ML20135E234

Forwards Request for Exemption from GDC 4.Nonproprietary WCAP-10584 & Proprietary WCAP-10585, Technical Bases for Eliminating Large Primary Loop Pipe Rupture As..., Encl. Proprietary Rept Withheld (Ref 10CFR2.790).Fee Paid

ML20135E234
Person / Time
Site:	McGuire, Mcguire
Issue date:	08/30/1985
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML19273A510	List: ML20135E234 ML20135E255 ML20135E269
References
TAC-59660, TAC-59661, NUDOCS 8509160333
Download: ML20135E234 (43)

Text

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HALILTUCKER Tmpnon VME.FSIDENT (704) 373-4531 August 30, 1985 mm...omm.

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Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation f,

U. S. Nuclear Regulatory Commission Washington, D. C. 20555

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h Attention:

Ms. E. G. Adensam, Chief Licensing Branch No. 4

Subject:

McGuire Nuclear Station Docket Nos. 5-369 and 50-370 Pipe Break Criteria Relief for Reactor Coolant Loop

References:

1) Letter from W. H. Owen (Duke Power Company) to W. J. Dircks (NRC), dated September 19, 1983

2) Letter from H. R. Denton (NRC) to W. H. Owen (Duke Power Company), dated October 17, 1983

3) NRC Generic Letter 84-04, dated February 1, 1984

Dear Mr. Denton:

Please find attached a request for an exemption (pursuant to 10 CFR 50.12 (a))

from General Design Criteria 4 to apply the " leak-before-break" concept to the McGuire Nulcear Station to eliminate postulated pipe breaks in the RCS primary loop from the plant structural design basis. Reference 1 informed the NRC that Duke Power Company was evaluating the technical feasibility and poten-tial benefits of eliminating postulated pipe breaks in the Reactor Coolant System (RCS) primary loop from the structural design basis of the McGuire Nuclear Station. As a result of efforts by Westinghouse, the NRC, and Duke Power, Duke has concluded that it is technically feasible to eliminate these postulated pipe breaks. In addition, Westinghouse has assured Duke Power Company that the generic information previously submitted to the NRC to justify the elimina-tion of RCS primary loop pipe breaks is applicable to the McGuire Nuclear Station.

Further, a safety balance in terms of accident risk avoidance versus safety gain will be demonstrated. As a result of the preceding developments, and in accordance with statements in References 2 and 3 that applications related to the leak-before-break pipe failure concept will be permitted prior to the NRC completing all of the changes in regulatory requirements, this exemption request (Attachment 1) is submitted.

The Westinghouse technical report (WCAP-10585) entitled " Technical Basis for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for McGuire Units 1 & 2" is included as Enclosure A to provide technical justification for elimination of RCS primary loop breaks for McGuire. As Enclosure A contains information proprietary to Westinghouse Electric Corp-oration, Attachment 6 is an application for withholding (CAW-84-60,non-proprietary) and the supporting affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information Jp) Y ffh K

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r Mr. Harold R. Denton, Director August 30, 1985 Page 2 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b) (4) of Section 2.790 of the Commission's regulations. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.790.

Correspondence with respect to the proprietary aspects of the Application for Withholding or the supporting Westinghouse affidavit should ref:*ence CAW-84-60, and should be addressed to R. A. Wiesemann, Manager, Regulatory ano ' taislative Affairs, Westinghouse Electric Corporation, P. O. Box 355, Pittsburgh, rennsylvania 15230. A non-proprietary version of the Specific Plant Applicability Report (WCAP-10584) is also included as enclosure B.

The impact on important design aspects of implementing leak-before-break on McGuire Nuclear Station has been evaluated by Duke Power and is summarized in.

A detailed list of affected pipe whip restraints is provided ih Attachment 3.. A summary of the potential benefits which can be realized specifically from the elimination of these pipe breaks for McGuire is provided in Attachment 4.

Note that these benefits total at least $2.6 million and involve an estimated 1064 person-rem dose reduction over the life of two Units.

Implementation of the leak-before-break concept will therefore be cost-effective as well as being technically justifiable while resulting in improved overall plant safety. Attachment 5 consists of the revised McGuire FSAR pages associated with the elimination of RCS primary loop breaks. Once approved, these changes will be incorporated into the next applicable annual update to the McGuire FSAR.

Pursuant to 10 CFR 170.3(y),170.12(c), and 170.21, Duke Power proposes that this exemption request constitutes a required approval for McGuire Units 1 and 2 subject to fees based on the full cost of the review (to be calculated using the applicable professional staff rates shown in 10 CFR 170.20) and must be accompanied by an application fee of $150, with the NRC to bill Duke Power at six-month intervals for all accumulated costs for the application or when review is completed, whichever is earlier. Accordingly, please find enclosed a check in the amount of $150.00.

It is requested that a resolution concerning implementation of this exemption on McGuire Units 1 and 2 be prior to October 30, 1985.

If there are any questions concerning this request, please advise.

Very truly yours, ff Hal B. Tucker PBN/hrp Attachments / Enclosures

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o Mr. Harold R. Denton, Director August 30, 1985 Page 3 cc:

(w/ attachments, Enclosure A)

Dr. J. Nelson Grace, Regional Administrator U. S. Nuclear Regulatory Commission Region 11 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. Darl Hood Division of Project Management Office of Nulcear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. W. T. Orders Senior Resident Inspector McGuire Nulcear Station

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Duke Power Company McGuire Nuclear Station RCS Loop Pipe Break Design Basis Exemption Exemption Request Pursuant to 10 CFR 50.12(a), Duke Power Company hereby applies for an exemption from the provisions to 10 CFR Part 50, Appendix A, authorizing alternative pipe break analyses utilized in resolution of generic issue A-2, "Asymmertric Blowdown Loads on PWR Primary Systems." The requested exemption is based upon the appli-cation of advanced fracture mechanics technology as evaluated in the Westinghouse technical report WCAP-10585 (Enclosure A).

Specifically, Duke Power Company requests the elimination of postulated circum-ferential and longitudinal pipe breaks in the reactor coolant system primary loop from consideration in the structural design basis of McGuire Nuclear Station.

Thepipe breaks are those identified in Westinghouse topical report WCAP 8172 for the RCS primary loop.

The impact on important design aspects of implementing leak-before-break on McGuire Nuclear Station has been evaluated by Duke Power and is summarized in Attachment 2.

A detailed list of affected pipe whip restraints is provided in Attachment 3.

The bases for the requested exemption are as follow:

1.

Extensive operating experience has demonstrated the integrity of the RCS primary loop including the fact that there has never been a leakage crack.

2.

In-shop, pre-service, and in-service inspections performed on piping for the McGuire Nuclear Station minimize the possibility of flaws existing in such piping.

The application of advanced fracture mechanics has demon-strated that if such flaws exist they will not grow to a leakage crack when subjected to the worst case loading condition over the life of the plant.

3.

If one postulates a through-wall crack, large margins against unstable '

crack extension exist for certain stainless steel PWR primary coolant piping when subjected to the worst case loading conditions over the life of the plant.

The application of advanced fracture mechanics technology has demonstrated that small flaws or leakage cracks (postulated or real) will remain stable and will be detected either by in-service inspection or by leakage monitoring systems long before such flaws can grow to critical sizes which otherwise could lead to large break areas such as the double-ended rupture of the largest pipe of the Reactor Coolant System. To date, use of this advanced fracture mechanics technology has been limited by the definition of a LOCA in Appendix A to 10CFR Part 50 as including postulated double-ended ruptures of piping regardless of the associated probability.

Application of the LOCA definition without regard to this advanced technology to large diameter thick-walled piping such as the primary coolant pipes of a FAR imposes a severe penalty in terms of cost and

occupational exposure because of the massive pipe whip restraints it requires which must be removed for in-service inspection.

This penalty is unreasonable because these pipes do not have a history of failing or cracking and are conservatively designed. Accordingly, for design purposes associated with protection against dynamic effects, Duke reauests this exemption from.the regula-tions to eliminate the need to postulato circumferential and longitudinal pipe breaks.

Implementation of the exemption will have the following effects on the structural design for McGuire Nuclear Station:

1.

Eliminate the need to postulate circumferential and longitudinal pipe breaks in the RCS primary loop (hot leg, cold leg, and cross-over leg piping).

2. ' Eliminate the need for associated pipe whip restraints in the RCS primary loop and eliminate the requirement to design for the structural effects i

associated with RCS primary loop pipe breaks including jet impingement.

3.

' Eliminate the need to consider dynamic effects and loading conditions associated with previously postulated primary loop pipe breaks.

These effects include blowdown loads, jet impingement loads, and reactor cavity l

and subcompartment pressurization.

The exemption would not eliminate pipe breaks in the RCS primary loop as a design basis for the following:

i 1.

Containment design 2.

Sizing of Emergency Core Cooling System 3.

Environmental qualification of equipment l

4.

Supports for heavy components l

The crack sizes from the leak-before-break analysis will be used when revising l

reactor cavity and subcompartment pressurization data, t

As stated above, Duke requests that de exemption authorize, with. respect to the plant structural design basis, the elimination of pipe breaks in the RCS primary loop. Thus, the use of advanced fracture mechanics permits a deter-ministic evaluation of the stability of postulated flaws / leakage cracks in piping as an alternative to the current mandate of overly conservative postu-lations of piping ruptures.

This exemption request is consistent with the provisions of footnote 1 to 10 CFR Part 50, Appendix A, which refers to the t

development of "further details relating to the type, size and orientation of postulated breaks in specific components of the reactor coolant pressure boundary."

As support for this request, in addition to the previously specified information, Duke would request consideration of the following:

1.

Letter from Darrell G. Eisenhut (NRC) to E. P. Rahe (Westinghouse) dated February 1, 1984.

2.

Memorandum from Darrell G. Eisenhut (NRC) to All Operating PWR Licensees,

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Construction Permit Holders and Applicants for Construction Permits, dated

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February 1,1984 -

Subject:

Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary i

Main Loops (Generic Letter 84-04).

3.

CRGR resolution of generic issue A-2.

4.

ACRS letter dated June 14, 1983, re:

" Fracture Mechanics Approach to Pipe Failure."

5.

Memorandum from William J. Dircks, EDO, to ACRS dated July 29, 1983, re:

" Fracture Mechanics Approach to Postulated Pipe Failures."

6.

Memorandum from Harold Denton (NRC) to Murray Edelman (AIF), dated May 2, 1983.

Safety Balance Further, pursuant to 10 CFR 50.12(a) Duke believes the requested exemption will not endanger life or property, or the common defense and security, and is in the public interest. The estimated increase in public accident exposure associated with omitting the RCS primary loop pipe whip restraints is only 0.6 person-rem per unit. This nominal estimate is based on the " Leak-before-Break Value-Impact Analysis" of Enclosure 2 to NRC Generic Letter 84-04, with adjustments made for the 40-year life and four loop design of McGuire.

