ML19319E092
| ML19319E092 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/09/1976 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8003310601 | |
| Download: ML19319E092 (7) | |
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DISTRIBUTION:
L PDR Dross Docket Fil CTrannuell Docket he. - $ 5-312 ORB #4 Rdg
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KRGoller TJCarter Sacramento: Municipal Utility l'istrict.
' Mr. J. 7.' _.Em t t imoe
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S A"CHO SECO NUCLI.AP.CD8MATIdC STATIGF LOn Octhber;16, 1973, we,.Inforced you 'of a potential safety question fwhich'has'been. raised regarding 'the. design of reactor pressure vessel.
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_g(suppossys;thei'Nirequeste(thatCyou revies thefdesign bases for
' lthe: reattdEvesseF adpport syst.e a for' your' facility to determine
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c.!J,whethe r "t hei trans lent load.s. described,~'in the Anclosure tonour letter
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" '*ere' apprppriMely',tahn into account 'in' the desI;n.
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' dif ferent'I AIL pressures ~in: the; annular rc;; ion' between the. react or vessuL7and the cavity-shield wall-a'ndiaerossTthe core barrel' vere E
not'. coridide red ' in 't he siteort des irn.'.
In;our letter'of' October 16, 1975i we indicated that on t.be basis
. o f your( init ial review,' = reassesswent' of th'e-vessel support desi n s
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.mightlbe} required.'Mk'e1.have now determined that such a reassessrent
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'x'.C. vendors and ', rob, ably. aware, Ee have be.endiscussinig.
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- <...:you are,p vari.-ous architect /en.gincer firds'the:,senerie aspects r
'.of f this' proSlersJJ Should you cobteiplate ut'iliziedorkanizations Me'ther tha,n.'your; PVit vendor for calcul.tLion of t% sub-cooled for-th. henefit.of a
" InternalT. loads, we su nest'youfcontact un
- brief'refiew of;our seneric discuysions to dat.e.
We icill conticoe these generic, discussions witn the vendors an.1 architect / crc,ineers, but such discussions are. net intended to pace your evaluation
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- this~ concern nor to elininate the possibility that we may have
-addit ional quest ions rdgarding' your evaluat ion af ter suf.mit tal.
While f the' esphasis -riven in this letter deal s with. the reactor veseel..
" cavity, for your informatio'n ant' guidance our generic revisv ".ay
' consider Loiher arcos"in the nuclear' stw( cupply system and furiher evaluat. ion.say'be required.
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l't ility nistriut Nd ei i> le a s.- i-fo r-us within inrs citer receint of LSis letter of ymr ichedule fer nrovifin-ns your -valuatfor of the 33equsey of the l o a '!s pres %re vessel so; porta when the sub enoted c.ticulated and t a' en into account in 1re n
you +1etur+ine best a, ener wh iel' represents these phenemena.
Sour evnluation should include the arsvurs to the attached i n fo r r-A t i on.
request for additionst This request for renerie number E-If402 2 5 '(R0072).inforration was approved by GAO blank clearance This clowrance expires Juiy 31, 1977.
Sincerely, O'l@*@51eedby t
/ohert W.
Teid, Chief Operatin.: Teactors Branch 44 Division of Opurnt in; 11eactors P.r,c l oattre t..
3eques t for. Addit.f ocal In fo rma t ion '.
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Dav.iet S. "ap..
s lan, Secretary an.1 Cene ra 1. Couns ee I fa01 3 Street Post Of fice Bor 13430
'.?c r a ne r t o, Califorqio " 5 l l
'hisiness and Municipal Pecart:ent 5,1cramento City-County I.ibrary
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- REOUEST FOR ADDITIONAL IflFOR ttTI0ft Recent analyses have shown that reactor pressure vessel supports may be subjected to previously underestimated lateral loads under the conditions that result from the postulation of design basis ruptures of the reactor coolant piping at the reactor vessel nozzles.
It-is therefore necessary to reassess the capability of the reactor coolant system supports to assure that the calculated motion of the reactor vessel under~the most severe design basis pipe rupture condition will be within the bounds-necessary to assure a high probability that the reactor can be brought safely to a cold shutdoan condition.
