ML20196C132

Forwards Util Responses to Staff Concerns Re NUREG-0824, Section 4.12, Design Codes,Design Criteria & Loading Combinations & Unresolved Items from SEP Topic III-7.B

ML20196C132
Person / Time
Site:	Millstone
Issue date:	02/08/1988
From:	Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-NUREG-0824, RTR-NUREG-824, TASK-03-07.B, TASK-3-7.B, TASK-RR A06423, A6423, B12811, NUDOCS 8802120257
Download: ML20196C132 (35)

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NORTHEAST UTILITIES o.n.,.i On4., . s,m.n sir..i e.,nn. Conn.ciicu, 9 2.C.YUU[$.T. P O. Box 270 k k J 7Cd'.7*CE.

.. w ww ~ u H ARTFORD, CONN ECTICUT 061410270 (203) 665 5000 February 8, 1988 Docket No. 50-245 B12811 A06423 Re: SEP Topic !!!-7.B ISAP Topic 1.19 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:

Millstone Nuclear Power Station, Unit No. 1 Design Codes, Criteria, and Load Combinations SEP Tonic III-7.B In a letter dated March 10, 1987(I) the NRC Staff requested that Northeast Nuclear Energy Company (NNECO) provide a schedule to rasolve the following unresolved items from SEP Topic 111-7.B concerning Hillstone Unit No. 1:

o Unstiffened Compression Members o Coped Beam Shear Connections o Column Splice Stress Reversal o Containment Bellows Design o Load Combination Pipe Break and SSE o Extreme Environmental Loads Due to Snow In a letter dated June 2, 1987,(2) NNECO committed to provide the requested information. Accordingly, attached are NNEC0's responses to the Staff's concerns. Since SEP Topic !!!-7.B is part of ISAP Topic 1.19, this submittal is also applicable to that ISAP topic.

(1) C. O. Thomas letter to E. J. Hroczka, "NUREG-0824, Section 4.12, Design Codes, Design Criteria and Loading Combinations," dated March 10, 1987.

(2) E. J. Hroczka letter to U.S. Nuclear Regulatory Commission, "Hillstone Nuclear Power Station, Unit No. 1, Design Codes, Criteria, and Load Combinations, SEP Topics !!!-7.B/Il-2.A," dated June 2, 1987. p.0 5 0802120257 000200 t 1 DR ADOCK O 25

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U.S. Nuclear Regulatory Consission B12811/ Pag 22 ,

February 8, 1988 '

If you have any questions, please feel free to contact my-staff.

wry truly-yours, >

NORTHEAST NUCLEAR ENERGY COMPANY

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J E.'"3fMroczka # '

Senior Vice President cc: W. T. Russell, Region I Administrator -

M. L. Boyle, NRC Project Manager, Millstone Unit No. 1 -

W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 i

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. .f u .. . e Docket No.- 50-245 011111 A06423 Millstone Unit No, I SEP Topic !!I-7.8 February 1988

h, Millstone Nuclear Power Station, Unit No. 1 Request for Additional Information NUREG-0824, Section 4.12 Design Codes, Design Criteria, and Loading Combinations l

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i INTRODUCTION In its safety evaluation xalating to NUREG-0824. Section 4.12 "Design Codes. Design Criteria, and Loading Eembinations (6)," NRC-Staff-has requested additional-information to resolve six open issues. These issues are:

In developing responses to these issues, Northeast Nuclear Energy Company (UNEco) has performed systematic evaluations-using the following general procedure. First, .the-relevant data were assembled for'each-issue as-it applies to the entire plant using-information in previous submittals, plant drawings, and walkdowns. Next, the data was. evaluated by performing: conservative, bounding analysss. Finally,-the.

evaluation results were summarized to facilitate resolution of each issue.

The results for each topic are presented in the following sections. Supporting details and photographs are contained' in tables and figures at.the ond. In each case the results are satisfactory, and it may be concluded that Hillstone 1 meets the requirements of the topic in question.