The projected population densities within a 50 mile radius of the McGuire site for years 1990 and 2020 are 236 and 369 people per square mile, respectively. The population density calculated for the year 2020 is conservative based on the termination of a 40 year plant life as compared to a constant population density of 340 per square mile as stated in NRC Generic Letter 84-04.

Therefore, the 0.6 person-rem risk to public health is conservative for the McGuire-specific case. After adjusting NRC Generic Letter 84-04 data for a 40-year plant life, the estimated increase in occupational accident exposure associated with ommitting the RCS primary loop restraints is also low--less than 0.16 person-rem per unit for the nominal case.

The net benefit in avoidance of exposures for McGuire associated with the requested exemption is 1064 person-rem of occupational exposure over plant life, based on Duke Power studies.

This eliminated radiation exposure is related to pipe whip restraint inspection tasks, restraint disassembly / reassembly for pipe weld inspections, improved personnel access for operation and mainte-Consequently, the savings in exposure by granting the exemption far nance.

exceed the potentially small increase in public risk and avoided accident exposure associated with omitting pipe restraint devices.

Duke Power. Company estimates the net cost savings for McGuire Nuclear Station of at least 2.6 million dollars, as presented in Attachment 4.

The above net benefits have considered the costs associated with disposal of the restraint devices.

Additionally, with removal of pipe restraint devices, a substantial improvement in the quality of in-service inspections is anticipated.

Also, simplified plant designs will result since removal of these restraints will eliminate potential interferences with other plant structures.

Reduced RCS heat loss to containment at whip restraint locations will result.

The risks of unantici-pated pipe restraint for thermal growth and seismic movement can be avoided.

Thus, tne exemption will lead to an overall improvement in plant safety.

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of 1064 person-rem, a net safety gain has been demonstrated for McGuire Nuclear i

Station.

Also, a cost savings of a least 2.6 million dollars has been shown, i

and a technical basis for elimination of RCS primary loop pipe breaks has been i

demonstrated. Therefore, Duke Power Company hereby requests NRC approval of an exemption to GDC-4 in order to apply the " leak-before-break" cor. cept to McGuire Nuclear Station to eliminate postulated pipe breaks in the RCS primary loop from the plant structural design basis.

A summary of the potential benefits which can be realized specifically from the elimination of these pipe breaks for McGuire is provided in Attachment 4.

Note that these benefits total at least $2.6 million and involve an estimated 1064 person-rem dose reduction over the life of two Units.

Implementation of the leak-before-break concept will therefore be cost-effective as well as being

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technically justifiable while resulting in improved overall plant safety.

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ATTACHMENT 2 Impact of Elimination of Postulated Circumferential and Longitudinal Pipe Breaks in the RCS Primary Loop STRUCTURES, SYSTEMS, COMPONENTS, 3

PROGRAM CONSIDERED FOR IMPACT IMPACT Primary Loop Pipe Whip Restraints Deleted from Design

• Reactor Cavity / Primary Shield Wall /

Reduction in pressurization Crane Wall / Operating Floor loading Steam Generator Sub-compartment No change RCS Component Supports / Heavy No change Component Supports j

Emergency Core Cooling Systems No change 1

Containment Design No change RCS Pressure Boundary Leakage 110 change Detection Systems 1

Environmental Qualification Program No change

• Due to small hot gaps, the hot leg pipe whip restraints currently receive relatively small loadings from postulated main steam pipe breaks.

It has been shown that the Steam Generator column supports are adequate to support the additional load in the absence of the hot leg pipe whip restraints.

Also, an analysis has shown that the reactor coolant loop loadings from the main steam pipe breaks will be acceptable without the hot leg pipe whip restraints.

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i ATTACHMENT 3 Postulated RCS Primary Loop Pipe Breaks and Associated Pipe Whip Restraints Per Unit Postulated Break Associated Whip Restraint Location Per Loop for Primary Loading 1.

Reactor Vessel inlet nozzle 1.

Cold Leg Nozzle Break Restraint (wagon wheel)

Crossover Leg Pipe Whip Restraint (horizontal run) 2.'

Reactor Vessel outlet nozzle 2.

Hot Leg Nozzle Break Restraint (wagon wheel)

Crossover Leg Pipe Whip Restraint (horizontal run) 3.

Steam Generator inlet nozzle 3.

Hot Leg pipe whip restraint 4.

Crossover leg horizontal run 4.

Crossover leg elbow restraints break Crossover Leg Pipe Whip Restraint (horizontal run) 5.

Steam generater outlet nozzle 5.

Crossover leg pipe whip restraint (vertical run)

Crossover leg elbow restraints 6.

Reactor coolant pump inlet 6.

Crossover leg elbow restraints nozzle (pump suction) 7.

Stear. generator inlet elbow 7.

Crossover leg elbow restaints split II 8.

Reactor Coolant Pump outlet 8.

None nozzle break

O ATTACHMENT 4 Summary of Benefits from the Elimination of Primary Loop Pipe Breaks on McGuire Nuclear Station, Units 1 & 2 1

Category Benefits 1.

Plant Design Simplifies plant design by elimination of q

potential interferences with piping, hangers, impulse tubing, etc.

2.

Relief of congestion 1064 person-rem reduction in radiation improving access for exposure over life of plant ($2.6M).

operation and mainte-nance.

3.

Reduction in RCS heat Not quantitatively accessed.

loss to containment Insulation can be installed on piping at at whip restraint current locations of RCS pipe whip restraints.

location.

4.

Improvement in over-Improvement in ISI quality.

Elimination all plant safety of potential for restricted thermal or seismic (NUREG/CR-2136) movement.

5.

Simplification of Pressurization loadings reduced on primary analysis' associated shield wall, cranewall, operating floor, and with dynamic effects subcompartment analysis.

and loading conditions.

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ATTACHMENT 5 Revision to McGuire FSAR For Leak-Before-Break Criteria Change l

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..: m ' CRITERION 4 - ENVIRONMENTAL AND MISSILE DESIGN BASES Structures, systems and components important to safety'shall be designed to

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accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and

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postulated accidents, including loss of-coolant accidents.

These structures,

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• systems and components shall be appropriately protected against dynamic A

_. effects, including the effects of missiles, pipe whipping, and discharging

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t. g r outside the nuclear power unit.

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function in a manner which assures public safety at jll times s

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by ' appropriate missile barriers, pipe restraints, and station layout.

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Reactor Building is capable of withstanding the effects of missiles The originating outside the Containment such that no credible missile can result J'

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in a loss of coolant accident.

The Control Room is designed to withstand such missiles as may be directed toward it and still maintain the capability of i

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Emergency core cooling components are austenitic stainless steel or equivalent corrosion resistant material and hence are compatible with the containment atmosphere over the full range of exposure during the post-accident conditions.

Reference:

Chapter 3.0 and Section 6.3 CRITERION 5 - SHARING OF STRUCTURES, SYSTEMS, AND COMPONENTS Structures, systems, and components important to safety shall not be shared between nuclear power units unless it is shown that their ability to perform

.M their safety functions is not significantly impaired by the sharing.

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Structures, systems, and components, which are either shared (a)'between the

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This design protects the ability of shared structures, systems and

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Chapters 3.0, 6.0, 8.0, 9.0 and 11.0 m '. : d

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including the effects of anticipated operational occurrences.

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Steam generator water inventory on secondary side (manual or automatic

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feedpump flow through feedwater control valves).

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average temperature in the reactor coolant during steady state operation and

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to insure that unit conditions do not reach reactor trip settings as the

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1 control assembly from its bank position.

4tO d'2 h monitors with visual and audible annunciation to. avoid. loss l.of_

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" positioning at the bottom of its travel.Each rod cluster control assembly is provided w This condition is also alarmed in the Ccntrol Room.

flux distributions indicative of rod misalignment.Four ex-cere 1:ng ion chamb

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J.7.' operational aids to the operator. Movable in core flux detectors and fixed in-e#.

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instrumentation and controls.

Chapter 7 contains further details on system provided to measure environmental activity and alarm high levels

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O rail reactivity control is achieved by the combination of soluble boron and rod cluster control assemblies.

accomplished by adjusting the concentration of boric acid in the reactorLo coolant.

^ the Rod Control System which automatically moves rod cluster c assemblies.

temperature, and turbine load.This system uses input signals including neutron flux, co N

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Chapters 7.0 and 11.0

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.c The reactor coolant pressure boundary shall be designed, fabricated, erected1.d. M, > ;y and tested so as to have an extremely low probability of abnormal leakage, or

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The reactor coolant pressure boundary is designed to accommodate the system

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.*I pressures and temperatures attained under all expected modes of plant within applicable stress limits. operation, including all anticipated transients, and to J$M}j In aodition to the loads imposed on the piping under operating conditions; consideration is also given to abnorma1 c.

loadings such as pipe rupture h na seismic loaoings as discussed P

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relieving devices as required by applicable codes.The piping is protected in Sections

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s 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE RUPTURE OF PIPING

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In the design of a pressurized water reactor, special provisions are made for 7....c ' ~,;.includ' ing a loss-of coolant or steam line break accident. protecting

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considered in these accidents, the various system necessary to recover from these accidents and the mechanical provisions which

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J-and mechanical restraints preclude the formation of plastic hinces for breaks postulated to occur in tt "c::t-

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/ocuNoas The Reactor Coolant System the main coolant loop pipi g and all branch connection nozzles out to th first butt weld. Dynamic. efree +s on ly cons,;;te,.e.cl -%' p/Pc bn-~les p* m i-rod. m+

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'modated directly by the supporting structures-of the reactor vessel, the steam' co generator and the reactor coolant pumpssiehdi ; +vn W

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'IYd Tnis subdivision discusses all piping systems excluding the Reactor Coolant.:p System as described in 3.6.1.1 and is in accordance with HRC Branch Technica iQ.

Position APSCSB 3-1 and Regulatory Guide 1.46 except as noted in Table 3 6

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, systems terminating at the main coolant loop piping nozzle.

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pressure less than or equal to 275 psig are not considered high energy W_e regar.dless of the temperature, and (2) for liquid systems other than water,

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One (1) percent of the normal operating lifespan of the plant, or

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Open ended vents and drains an,d

(

. Systems or appropriate portions which fall in either or both o categories are analyzed as described in 3.6.4.2 and protected in accordance

'ith 3.6.5.2.

w Table 3.6.1-1 lists Hign Energy Systems or portions thereof and

... w.4

Table 3.6.1-2 lists mode ~ rate energy systems or. portions -thereof in accordance;.

t with the above definitions that are analyzed for the station.

.,.[f$* a

..e p ;,y...

~.3

..',,,,. ' Sub. division 3.9.2.8 discusses Containment integrity with respect to breaks Jy',Qf9:

.=-

2.

involving mechanical penetrations.

.:.Q g-jg

" w e.w s

3.6.2 DESIGN BASIS PIPING BREAK CRITERIA e,:y 3.6.2.1

%;M */

Postulated Pipino Break Location Criteria for the Reactor Coolant

f f-ir;?. u Q.}

System

. y@j@,

~

,;?::i:.s

- ~The design basis for postulated pipe breaks c.

