ML19317G748

Forwards Revised Request for Addl Info Re Damage Potential to NSSS Component Supports,Fuel Assemblies,Control Rod Drives & ECCS Piping Due to Loadings Associated W/Postulated Coolant Sys Piping Breaks

ML19317G748
Person / Time
Site:	Rancho Seco
Issue date:	01/25/1978
From:	Stello V Office of Nuclear Reactor Regulation
To:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
References
GL-78-2, TAC-8597, NUDOCS 8004010578
Download: ML19317G748 (17)

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UNITED STATES

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January 25, 1978

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All PWR 'icensees (Except for Trojan, North Anna, Indian Point 3,

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Beaver Valley and St. Lucie 1)

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In October of 1975, the NRC staff notified each licensee of an operating 1.

PWR facility of a potential safety problem concerning the design of the

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reactor pressure vessel support system.

Those letters requested each

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licensee to review the design basis for the reactor vessel support system

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' for each of its PWR facilities to detennine cather certain transient

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loads., which were described in the enclosure to the letter, had been

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appropriately taken into account in the design. Furthennore, these

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letters indicated that, on the basis of the results of licensees' reviews, a reassessment of the reactor vessel support design for each operating PWR facility may be required.

Licensee responses to that request indicated that these postulated F

asymmetric loads have not been considered in the design basis for

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the reactor vessel support system, reactor internals including th fuel, steam ger.erator supports, pump supports, emergency core cooling

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system (ECCS) lines, reactor coolant system piping, or control rod gT drives.

Subsequently in June 1976, the NRC staff informed each PWR licensee 5

that a reassessment of the reacto-vessel support system design for

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each of its facilities was requiree. While the emphasis of these letters was primarily focused on the aeed to reassess the vessel

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support design for transient differential pressures in the annular

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region between the reactor vessel and the cavity shield wall and

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across the core barrel, we indicated that our generic review may

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extend to other areas in the nuclear steam supply system (NSSS) and

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that further evaluation may be required.

For your infonnation, Enclosure 1 is a summary of the background

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and current status of our review efforts related to this generic

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concern.

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f G =w All PWR Licensees January 25, 1978 Th:.,

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We have now detemined that an assessment of the potential for damage to other NSSS component supports (e.g., steam generator and pump supports), the fuel assemblies, control rod drives, and ECCS piping attached to the reactor coolant system due to loadings i

associated with postulated coolant system piping breaks is required.

Our request for additional infomation transmitted to you in June 1976 has been revised both to clarify our original request and to identify the extension of our concerns to other areas in the NSSS, as -

identified above. A copy of this revised request for additional 57==

infomation is provided as Enclosure 2.

The revised request for additional infomation identifies a requirement -

that your assessment of potential damage to the reactor vessel and other

=J NSSS component supports, reactor vessel, fuel and internals, attached ECCS

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lines and the control rod drives shou!d include consideration of breaks

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both inside and outside of the reactor pressure vessel cavity. This

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assessment should be made for postulated breaks in the reactor cociant

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piping system, (secondary systems are not to be included), including the

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following locations:

?e a) Reactor vessel hot and cold leg nozzle safe ends I

b) Pump discharge nozzle c) Crossover leg j

d) Hot leg at the steam generator (B&W and CE plants only)

A number of licensees, have presented to the NRC staff alternate proposals, other than to conduct a detailed analyses, to resolve this concern.

Based upon our review of these proposals, we have concluded that these alternative

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proposals do not establish an acceptable basis for long tem operation m;.;. =

without a detailed assessment of the risk resulting from these postulated

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cransient loading conditions. We have, however, concluded that the low

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probability for occurrence of an event which could result in these loads

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x establishes an adequate basis to justify continued operation for a short

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term period.

The NRC staff will consider an analysis that is applicable to more than

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one specific plant if it can be adequately demonstrated that such an analysis is either representative or bounding for each plant concerned.

