ML19317D473
| ML19317D473 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/15/1975 |
| From: | Purple R Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| Shared Package | |
| ML19317D475 | List: |
| References | |
| NUDOCS 7912060739 | |
| Download: ML19317D473 (7) | |
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f.0-270 ent ! 0-?fi7 Euke fuser C m peny WIU Pr. Uilliar O. Parker, Jr.
Vice frecident Stear Productico F. C. Por 237f 127 fcuth Church Ttreet Cherlotte,I;crth Caroline 2E242 Ccnticcens FP Ocor.ec l'uclear ftetion Cnits I, 2, and 3 The purpre of this letter is to intcrr. you of c pctcntial refety cucction which han teen reined renarding tre 6ccien of recctor rreccure veccc]
sigoort ryrterr for cressurized veter reactors (M. 's).
T On l'ey 7,1975 the Frc vec inferrtf by e licentre thtt certain trencient loats en tbc reacter vcesel sucrort rrcrivro that woule recult frm e tortuletet reactor coolant pirae ructure irredictc3y rfjcccrt to tre rccctor wrrel had been urterestirated in their cricined ecsic.n analyser.
It in the tsTc rtaff'r ceirion thot the cuesticn rcleted tc tbc trcatrent cf transient 1cw'r in tk desion of reccter vecsel currett ryr.tcire rty trp3y to otkr N F freilitics, errceir13y ti.cce for W.ich the ((:cjer crelycce W rt 5c hcvc therefore initiete a tyrterctic reviev perforr e. Fore tire coo.
of thic retter to 6eterrine hev thcre Icecs were trVen irto cecount or other nn facilition, en6 vbet, if erv, correcrim r ersures rey tr recuirc(
for crecific freilitics.
':1c reruits of licenece stu63c' rcrcrtcd tr cerc ir(icrte thrt, clthcuch tbc rercinc of rnfety rav te Icre than cricirelly irtcrrr?, tre reacter vercel ru m rt rynter voulf retain rufficicnt rtri.cturel fotrcrity to surpcrt the versel and thet tbc ultirate consecuercer cf thic rettelatc<9 uccifent ubicb eculf zffect the cercrel rttlic cre re verre tPcr cricinr13y ttttc6.
re bmr ret corplcret our infeccreent evolucticn cf therr ntu6 irs. Icsevct, bared cn the rerults of cur evoluctico cf thic rhrneceren to f.etc cnd in l
reccenitico of the Icu prebebilit/ cf tre r.crticu3rr Fir.c recture stich could Icad to afeitionel trencient 3czdn cn the runcrt systern, we cerclude the,t centinuM reacter oceratien era certirec6. 31ccesfrn cf facilities for crerttien cre ccccptchle while we conduct our concric review.
- Ue rcqocct thet ycu review the terien bares for tk reccter versc1 rurecrt svitem for veur facilitics to /eterdire stetrer tFc trrnsf ert loadr l
S descrits in the enciccure were tehn into r.cccunt rpprcrrietely in the
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3-Tmke Iover Ccmny 2-00T 151975 desion. Pleere inferr us cf the rerults of your review within 20 days.
7he ettachtertn to the enc]csure are provided to indicate the inforretion that could te nee <*co, thould we deterrine, on the basic cf your review, ther e rearrecreert of the vertel curiport desicn ir recuired.
Fe ere continuf m to evaluate enn review the rethodolocy fer calculatino the tutecoled bicstbr loads with the nuclear etcer rytter cupplicts.
Ycu chould centret your nuclear steer syrter cuppifer for inferration recertino thene cciculations if recerrery te cerclete your review.
'Ibic recuert for nrperic irforraticri van ecproved by CAC under a. blarket cicarence nurter F-1F0225 (FC072). 7hir clearance cypirer July 31, 1977.
Sincerely, i
Original signed by.
R.A.l'urple rebert t.. Purple, Chief Operating reactors Erench fl Pivision of Feactcr Licensino Trelerure Etatenent cf the Frcbler cc w/ enc]crure:
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Duks Pow 2r Comp:ny October 15, 1975 cc:
Mr. William L. Porter Duke Power Company P. O. Box 2178 422 South Church Street Charlotte, North Carolina 28242 Mr. Troy B. Conner Conner, Hadlock & Knotts 1747 Pennsylvania Avenue, NW Washington, D. C.
20006 Oconee Public Library 201 South Spring Street Walhalla, South Carolina 29691 i
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STATEME?E OF TIE PFCBLEM L
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l In the unlikely event of a WR primary coolant system pipe rupture in the imediate vicinity of the reactor vessel, transient loads criainatina from three principal causes will be exerted on the reactor vessel support system.
%ese are:
1.
Blowdown jet forces at the location of the rupture (reaction forces),
2.
Transient differential pressures in the annular region between the vessel and the shield, and 3.
