ML19319E042
| ML19319E042 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/17/1975 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8003310509 | |
| Download: ML19319E042 (6) | |
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OCT ] 71;7~,
Docket No.80-312 4
i Sacra::: ento thinicipal titility District ATTW: Mr. E. K. Davis s
General !!anager 6201'S Street
~ Post Office Box 15830 Sacramento, California 95813 Gentlenen:
Py our letter to you dated October 16, 1975, regarding the design of reactor pressure vessel support sy.< tens for pressuri cd water reactors, t.'e inadvertently left out the enclosure. Transmitted herewith is the enclosure to that Ictter.
Sincerely, Robert W. Reid, Chief Operating nesetors Branch #4 1
Division of Reactor Licensing
Enclosure:
DISTRIBtJIION Statencnt of the Peoblen Docket File NRC PDR cc v/ enclosure:
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October 17, 1975 Da id S. Kaplan, Secretary and cc:
General Counsel 6201 S Street Post Office Box 15830 D g g
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EEG CSCPE STATEPENI CF THE PPCBLFM In the unlikely event of a WE primary coolant system pipe rupture in the irrediate vicinity of the reacter vessel, transient loads crioiratino from three principal cauces will be exerted on the reactor vessel supocrt system.
1 These are:
Blowdown jet forces at the location of the rupture (reaction forces),
1.
Transient differential pressures in the annular region between the vessel 2.
and the shield, and Transient differential pressures across the core Perrel within the reacter 3.
vessel.
The blowdown jet forces are edecuately understood and 6esion precedures are available to account for them. Both of the " differential pressure" forces, however, are three-dimensional and tine dependent end recuire sephisticated analytical procedures to translate them into loads actino on the reactor All of the loads are resisted by the inertia and vesnel support system.
by the support members and restraints of other components of the primary ecolant system includino the reactor pressure vessel supports.
The transient differential pressure actino externally en the reactor vessel
'Ihe is a result of the flow of tre blowdown effluent in the reactor cavity.
l maonitude and the time dependence of the resultino forces depends on the nature end the size cf the pipe rupture, tbc clearance between the vessel and tbe shiel6 and the size end location of the vent openings leadino from For scre tire refined analytical
-o the cavity to ti e containrent as a whole.
- i rethods have been available for calculatina these transient differential
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'Ibe results of such analyses indicate pressures (gulti-node analyses).
that the consecuent loads on the vessel support syster calculated by less sephisticated nethods nay net te es conservative as cricinally intended for to this enclosure prcvides for your information earlier desiens.
a list of information recuests for which respcnses could be needed for a proper assessrent of the impact of the cavity differential cressure en the desion adecuacy cf the vessel support syster for e power plcnt.
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- The contro11 ira loads for desian purposes, however, appear in typical cases to be those associated with the internal differential pressures across the core barrel. The internally cenerated loads are due to a rcrentary differential pressure which is calculated te exist across the core bcrrel
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when the pressure in the reactor annular recion between the core barrel and vessel well in the vicinity of the ruptured pipe is assured to rapidly decrease to the saturation pressure of the primary coolant due to the cutflow of water. Althouab the depressurization wave travels rapidly around the core barrel, there is a finite period of tire during whidi the pressure in the annular region opposite the break Iccation is assured to rerain at, or near, the original reactor cperating pressure. Thus, transient asymretrical forces are exerted on the core barrel and the vessel wall which ultimately result in transient leads on the support systers. These are the loads which were underestimated by the licensee originally reporting this probler end whidi ray be underestimated in other cases. They are therefore of generic concern to the staff. Attachnent 2 to this enc 1csure provides for your information a list of inforration reauests for which responses would be needed for a procer assessment of the irpact that the vessel internal differential pressure, in conjunction with the other concurrent Icads, could have en E
the desion adeouacy of the support syster.