The following information should be included in your reassessment of the reactor vessel supports and reactor cavity structure.
1.
Provide engineering drawings of the reactor support system sufficient' to show the geometry of all principle elements and materials of construction.
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 considering a pos.tulated break at the design basis locatio'n that results in the most severe loading condition for.the reactor pressure vessel supports.
Include
'l a summary of the analytical methods employed and specifically state the effects of asymmetric pressure differentials across the core barrel in combination with all external loadings including asymmetric cavity pressurization calculated to result frcm the required postulate.
This analysis should consider:
(a) limited displacement break areas where applicable
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(b) consideration of fl.uid structure interaction (c) use of actual time dependent 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 exceeding previous design limits provide an evaluation of the following:
(a)
Inelastic behavior (including strain hardening) of the material used in the reactor support design and the effect on the load transmitted to the reactor coolant system and the backup structures to which the reactor coolant system supports are attached.
5.
Address the adequacy of tfie reactor coolant system piping, control rod drives, steam generator and pump supports, structures surrounding the reactor coolant system, [ core support structures, fuel assemblies, other reactor internals....] and ECCS piping for both the elastic and/or inalastic analyses to-assure that the reactor can be safely brought to cold shutdown.
For each item include the method'of f
analysis, the structural and hydraulic ccmputer codes employed, drawings of the models employed and ccmparisons of the calculated to allowable stresses and strains or deflections with a basis for the allowable values.
The compartment multi-node pressure response analysis should include the following information:
6.
The results of analyses ~of the differential pressures resulting from hot leg and cold leg (pump suction and discharge) reactor coolant system pipe ruptures within the reactor cavity and pipe penetrations.
7.
Describe the nadalization sensitivity study performed to determine the minimum number of volume nodes required to conservatively predict the maximum pressure within the reactor cavity.
The nodalization sensitivity study should include consideration of spatial pressure variation; e.g., pressure variations circumferentially, axially and radially within the reactor cavity.
8.
Provide a schematic drawing showing the nodalization of the reactor cavity.
Provide a tabulation of the nadal net free volumes and interconnecting flow path areas.
9.
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.
.- 10.
Provide and jus'tify the break type and area used in each analysis.
11.
Provide and justify values of vent loss coefficients and/or friction factors used to calculate flow between ncdal volumes.
When a lo,ss coefficient consists of more than one compone.nt, identify each component, its value and the flow area at which the loss coefficient applies.
12.
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 area.
Provide justification that vent areas will not be partially or completely plugged by displaced objects.
13.
Provide a table of blowdownmass flow rate and energy release rate as a function of time for the reactor cavity design basis accident.
14.
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.
15.
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 dif'ferential pressure is unifonnly* applied to the reactor cavity or whether it is spatially varied.
In order to review the methods employed to compute the asymet'rical pressure differences across the core support barrel during the subcooled portion of the blowdown analysis, the following information is requested:
16.
A complete description of the hydraulic code (s) used including the
l I development of the equations being solved, the assu ptions and simplifications used to solve the cquations, the limitations resulting from these assumptions and simplifications and the numerical methods used to solve the final set of equations.
17.
In support of the hydraulic code (s) used provide comparisons with the code (s) to cpplic:ble experim:ntl tests, including the following:
(a). CSE tests B-63 cnd B-75 (b). LOFT test L1-2 (c).SemiscaletestsS-02-6 ands-02-8 The models developed should be based on tha assu ption_s preposed for the analysis of a P'n'R.-
18.
Provide a detailed descriptio.n of the model proposed for your plant and include a listing of the input data used and a time zero edit.
Identify the assumptions used in developing the model, specifically the treatment of area, length and volume.
19.
Typically the current generation of hydraulic subcooled blowdown analysis codes solve the one-dimensional conservation equations.
However, they are used to model the multi-dimensional aspects of the reactor system (i.e. the downcomer annulus region).
Provide justification for the use of the code (s) to model multi-dimensional regions, including the equivalent representation of the region as modelled by the code (s).