'e

...m

.m-.

r'~

J-shMd include not on!y the break criteria, but also the criteria to protect 12:p p y g

(:M($@

other piping and vital systems from the effects of the postulated break.

i

"'h j A loss.of reactor coolant accident is assumed to occur for a pipe break down to the restraint of the second normally open automatic isolation valve (Case 11 c

?i 9

'.. p &.'.,

.g

.93 3.6-2 12/83

~6D 4%

y;..

-g W-p.

y4 w.. "f 0

-,n.

~, s...-? *<. ?0 '.

a.

~

3:is? r s

.~

N,.,

C";n check valve (Case 111 Figure 3.6.2-1) on incoming lines O..

pipe break beyond the restraint or second check valve does not result in an A

uncontrolled loss of reactor coolant if either of the two valves in the line M.N close.

"X...

i.c

... Q,9 Accordingly, both of the automatic isolation valves are suitably protected and restrained as close to the valves as possible so that a pipe break beycnd the

,d.d g,3 restraint does not jeopardize the integrity and operability of the valves.

i,.g.i 'i~

.4

..v.fy;.

M; f function is essential.Furttier, periodic testing capability of the valves to perform their int

' O' g.['I valves performing its intended function.This criterion takes credit for only one of the two j

/

? ;-

f incoming check valves (Case I and IV in Figure 3.6.2-1) a loss of reactorFor n

" G.s ;

O coolant accident is assumed to occur for pipe breaks on the reactor side of 4E@W-M/-

d

.; w.. the valve.

-WA:

.,s t'ry$hq..

4

.7

..a c-2.

' an oiping system, there is a limited number of locations dich are u.

- ;a.

e more susceptible i '

• a bv virtue of stress nv- ':tige t. nan the remainder Qw,j-(

of the system.

These postulaton considerina nne-M W upset operating conditionsions are defined in 3.6.2.1.1

.. :3.W L.w specification as required by ASME Code,Section III).b the applicable

^O

, e.y.

.u g h, Engineered Safety Features are provided for core cooling and boration pressur

.y reduction, and activity confinement in the event of a loss of reactor coolant I'4 or steam or feedwater line break accident to ensure that the public is protected in accordance with 10CFR100 guidelines.

to provide protection for a Reactor Coolant System pipe rupture of a s C

and including a double ended severance of the Reactor Coolant System ma o

.c

d,.;

Branch lines connected to the Reactor Coolant System are defined as "small" if they have an inside diameter equal to or less than 4 inches.

that no clad damage is expected for a break area of up This size is such t.e;,, Q

^

corresponding to 4 inches inside diameter piping.

m, n

$.f &

. h.,In order to assure the continued integrity of the vital componen DIS M,.

.. o. ?

' neered safety systems, consideration is given to the consequential effects of y,'

the pipe break itself to the extent that:

'Q'j,yNqm m

a.

not reduced below that iequired to protect agains

.<wq

..gymf3 u s.I -

NSY.D$.i y b.

.The Containment leaktightness is~not decreased below the design valu

- 'Ep2H2 4

the break leads to a loss of reactor coolant; (**) and gj

. n,:

.5

.w.

. ~ * -

{Nf.v.,

(*) It is assumed that motion of the unsupported line containing the isolatio 5 E.'I5M valves could cause failure of the operators of both valves.

n

'~T

. g.

(**) The Containment is here defined as the Containment vessel and'penetrat and the steam generator shell, the steam generator steam side instrumenta-

ons,

di cW:

tion connections, the steam feedwater, blowdown and steam generator dr W

pipes within the Containment structure.

i Q

, q' g y w

3.6-3 12/83 NMNN ydh M"

N

..g.w w. '

y.

4..

. y, m

T.:. ', c.

49.0 Propagation of damage is limited in type and/or degree to the extent that:

1 1)

A pipe break which is not a loss of reactor coolant does not cause Kit a loss of reactor coolant or steam or feedwater line break, and f,1;.)q.

.g:

2)

-f '

A reactor Coolant System pipe break does not cause a steam-feedwater i

9e system pipe break and vice versa.

i

.p

?p

%j @Y.~I In the unlikely event that one of the small pressu

. ' m @y.

'^

result in a loss of reactor coolant accident, the piping is restrained or

' F ;'c.) above:.. arranged to' meet the' followihg. additional l~ criteria in addition to (a. through :j.@

Nu.?..

.. ~ ~

7

'.2

~ ' ip f e.

Break propagation must be limited to the affected leg, i.e., propagation

..llX

.s; Q.;4..y to the other leg of the affected loop and to other loops is prevented; NM

.s

~'p g{

r The RTD Bypass line are exceptions to this particular criteria.

A break in

• ,,, one of these line is allowed to propagate to either of the other RTD lines in the affected loop.

.d5

.r

.This is permitted since:

. x., j.e.

e.

,'.. v.'. -1)

A break in a RTD line effectively results in propagation to the other d!.:M.-

~

.n.

2)

... legs by flow through the unbroken RTO lines,

~

An analysis of dual aperture breaks has been perforced and the results break cases 'or the RTD lines. indicate that the limiting small break case bo (Reference 2)

/"

b.

Propagation of the break in the affected leg is permitted but is limited k

to. a total break area of 12.5 square incht. (4 inch inside diameter).

head safety injection line.The exception to this case is when the init.iating sm

. g Further prcpagation is not permitted for this case;

' of the affected loop or to the other loops is prevented; and k:b c.

M%[j d.

Propagation of the break to high head safety injection line connected to k.G Q*

?

e?

h,@

affected leg is prevented if the line break results in a loss of core cooling capability due to a spilling injection line.

.dll",

'3.6.2.1.1

.9$.....

w 1

Postulated Piping Break Locations and Orientations Ml'%y ch leg of the Reactor Coolant System, a minimum of three postulate

.m

. rupture tions shall be selected in the following manner:

- ~..i, f'

,j Breaks shall be postu.

d at the terminal points all locations in a run or branch in which the c ative usaoe #

1.G upset operating conditions or in wn or exceeds 0.2 for normal and d.

stress intensity for normal a

..e range of primary plus seconoary M

set oper of the ASME Section II e allowable on an elao conditions exceeds 80 percent

'y event that a 1

- on between the terminal points cannobasis (2.4 S,).

In the d

point cf highest fatigue usage shall be used to o ochosen in this manner. +

break locations.

a total of

,yj

-4 An.fe.ren e.

3.6-4 r:

I de.r;<,e se ori 12/83 P'i'M:<y hop. Refe.,<, enc +e i.e.yal was /,< p.shalalig pipspro ider the. hasi.s & e.

ess re-elor o.o/as,t. &

b@ :n acu ter.wsrem pm

, al-sna,y pnielated ecsc)er~.

% ble 3.C,. 2-t o,a f;w we. 3. t asea.xr oin tre. e.nceMion.J isa.re posetual

~

~

we, e.s e.nr>ac.+io

--. ~ ~ * ~ -:-

n s. s*<-~~

• s*

Ng+

'u 6.y.

3

,y lp Q,~

.J At' eacN possible break location, consideration must be given to the occurre e

f either a circumferential or longitudinal break. As discussed in WCAP-72 La ircumferential rupture is more likely than a longitudinal rupture fo reactor

, coo nt piping. Only in the case of one elbow is a longitudinal rupt 4 pn W late.

e postu-n Q!k90n v N(@EN!lY'I QCircumfer tial breaks are perpendicular to the longitudinal ax' of the pipe.

4 c

-p A k dn, w w.p ; ~itudinal reaks are parallel to the longitudinal axis of he pipe. Certain ptd.dlong

@ T.S.'Jg # ? longitudinal b ak orientations may be excluded on the bas' of the state of i

j M N M n;h. stress at the lo tion considered.

Y @b # i 4

a-y p $:=][mN.Forthemainreacto coolant piping system, eleven 6

were determined by st ss and fatigue analyses. :T locations are given in M. h7 % Table 3.6.2-1 and shown in Figure 3.6.2-2.

The stulated locations conform bc ? to the criteria stated a ve and are discussed n WCAP-8172.

q h.. "

m C

!)3lg ['M 23621 N. ' 18reak orientation at each di crete break lo tion is presented in Table

?h f.

The results of the alyses whi lead to the break orientations are

'h

W A ; discussed in WCAP-8172. 4 W h- ~ ' W ' g' 'i'u a' ' m 7,.

• nw

' =

s 1 %-

. fy:& +

':=

d, s '

h@M S P 4l' tin Table 3.6.2-2 for a referenc w dbp?HThe primary plus secondary stress i ensity ranges' and the fatigue cumulative y

9

' qusage factors at the design break a tions specified in WCAP-8172 are given fatig analysis. =The reference analysis has

. J 6~

3.been prepared to be applicabl for many ants.

It uses seismic umbrella moments which are higher th those used i WCAP-8172 such that the primary N

. stress is equal to the li 'ts of equation 9 'n NB 3650 (Section III of the ASME code) at many loca 'ons in the system.

erefore, the results of the

. reference analysis ma differ slightly from WCA -8172, but the philosophy and v-2 conclusions of the AP are still maintained.

T re are no other locations in

.the model used in he reference fatigue analysis, nsistent with WCAP-8172,

}

.where the stres intensity ranges and/or us' age fact s exceed the criteria of 710l' M 2.4 S, and O., respectively.

' Actual mom ts for the McGuire units are also given in le 3.6.2-2 so that the refe nce fatigue analysis can be shown to be applica e for McGuire.

By y

'showin actual moments to be no greater than those used in e reference

_ anal is, it follows that the stress intensity ranges and usa e factors for

~

.McG ire plant will be less than those for comparable locations 'n the reference

-m el.

By this means it is shown that there.are no locations ot r than those dentified in WCAP-8172 where the stress intensity ranges and/or f tors for McGuire might exceed the criteria of 2.4 S and 0 us, the applicability of WCAP-8172 to McGuire has 8een ver.2, respectively, ified.

3.6.2.1.2 Postulated Piping Break Sizes For a circumferential break, the break area is the cross-sectional area of the pipe at the break location, unless pipe displacement is shown to be less by analysis, experiment or physical restraint.

For a longitudinal break, a break area less than the cross sectional area of the pipe may be assumed when analytically or experimentally substantiated.

In 3.6-5 1984 Update vi,g-y-

wq ow---

- +--

9

,-+g-t.s.

-il.ey.-'

g

-r-i m--w'.r

-w++e---

• ----i-y e-

,--.,--ai$

...y },.

the absence of this data, the break area shall be assumed to be the cross-

.C '*:

sectional area of the pipe and the break length shall be assumed to be two

}.

pipe diameters.

7

(

3.6.2.1.3 Line Size Considerations for Postulated Piping Breaks p sAAb Branch lines connected to the Reactor Coolant System are defined as "large" for wae,e.

the purpose of this criteria as having an inside diameter greater than 4 inches

.t up to the largest connecting line, gere- "y "'c p"ers"#her crge ' Se.

break"of these lines results in a rapid blowdown of the Reactor Coolant System Pipe

..: t.. -

g he,.f Jand pr'ote~ction is basically provided by the accumulators and the low head safety.