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1. COMPRESSION MEMBERS Question

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Current Code Requirements for Compression Herbers (AISC 1980 vs AISC 1985): The staff 's :oncern is tha t the web of most standstd tee sections when used as a compression renber does not =cet the current code require ents, i.e. buckling of the member would be expected, as discussed in the attached TER.

A sampling approach was chosen by the licensee to address this issue and the evaluation results showed that one of the sa:ples did not meet code limit. The licensee argued that this non-conformance occurred in members used as zanhole pit cover stiffeners which are not the rain load carrying

=erbers and concluded that the applications of standard tee sections a t Hi11s tone Unit 1 are acceptable for this code change. The staff agreed that the sa:pling approach wculd be a reasonable way to address this issue. However, the details cf sa pling procedure, sa pling basis, sa ple si:e, evalua tion results, and basis for the conclusion should be submi t ted for review. In addition, the licensee should expand sample si:e to provide assurance that the remaining "standard tee section" cc pression members will satisfy the code requirements. As an alterna tive, the licensee may demonstrate that the failure of these "s tandard tee section "

cc:pression nerbers will not degrade the safety targins of structures.

Response

This issue concerns new requirements in the 1980 AISC Code providing stress reduction factors for unstiffened compression elements with one free edge parallel to the compressive stress. The new requirements were more -

conservative only for the stems of tees. In the previous 3

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-submittal (2].a sample of tee sections was checked to see whether the new code would have required a reduction in

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allowable stress. One section in the sample would have had a

r. eduction.

In order to resolve this issue, a comp 1cte. review of the structural steel drawings, supplemented by a walkdown,,was carried out to identify the size, location, and functionHof all tee sections used in the plant. It was found that the

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original sarple included-all of the tee sections used in the plant.

Table 1 provides a summary of the evaluation of each tee section in the plant. Several of the sections were found to have their stems welded to adjacent steel, =aking the new requirement not applicable. The case previously identified as not meeting the new requirement is one of these (see Figures 4 and 5). It was seen that no tees are used as main me er s and ncne are affected by the new requirement.

Conclusion From the results of the investigation of this topic, it may be concluded that the structural steel sections used at Millstone 1 neet the current code requirements for compression ntmbers in AISC-1980, Section 1.9.1.2 and Appendix C.

2. COPED BEAM CONNECTIONS Question Shear Load Capacity of Coped Beam Connections (AISC 1980 vs AISC 1965): The licensee evaluated the significance of this code change by a sampling approach. Based on the calculations performed (R ef . 2), the licensee concluded that all of the bolted or riveted beam connections in the plan t where the top flange (or both top and bottom flanges) of steel bea s was coped are adequate for the new code provisions. However, the details of evaluation and basis for the conclusion are not clear to the staff.

In order to close this open item, the licensee should submit its sampling procedure, sampling basis, sample size, evaluation results and basis for conclusion for review. In a ddi ti on , the licensee should calculate the nost severe shear loads imposed on the beam in the actaal service condition and evalua te the adequacy of the connections.

Response

This issue concerns a new requirement in the 1980 AISC Code limiting the shear in bolted coped beam connections. This requirement applies to connections in which bolts pass through the web of the coped beam, In the previous submittal, a sample of coped bear connections was checked to see if the new requirement, allowable block shear, was ret.

The evaluation assumed standard AISC connection details and standard design loads.

In order to provide the requested additional information, a complete review of the structural steel drawings, supplemented by a walkdown, was carried out to identify the size, location, and function of all coped beams in the

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1 plant. In particular, the walkdown identified the connection details and the actual service conditions. It was found that all of the coped beams in the. plant were included in the

, original sample.

Table 2 provides a summary of each coped beam connection in the plant. Connection details are shown in Figures 1 to 9.

All of the beams are primary steel in floor and roof fraring systems. Most of the connactions were found to be wel'ded, making the new requirement not applicable. Where the requirement does apply, the adequacy of the connection was checked using the most severe shear loads imposed on the beam in the actual service condition. All of the beams meet the new requirement by a substantial margin.