Transient differential pressures across the core barrel within the reacter vessel.
%e blowdown iet forces are adequately understood and desian procedures are available to account for them.
Both of the " differential pressure" forces, bowever, are three-dimnsional and tine dependent and require sophisticated analytical procedures to translate them into loads actino on the reactor vessel support system. All of the loads are resisted by the inertia and by the support members and restraints of other components of the primary coolant system includino the reactor pressure vescel supports.
ne transient differential pressure actina externally on the reactor vessel is a result of the flow of the blowdown effluent in the reactor cavity. 'Ihe magnitude and the time dependence of the resultina forces depends on the nature end the size of the pipe rupture, the clearance between the vessel and the shield and the size and location cf the vent openings leading from the cavity to the containment as a whole.
For some tine refined analytical methods have been available for calculating these transiet differential pressures (multi-node analyses). 'Ihe results f such analyses indicate that the consecuent loads on the vessel support system calculated by less sephisticated nethods nay net be as conservative as originally intended for earlier desicns. Attachment 1 to this enclosure provides for your information a list of information request.s for which resconses could be needed for a proper assessment of the irpact of the cavity differential pressure on the design adecuacy of the veswi =pport system for a power plant.
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- 'Ihe controllira loads for desian purporcs, however, appear in typical cases to be those associated with the internal dif ferential pressures across the core barrel. The internally cenerated loads are due to a recentary j
l differential pressure which is calculated to exist across the core barrel when the pressure in the reactor annular reaion between the core barrel ar# vessel wall in the vicinity of the ruptured pipe is assured to rapidly decrease to the saturation pressure of the primary coolant due to the outflow of water. Althouch the depressuri7ation wave travels rapidly around the core barrel, there is a finite period of tire during which the pressure in the annular reoion epy> cite the break location is assurced to remain at, er near, the original reactor operatirq pressure. Thus, transient asyneetrical forces are exerted on the core barrel and the vessel wall w' ich ultirately result in transient 1 cads on the support systers. These are the loads which were underestinated by the licensee originally reportino this croblem and which may be underestimated in other cases. They are therefore of ceneric concern to the staff. Attachnent 2 to this enclosure provides for your information a list of information recuests for which responses would be needed for a proper assessment of the inpect that the vessel internal differential pressure, in conjunction with the other concurrent leads, could have en the desion adecuacy of the support syster.
In that there are considerable differences in the reactor support syster desions for various facilities and probably in the design rarains provided by the desianers of older facilities, the underestimation of these " differ-ential pressure" loads may or reay not result in a determinatien that the adecuacy of the vessel support syster for a specific facility is auestion-Since local failures in the vessel supports (such as plastic deformation) able.
do not necesserily lead to the failure of the supports as an inteoral syster, there may be sone limited reactor vessel notion provided that no further sienificant conceauences sould ensue and the eneroency core coolina systers (FCCS) would be able to Wrform their design functions.
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l ATTACK 1E*lT 1 C0tlTAlll!4EllT SYSTEMS BRAllOf I
REQUEST FOR ADDITIONAL IllFORMATI0!l I
I In the unlikely event of a pipe rupture inside major component subcompartments, e
I the initial blowdown 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 perform a compartment multi-node pressure response analysis to provide the following information:
(a) The results of analyses of the differential pressures resulting from hot leg and cold leg (pump suction and discharge) reaccor coolant system pipe ruptures within the reactor cavity and pipe penetrations.
(b)
Describe the nodalization sensitivity study performed to detennine the minimum number of volume nodes required to. conservatively predict the maximum pressure withi,n the. reactor cavity.
The nodalization sensitivity study should include consideration of spatial pressure variation; e.g., pressure variations circumferentially, axially and radially within the rea'ctor cavity.
l (c)
Provide a schematic drawing showing the nodalization of the mattor cavity.
Provide a tabulation of the nodal net free volumes and interconnecting flow path areas.
l (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, to permit verification of the reactor cavity nodalization and vent locations.
(e)
Provide and justify the break type and area used in each analysis.
i (f)
Provide and justify values of vent loss coefficients and/or friction
, factors used to calculate flow between nadal volumes.
When a loss coefficient consists of more than one component, identify each component, its value and the flow area at which the loss coefficient applies.
(g)
Discuss the manner in which movable obstructions to vent flow (such as insulation, ducting, plugs, and seals) were treated.
Provide analytical justification for the removal of such items to obtain vent Provide justification that vent areas will not be partially or area.
completely plugged by displaced objects.
(h)
Provide a table of bloddown mass flow rate and energy release rate as a function of time for the reactor cavity design basis accident.
(i)
Graphically show the pressure (psia) and differential pressure (psi) responses as functions of time for each node.
Discuss theihasis 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,desiga values, for determining the acceptability of the calculated results.)