.h injection pumps (residual heat removai pumps).

F

~;;. pp9.};;.

s

,e c.O c W ; ~ 3.6.2.2

• ~ General Desian Criteria for Postulated Pioina Breaks Other Th Y/M.-

3.

'./m Reactor Coo 1 ant Svstem w.V~.%.

:w w'

..,a

'a. - Station design considers and accommodates the effects of postulated pipe

~

... :s.

breaks with resoect to pine whip, jet impingement and resulting reactive

_ ;.,a forces for piping botn inside ano outsioe t ontainment.

j-M_

The anaiytical method utilized to assure that concurrent single active component failure.

2,.

T y

.l -are outline in Figure 3.6.2-3.and pipe break effects do not jeopardize the saf 9'5

? WEI b.

Station general arrangement and layout design of high energy systems utilizes the possible combination of physical separation, pipe bends, pipe

,7 whip restraints and encased or jacketed piping for the most practical design of the station.

lated piping break consequences to minimum and acceptable levels.Th e

In all cases, the design is of a nature to mitigate the consequences of the f'

break so that the reactor can be shutdown safely and maintained in a

(

" ' safety shutdown condition.

The environmental effects of pressure, temperature and flooding are con-c.

trolled to acceptable levels utilizing restraints, level alarms and/or J,.

other warning devices, vent openings, etc.

'~

' } d.

Plant Operating Conditions

- Q{\\'$

.. a..n 1)

Power Level - At the time of the postulated pipe break, the plant is

.f?.Q j,.

assumea to be in the normal mnde of plant operation, in which the

' %'.:yd piping under investigation experiences the maximum conditions of pres-

'MM.cM sure and tecperature.

' tion, th,e power level assumed is that assumed in the evaluat i

,. -j the lo'ss-of ccolant acciaent, steannine creak accicent. or fecowater

, ;,n line b'reak ai:cicent,

- in Chapter 15 of the safety analysis report.

e.

steam line creak accident,If the pipe break results in a lo c.

2) Offsite Power -

offsite power is a:suned to occur suasecuent to the pipe ruptu

~

yE 3)

Seismic Loacinos equivalent to either tne Safe Shutdown Earthquake (SSE) or tna coerating Basis Earthoucke (OP.E), as appropriate

[.ti' be used in r.na snalysis of piping, v;uiment, protective devices, etc

, will 2.

-4 4

3. 5-ri 7.;.

12/83

m...~, a.3 M i "*"~'Mit.AU m. W.&-@; p..

s....> J &. 2. :w.

.. w :

s 5s Q

f

. ' ($

• 1-t 4

A - l (I0)2 s

J ~ Through-wall crack pipe break areas are based on length equal to one-half the nominal outside diameter (1/2 ID) and a width equal to one-half the minimum g ;x y ' wall thickness (1/2 t) of the system piping materials, i.e.,

?,g. ml {j. f' V

ID A=gt g,y 2O ;M Mp y_y

>cyq[r' h

13.6.2.2.3 Line Size Considerations for Postulated Piping Breaks

$ m u $.:M For high energy systems, piping larger than 1" nominal pipe size (NPS) is mo 3& l ;

Q"7; ' (* reviewed for the consequences of a double ended break.

hwa;N Y.EFor high' energy systems, piping 4" NPS and larger is reviewed f n

.c, My

~

[MW e, Shquencesofdoubleendedandequivalentarealongitudinalbreaks.

i}$ l

<=

..For moderate energy system, piping larger than 1" NPS is reviewed for the rp.y

• l

. J. e"r ;.i -.. e ; consequence of through-wall cracks.

~

l l

s %

f., yl N, 4p, s3.6.2.3 Analysis and Results sf a

The results of analyses of failure in fluid systems occurring inside and s

Jyp+

ioutside containment for McGuire are presented in the " Summary Evaluation of

.g

.the Effects of Postulated Pipe Failures", Report No. NE-1019.

,.. g 3.6.3 DESIGN LOADING COMBINATIONS

3. g t

'3.6.3.1 Reactor Coolant System Design Loading Combinations we owwe eencu Assaump wosu enrumo pracm ruews ens ~cx couweerrou sturrunt mee

~

As described in Section 5.2.,"th; force " ~ 4eted i th rupturc of r;=tc-

-pipir.g Ustcm3 cre considered in the design of supports and restraints in order to assure continued integrity of vital components and Engineering Safety Features.

Reaction forces used in the design of supports and restraints are computed on a

c the basis of an assumed break equal to the cross sectional flow area of the pipe.

The design stress limits applicable to postulated reactor coolant piping breaks and supports are discussed in WCAP-8172 and are listed in Table 5.2.1-3.

3.6.3.2 All Other Mechanical Piping Systems h an loadino Combinations Since all locations of consequences are reviewed and as detailed stress analysis

'information is extremely extensive, stress analysis information is only l

reviewed for special identified problem areas which might require additional restraints.

These additional consequential piping breaks posing safety-related probitms to structures, systems or components in the immediate area are either restrained to mitigate the consequences of the break or reviewed in detail against existing stress analysis.

If the stress allowables discussed in 3.6.2.2.1 are not exceeded, then the break is not considered to occur.

z

~.

m*

3.6-14 1984 t*pdate

,.,..y, e-.

--.-------n

,m-

i 8

. )lg h4

,(

I 1

f Loading and stress criteria for pipe whip restraints is fully described in

3. 9.

Postulated pipe breaks are considered a faulted condition with respect t{

p."lto the pipe whip restraint design and allowable restraint stresses.

.o

..f--

1 3.6.4 DYNAMIC ANALYSIS e -

'@ w _Nd

' k } !.r

. Q V > ; N ;.3. 6. 4. 1 Reactor Coolant System Dynamic Analysis

-Q WW

?:p This section summarizes the dynamic analysis as it applig to the LOCA resulting t

yyyl a m from tge ogty @ lated design basis pipe breaks 4 th: 9 reactor coolant p+pmg

73 h8[i"y J

WE" rYh

' discussion of the dynamic analysis methods used to verify the

% :..:' J > M' design adequacy of the reactor coolant loop piping, equipment and supports is

.l$$, [D[ given in WCAP-8172As or PGATAsk) To Posr4LarEO BRLAkJ Ar BRAN 4i CONN

~ u

%g:e c,

i gb W@MlO@i,c The particular arrangement of the Reactor Coolant System f V Station is accurately modeled by the standard layout used in WCAP-8172 and the h

M h*$j.. l postulated break locations do not change from those 4

• • • b """T'"

- M2 N

t (M k f y h s44 break location the motion of the pipe e rom.teno sanucncoauterich - li 6 rIn addition, an analysis will be performed to demonstrate that at eachv4 W m M N%,(U ', ~ ' unacceptable damage due to the effects of pipe whip Q

Wd2,1 3

i 9p

. major components. The loads employed in the analysis will be based on full.Q bTW

. pipe areas discharge except where limited by major structures. The effects of

'M'

-jet discharges will be analyzed to demonstrate that any structure, system or

- component required to safety shutdown the reactor or mitigate the consequences of an accident will not be impaired.

The dynamic analysis of the Reactor Coolant System employs displacement method, lumped parameter, stiffness matrix formulation and assumes that all components behave in a linear elastic manner.

_ The analysis is performed on integrated analytical models including the steam

~,

generator and reactor coolant pump, the associated supports r d r= tr i t:,

and the attached piping. An elastic-dynamic three-dimensional model of the Reactor Coolant System constructed. The boundary of the an.lytical model is,

_ in general, the foundation concrete / support structure interface.

The antici-pated deformation of the reinforced concrete foundation supports is considered

.g where applicable to the Reactor Coolant System model.

The mathematical model is shown in Figure 3.6.4-1.

The steps in the analytical method are:

The initial deflected position of the Reactor Coolant System model is

~

~

a.

. defined by applying the general pressure analysis; ennuce eeuwemos b.

Natural frequencies and normal modes of the broken 4eopvare determined; The initial deflection, natural frequencies, normal modes, and time-history c.

forcing functions are used to determine the time-history dynamic deflec-tion response of the lumped mass representation of the Ractor Coolant System; d.

The forces imposed upon the supports by the loop are obtained by multi-plying the support stiffness matrix and the time-history of displacement l

vector at the support point; and 3.6-15 12/83 s

--_.,,,.,-m.

L-

, w ua-The time-history dynamic deflection at mass point g tgatg as an e.

imposeddeflectionconditionontheruptured16TkReacfortoolantSystem model and internal forces, deflections, and stresses at each end of the Wh R; b,. omembers.of the reactor coolant piping system are computed.

' N. T C The results are used to verify the adequacy of the restraintst The general SW.

nrr4 s*AucHco""Ecv*"J-7 Af l j 'W dynamic solution process is shown in Table 3.6.4-1.

Wl.f -

l J' In order to determine the thrust and reactive force loads to be applied to

- R,

+

? the Reactor Coolant System during the postulated LOCA, it is necessary to have

• K f * ~, q a detailed of the hydraulic transient.

Hydraulic forcing functions are W'.,

i calculated for the ruptur:d rd ".txt reactor coolant loops as a result of a lif 73

< postulated loss of coolant accident (LOCA)

These forces result from the

@ i i (i etransient flow and pressure histories in th Reactor Coolant System.

The d

i calculation is performed in two steps.

Th first step is to calculate the N

?-

ll transient pressure, mass flow rates, an ther hydraulic properties as a QQMy M function of time.

The second step us the results obtained from the s

M%..s - -

hydraulic analysis, along with inp of areas and direction coordinates and is T M 5: f to calculate the time history of orces at appropriate locations in theg ( reactor coolant loops. m n p u, y y,, w s n o snipx n ce weteno a SacA g, e The hydraulic model represents the behavior of the coolant fluid within the 6 entire reactor coolant system. Key parameters calculated by the hydraulic model are pressure, mass flow rate, and density. These are supplied to the thrust calculation, together with appropriate station layout information to , cetermine the concentrated time-dependent loads exerted by the fluid on the loops. In evaluating the hydraulic forcing functions during a postulated LOCA, the pressure and momentum flux terms are dominant. The inertia and gravitational terms are taken into account only in.the evaluation of the local fluid conditions in the hydraulic model. ~ TheblowdownhydraulicanalysisisrequireNtoprovidethebasicinformation concerning the dynamic behavior of the reactor core environment for the loop - forces, reactor kinetics and core cooling analysis. This requires the ability to predict the flow, quality,.and pressure of the fluid throughout the reactor system. The SATAN-V code was developed with a capability to provide this information. The SATAN-V computer code performs a comprehensive space-time dependent analysis of a loss of coolant accident and is designed to treat all phases of the blow-down. The stages are: (i) a subcooled stage where the rapidly changing pres- ~ sure gradients in the subcooled fluid exert an influence upon the Reactor Coolant System internals and support structures; and (ii) a two phase depres-surized stage. The code employes a one-dimensional analysis in which the entire Reactor Coolant System is divided into control volumes. The fluid properties are considered uniform and thermodynamic equilibrium is assumed to each element. Pump characteristics, pump coastdown and cavitation, core and steam generator heat transfer including the W-3 DNB correlation in addition to the reactor kinetics are incorporated in the code. l The blowdown hydraulic loads on primary loop components are computed from fluid r transient information calculated using the following time dependent forcing function: 3.6-16 12/83

y Tihn g'- ,4,_ ' q;, w ; + 52.nt.

u m l'

S-S ll

y m,.

