve can coepicte our review of the subject application. These requests, prepared by the L Mechanical En:;ineering Branch, concern the reactor coolant pressure boundary, reactor internal structures, safety related cechanical syntens, seismic design criteria and pipo whip criteria subuitted in Volumes I throu? IV of the FSAR. This request cupersedes our previous request dated 4-27-72 and reflecto the results of the reevaluation discussed in our letter dated 5-27-72 t
}
The applicant's description of ceic=ic qualification testing for instrumentation and equip =ent (Section 7A.2) appears to be extracted from B & W Topical Report 1
BAV-10003, " Qualification Testing of Protection Systes Instrenentation" (March 1971). The additional it. formation required to complete our review of thia report was forwarded in our requent of March 13, 1972 for the nree Mile Island Station Unit No.1, Docket No. ;0-289. Our review of the Oconee 2/3 application can therefore be expedited if the applicant will agree to reference BAW-10003 and provide the requested information.
The applicant has referenced (Section 14) B & W Topical Report BAU-10008, " Reactor Internals Stress and Deflections due to Loss-of-Coolant Accident and Maximun Eypothetical Earthquaka" (June 1970). The acetions of the report applicable to Oconee are currently under review by the MEB and additional information may be required prior to completion of this review.
% %*2f tiy R. R. *.f a ca r y C###
R. R. Maccary, Assistant Director for Reactor Safety Directorate of Licensing cc w/encit E. J. Bloch, DDR S. H. Uanauer, DRTA J. M. Hendrie, L
}
g F. Schroeder, L pa
)
!1 A. Schwencer, L f
Lu D. F. Lange, L I. A. Peltier, L V. H. Davison, L
_J S. N. Eou, L g
P K. K. Kapur, L J. P. Enit;ht, L
. LIME..
..L2E.. -.
_L E.... -.
LIME-
-LtRS H. E'." Tram.er,--I, -
?
cc54/trWels-- -
H... Aramert; m..S..Hou......
..K...Kapur...... D... Lan ge ---.._ --R.- Maccay ;
A. G
- hso, ron. N~~9@6~UMOA83d. L....
. 6/...].72............ 6 /..../ 7 2...
.. 6/..../72..._.
6f...172..... -6/ /72' EIno
- -'5-"*~'
e 0 01130 6Y/
g
(
i REQUEST FOR ADDITIONAL INF0PJ'.ATION OCONEE NUCLEAR STATION UNITS 2 & 3 DOCKET NOS. 50-270/287 3.6 CRITERIA FOR PROTECTION AGAINST DYNA?iIC EFFECTS ASSOCIATED WITH A LOCA 1.
Provide a more detailed description of the measures that have been used to assure that the contcinnent liner and all essential equip-ment within the contain=ent, including co=ponents of the primary and secondary coolant systems, engineered safety features, and equipment supports, have been adequately protected against blev-1 I
down jet forces, and pipe whip resulting from a loss-of-coolant k
accident. The descriptien should include:
^
Pipe restraint design require =ents to prevent pipe whip i= pact.
a.
b.
The features provided to shield vital equipment from pipe whip.
c.
The neasures taken to physically separate piping and other components of redundant engineered safety features.
d.
A description of the type of pipe whip restraints and the location of all restraints.
2 Describe the dynamic system analysis eethods and procedures that were used to confirm the structural design adequacy of the reactor coolant system (unaf fected loop) cad the reactor internal loadings.
The following information should be included:
Of
(.
the locations of the postulated double ended pipe rupture on a.
which dynamic analyses were based.
b.
the rupture type (s), such as circumferential and/or longitudinal break (s), for each postulated rupture location.
the description of the forcing functions used for the pipe whip c.
dynamic analyses. The function should include direction, rise tice, -agnitude, duration and initial conditions. The forcing function should adequately represent the jet stream dynamics and the system pressure differences.
i d.
a description of the mathematical =odel used for the dynamic analysis.
analyses perfor=ed to demonstrate that unrestrained cotion of e.
ruptured lines will not sever adjacent icpacted piping or pierce impacted areas of containment liner.
1 f
1 I
I l
I O
O;
,1_ d :
. 3.7.2 SEISMIC SYSTEM ANALYSIS 1.
Confirm the validity of a fixed base assumption in the mathematical models for the dynamic system analyses by providing summary analyt-ical results that indicate that the rocking and translational responses are insignificant. A brief description should be included of the cethod, mathe=atical =odel and damping values (rocking verti-i cal, translation and torsion) that have been used to consider the soil-structure interaction.
l 2
Describe the cethod employed to consider the torsional modes of I
vibration in the seismic analysis of the Category I building structures.
If static factors are used to account for torsional accelerations in the seismic design of Category I structures, justify this procedure in lieu of a corbin2d vertical, horizontal, and torsional =ulti= ass system dyna =ic analysis.
3.
The use of both the modal analysis response spectrum and time history rethods provides a check on the respenses at selected 4
points in the station structure.