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Additional guidance regarding loading combinations (safe shutdown earthquake aEi.;.t EE loads, loss of coolant accident ioads), will be provided by cbout March 1, 1978, following the conclusion of staff investigations in this area.

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All PWR Licensees January 25, 1978.

Please respond within 30 days of receipt of this letter, indicating your

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intent to proceed with an evaluation of the overall asymmetric l'oss of

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coolantaccident(LOCA)loadsasdescribedherein.

In addition, please

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submit to us, within 90 days, your detailed schedule for providing the

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required evaluation.

Your schedule should be consistent with our desire

-5:!:iiE to resolve this problem within two years and should clearly state your intent to demonstrate the safety of long term continued operation.

We are transmitting information copies of this letter to the Westinghouse, Combustion Engineering and Babcock & Wilcox Companies.

If you have any

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questions or want any clarification on this matter, please call your NRC Project Manager.

Copies of this letter are being sent to all addressees on the current service lists for each docket.

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Sincerely, I...[1 Victor Ste 10,

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Division of Operating Reactors Office of Nuclear Reactor Regulation

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Enclosures:

1.

Background and Current Status 2==

2.

Revised Request for Additional rs=.

Info rmation

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cc w/ enclosure:

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See attached listing

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January 25, 1978

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Y ENCLOSURE 1 f':1. #

BACKGROUND AND CURRENT STATUS OF THE NRC STAFF REVIEW

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OF ASYMMETRIC LOCA LOADS FOR PWR FACILITIES

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On May 7,1975, the NRC was informed by Virginia Electric & Power Company 7~

that an asymmetric loading on the reactor vessel supports resulting from a postulated reactor coolant pipe rupture at a specific location (e.g.,

EM the vessel nozzle) had not been considered by Westinghouse or Stone &

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Webster in the original design of the reactor vessel support system for North Anna, E 4.s 1 and 2.

It had been identified that in the event of a postulated...;tantaneous, double-ended offset LOCA at the vessel nozzle, asymmetric loading could result from forces induced on the reactor inter-nals by transient differential pressure across the core barrel and by forces on the vessel due to transient differential pressures in the reactor cavi ty. With the advent of more sophisticated computer codes and the accompanying more detailed analytical models, it became apparent that such differential pressures, although of short duration, could place a signi-ficant load on the reactor vessel supports and on other components, there-by possibly affecting their integrity. Although this potential safety concern was first identified during the review of the North Anna facilities, f.....~

E it has generic implications for all PWRs.

Upon closer examination of this situation, it was determined that postu-2;==

lated breaks in a reactor coolant pipe at vessel nozzles were not the only area of concern but rather that other pipe breaks in the reactor coolant system could cause internal and external transient loads to act upon the i*

reactor vessel and other components.

For the postulated pipe break in the cold leg, asymmetric pressure changes could take place in the annulus between the core barrel and the vessel. Decompression could occur on the

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side of the vessel annulus nearest the pipe break before the pressure on the opposite side of the vessel changes.

This momentary differential 7j pressure across the core barrel could induce lateral loads both on the core barrel and on the reactor vessel. Vertical loads could also be applied to the core internals and to the vessel due to the vertical flow

.y resistance through the core and asymmetric axial decompression of the y

vessel. Simultaneously, for vessel nozzle breaks, the annulus between the reactor and biological shield wall could become asymmetrically pressurized resulting in a differential pressure across the vessel

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causing additional horizontal and vertical external loads on the vessel.

In addition, the vessel could be loaded by the effects of initial ten-

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sion release and blowdown thrust at the pipe break.

These loads could W'5 occur simultanecusly. For a reactor vessel outlet break, the same type M

of loadings could occur, but the internal loads would be predominantly vertical due to more rapid decompression of the upper plenum.