( faither a total compartment volume or an arbitrary control volume assumed for

,, computational purposes.

y m e N, 1The results of the pipe rupture analysis for Category 1 piping systems other .!_:than the Reactor Coolant Loop are presented in Appendix 3P and Report No. f$ce?"ft;MDS/PDG-77-1forinsidecontainmentandoutsidecontainment,respectively. %, D.

g

-s, %em.

p. N,3.6.4.3..

' Structural Analysis of Postulated Piping Breaks ~ f .,s.e n w x ff I, !.4] TS ', hY E ??

x. Evaluation utilized to demonstrate the adequacy of or in the design of M$d[?iD! Category 1 structures subject to loadings of postulated piping breaks "T

$ ?Q 6 '1 d 41nclude: a.. - If M7%?'. M + .i 6:/ 4@%W a;;c.a..J;p g lwgMethod of evaluating stresses; y 73.c j L -lg. W,O dG M M b.'" 1 Allowable design stresses and/or strains; - a '- M: . G &r;;e. ~ Pn m e.... 4 s.. .' Load factors and combinations; mau D, c. - n c c:ggy. p "hN]Q,-(d. Design loads' including pressure and temperature transients; w .c c, g #;g - A so J e. ' Load reversal effects. ~.1 ; ..;g it :4 2 ~4 O ^

9. ' "

U Details of the structural analysis involving the above combinations are d i l@' b' Wdiscussed in 3.8.1. ,y . w%:P 3.6.5 PROTECTIVE MEASURES i s,w. A.~ 3.6.5.1 Reactor Coolant System 7: q' w Pos74LATEo POPE BREAKS AT EMAucit couwssroows The fluid discharged from br:h n ":::ter C:: M t Spite piping will produce v . reaction and thrust forces in branch line piping. The effects of these loadings :c _ are considered in assuring the continued integrity of the vital components and .the engineered safety features. - 1 The accomplish this in the design, a combination of ccmponent restraints, ~- barriers, and layout are utilized to ensure that for a loss of coolant or steam-feedwater line break, propagation of damage from the original event is limited, and the components as needed, are protected and available. 3.6.5.1.1 Postulated Pipe Break Restraint Design Criteria for Reactor Coolant System p,pon cowsereo T* M csox ouensonswAu on MMen) -bw-ga Reactor Coolant System"p+ pag and all connecting pipirg out to the LOCA boundary valve (Figure 3.6.2-1) is restrained to meet the following criteria. Propagation of the break to the unaffected loops is prevented to assure a. the delivery capacity of the accumulators and low head pumps; b. Propagation of the break in the affected loop is permitted to occur but must not exceed 20 percent of the area of the line which initially failed. This criterion is voluntariiy applied so as not to substantially increase the severity of the loss of coolant; and 3.6-21 12/83 3 4 m q. .__,.___y .,,,,__.._.,,,__m

pp ,x E f T ,3;,, QQ c. Where restraints on the lines are necessary in order to prevent impact on ~ and subsequent damage to the neighboring equipment or piping, restraint j type and spacing is chosen such that a plastic hinge on the pipe at the two support points closest to the break is not formed. ( 13.6.5.1.2 Protective Provisions for Vital Equipment ,im a e~ a.W ' ( In addition to pipe restraints, barriers and layout are used to provide 1 7 qg. n * - protection from pipe whip, blowdown jet and reactive forces for postulated 4 n. n___ r__,__. e.. breaks. 3 %'. PAPd 5' ~ u Some of the barriers utilized for protection against pipe whip are the MM w following. The polar crane serves as a barrier between the reactor coolant N$M-loops and the Containment liner. In addition, the refueling cavity walls, ' fi % Jc 4 y various structural beams,-the operating floor, and the crane wall enclose each e ' reactor coolant loop into a separa p,g g thereby preventing an QRM% ' accident, which may occur in any 1 pi; rom f ting another loop or the t "4 M R . Containment. The portion of the steam and feedwater lines within the

M.E :

. Containment have been routed behind barriers which separate these lines from 1 SE O all reactor coolant piping. The barriers described above will withstand Q.,;,- loadings caused by jet forces, and pipe whip impact forces. ~. ,Gm. uM a ~ .0ther than Emergency Core Cooling System lines, which must circulate cooling 7 water to the vessel, Engineered Safety Features are located outside the crane , S. _ . all. The Emergency Core Cooling System lines which penetrate the crane wall w

• ~ ~

'are routed around and outside the crane wall to penetrate the crane wall in the vicinity of the loop to which they are attached. In revi, ewing the mechanical aspects of these lines, it has been demonstrated l .by Westinghouse Nuclear Energy System tests that lines hitting equal or larger size lines of same schedule do not cause failure of the line being hit, e.g., a'one-inch line, should it fail, does not cause subsequent failure of a one-s- i a inch or larger size line. The reverse, however, is assumed to be probable, discharged through the line, could break smaller size lines such as neighboring i three 'ach or two-inch lines. In this case,- the total break area shall be _'less t.an 12.5 square inches. Alternately, the layout is planned such that whipping of the two free sections cannot reach equipment or other pipes for which protection is required; plastic hinge formation can be allowed to form. As another alternative, barriers can be erected to prevent the whipping pipe from impacting on equipment or piping requiring protection. Finally, tests and/or analyses may be performed to demonstrate that the whipping pipe doe not cause damage in excess of the acceptable limits. t Whipping in bending of a broken stainless steel pipe section such as used in the Reactor Coolant System does not cause this section to become a missile. This design basis has been demonstrated by performing bending tests on large and small diameter, heavy and thin walled stainless steel pipes. 3.6.5.1.3 Criteria for Separation of Redundant Features There are no redundant features associated with reactor coolant piping system. Redundant features of other mechanical piping systems are discussed in 3.6.5.2. 3.6-22 198?, tipdate j h

^%q. sa 4; [ ,n {c? .g a--- t 3.6.5.1.4 Separation of Piping 7 + ~iThe Reactor Coolant System is separated from other piping systems and components by barriers, as discussed in 3.6.5.1.2. .G - ' " ' ' 3,0, ;,1, ; ajp; asa3traint and L;c;ticr,; hg:, >x 3 Pe.;- f:r th: ";;;ter Cecient System, : pipa rest" Mat is !cc:t;d at ;;;h of the M'L-j p % M, H lbOMS ^f t% cr000-0V0F !^g--- h[; o, I),,3.6.5.2 All Other Mechanical Piping Systems g. Ek;I 'M' easures to protect against pipe whip, jet impingement and resulting reactive 9L s 1 Om, ' N forces to meet established criteria outlined in 3.6.2.2 are as follows: h,pim s c " a.' Separation and remote location of fluid system piping from essential [ f ', structures and equipment. m' n. . [7, ' ' b. Structural enclosure of the fluid system piping with access provided for b, -inservice inspection; or, alternatively, enclosure of the essential A/., equipment. n I c. Provision of system redundant design features separated, or otherwise protected, from the effects of the postulated pipe rupture; or additional N protection features such as restraints and barriers. 7 d. Design of essential structures and equipment to withstand the effects of ~i the postulated pipe rupture. + \\ l e. Addition of guard piping for the main purpose,of diverting or restricting ' blowdown flow. e i 1 . f. In areas where none of the above can be met, or where unacceptable, more severe problems may be creased, augmented inservice inspection may be used on a case by case basis to reduce the probability of failure to acceptable levels, and not postulate the failure. The augmented inservice t inspection is in accordance with the guidelines presented in NRC MEB Branch Position No. 4 " Augmented Inservice Inspection and Secondary Pro-tective Measures." ~ Table 3.6.5-1 identifies all. cases where exceptions to the criteria of Section r 3.6 have been taken, r -i See Table 3.6.1-1 and 3.6.1-2 for protection methods on a system basis. Curbs are provided around passageways to the Auxiliary Building from the Turbine Building. These curbs are of adequate height to contain flood water caused by the break of the main consenser circulating water expansion joint, or the most severe Condensate System failure for a minimum of fifteen minutes. There are no pipe or cable chase entrances below the elevation of the top of the curbs. This flooding condition does not render any essential system or component inoperable.

1 d

a j;; + v. 3.6-23 12/83 -V

1 l The structure is that feature of the building which is a necessary part e. of the building but also is designed to accommodate the loads transmitted through the anchorage caused by the postulated pipe break. It may be either a steel or concrete component and is characterized by being y . '. n. .relatively stiff and massive when compared to the pipe break restraint. m 97 ~ ,uy. . " t,' _ f,' Allowable stresses used in the design of the pipe break restraint compo-d{& s '7 1; W nents are consistent with the component function. In general, the 0 allowable stresses associated with the total reaction force, including 'hs I impact, on the structure extension, anchorage and structure is taken as m. [' ~ the minimum yield stress for structural steel and concrete embedments. m W. For those situations where structure load limiting features cannot be D provided to' maintain the allowable stresses to within yield, plastic 1 deformation in structural components is tolerated so long as the struc-i e ture is capable of continuing its functional requirement after the

,A.ui 4

w !ky Yl} c ' deformation occurs..The upper design limit for pipe break restraint up material ultimate strain. 4 MN,h;,,'.3.6.5.5.1 . Typical Pipe Whip Restraints p g; : - P ' . A description of the typical pipe whip restraints and a summary of number and p' f . location of all pipe ruptures requiring restraints in each system is presented 1 in 3.6.4.2 and Appendix 3P. ,gh. t - 1 "4 3.