In that there are considerable differences in the. reactor support syster desions for various facilities and probably in the design rargirs provided by the desioners of older facilities, the underestimation of these " differ-ential pressure" loads rey or ray not result in a deterrinatien thet the edecuacy of the versel support syster for a specific facility is cuestion-able.
Since local failures in the vessel supports (such as plastic deforration) do not necessarily lead to the failure of the supports as an intecral syster, there ray be sore lirited reactor vessel notion provided that no furtber sionificant consecuences would ensue and the energency core coolina systers (ECCS) would be able to perforra their design functions.
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ATTACir1EllT 1 C0tlTAlflMEllT SYSTEMS BRAllCH REQUEST FOR ADDITIONAL IllFORMATIO!1
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In the unlikely event of a pipe rupture.inside major component subcompartments, the initial bicwdown transient would lead to non-uniform pressure loadings on both-the_ structures and enclosed components.
To assure the integrity of these design features, we request that you pefform a compartment multi-node pressure response analysis to provide the following information:
(a) The results of analyses of the differential pressures resulting fromhotlegandcoldleg(pumpsuctionanddischarge)reactorcoolant system pipe ruptures within the reactor cavity and pipe penetrations.
(b)
Describe the nodalization sensitivity study performed to determine the minicum number of volume nodes required to conservatively predict the maximum pressure within the reactor cavity.
The nodalization sensitivity study should include consideration of
- spatici pressure variation; e.g., pre sure variations circumferentially, axially cnd radially within the rea'ctor cavity.
(c)
Provide a schematic drawing showing the nodalization of the reactor cavity.
Provide a tabulation of the nodal net free volumes and interconnecting flow path areas.
-(d)
Provide sufficiently detailed plan and section drawings for several views showing the arrangement of the reactor cavity structure, reactor vessel, piping,'and other major obstructions, and vent areas,
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to permit verification.of the reactor cavity nodalization and vent
-l ocati c r.s.
(e)
Provide and justify the break type and area used in each analysis.
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(f)
Provide and justify values of vent loss coefficients and/or frictka
, factors used to calculate flow between nodal volumes.
When a loss coefficient consists of more than one component, identify each 3.,
component, its value and the flow area at which the loss coefficient applies.
(g)
Discuss the manner in which movable obstructions to vent flow (suchasinsulation, ducting, plugs,andseals)weretreated.
Provide
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analytical justification for the removal of such items to obtain vent area.
Provide justification that vent areas will not be partially or completely plugged by displaced objects.
(h)
Provide a table of blowdown mass flow rate and energy release rate as
, a function of time for the reactor cavity design basis accident.
i (i) Graphically show the pressure (psia) and differential pressure (psi) responses as functions of time for each node.
Discuss the basis for establishing the differential pressures.
(j) Provide the peak calculated differential pressure and time of peak pressure.for each node, and the design differential pressure (s) for the reactor cavity.
Discuss whether the design differential pressure is uniformly applied to the reactor cavity or whether it is spatially varied.
(Standard Review Plan 6.2.1.2, Subcompartment Analysis attached, provides ' additional guidance in establishing acceptable design values, for determining the acceptability of the calculated results.)
U.S. NUCLEAR REGULATORY COMMISSION February, 1975 STANDARD REVIEW PLAN OFFICE OF NUCLE'AR REACTOR REGULATION SECTION 6.2.1.2 SUBCOMPARTME..( ANALYSIS REVIEW RESPONSIBILITIES Primary - Containment Systems Branch (CSB)
Secoridary - Mechanical Engineering Branch (MEB)
Core Performaryce Branch (CPB)
Auxiliary and Power Conversion Systems Branch (APCSB)
I.
AREAS OF REVIEW The CSB reviews the infonnation presented by the applicant in the safety analysis report concerning the determination of the design differential pressure values for containment sub-compartments. A subcompartment is defined as any fully or partially enclosed volume within the primary containment that houses high energy piping and would limit the flow of fluid to the main containment volume in the event of a postulated pipe rupture within this volume.