Submit the responses obtained from both of these =ethods at selected points in the Category I structure to provide the basis for checking the seismic system l
analysis.
)
4.
Provide the dynsmic eethods and procedures used to determine Category I structure overturning moments.
Include a description i
...j h I
33 s
l
-oy, p__
~.
g g
i of the procedures used to account for soil reactions and vertical 4,
carthquake effects.
5.
Provide the analysis procedure followed to account for the da mping in different elements of the rodel of a coupled system Include the criteria used to account for co=posite damping in a coupled system with different structural elements.
4 9
I i,
i-1 1
l i
I l
l I
1 i
nO b
g g y+,
_ _ _ _ _ _ 3.7.3 SEISMIC SUBSYSTE'i ANALYSIS 1
Provide the criteria for combining modal responses (shears, moocnts, stresses, deflections. and/or accelerations) when modal frequencies are closely spaced and a response spectrum todal analysis method is used.
2 With respect to Category I piping buried or otherwisc located out-side of the containment structure, describe the seisnic design criteria employed to assure that allowable piping and structural stresses are not exceeded due to differential movecent at support i
j points, at containment penetrations, and at entry points into other structures.
3 Describe the evaluation performed to determine seisnic induced ef fects of Category II piping systems on Category I piping.
4.
Provide the criteria employed to determine the field location of seismic supports and restraints for Category I piping, piping system components, and equipment, including placceent of snubbers and danpers. Describe the procedures followed to assure that the field location and characteristics of these supports and restraining devices are consistent with the assumptions made in the dynamic analyses of the system.
i 3.7.4 CRITERIA FOR SEISMIC INSTRL71ENTATION Provide the following information with respect to the us'e of the seismic instrumentation for the facility:
1.
Discuss the seismic instrumentation provided and compare the proposed seismic instrumentation program with that described in AEC Safety Guide 12, "Instru=entation for Earthquakes." Submit the basis and justification for elements of the proposed procram that differ substantially from Safety Guide 12, 2.
Provide a description of the seismic instrumentation such as peak recording accelerographs and peak deflection recorders, that will be installed in selected Catecory I (Class 1 Seismic) structures and on selected Category I (Class 1 Seismic) components.
Include the basis for selection of these structures and components, the basis for location of the instrumentation, and the extent to which this instrumentation will be employed to verify the seismic analyses following a seismic event.
3.
Describe the provisions that will be used to signal the control room operator the value of the peak acceleration level experienced in the tendon access gallcry of the reactor containment structure to the control room operator within a few minutes af ter the carth-quake.
Include the basis for establishing the predetermined values for activating the readout of the accelerograph to the control room operator.
D**D
- D Wy
4 Provide the criteria and procedures that will be used to comparc measured responses of Category I (Class 1 Seismic) structures and the selected components in the event of an earthquake with the results of the system dynamic analyses.
Include consideration of-different underlying soil conditions or unique structural dynamic characteristics that may produce different dynamic responses of Category I (Class 1 Seismic) structures at the site.
- 9%
i d.
p e
i i
t s
s i
D""2D *AL3Ju@,[a FWi g
i
~
t 1
i
8-3.7.5 SEISMIC DESIGN CONTROL MEASURES Describo the design control neasures (as specified in Appendix B -
" Quality Assurance Criteria for Nuclear Power
~
Plants") implemented to, assure that appropriate seismic input data (as derived from seismic system and subsystem analyses including any necessary feedback from such analyses) are correctly specified to the manufacturer of Category I components and equipment to constructors of other Category I structures and systems.
The i
responsible design groups or organizatiens that will verify the adequacy and validity of the analyses and tests employed by manu-l facturers of Category I couponents cnd equipecnt and constructors of Category I structures and systens should be identified.
A description j
of the review proced'res ceployed by each group or organiration should u
j be included.
4 O
w i
i i
1 1
or r-w r
=
,ee-
3.9.1 DYNAMIC SYSTEM ANALYSIS AND TESTING 1
Paragraph 1701.5.4 of the ANSI B31.7 Nucicar Power Piping Code requires that piping shall be supported to prevent excessive vibration under startup and initial operating conditions.
Submit a discussion of your vibration operational test program which will be used to verify that the piping and piping restraints within the reactor coolant pressure boundary have been designed to withstand dynamic cf fects due to valve closures, pump trips, etc.
Provide a list of the transient conditions and the associated actions (pump trips, valve actuations, etc.) that will be used in the vibration oper1tional test program te verify the integrity of the system.
Include those trans'ients introduced in systers other than the reactor coolant pressure boundary that will result in significant vibration respense of reactor coolant pressure boundary systems and components.
2.
Discuss the testing procedures used in the design of Category I mec.'anical equipment such as fans, pumps, drives, valve operators and heat es. changer tube bundles to withstand seismic, accident and operaticnal vibratory locding conditions, including the manner in which the methods and procedures crployed will consider the frequency spectra and amplitudes calculated to exist at the equipment supports."