6.6 REFERENCES

1. Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant, Loop, WCAP-8082, June,1973 (Westinghouse NES Proprietary), and WCAP-8173, July, 1973. " Multiply Aperture Small Break ECCS Analysis,' L tter from W. O. Parker, Jr. 2. to Harold R. Denton, June 6, 1980. In. L ETitM FRon y_s,7uckfA (DPc) to H.R,0EN70y (sonc) D/11To th,ggy-3 o,q y y g u3 g,g,g, '^ D Con / Jo A yj j Y "'? P'?5 0#Ehk5 Y A ' 4 ~ I4 p ~ r 4 3.6-26 12/83 .--a, .e, -m-y.-. .m .e,_,,,-.-m.-_. g,.. ,-p, 9~ g,, -. -. -w .ev .--,.,.,._y,mye,9

. -My..;;.. o m i.s.:';:tn S q ( "* Tt; 3, ~ ~. .g '%g a Table 3.6.1-3.* .~ Comparison of Duke Pipe Rupture Criteria And fikC Requirements of Branch Technical Positfon' APCSB 3-1 Tilarch 1973) and NITC Regulatory Guide 1.46~(May 1973) NRC Criteria Duke Criteria APCSB 3-1, Section B.2.c SAR Section 3.6.2.2 i i s. p Section B.2.c. requires that piping between containment f' isolation valves be provided with pipe whip restraints Duke criteria is roughly equivalent to NRC c'riteria p capable of resisting t,ending and torsional moments pro-as clarified below: duced by a postulated failure either upstream or down-stream of the valves. Also, the restraints should be The containment structural integ'rity is provided for designed to withstand the loadings from postulated all postulated pipe ruptures. In addition, for any failures so that neither isolation valve operability postulated rupture classified as a loss of coolant nor the leaktight integrity of the containment will be accident, the design leaktightness of the containment impaired. fission product barrier will be maintained, 3 Terminal ends should be considered to originate at a Penetration design is discussed in SAR Section 3.9.2.8. point adjacent to the required pipe whip restraints. This section also discussed penetration guard pipe design criteria. Terminal ends are defined as piping originating at structure or component that act as rigid constraint to the piping thermal expansion? APCSB 3-1. Section 8.2.d SAR Section 5.2.8

j. (1) The protective measures, structures, and guard h

pipes should not prevent the access required to Duke criteria is different than the NRC criteria due

y conduct inservice inspection examination.

to the code effective date as described below:

o

![ (2) For portions of piping between containment isola-ASME Class 2 piping welds will be inspected in accor-y.. tion valves, the extent of inservice examinations dance with Tables ISC-251 of Section XI (1971), completed daring each inspection interval should through Winter 1971 Addenda, of the ASME Code, as t' i '( provide 100 percent volumetric examination accessibility permits. Inservice inspection program ii circumferential and longitudinal pipe welds,of requirements are given in SAR Section 5.2.8. s ,t .t 1

g.

..g. p;p b ra, k s in the ACS Pm='Y oop arc. nef" pasivlated h' c*nsider' <Nion in cerfain aspeeds l (, of p/ nt design, c s &cAnad in A*l "nCc l* %

6 c.

{ l:,. r .,:. 25 A 12/83 m (T m.. nl>. :.,.~=. %n.._w. e; .3 M n ~ = L L*,. n L n qm. _g. ~.. 3_.c.. ..m

4 ~) - Table 3.6.2-1 _ Postulated Break Locations in Reactor Coolant Locos .!n..G. iy.y,,; Location of Postulated Ruoture Type

3. n,~ M 1.

Reactor Vessel Inlet Nozzle 3 % ).- Circumferential .. ~ = s,..,U M 2. Reactor Vessel Outlet Nozz.le. y ~- Circumferential i-M 3. ' Steam Generator Inlet Nozzle f-Circumferential T 7 . f.,p 4. Steam Generator Outlet. Nozzle Circumferential W M. % 5. -Reactor Coolant Pump Inlet Nozzle Circumferential M 6. Reactor Coolant Pump Outlet Nozzle Circumferential .e-

• 7.

50 Elbow on the Int.rados Longitudinal .,,.,W .* 8. Loop Closure Veld in Crossover Leg Circumferential 9. Residual Heat Removal (RHR) Line/ Primary Coolant Loop Connection Circumferential (Viewed from the RHR line) 10. Accumulator (ACC) Line/ Primary Coolant Loop Connection Circumferent'ial (Viewed (' from ACC line) ( 11. Pressurizer Surge (PS) Line/ Primary Coolant loop Connection Circumferential (Viewed from the PS line) 'M-x 4 %farence. I Jsfinez ~ c::.st .. u ;Q,. ' +he o<, yin } inris A< pns/J;,y pipa-bee = ke in '** *fer coolant sys/em primary loop. f.}..

1. " c Nm inaling +his Refa<enee /.s, peavider He pav:ausly paduMed bs.ris ' ':4

SPcc Is a dessyn co,,;;s,.J;,n, p spe. 6,ce k f<o m ce,-hin - e ~~ " = ... C+$ .'. rik. e WMyrj 6.&'.i'[h. ..? ;.-s>. ..h. .y. ..v 3 . MN

u

.w ug' k ? d.} ? .....i,1 +'

• '12/83

... ;T'e ' 'f ,..:c. m44..w*2dh& -e& 6Nd

.].) y Jet impingement forces are established and defined as follows: t 1. ~ ',.. Pipe break size and locations are as described in Section 3.6 of the a. ,f.: FSAR. - ~ - -- : pgijy.i 'l ,m. b.

>; a, Jet force orientation was' established as follows

2 ER 1) For circumferential breaks, the jet orientation is limited to 3. ".4}{.,'.,,%' elastic deflections at the ruptured end of the pipe relative to f.', the nearest anchor point if the pipe section remains elastic at TN', ' t the anchor when a hypothetical load equal to the jet load is' c.9. A..~... - applied perpendicular to the axis of the pipe at the ruptured '? I T .GC. T end. If' the pipe section is partially yielded but a full i ~.,4 plastic hinge does not form, the jet orientation is established j by a hypothetical lateral displacement of one pipe diameter i.'; p *h, relative to the axis of the pipe. If a plastic hinge does form at the anchor, lateral restraints are provided or the jet orient- } ation will be estooiished considering 6;l orieniations relative "J 1 to the hinged anchor. e 2) t For longitudinal breaks, the jet orientation shall be'estab-lished at the break location in a direction perpendicular to V'T the axis of the pipe. I j Pipe sleeves, physical restraints and obstructions have been considered in evaluating jet orientation and impingement area. J Jet impingement pressures were established as defined in Reference 10 except that the jet dispersion angle has been a ten degree one-half angle relative to the orientation axis. To account for the dynamic response of an object subjected to a i .- r jet force, a dynamic load factor (DLF) based upon the ratio of } natural frequency of the model to the duration of the jet-force i .4 mt .m as defined in Reference 3, has been used to establish an $.3 i f.) equivalent static load.

1. .@ (

Theoveral5interiorstructureisdesignedforthemaximumup- ..: W:q y > . v.wd r lift, horizontal shear and overturning moment. .' % } 4 ,c i 9..:g'- l h Each break location in the lower compartment has been evaluated a .a, to establish the maximum uplift, horizontal shear and over- ' ~ ~s..g + turning mcments on the interior structure. ' ;J i a tabulation of the maximums, the time at which the maximumTable 3.7.2 I T. i .t occurs, and identifled the break producing the' maximum. ,3-iW 7 + related forces occurring at the same time as the maximums are, JCh7Ik The l" combined with the maximum for the final design. z.g : X MsiM T # h:? 3.8.3.3.2 Dynamic Analysis ( i i I '^ j ation are determined by a dynamic analysis as descri nd.y, ! ".( 3.7.2.1(B). j 2 1 1 m l i I ' C; i 3.8-20 e @.uhF

d. M I C + w + d M...M,.

12/83 . z. 2 s..L. s._

Th e loa c!:ny c esc <,' he d l a.bove-u.4.'/; acel ;n -oe sa c.;,, s f - w e<e +^e in h <i < s&va we. so 6 aeg va,,6 o to e,ic des.yn Jfa bre* k ce,'/eeic~ ,cv,.,,a p, sis /cs w sa ;,, ),, y,,, g,, y 9,., m 3g, yg, f,, g,.as, iA,,,f' '

• • d d;ff,,,,, 4;o.}

p,,,,,,,, ,,,,cn t(e, ,n ,,,,. 3, ,,3 ,,, 3,, _ a. ". ,glb'l**diqfSc'P'cIEuc ty,,,e ,up "p? ic l l=/c 4s le d boL ecpecer,d upps-havnd 'c*0 Nin) -bu u ~ p/. /e 4 p.pe m 4,a,r 7se Ll dif>G er. lia l { * ~ ~T Co~Pa'Emen 21 diffe<en f ;. I _ p<essuec g ye, ;,,, jj g,,,, & <lesig n. j,,, ' ),; y,,,, ma x a. .? '.'.lw Se 'O O w e g 4 e 6 "" 9 O ,g $6 p 5 J 4.. . *;p;::q 4Pp;3'..g - [9' % a g ..+. a, .s %4 'i '#.] 4 s*.* .. ? ',. 9

• lskhE*}}

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)

4 .% 4 1 f3 *. . 4,- c? 7 t . 9-i, .t' e s {." A s e '. a y gy 4.* ,,,e,, ', "fg',,,,',

The design of the Internal Structure is in accordance with the requirements of the applicable codes as shown in Table 3.8.3-3. The forces and moments of each of the Internal Structure components are determined from established analytical procedures and computer structural analysis programs. outline of the analysis and design methods are as follows: A brief h.m. Containment Floor Slab: -l* (1) 'The Containment floor slab is a fill slab, therefore, minimum rein- ~ ...,forcement i,s.provided as,specified in the ACI-318-63 Code. 'T g r ;'n .. (2) Reactor Vessel Cavity Wall: /' C 4*" d'3 #" c'" The reactor vessel cavity wall is designed to accommodate loads generated by 10t$ nd support. The reactor cavity wall is analyzed as a thick .n. . cylindrical wall and the steel reinforcement is basically a hoop rein-forcement pattern. (3) Vocer Reactor Cavity: The upper reactor cavity is analyzed as a space finite elementi model together with the refueling canal walls and floor. is designed to accommooate loads due to tiR which consist of internal and The upper cavity wall external pressures. %f ju j,./g p j,4 (4) Refuelina Canal: The refueling canal walls and floor are analyzed as an integral part with the upper reactor cavity space finite elecents model. (5) Crane Wall: sLJed P'P' .n j A space finite elements mathematical model is utilized in order to per-form the analysis on the crane wall. The crane wall is designed as a e, secondary barrier for th GOA differential pressuresCacross the various .t; compartments of the Interior Structure. are inouced on the crane wall. Asymmetric loadings due to

• p g /.dcJ.1' 4 <,.. hs For this type of loading, the computer' pipe.

program No. 2 of Subdivision 3.8.3.4.2 of the FSAR is utilized to perform J3 7.yt, the crane wall analysis as a shell of revolution subject to asymmetric '.f T loads. The polar crane is designed for the loadings as defined in Table.,., pM* 3.2.1-2. The polar crane is not in use during unit operation and does

• W

,not have an influence on safe shutdown of the reactor. . ',I.}.ry (6) Steam Generator Comoartments: ..G& The steel dome of the steam generator ccmoartment is analyzed as a thin es'@O shell of revolution employing Kalnin's orogram (flo.1) described in Sub- '. r,s division 3.8.3.4.2. compartment are idealized together as a plane finite eleme S steel dome, the side walls and the cylindrical steel shell of the steam The due to main steam line rupture. generators ccmoartments are designed to ac .t 'l includes the high strength anchor bolts meets the requirement ASME Code,1968 Edition, St.bsection 8. The connection between the steel 3 a 3.8-22 12/83 .C fe , s u. ,%*w.,'. m: ~

• 4.