5
U.O. NUCLEAR REGULATCRY CCMMISSISN Fcbrua ry, 1975 STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION O
SECTION 6.2.1.2 SUBCOMPARTMENT ANALYSIS REVIEW RESPONSIBILITIES e t Primary - Containment Systems Branch (CSB)
Secondary - Mechanical Engineering Branch (MEB)
Core Performance Branch (CPB)
Auxiliary and Power Conversion Systems Branch (APCSB)
I.
AREAS OF REVIEW The CSB reviews the information 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 volame. This pressure transient produces a pressure differential across the walls of the subcompartment which reaches a maximum value generally within the first second after 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 meets the above definition.
The CSB review includes 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 pressure 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 within each subcompartment containing a high energy line and the analytical procedure for predicting the short-term mass and energy release rates.
The CSB review of the subcompartment model includes the basis for the nodalization within each subcompartment, the initial themodynamic conditions within each subcompartment, the nature of each vent flow path considered, and the extent of entrainment assumed in the vent flow mixture. The review may also include an analysis of the dynamic characteristics of components, such as doors, blowout panels, or sand plugs, that must open or be removed to USNRC STANDARD REVIEW PLAN
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I previde a vent flow path, and the methods and results of components tests performed to 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 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 APC58, the CSB evaluates or performs pressure response analyses for subcompartments outside containment.
The MEB 15 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 breaks postulated to occur within subccmpartments (See Standard Review Plan 3.6.2).
II.
ACCEPTANCE CRITERIA 1.
The subcompartment analysis should incorporate the following assumpticas:
a.
Break locations and types should be chosen according to Regulatory Guide 1,46 for subcompartments 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 subcompartment 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 par: meter would result in a less severe blowdown profile.
2.
The analytical approach used to compute the mass and energy release profile will be accepted if both the computer program 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 (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 acceptable model would be to assume air at the maximum allowable temperature, minimum absolute pressure, and zero percent rel-l ative humidity. If the assumed initial atmospheric conditions differ from these, the selected values should be justified.
6.2.1.2-2 4
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Another model that is also acceptable, for a restricted class of subcompartments, in-volves simplifyltg the air model outlined above. For this mooel, the initial atmos-phere within the subcompartment is modeled as a homogeneous water-steam mixture with 8
an average density equivalent to the dry air model. This approach should be limited to subccanpartments that have choked flow within the vents. However, the adequacy of this simplified model for subcompartments having primarily subsonic flow through the vents has not been established.
4 Subcompartment nodalization schemes should be chosen such that there is no substantial pressure gradient within a node, i.e., the nodalization scheme should be verified by a sensitivity study that includes increasing the number of nodes until the peak cal.
culated pressures converge to small resultant changes, 5.
If vent flow paths are used which are not immediately available at the time of pipe rupture, the following criteria apply:
The vent area and resistance as a function of time af ter the break should tu a.
based on a dynamic analysis of the subcompartment pressure response to pipe ruptures.
b.
The validity of the analysis should be supported by experimental data or a testing program should be proposed at the construction permit stage that will support this analysis.
The effects of missiles that may be generated during the transient should be c.
considered in the safety analysis.
6.
The vent flow behavior through all flow paths within the nodalized compartment model should be based on a homogeneous mixture in thermal equilibrium, with the assumption of 100% water entrainment. In addition, the selected vent critical flow correlation should be conservative with respect to available experimental data. Currently accept-able vent critical flow correlations are the " frictionless Moody" with a multiplier of 0.6 for water-steam mixtures, and the thermal homogeneous equilibrium model for air-steam-water mixtures.
7.
At the construction permit stage, a factor of 1.4 should be applied to the peak differential pressure calculated in a manner found acceptable to the CSB for the subcompartment. The calculated pressure multiplied by 1.4 should be considered the design pressure. At the operating license stage, the peak calcult.ted differential pressure should not exceed the design pressure. It is expected that the peak calcu-I lated differential pressure will not be substantially different from that of the construction permit stage. However, improvements in the analytical models or changes in the as-built sub:ompartment may affect the available margir..
III. REVIEW PROCEDURES The procedures described below are followed for the subcompartment analysis review. The reviewer selects and emphasizes material from these procedures as may be appropriate for a
6.2.1.2-3
a particular case. Portions of the review e.ay lie carried out on a generic basis or by adopting the results of previous reviews of plants with essentially the same subcompartment and high pressure piping design.
The CSB reviews the initial conditions selected for determining the mass and energy release rate to the subcompartments. These values are compared to the spectrum of allowable opera-ting conditions for the plant. The CBS will ascertain the adequacy of the assumed conditions based on this review.