Where tests or analyses do not include evaluation of the equipment in the operating mode, describe the bases for assuring that this D~~
'Y M
w@
a
/
A i
4 e
e.
_-ng -
i equiprxnt will function when subjected to seismic and accident Irmd ings.
3.
Provide a brief description of the dynamic system analysis methods and procedures used to determine dynamic responses of reactor internals and associated Class I components of the reactor coolant pressure boundary (e.g., analyses and tests).
The discussion should include the preoperational test program elcments described in Safety Guide 20, Vibration Measure =ents on Reactor Internals.
In the event elenents of the program differ substantially from the require-ments of Safety Guide 20, the basis and justification for these differences should be presented.
4 Provide a discussion of the preoperational analysis and testing results that will be used to augment the LCCA dynamic analysis methods and procedures, i.e., barrel ring and beam modes, guide tube respenses, water mass and compliance effcets, damping factor selection, etc.
s i
11 -
3.9.2 ASME CODE CLASS 2 AMD 3 COMPONE'iTS 1
The FSAR states that faulted operating condition categories have been applied to certain reactor coolant system components.
Identify any other co=ponents or systems that are not a part of the reactor coolant pressure boundary for which the design stress limits associated with' faulted conditions were applied.
If faulted conditions are used for such cases, then provids justification for applying such conditions, including the bases for the loading con-ditions and combinations, and associated design stress limits which i
were applied.
t In addition, for all components and systens co= parable to ASME Code Class 2 and 3, provide the design condition categories (normal, upset or energency), the casociated design loading corbinatiens and the design stress limits which will be applied for each leading combination.
This inforestion may be submitted in tabular form as suggested below:
System Desien Londing Design Conditicn Desien and/or Combinations Categories (Nerral, Stress Component Upset, or Eccrgency)
Limits 2.
If any design stress limits allow inelastic deformation (or are comparable to the faulted condition limits defined in ASME Section 111 for Class 1 conponents) then provide the bases for the use of inelastic design limits by deconstrating that the component will maintain its functional or structural integrity under the s'pecified J
_m design loading combination.
Include a brief description of the methods and design procedures that were used in such ases.
3.
Describe the design and installation criteria applicable to the counting of the pressure-relieving devices (safety valves and relief valves) for the overpressure protection of systems with Class 2 corponents.
In particular, specify the desir,n criteria used to take into account full discharge leads (i.e., thrust, bending, torsion) imposed on valves and on connected piping in the event all the valves are required to discharge.
Indicate the provisiont cade to acconrodate these loads.
,5.2.1 DESIGN CRITERIA, METilCDS AND PROCEDURES (REACTOR OOj. ANT PRESSURE BOIT:iDARY) 1 Categorize all transients or combinations of transients listed in Table 4-8 of the FSAR with respect to the conditions identified as "norr.s1", " upset", "energency", or " faulted" as defined in the ASME Section III.iuclear Component CcJe.
In addition, provide the design loading combinations and the associated stress or deformation criteria.
2.
Table 4-20 of the FSAR includes faulted crindition stress limits for pump casings. Describe the criteria ceployed to assure that active *
{
co=ponents will function as designed in the event of a postulated piping rupture (Faulted Condition) within the reactor coolant pressure boundary (e.c., allevable stress limits established at or near the yicId stress calculated on an elastic basis).
Where e=pirical methods (tests) are c= ployed, provide a su==ary description of test methods, loading techniques and results obtained including the bases for 1
extrapolations to components larger or smaller than those tested.
3.
The design criteria which was used to account for full d,ischarge loads (i.e., thrust, bending, torsion) i= posed on safety and relief valves and connected piping in the event all valves are required to discharge, including the provision made to accon odate these loads should be specified.
- Active co=ponents of a fluid system (e.g., valves, pumps) are those whose oper-ability is relied upon to perform a safety function such as safe shutdcwn of the reactor or citi;stica cf the conse:;uences of a postulated pipe break in the reactor j
coolant pressure boundary.
i 4
%Q$ 'Okku
I 5.2.2 OVERPRESSURIZATION PROTECTION 1.
To facilitate review of the bases for the pressure relieving capacity of the reactor coolant pressure boundary, submit (as an cppendix to the FSAR) the " Report on Overpressure Protection" that has been pre;,ared in accordance with the requirements of l
the ASME Section III Nuclear Power Plant Conponents Code or, i
if the report is not available, indicate the approxinate date for submission.
In the event the report is not expected to be i
available until either the Operating License review or late in d
1 I
the construction schedule for the plant, provide in the FSAR the bases and analytical approach (e.g., prelininary analyses) being utilized to establish the overpressure relieving capacity 4
required for the reactor coolant pressure boundary.
1
+
J i
'y 4
?
l 1
i e
s S
l QD
' W d\\
e e JB e Ju.AJ\\.
A D "90
.:2
,,r..
-,-t
_,r--