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Although the tiRC staff's original emphasis and concern were focused primarily on the integrity of the reactor vessel support system with respect to postulated breaks insid9 the reactor cavity (i.e., at a nozzle), it has since become apparent that significant asymmetric forces can also be gen'erated by postulated pipe breaks outside the cavity and that the scope of the problem is not limited to the vessel support

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system itself.

For such outside-cavity postulated breaks, the

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aforementioned concerns, such as the integrity of fuel assemblies and "E=

other structures, need to be examined.

In June 1976, the f4RC requested all operating PWR licensees to evaluate the adequacy of the reactor system components and their supports at their facilities with respect to these newly-identified loads.

Q In response to our request, most licensees with Westinghouse plants Hit:

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proposed an augmented inservice inspection program (ISI) of the R==

reactor vessel safe-end-to-end pipe welds in lieu of providing an evaluation of postulated piping failures.

Licensees with Combustion Engineering plants submitted a probability study (prepared by Science Applications, Inc.) in support of their conclusion that a break at a t..

particular location (vessel nozzle) has such a low probability of 55=

occurrence that no further analysis is necessary.

A similar study has E

been recently submitted by Science Applications, Inc. (SAI) for B&W g.).yl.{

plants.

When the Westinghouse 'and CE owners group reports were received in s1=d September 1976, the f4RC formed a special review task group to evaluate

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these alternative proposals.

In addition, EG&G Idaho, Inc., was

~ :=e contracted to perform an independent review of the SAI probability

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study submitted for the CE ewners group.

y This review effort resulted in a substantial number of questions 5

1 which previously have been 'provided to representatives of each group.

Based on the nature of these questions and other factors to be ji.::.r,

_..:q discussed later in this report, we cannot accept these reports in

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their present form as a resolution for the asymmetric LOCA load generic issue. Based on our review, we have concluded that a sufficient data

==Em base does not presently exist within the nuclear industry to provide satisfactory answers to these information needs.

Several long-term

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experimental programs would be required to provide much of this

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information. Although the probability study recently submitted by SAI for certain B&W owners does respond to some of the informal questions raised during our review of the SAI report prepared by

.s. J CE plants, the more fundamental questions remain. Therefore, this

~===a conclusion also applies to the SAI topical report for B&W plants

i (SAI-050-77-PA).

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A second - and equally important - reason for not accepting probability /ISI approaches as a solution at this point concerns our need and industry's need to gain a better understanding of the problem. We consider it essential that an understanding of the important breaks and associated consequences be known before applying any remedy ~ be it pipe restraints, probability, ISI, or some combination of these measures. Only in this way will we have a basis on which to judge the importance of the remedy with respect to what it is designed to prevent.

Although.we have many questions on each of these topical reports, this

,.x x does not mean that we view the probabilistic/ISI approach as completely

-ll without merit.

In fact, the results of a probabilistic evaluation serves

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as the basis for continued operation and licensing of nuclear plants during this interim period while additional evaluations can be perfonned

=13 by vendors and licensees.

""g We believe that the justification for continued plant operation has as its basic foundation the fact that the event in question, i.e., a hypothetical double-ended instantaneous rupture of the main coolant pipe at a particular

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location, has a very low probability of occurrence.

m The disruptive failure probagility of a reactor vessel itself has been estimated to lie between 10-and 10-7 per reactor year - so low that it

.. 4 is not considered as a design basis event.

The rupture probability of F

f) pipes is estimated to be higher. WASH-1400 used a median value of 10-4 for LOCA initiating ruptures per plant-year gor all pjpes sizes 6" and

4 greater (with a lower and upper bound of 10- and 10

, respectively).

We believe that considering the large size of the pipes in question (up to 50" 0.D. and 4-1/8" thick), the lower bound is more appropriate since these pipes are more like vessels in size.

In addition, the quality con-E="""!!-

trol of this piping is the best available and somewhat better than that

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of the piping used in the WASH-1400 study.