A ~

. c.' i and concrete portions of the steam generator compartment is designed to -IC permit removal of the steel portion. The connections consist of embedded anchor bolts and exposed nuts which can be removed when desired. w. ' ' (7) _ Pressurizer Comoartment: p i The pressurizer compartment is designed for the internal pressure due to ~c- - \\... s l p/ pipe rupture-inside-the compartment. A plane finite element model is ' ;7 utilized in analyzing the pressurizer compartment. (., i z,3m p(8) .g;g ! (..-... Divide'r Deck: .3_..- ~ ,).y; Q,g .A plane" finite'e1.ementsmlate bending model is utilized to perform the

• 5 D V.P ;

analysis of the deck. The divider deck, which is the main pressure s[ .(: , barrier betwee.n.the lower and upper. Containment interior compartments, is n W kk [ designed to accommodate the D86 gr~=Ws and the jet impingement loads. i 3 ", I (9) Ice Cendenser Floor: " '# E (__.' p a."bre.'ks M" The ice condenser floor is designed for theM4r differential pressureg &c. as well as the weights of the ice condenser components, and the reaction p, pe. b,.. k s loads from the ice condenser lower support structure.

A v A plane finite element model is utilized in performing the analysis of the ice condenser floor.

i (10) Eouloment Floor: L l ) A plane finite elements mesh is utilized to represent the equipment floor. i The analysis is performed by employing some of the computer programs out-lined in Subdivision 3.8.3.4.2. t % and the weight of the equipment.The equipment floor is designed for the i i \\. (11) f c pa<M en t .//fe<ea.4,<./ pess. e- &m pesMdec' p.A. brec.kr i Accumulator Wino Walls ~ 1 A plane finite element model is utilized in performing the fr., pasivlahL i the walls. The adequacy of the walls has been demonstrated using the ~ ^ ultimate strength design method as outlined in Standard Review Plan 3.8.3 -TOI# for the load combination: U=0+L+Ta+Ra+1.5Pa. ' Tdp; 3.8.3.4.2 /,.dife Computer Programs for the Structural Analysis

.yggfigj ;

.. 3....fj The following comauter programs are employed in the analysis of Category 1 structures: 6'.~n. s

1. • ~ 1.

For the stresses, stress resultants and displacements produced in a thin M-i shell of revolution due to static and seismic loads: 't* written by Professor A. Kalnins of Lehigh University, Bethlehem,A computer prog Pennsylvania. description of the program. Refer to Subsection 3.7.2 and Subdivision 3.8.2.4 f s., I!

, l

,u

• ?',I l -f.~

l .u., 3.8-23 + 0 12/83 ..'..O..

• , Q!,{.?l,D{p u ; /.: y,4*,g.N

?- c m p..t *

f. 'J '..

8: 'I.d #., 'ed .-an/a4y( M"/J.JaT-- *

• N W'M N

Tha tubes in the steam generator are subject to a possible flow-induced vibration that does not exist in the primary coolant loop. could result from flow across the tubes due to vortex sh This vibration To ensure wide-frequency separation between the vortex frequency of the fluid and the , there is a beam frequency of the tube. Parallel flow vibration is analyzed using the correlations of Burgreen, and the amplitude of vibration is shown to be low /, enough that neither stresses, banging, nor fatigue is a problem. discussed in greater detail in Subdivision 5.5.2.3.5. Analyses are . p,, i, y, 3.9J1r3----Dynamic System Analvsis Methods for' React , l ':# + or Internals ... '. h 'The r'eactor intern ~als Ire modeled dynamically for: double ended pipe rupture of the reactor coolant loop,. (the Design Basisa) loa ~ Accident', DBA),' for both' cold'ahd~ hot leg' breaks;'b) response due to an 4 40' Operating Basis Earthquake (OBE); and c) for the most unfavorable combina of DBA and OBE. described in Sections 3.7.2 and 3.7.3. Seismic analysis of the reactor vessel and i ~4: M 4sM T 1 The upper internals support structure is made of two plates e connected by hollow columns bolted to the plates, with The upper to the core plate. This structure compresses the fuel assemblies and the annular hold-down spring during assembly and is subjected to vertical upwa forces from these springs. During operation, normal and transverse flow drag forces are applied to the columns and guidetubes and differential pressure e plates. the outlet nozzles. Because of the complexity of the upper package geometry and loading conditions, the modeling of the reactor internals was pe using the method of analysis based on the finite element idealization of t structure and matrix displacement for each finite element. . ~. finite element structural analysis computer program permits static eld ti This plastic analysis, steady state and transient heat transfer, dynamic mode s c and e Q" ;. analysis, linear and nonlinear dynamic analysis, and plastic dynamic ape analysis. the internals follow. Descriptions of the techniques used to model the various p , g. g '. '3'.e?"? The top structure, deep beas 'and the upper core plate have been modele A m. flat shell elements, the support columns with "three dimensional" beam with sM elements and the fuel assemblies and hold-down spring with "three dim , Usi spring elements. Because of symmetry, a one eigth slice of the upper package nal" 'E has been modeled. The core plate is perforated and is modeled as a according to the theory of perforated plates. geometrically equiva .. q g ,q,f, TV.. ~ ,, 14 Columns of two different lengths are modeled, the long columns connect and the short columns connecting the beam grid with the upper core pla ng the plates e. Under the loads used for design, according to the o ,. p,, on under deflections at all nodal points. u . ' ' ^ i.) 3.9-4 12/83 ?*7~$ .n.. x.wim. MMWSEhd

DUKE POWER COMPANY Form 00184 (6 81) o \\ l Dev./ Station Unit File No. Subject By Date Sheet No._of . Problem No. Checked By . Date N ^

• Mparal%-fcueyn,oA on-fqyc 2LAl

/b_ _7 racl 7 /1

• Gwre aJs rin./cesals-s /,

lure L(auc dec< e/7 4el neO xisLa A II]A 1 ~~ Abs.s And de.sdLbvaL a ao.ra d .Lekan 1- .+aL DLJau- -ecs.d ~Q8spk.bA.l$,1a n 4A A)~ks - w.ta aa ~ tp a LaLa_4g.e), gl,1at A,- Jp,,L ykG+(: x a L awaLLw v. t e

m' e 14

• }

• 'd
• i,

i

d. Stress levels for the transfer tube meet requirements of ASME III , Class 2. Table 3.9.2-2. Stress levels for the anchor ring component meet the requiremen e. an Design codes applicable to the Fuel Transfer Tube Penetrations are as follow h ,f The penetration is in accordance with ASME III Subsection NC. a. transfer tube is designed, fabricated, and inspected to ASME III l The fuel NC with the allowable stresses as defined above. , Subsection ,yed j -/"'@ b. Attachment welds to the fuel transfer tube and field welds betwe cy Containment and the anchor ring component meet and are inspected to ASM ~ ..'.?[ III, Subsection NC. "N" RQN& 3.9.3 COMPONENTS NOT COVERED BY ASME CODE o Pressure Vessel Code are tabulated in Table 3.2.2-2. Safety- .y:. s ~. components are as stated in the various system descriptions containing these " P '- components in Chapters 6.9, 9.0, 10.0, 11.0, and 12.0. be determined from Table 3.2.2-2, components requiring seismic reliabilitS r.?"Q designed to the same criteria as an ASME comoonent y are - f. the manufacturer has the option of performing detailed seismic desiAs describe j or conducting seismic testing. gn calculations components must undergo successfully a quality assurance a as described in Chapter 17.0, and appropriate QA documentation in strict accordance of Duke' e ( which fully describe necessary NDE and QA documentation requirements s structural integrity of functional capability is pr,esen . A summary ^' r a topical report once these components are ordered and the work ha 4 rmed. 3.9.3.1 Heatino, Ventilation and Air Conditioninq i I ~ ' HVAC equipment is designed as described in Section 3 9 3 and in acco d

e,p Table 3.9.3-1.

m.i; r ance with .,S;,, g"{ t 3.9.3.2 Reactor Core and Vessel Internals Not Covered By ASME Code ..o The response of the reactor core and vessel internals under excitatio . - Q'/ ' a ;:g . by a simultaneous complete severance of a reactor coolant pipe and se

l u,

ed excitation for a typical Westinghouse Pressurized Water Reactor Unit

' ' '(j has been determined. The following mechanical functional performance criteria s .. ~. V/" apply: s pn g g W$.)'. criterion to be met for the reactor internals is that a. ':h shutdown and cooleo in orderly fashion so that fuel cladding tempera is kept within specified limits. et .ation of certain critical reactor internals must be kept suf fici i W small to allow core cooling, '], rje bue.1 een u m Le/ ,,g,,, jg f /

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b. For large breaks, the reduction in water density greatly reduces the reactivity of the core, thereby shutting down the core whether the rods are tripped or not. Cooling System uses borated water to maintain the core i state. Therefore, the main requirement is to assure effectivenesss of the Emergency Core Cooling System. Insertion of the control rods, although a not needed, gives further assurance of ability to shut the unit down and Keep it in a safe shutdown condition. ~- . ~ r ' accident are shown in Table 3.9.3-2.The functional requirements ,c. 7 are controlled to insure no contacting of the nearest rod cluster cont guide tube. t to maintain a.The outward upper barrel deflections are controlled in order ..Q diameter and core barrel outer diameter.n adequate annulus for the coolant - N ". ' . u... ~ . f.. ~ W.M'E' d.- The rod cluster control guide tube deflections are limited to insure operability of the control rods. GU To insure no column loading of rod cluster control guide tubes e. core plate deflection is limited to the value shown in Table 3.9.3-2 , the upper a fy f. design core and internals and to assure that the co design basis accident operating conditions (1, 5) acceptable operating basis earthquake, safe shutdown earth g. ~ (1, 2, 3, 5). 3.9.3.2.1 Faulted Conditions The following events are considered in the faulted conditions category .m pew tMel t basis accident, for both cases: Loads produced by a/ double ended ..? a. esign cold and hot leg break. The methods of -.W # 7 break or hot leg break). analysis adopted are related to the type of a 'YNE Response due to a safe shutdown earthquake. 19.m b. ~ e,3.:.3 c. Most unfavorable combination of safe shutdown earthquake and desi '/iQ accident. Maximum stresses obtained in each case are added in gn basis - ' 'l . conservative manner. . E;p QjG6 Table 3.9.3-3 for each of the above conditions. Maximum stress .m..mo ven in establish the usage factor.the applicable stress concentrations factors and p j[ i of the structure and the stress analysis on each component is used to ,e an el'astic basis. For these cases only, when deformat' ion requireme 'M locations. A ,a 3.9-30 . 3. 12/83 ,i ) I' '. l ~. 6 _ _,,. u &b N

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is plastic analysis is independently performed to ensure that functional requ ments are maintained (guide tubes deflections and core barrel expansion) t elastic limit allowable stresses are used to compare with the result of the The analysis. No inelastic stress limits are used. The above-described analyses show that 'the stresses and' deflections wh result following a faulted condition are less than those whicn would advers affect the integrity of the structures. quencjes are such that resonance problems should not occur.Also, the natural and ?;;. 3.9.3'.'2.2 Reactor Internals Response Under' Blowdown and Seismic Excitation w. ~ pe,M aJed.