A short-term pressure pulse would exist inside a containment subcompartment following a pipe rupture within this volume. This pressure transient produces a pressure differential across the walls of the subcompartment which reaches a maximum value generally within the first second af ter blowdown begins. The magnitude of the peak value is a function of several parameters, which include blowdown mass and energy release rates, subcompartment volume, vent area, and vent flow behavior. A transient differential pressure response analysis should be provided for each subcompartment or group of subcompartments that rueets the above definition, o
The CSB review encludes the manner in which the mass and energy release rate into the break compartment were determined, nodalization of subcompartments, subcompartment vent flow behavior, and subcompartment design oressure margins. This includes a coordinated review effort with the CPB. The CPB is responsible for the adequacy of the blowdown model.
The CSB review of the mass and energy release rates includes the basis for the selection of the pipe break size and location witnin each subcompartment containing a high energy line and the analytical procedure for predicting the short-term mass and energy relear rates.
The CSB review of the subcompartment model includes the basis for the nodalization within each subcompartment, the initial thermodynamic conditions within each subcorpartment, the nature of eac5 vent flow path considered, and the extent of entrainment assumed in the vent flow mixture. The i ics of corponents such as oved to DUPLICATE DOCUMENT
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f provide a vent flow path, and the methods and results of components tests performed to r
demonstrate the validity of these analyses. The analytical procedure to determine the loss
. coefficients for each vent flow path and to predict the vent mass flow rates, including p.
,p, flow correlations used to compute sonic and subsonic flow conditions within a vent, is re-viewed. The design pressure chosen for each subcompartment is also reviewed. On request from the APCSB, the CSB evaluates or performs pressure response analyses for subcompartments outside containment.
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_... Er The MEB is responsible for reviewii the acceptability of the break locations chosen and of the' design criteria and provision.
methods employed to justify limited pipe motion for breats postulated to occur within wbccmpartments (See Standard Review Plan 3.6.2).
ACCEPTANCE CRITERIA 1.
The subcompartment analysis should incorporate the following assumptions:
a.
Break locations and types should be chosen accc7 ding to Regulatory Guide 1.46 for subcomnartments inside containment and to Branch Technical Position MEB 3-1 (attached to Standard Review Plan 3.6.2) for subcompartments outside containment.
An acceptable alternate procedure is to postulate a circumferential double-ended rupture of each high pressure system pipe in the subcompartment.
b.
Of several breaks postulated on the basis of a, above, the break selected as the reference case for subcomp3rtment analysis should yield the highest mass and energy release rates, consistent with the criteria for establishing the break location and area.
c.
The initial plant. operating conditions, such as pressure, temperature, water inventory, and power level, should be selected to yield the maximum blowdown conditions. The selected operating conditions will be acceptable if it can be shown that a change of each parameter would result in a less severe blowdown profile.
2.
Tne analytical approach used to compute.the mass and energy release profile will be accepted lf both the computer progran and volume noding of the piping system are similar to those of an approved emergency core cooling system (ECCS) analysis. The computer programs that are currently acceptable include SATAN-VI (Ref. 24), CRAFT
-j (Ref. 23), CE FLASH-4 (Ref. 25), and RELAP3 (Ref. 21), when a flow multiplier of 1.0 is used with the applicable choked flow correlation. An alternate approach, which is also acceptable, is to assume a constant blowdown profile using the initial conditions with an acceptable choked flow correlation. When RELAp-4 is accepted by the staff as an operational ECCS blowdown code, it will be acceptable for subcompart-ment analyses.
3.
The, initial atmospheric conditions within a subcompartment should be selected to max-imize' the resultant differential pressure. An acceptatile model would be to assume air at the maximum allowable temperatura, minimum absolute pressure, and zero percent rel-q ative humidity. If the assumed initial atmospheric conditions differ from the;e, the selected values should be justified.
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