These factors, coupled with the facts that (1) the break of primary con-cern must be very large, (2) it must occur at a specific location, (3) the break must occur essentially instantaneously, and (4) these welds are currently subject to inservice inspection by volumetric and surface techniques in accordance with ASME Code Section XI, lead us to conclude

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that the probability of a pipe break resulting in substantial transient leads on the vessel support system or other structures is acceptably small such that continued reactor operation and continued licensing of GE:

facilities for operation can continue while this matter is being resolved.

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In support of the above, the staff has developed a short-term interim cri-

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terion to determine if an acceptable level of safety exists for operating

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PWRs under conditions of a postulated pipe break. This interim criterion

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is based on a simplified probabilistic model that incorporates elastic frac-

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ture mechanics techniques to estimate the probability of a pipe break.

Critical flaw size and subcritical flaw growth rates were determined assuming the presence of a surface flaw located in a circumferential weld of a

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thick-walled pipe, Determination of the critical flaw size was based on iE=E:

an estimated fracture toughness value of KIC at a minimum temperature of

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200 F and a uniform tensile stress equal to the consideration of various

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operating conditions producing elastica 11y calculated stresses rangina in value from 1 to 3 times the material minimum yield strength.

Then, using the calculated critical flaw size, the subcritical growth rate, and ar. estimated probability distribution of an undetected flaw in thick-walled pipe w E: M estimated to be 10-gids, the upper bound probability of pipe break wasThis value is also

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confirn rates of 10 gsh* whigh states that actual failure statistics tion by Dr. S. H. B to 10- per reactor-year in large pipes, with higher rates as the pipe size decreases. Considering these analyses, we conclude that our conservative estjmate on a pipe break in the primary coolant system is in the range of 10-to 10-6 This estimated pipe

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break probability is considered acceptably low to justify short-term operation of nuclear power plants.

In view of all previous discussions concerning this issue, the NRC staff

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has concluded that an evaluation must be undertaken to assess the design adequacy of the reactor vessel supports and other affected structures and

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inside and outside the reactor cavity. On performing these evaluations the staff will permit the grouping of plants, where adequate justifica-tion for such grouping exists, in order to limit the number of plants E. _. _

to be analyzed.

Alternatively, the staff will permit the analyzing of F====

a " prototypical" plant, which is sufficiently representative of a group of plants, to provide the necessary i 'ormation.

Both of these concepts have been discussed with the West' jhouse and CE Owners Groups,

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and we believe that such approaches could save a significant amount of time and effort in obtaining results on which to base any needed

g corrective measures. The NRC staff is prepared to meet with PWR licen-e

sees to discuss such approaches, and has already done so.

For example,

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we met with the Westinghouse owners group on October 19, 1977 for the

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purpose of discussing a generic solution for breaks outside the reactor cavity.

It is expected that a similar meeting will be held in the near

• " Critical Factors in Blowdown Loads in the PWR Guillotine Nozzle Break (Volume 2 - the Asymmetric Load Problem)" dated June 6,1977

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future to address breaks located inside the cavity. This " phased" approach is acceptable to us, provided that it sheds light on and

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serves to expedite consideration of the more limiting inside-

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cavity breaks.

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For your information, the NRC has a technical assistance contract with T=i EG8G Idaho, Inc., to independently model representative Westinghouse, B&W, and CE plants for the purpose of assessing the loads on all major E.:.:= "]

structures and components resulting from asymmetric LOCA loads. We

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believe that the results of this program which will include sensitivity

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studies, will provide significant confirmatory information related to this generic safety concern.

' Although, as stated earlier, we believe that continued operation and

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licensing of facilities for the short-term is justified, we also believe

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that efforts to resolve this issue should proceed without delay, with the objective of both completing the necessary assessments and installing

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In making this state-

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ment., de wish to make it clear that plant modifications, if indicated by licensee assessments, is the preferred approach.

At the same time, we

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recognize that there may be cases wherein appropriate modifications may

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be judged to be unwarranted based on the consideration of overall risk.