- A loss of coolant accident would result from a4 rup

'~ ,; ) to vertical and horizontal excitation as a result of rarefaction wave e 6; inside the reactor vessel. gre'atly reduces the reactivity of the core, thereby shutting down the whether the rods cra tripped Or not. The cubsequent refilli.; :f the c;r: the Emergency Core Cooling System uses borated water to maintain the core i by subcritical state. of the Emergency Core Cooling System.Therefore, the main requirement is to ass a although not needed, gives further assurance of ability to shut the plan ~ ' < and keep it in a safe shutdown condition. The pressure waves generated within the reactor are highly dependent on the location and nature of the postulated pipe failure. A one millisecond severance time is taken as the limiting c components. In the case of the hot leg break, the vertical hydraulic forces produce an initial upward lift of the core. reactor hot le'g nozzle into the interior of the upper core barrel.A rarefaction wave has not reacneo the flow annulus on the outside of the barrel Since the barrel is subjected to an -impulsive compressive wave. , the upper (buckling) or large deflections of the upper core barrelThus, dynamic instability response of the barrel during hot leg blowdown. , or both, is the possible .Y!P the fluid exits the hot leg nozzle.the hot leg break results in transverse p,.. omponents as - j,l$, In the case of the cold leg break, a rarefaction wave propagates along 7.l inlet pipe, arriving first at the core barrel at the inlet nozzle of th a reactor loop. impulse which changes as the rarefaction wave propag e broken ,egh ' V. ' and down the outer flow annulus between vessel and barrel. e barrel ~~ break, the initial steady state hydraulic lift forces (upward) decreaseAfter the My (within a few milliseconds) and then increase in the downwar rapidly M.QQ cause the reactor core and lower support structure to move initially downward These D." N If a simultaneously seismic event with the intensity of the safe shutdown quake is postulated with the loss of coolant accident, the imposed loading earth-U the internals component may be additive in certain cases and theref combined loading must be considered. .g ore the by the earthquake is small compared to the blowdown loadingIn general, n ""~^M,4 .~. 3.9-31 12/83 f 7.Y.?,. .w., .:.,z. A.l%2EEUY

e 3.9.3.2.3 Acceptance Criteria =. The criteria for acceptability in regard to mechanical integrity analy that adequate core cooling and shutdown must be assured ses is the geometry remains substantially intact.the deformation of the This implies that ~ so that established on the internals are concerned principally with the maximumConse ,,e allowable deflections and stability of the parts in addition to a stress criterion to assure integrity of the components. x I.llowable Deflection and Stabil,ity Criteria . f t,w., a. ^ e,.n. ~~ ..} deflections of critical internal structures are limited t w . - p., 4s N. J: given in Table 3.9.3-2. of the internals, energy absdr'bing devices limit the displacem - ' N. M,. es . inches by contacting the vessel bottom head. /Md py-W Uccer barrel. The upper barrel deformation has the following limits: j'?" ,J 1. To insure a shutdown and cooldown of the core during blowdown locations of the inlet nozzles connected to the uncroke , the basic .,y e outward deflection of the barrel in front of the inlet nozzles A large with permanent strains, could close the inlet area and stop th , accompanied coming from the accumulators. e cooling water in front of the unbroken inlet nozzles, larger than a certain' limitCon the " hot-loss of function" limit, could impair the efficienc called Core Cooling System. y of the Emergency 2. To assure rod insertion and to avoid disturbing the control rod cl guide structure, the barrel should not interfere' with the guide tubes uster This condition also requires a stability check to assure that th barrel does not buckle under the accident loads. e Control Rod Cluster Guide Tubes. ~ package nouse the control roos. The guide tubes in the upper core support . Ls tests and are provided in Table 3.9.3-2.The deflection limits are established from ' ~ J'?,@If. Fuel Assemb1v. . "# 2 of tne tnimoles in the upper end.The limitations for this case are related to th buckling of the upper end of the thimbles due to axi g.h Any 4* distort the guide line and thereby affect the free fall of the n could d# control rod. Upoer Packace. wnere a gutoe tube is located, shall be below 0.100 in.The local v , z.J;,,; ~ LM;.?lnn plate and guide tube is 0.100 in.causes the plate to contact the guide tube -...S from undergoing compression. This limit will prevent the guide tubes ~ the guide tube is compressed and deformed transversely to thFor a pla previously established; consequently, the value of 0 150 in e upper limit . is adopted as O[ 3.9-32 12/83 ...j ,... / a, m A. -- .,2.ws.b.DdN.L-&

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s - 3.9.4 References 1. Bohm, G. J., Indian Point Unit No. 2 Internals Mechanical Analysis for Blow-down Excitation, WCAP-7822, Westinghouse Electric Corporation, (December,19 2. house Electric Corporation, (February, 1972).Kraus, S., Neutron ~ ~~" -.. 3. WCAP-7879, Westinghouse Electric Corporation, (July, 1972 R., et. al., Four Loop PWR Internals Assurance and Test Program, s 3.y. Moore, J. S., Westinghouse PWR Core Behavior Fo'11owing a loss-of-Coolant .. n _....h;r, 4. Accident, WCAP-7422, Westinghouse Electric Corporation, (September, 1971). e.,., _t., 5.i r- ,.. a 5. ' Accident, First Intl. Conf. on Structural Mech.Bohm, G. J. and J. P -(Sep temoer 20-24, 1971). in Reactor Tech., Berlin, ....:e. 6. Westinghouse Electric Corporation, (October,1970).Fabic, S., 7. Force Transients in Liquid Filled Piping Networks, Kaiser Computer Program WHAM for Calculation of Pressure Velocity, and'- /, No. 67-49-R, (November, 1967). 8. Fabic, S., Loss of-Coolant Analysis: Comparison Betweeen BLODWN-2 Code Results and Test Data, WCAP-7401, Westinghouse Electric Corporation, (November,1969 ' 9. WCAP-8252, Westinghouse Electric Corporation, (March {

10. Prediction of the Flow-induced Vibration of Reactor Internals by Scale Model Tests, WCAP-8317-A, July, 1975.

1

11. UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Functional Examinations, WCAP-8517, July,1975.

,','... [., e; 'M . 12. Bloyd, C. H., Singleton, N. R., and Claramitaro Pad and 17 x 17 Guide Tube Desian by Preocerational Tests on the Trojan 1W .Qf$Z Power Plant, WCAP-e780, May, 1976. ~ c"W M,*\\,. '.i - 13. (Sequoyah Preliminary Results Letter), Verification of Uooer Head Injection Reactor Vessel Internals by Precoerational Test on Seouoyan 1 Power - -M$.1 Plant, ',iCAP-9945, March, 1981. 6* - ) i = ,. fe4&r. l'l. $ctYce h.:,, Y A l~dN (DPC) /o //.R Den h,q GyAC),da /cd , /t,1 L-4ran sm,I{ing uksdos), Lu: e-4 re urf p f;f..y" e./.m.n a locn f y of Acs p<, v.,y be,a s c m ,s,. ec,. x,,, y i 3.9-40 12/83 l , n.,s_., s.4.a -4 h E * .! + ._._,_...___..'._JC'-.. _. _, _ _

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Equips nt supports are designed in a way to allow virtually un lateral thermal movement of the loop during normal operating c restrained onditions. 5.5.14.2 Desion Description a. Steam Generator - - ~ Trie steam generator support system consists of vertical steel col lower and upper lateral steel frames.

umns, 5 5.14-3 for outline of the steam generator support systemSee Figures 5.

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- 'k. The reactor coolant pump su ~ and a lat ral~ steel frame. pport system consists of vertical steel columns ~ " J' ~ . outline of the support system of the reagtp_r__ coolant _ pump. .g.. c.. Pressurizer The pressurizer (Figure 5.5.14-6) is supported by: 3 l. wall and to vertical steel columns which in turn n e_ operating floor, and 2. An upper steel ring anchored to the crane wall and the pressuriz enclosure walls (Figure 5.5.14-8), er d. _ Reactor Vessel Figure 5.5.14-9 shows an outline of a typical vessel support 9 vessel nozzles and anchored to the primary shield wal The supports support shoe into the vessel support by direct beari Loads in both the a ' A cross uver Leo e me ..e The cros, r leg pipe is restrained as shown in Figures 5.5.14-5.5.14-13. . e.x, The straining the pipe in sort consists of a reinforced concrete s rough f7' crane wall. lateral direction and stee ure re- '7. es anchored to the 72% Qg f. Hot and Cold Leo Picino The hot and cold len-r the primary loop are restran Figures 5.5. - and 6.2.1-103. or Building Floor. primary shield wall sleeves and steel pipe attached toeel -h5 ' s shown in These restraints consist o ings i i a (\\ ,Q 5.5.14.3 } Structural Accentance Criteria The equipment supports are designed such that the stress levels i 3: ?., i supports are below tne yield stress and for the supports to maintain th i n these elastic behavior if subjected to any of the loading combinations er Oc /e/c r e n /~y 4 5-36 s.s. M -/0 0 v. S 5 N~N and 12/83 e

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4 2) All structural heat sinks were at temperatures consistent with the current analyses. For this case the ice bed melted out at approximately 67500 seconds. The peak pressure following ice bed meltant was less than 4.0 psig. Since design with respect to demonstration of containment structural integ 4 6.2;1.3.7 Containment Environment, Safety Features Performance and . ~ - P,. Energy Input Curves The pressure curves indicate the Containment total pressure as a function of ..N C time after the accident. The temperature curves show the temperature of the - Heat removal rates for the Containment s'afety features ar ~ 0 i discussions of the short and long-term analyses. The energy input curves and tables show the input rate of energy. l The curves, Figures 6.2.1-24 and 6.2.1-25, present the complete upper and low compartment temperature transients to 105 seconds. The transient was generated 7, by the LOTIC code. i 6.2.1.3.8 Subcompartments Consideration is given in the design of the Containment internal structures to localized pressure pulses that could occur following a loss of-coolant acci-dent. If a loss of-coolant accident were to occur due to a pipe rupture in these relatively small volumes, the pressure would build up at a rate faster than the overall Containment, thus imposing a differential pressure across the walls of the structures. These compartments include the steam generator enclosure, pressurizer en j sure, and the reactor cavity. blowdown flow resulting from the severance of the largest conn 4 q' p. within the enclosure or the blowdown flow into the enclosu adjacent region. N ^:1, + hud C. The following paragraphs summarize the design basis calculations: _ Steam Generator v; y The largest break possible in the upper cavity of the steam generator e .i w is a double-ended break of the steam line pipe at no-load conditions worst break location is considered to be at the steam generator nozzle-The L'p junction which is upstream of the flow-limiting nozzle. location in the upper cavity where the steam line is not enclosed by a conti This is the only I ous guard pipe. A full guillotine break is assumed at this location. full flow area is not developed because pipe rupture restraints and guides ar

However, used to restrict the movement of the ruptured pipe after rupture absorbing restraint system involving the process pipe, a continuous guar An energy internal pads between the process pipe and the guard pipe, load transfer structure on the guard pipe, an energy absorber and load transfer structure on 6.2-32

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