The CSB confirms with the MEB the validity of the applicant's analysis of subcompartments containing high energy lines and pnstulated pipe break locations, using elevation and plan drawings of the containment showing the routing of lines containing high energy fluids. The CSB determines that an appropriate reference case for subcompartment analysis has been identified, in the event a pipe break other than a double-ended pipe rupture is postulated by the applicant, the MEB will evaluate the applicant's justification for assuming a limited displacement pipe break.
The CSB may perform confirmatory analyses of the blowdown mass and energy profiles within a subcompartment. The analysis is done using the RELAP3 computer program (See Reference 21 for a description of this code). The purpose of the analysis is to confirm the predic-tions of the mass and energy release rates appearing in the safety analysis report, and to confirm that an appropriate break location has been considered in this analysis. The use of RELAP3 will continue until the RELAP4 computer code has been approved by the staff as an acceptable blowdown code. At that time, the CSB will replace RELAP3 with RELAP4 for all subsequent analyses.
The CSB determines the adequacy of the information in the safety analysis report regarding subcompartment volumes, vent areas, and vent resistances. If a subcompartment must rely on doors, blowout panels, or equivalent devices to increase vent areas, the CSB reviews the analyses and testing programs that substantiate their use.
The CSB reviews the nodalization of each subcompartment to determine the adequacy of the calculational model. As necessary, CSB performs iterative nodalization studies for sub-compartments to confirm that sufficient nodes have been included in the model.
The CSB compares the initial subcompartment air pressure, temperature, and humidity condi-tions to the criteria of II, above, to assure that conservative conditions were selected.
The CSB reviews the bases, correlatior.5, and computer codes used to predict subsonic and i
sonic vent flow behavior and the capability of the code to model compressible and un-compressible flow. The bases should include comparisons of the correlations to both experimental datt and recognized alternate correlations that have been accepted by the staff.
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- Using the nodalization of each subwartment as specified in the saf"ty analysis report, the C58 perfoms analyses using one of several available computer programs to determine the adequacy of the calculated peak differential pressure. The computer program used will depend un %) the subcompartment under review as well as the flow regime. At the present time, the two programs used by the CSB are RELAP3 (Ref. 21) and CONTEMPT-LT (Refs. 7, 8, and 9). A multi-volume computer code is currently under development.
At the construction permit stage, the C58 will ascertain that the subcompartment design pressures include appropriate margins above the calculated values, as given in II, above.
IV. EVALUATION FINDINGS The conclusions reached on completion of the review of this section are presented in Standard Review Plan 6.2.1.
V.
REFERENCES The references for this plan are those listed in Standard Review Plan 6.2.1 together with the following:
la. Regulatory Guide 1.46, " Protection Against Pipe Whip Inside Containment."
2a. Standard Review Plan 3.6.2, " Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping," and attached Branch Technical Position MEB 3-1, " postulated Break and Leakage Locations in Fluid System Piping Outside Containment."
6.2.1.2-5 I
ATTACH!!E!lT 2 MECHANICAL EllGINEERING BRAflCH REQUEST FOR ADDITIONAL INFORMATION l
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Recent analyses have shown that reactor pressure vessel supports may be subjected to previously underestimated lateral loads under the conditions that would exist if an instantaneous double ended break is postulated in the reactor vessel cold leg pipe at the vessel nozzle.
It is therefore necessary to reassess the capability of the reactor coolant system supports to limit the calculated motion of the reactor vessel during a postulated cold leg break within bounds necessary to assure a high probability that the reactor could be brought safely to a cold shutdown condition.
The following information is required ~for purposes of making the necessary reassessment of the reactor vessel supports:
1.
Provide engineering drawings of the reactor support system sufficient to show the geometry of all principle elements and materials of con-struction.
2,'.
Specify the detail design loads used in the original design analyses of the reactor supports giving magnitude, direction of application and the basis for each load.
Also provide the calculated maximum stress in each principle element of the support system and the corresponding allowable stresses.
3.
Provide the information requested in 2 above for the RV supports con-sidering a postulated break at the cold leg nozzle.
Include a summary of the analytical methods employed and specifically state the effects of short tem pressure differentials across the core barrel in combination.
-2, with all external loadings calculated to result from the required
, postulate.
This analysis should consider:
(a) limited displacement break areas where applicable (b) consideration of fluid structure interaction (c) use of ac,tual time dep.endent forcing function (d) reactor support stiffness.
4.
If the results of the analyses required by 3 above indicates loads leading to inelastic action in the reactor supports or displacements t
exceeding previous design limits provide an evaluation of the following:
(a)
Yield behavior (ef,fects of possible strain energy buildup) of the material used in the reactor support design and the effect on the loads transmitted to the reactor coolant system and the backup structures to which the reactor coolant system supports are attached.
(b) The adequacy of the reactor coolant system piping, control rod drives, steam generator and pump supports, structures surrounding the reactor coolant system, reactor internals and ECCS piping to assure that the reactor can be safely brought to cold shutdown.
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