In such cases, and only in such cases, we will be prepared to give further F====

consideration to alternate approaches, such as probability /ISI. We feel, however, that ISI techniques as they exist today could be considerably improved, and, to the extent that such improvements could have a direct bearing on this problem as well 7.s an impact of. nuclear safety in general, we would welcome their developmer..

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ENCLOSURE 2 REVISED REQUEST FOR ADDITIONAL INFORMATION

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Recent analyses have shown that certain reactor system components and n..h[

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their supports may be subjected to previously underestimated asynnetric loads under the conditions that result from the postulation of ruptures

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of the reactor coolant piping at various locations.

It is therefore necessary to reassess the capability of these reactor system components

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to assure that the calculated dynamic asymmetric loads resulting from

==i these postulated pipe ruptures will be within the bounds necessary to

==F provide high assurance that the reactor can be brought safely to a cold shutdown condition.

For the purpose of this request for additional infor-mation the reactor system components that require reassessment shall include:

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Reactor Pressure Vessel

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Fuel Assemblies, Including Grid Structures c.

Control Rod Drives

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ECCS Piping that is Attached to the Primary Coolant riping

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Primary Coolant Piping f.

Reactor Vessel, Steam Generator and Pump Supports 9

Reactor Internals h.

Biological Shield Wall and Neutron Shield Tank (where applicable) i.

Steam Generator Compartment Wall Q'

The following information should be included in your reassessment of the zy.

effects of postulated asymmetric LOCA loads on the above-mentioned reactor systen components and the reactor cavity structure.

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1.

Provide arrangement drawings of the reactor vessel, the steam generator and pump support systems to show the geometry of all principal elements and materials of construction.

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2.

If a plant-specific analysis will not be submitted for your plant,

_3 provide supporting information to demonstrate that the generic plant

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analysis under consideration adequately bounds the postulated accidents at your facility.

Include a comparison of the geometric, structural, nachanical and thermal hydraulic similarities between your facility e-..

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and the case analyzed.

Discuss the effects of any differences.

b 3.

Consider postulated breaks at the reactor vessel hot and cold leg nozzle safe ends, pump discharge nozzle and crossover leg that re-

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suit in the most severe loading conditions for the above-mentioned

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systems.* Provide an assessment of the effects of asymmetric pres-7E-"

sure differentials on these systems / components in combination with all external loadings including asymmetric cavity pressurization for

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both the reactor vessel and steam generator which might result from

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limited displacement break areas where applicable E

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consideration of fluid-structure interaction ER c.

use of ' actual time-dependent forcing function rW -

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reactor support stiffness e.

break opening times.

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If the results of the assessment required by 3 above indicate loads l=

leading to inelastic action in these systems or displacement exceeding E

previous design limits provide an evaluation of the foll,owing:

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a.

Inelastic behavior (including strain hardening) of the material used in the system design and the effect on the load transmitted

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to the backup structures to which these systems are attached.

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For all analysis performed, include the method of analysis, the struc-tural and hydraulic computer codes employed, drawings of the models

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employed and comparisons of the calculated to allowable stresses

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and strains or deflections with a basis for the allowable values.

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Provide an estimate of the total amount of permanent deformation i==F.

sustained by the fuel spacer grids.

Include a description of the

.:151 impact testing that was performed in support of your estimate.

.E9 Address the effects of operating temperatures, secondary impacts,

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and irradiated material properties (strength and ductility) on the

=nTG amount of predicted deformation.

Demonstrate that the fuel will remain coolable for all predicted geometries.

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Demonstrate that active components will perform their safety function

.==r when subjected to the postulated loads resulting from a pipe break

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8.

Demonstrate functionability of any essential piping where service level B limits are exceeded.

In order to review the methods employed to compute the asymmetrical pressure differences across the core support barrel during subcooled

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portion of the blowdown analysis, the following information is requested:

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• B&W and CE plant licensees should' also consider breaks in' the hot leg l

at the steam generator inlet.

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January 25, 1978

:=51 ENCLOSURE 2 I.T

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REVISED REQUEST FOR ADDITIONA'. INFORMATION Recent analyses have shown that certain reactor system components and^

their supports may be subjected to previously underestimated asymmetric loads under the conditions that result from the postulation of ruptures

j of the reactor coolant piping at various locations.

It is therefore

...=.=g necessary to reassess the capability of these reactor system components to assure that the calculated dynamic asymmetric loads resulting from

..."My these postulated pipe ruptures will be within the bounds necessary to

_ ac=;.;g provide high assurance that the reactor can be brought safely to a cold 3

shutdown condition.

For the purpose of this request for additional infor-

=s.=.=j mation the reactor system components that require reassessment shall

...d include:

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a.

Reactor Pressure Vessel

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Fuel Assemblies, Including Grid Structures

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Control Rod Drives

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ECCS Piping that is Attached to the Primary Coolant Piping

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Primary Coolant Piping J""'"9 f.

Reactor Vessel, Steam Generator and Pump Supports ra 9

Reactor Internals

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Biological Shield Wall and Neutron Shield Tank (where applicable)

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Steam Generator Compartment Wall

=].j The following information should be included in your reassessment of the 7

effects of postulated asymmetric LOCA loads on the above-mentioned reactor

....1 system components and the reactor cavity structure.

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Provide arrangement drawings of the reactor vessel, the steam generator 3 =I and pump support systems to show the geometry of all principal elements-E- --s and materials of construction.

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If a plant-specific analysis will not be submitted for your plant,

+ 5 =E provide supporting information to demonstrate that the generic plant

=5HEd anC/ sis under consideration adequately bounds the postulated accidents

===El at your facility.

Include a comparison of the geometric, structural,

="=1 mechanical and thermal hydraulic similarities between your facility

~ ~ ~""1 and the case analyzed.

Discuss the effects of any differences.

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3.

Consider postulated breaks at the reactor vessel hot and cold leg

...j nozzle safe ends, pump discharge nozzle and crossover leg that re-

' ::1 salt in the most severe loading conditions for the above-mentioned

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systems.* Provide an assessment of the effects of asynnetric pres-9N sure differentials on these systems / components in combination with u=

all external loadings including asymmetric cavity pressurization for both the reactor vessel and steam generator which might result from

-the required postulate. This assessment should consider:

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limited displacement break areas where applicable

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consideration of fluid-structure interaction c.

use of actual time-dependent forcing function

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reactor support stiffness

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break opening times.

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If the results of the assessment required by 3 above indicate loads

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leading to inelastic action in these systems or displacement exceeding

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previous design limits provide an evaluation of the following:

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a.

Inelastic behavior (including strain hardening) of the material used in the system design and the effect on the load transmitted to the backup structures to which these systems are attached.

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For all analysis performed, include the method of analysis, the struc-E =(.if.!

tural and hydraulic computer codes employed, drawings of the models

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employed and comparisons of the calculated to allowable stresses and strains or deflections with a basis for the allowable values.

6.

Provide an estimate of the total amount of permanent deformation

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sustained by the fuel spacer grids.

Include a description of the impact testing that was performed in support of your estimate.

Address the effects of operating temperatures, secondary impacts, g4..;

and irradiated material properties (strength and ductility) on the

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amount of predicted deformation.

Demonstrate that the fuel will

sis remain coolable for all predicted geometries.

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Demonstrate that active components will perform their safety function when subjected to the postulated loads resulting from a pipe break

=iiE in the reactor coolant system.

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Demonstrate functionability of any essential piping where service level B limits are exceeded.

In order to review the methods employed to compute the asymmetrical

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pressure differences across the core support barrel during subcooled portion of the blowdown analysis, the following infonnation is requested:

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• B&W and CE plant licensees should also consider breaks in the hot leg

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at the steam generator inlet.

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A complete description of the hydraulic code (s) used including.the

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= a development of the equations being solved, the assumptions and -

simplifications used to solve the equations, the limitations re-sulting from these assumptions and simplifications and the numerical

====i methods used to solve the final set of equations.

Provide comparisons

~5 with experimental data, covering a wide range of scales, to demonstrate 54

'===J the applicability of the code and of the modeling procedures of the subcooled blowdown portion of the transient.

In addition, discuss application of the code to the multi-dimensional aspects of the

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reactor geometry.

If an approved vendor code is used to obtain the asymmetric pressure

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difference across the core support barrel, state the name and version of the code used and the date of the NRC acceptance of the code.

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If the assessment of the asymmetric pressure difference across the

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core support barrel is made without the use of a hydraulic blowdown M

code, present the methodology used to evaluate the asymmetric loads

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and provide justification that this assessment provides a conservative

== j estimate of the effects of the postulated LOCA.

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A compartment multi-node, space-time pressure response analysis is necessary to determine the external forces and moments on components.

Analyses should be performed to determine the pressure transient resulting

- 7;=

from postulated hot leg and cold leg reactor coolant system pipe ruptures

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within the reactor cavity and any pipe penetrations.

If applicable,

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similar analyser aould be performed for steam generator compartments

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that may be sut

.;t to pressurization where significant component support loads may resui..

This information can be provided to encompass a group

.:=3 of,itilarly designed plants (generi~c approach) or a purely plant specific

=

(custe, plant) evaluation can be developed.

In either case, the proposed 5MEE method of evaluation and principal assumptions to be used in the analysis

==:

should be provided for review in advance of the final load assessment.

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For generic evaluations, perform a survey of the plants to be included

===5 and identify the principle parameters which may vary from plant to plant.

For instance, this should include blowdown rate and geometrical varia-tions in principle dimensions, volumes, vent a eas, and vent locations.

A typical or lead plant should be selected to perform sensitivity and envelope calculations.

These analyses should include:

T 5.1 (1) nadal model development for the configuration representing the

==q rost rest.rictive geometry; i.e., requiring the greatest nodalization; (2) the most restrictive configuration regarding vent areas and

~~

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obstructians to flow should be analyzed; and, 22=.1 (3) sensitivity to code data input should be evaluated; e.g., loss g=$

coefficients, inertia terms, vent areas, ncdal volumes, and any L.,,.--

other input data where there may be variations from plant to

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plant or uncertainty for the given plant.

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These studies should be directed at evaluating the maximum lateral and

2. U.h vertical force and moment time functions, recognizing that models may

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be different for lateral as opposed to vertical load definitions.

.j The following is the type of information needed.for both generic and m...,..

custom plant evaluations. Although this request was primarily developed

_ =

for reactor cavity analyses it may be applied to other component sub-compartments by general application.

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(1)

Provide and justify the pipe break type, area, and location for i

each analysis.

Specify whether the pipe break was postulated for

=q the evaluation of the compartment structural design, component

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supports design, or both.

(2)

Fo'r each compartment, provide a table of blowdown mass flow rate

]

and energy release rate as a function of time for the break which results in the maximum structural load, and for the break which was used for the component supports evaluation.

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(3)

Provide a schematic drawing showing the compartment nodalization for the determination of maximum structural loads, and for the

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component supports evaluation.

Provide sufficiently detailed plan and section drawings for several views, including principal

==

dimensions, showing the arrangement of the compartment structure, major components, piping, and other major obstructions and vent c:..

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areas to permit verification of the subcompartment nodalization and vent locations.

(4)

Provide a tabulation of the nodal net-free volumes and inferconnecting flow path areas.

For each flow path, provide an L/A (ft- ) ratio,

=

where L is the average distance the fluid flows in that flow path

. c=.

and A is the effective cross sectional area.

Provide and justify es values of vent loss coefficients and/or friction factors used to

-==

calculate flow between nodal volumes. When a loss coefficient con-sists of more than one component, identify each component, its

==E value and the flow area at which the loss coefficient applies.

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(5)

Describe the nodalization sensitivity study performed to detennine

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the minimum number of volume nodes required to conservatively predict

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the maximum pressure load acting on the compartment structure. The

===?

nodalization sensitivity study should include consideration of spatial pressure variation; e.g., pressure variation circumferentially,

. 5..1

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axially and radially within the compartment. The nodal model development studies should show that a spatially convergent differen-tial pressure distribution has been obtained for the selected evalua-

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tion model.

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1 Describe the justify the nodalization sensitivity study performed for the major component supports evaluated, it different from the structural analysis model, where transient forces and moments acting

=+

on the components are of concern. Where component loads are of'

.=

primary interest, show the effect of noding variations on the transient forces and moments.

Use this information to justify E=

the nodal model selected for use in the component supports evaluation.

~

If the pressurization of subvolumes located in regions away from the

=

break location is of concern for plant safety, show that the selec-tion 'of parameters which affect the calculations have been conserva-

"=

tively e. valuated. This is particularly true for pressurization of the volume beneath the reactor vessel.

In this case, a model which predicts the highest pressurization below the vessel should be selected for the evaluation.

.;j; NOTE:

It has been our experience that for the reactor cavity, three regions should be considered (i.e., nodalized) when developing

.. :EE a total model.

These are:

(1) the volume around or in the vicinity of the break loca-EE-tion out to a radius approximated by the adjacent nozzles, and including portions of the penetration volume for some plants;

=

(2) the volume or region covering the upper reactor cavity, primarily the RPV nozzles other than the break nozzle;

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(3) the region encompassing the lower reactor cavity and other portions of the reactorecavity not included in Items (1) and (2).

(6)

Discuss the manner in which movable obstructions to vent flow (such

=

as insulation. ducting, plugs, and seals) were treated.

Provide

=

~~

analytical and experimental justification that vent areas will not be partially or completely plugged by displaced objects.

Discuss how insulation for piping and components was considered in determining

==

volumes and vent areas.

=

=g (7)

Graphically show the pressure (psia) and differential pressure (psi)

==

response as functions of time for a representative number of nodes to indicate the spatial pressure response. Discuss the basis for

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establishing the differential pressure on structures and components.

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(8)

For the compartment structural design pressure evaluation, provide the peak calculated differential pressure and time of peak pressure

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for each node.

Discuss whether the design differential pressure is uniformly applied to the compartment structure or whether it is

.y g spatially varied.

If the design differential pressure varies depending on the proximity of the pipe break location, discuss how Mis the vent areas and flow coefficients were determined to assure

e

that regions removed from the break location are conservatively designed, particularly for the reactor cavity as discussed above.

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(9)

Provide the peak and transient loading on the major components used to establish the adequacy of the support design. This should include roments (e.g., Mx(t), My(t), Mz(t))x(t), f (t), f (t)) and transient the load forcing functions (e.g., f y

z

~

as resolved about a specific, identified coordinate system. The centerline of the break nozzle is recommended as the X coordinate and the center line of the E~ ~

vessel as the I axis. Provide the projected area used to calculate f.T......

- " =

these loads and identify the location of the area projections on plan and section drawings in the selected coordinate system. This

.,,. =.,

infomation should be presented in such a manner that confirmatory evaluations of the loads and moments can be made.

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.Sacranento IMnicipal Utility District

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General. Counsel

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.cc: Gavid S'. Kaplan, Secretary'and.

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6201'S Street f".

Post Of fice Box ~15830 Sacranento,. California ' 95813

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Business and flunicipal.Departraent

' Sacramento City-County Library

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828 1 Street

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Sacramentio, California 95814:

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