Conclusion From tne results of the investigation of this topic, it may be concluded that the shear load capacity of coped beam connectione used at Millstone 1 meet the current code requirements of AISC-1980, Section 1.5.1.2.2.

3. COLUMN SPLICES Question Column with Spliced Reinforcements Subject to Stress Reversal (ACI 349-76 vs ACI 318-63): the licensee presented a discussion of the ~ "Fa tigue Environment" of concre te columns in Scismic Category I buildings, expressing the opinion that fatigue usage could be found small and concluded that column reinforcement splices designed to ACI 318-63 code will be adequate for alternating stresses. This is not acceptable to the staff. The licensee should evaluate the adequacy of the column splices for the most severe load reversal either by using the existing test data or by calculation.

Response

This issue concerns new requirements in the ACI 349-76 Code pertaining to reinforcement splice details. ACI 349-76 requires that splices in each face of a column, where the design load stress in the longitudinal bars varies from fy in compression to 1/2 fy in tension, be developed to provide at least twice the calculated tension in that face or at least 1/4 of the yield capacity.

In order to evaluate the adequacy of the column splices, a complete review of the plant concrete drawings, supplerented by a walkdown, was carried out to identify the size, reinforcement, location, and function of concrete columns in the plant. The only heavy columns in the plant are in the reactor building secondary containment structure.

Table 3A provides a summary of the data for the reactor building columns. Table 3B presents results of a stress evaluation of three of the column sizes which are i

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representative of the full population. The loads used are dead load and safe shutdown earthquake. The safe shutdown earthquake loads are~ from [7). It may be seen that tensile

. stresses in the columns do not exceed the modulus of rupture of the concrete and tensile stresses in the steel are very low.

It may be see.1 that the splices provided during construction develop sufficient tensile capacity to provide both (a) twice the calculated tension in any column face and (b) at least 1/4 of the yield capacity. Thus, the column splices are adequate with respect to the requirements of ACI 349-76.

Conclusion From the results of the investigation of this topic, it may be concluded that the c'olumn splices used at Millstone 1 meet the current code requirements for columns with spliced reinforcerent subject to stress reversal in ACI 349-76, Section 7.10.3.

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4. CONTAINNENT PENETRATIONS Ouestion

, Design of Containment Penetration Bellows: The licensee submitted Tube Turns Report No. TT-1-66 (Ref. 5), for the qualification of containment penetration bellows assemblies X-7A through 70. As a result of review, two concerns were raised by the staff:

1. The procedure used in evaluating the bellows assemblies acceptable per NE-3365.2(e)(1) of the 1983 ASME Code, provided that the testing require-ment made there to ensure validity of the calculations performed has been met. The statement made on page 6 of the report, "Verification of these factors is found in Tube Turns Report No. S.122-6", is not sufficient to ensure compliance. The Ifcensee should either provide a summary of the report sufficiently detailed to determine that the requirements as to the number, type, and outcomes of the tests are met, or provide the report itself for review.

2. The report does not include evaluation of the bellows assemblies for the seismic Inads. Such evaluation should consider two factors: (1) the effect of the relative displacements of the bellows assembly ends resulting from the seismic response of the attached components (2) and the effect of simultaneous direct seismic responses of the bellows.

! Response

1. A copy of Tube Turns Report No. 5.122-6A has been obtained and is enclosed as requested.

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2. An evaluation of the bellows for seismic loads has been performed. Seismic movements for the primary containment were obtained from [7). Thermal and seismic movements of the main steam piping were obtained from the latest piping analysis. Direct seismic response of the bellows was based on the safe shutdown earthquake spectra of (7).

Table 4 presents the results of the evaluation. All loads have been converted to equivalent axial compression in inches per convolution in order to compare with the allowable of .35 inches per convolution in (5). It may be seen that the LOCA condition uses about 35% of the allowable, thermal piping movements use about 10%, and seismic loads use about 8%. Seismic piping movements are insignificant since there are major piping restraints located at the containment penetrations. Combining all of these loads directly shows that there is substantial margin against failure of the bellows.

Conclusion From the results of the investigation of this topic, it may be concluded that the design of the containment penetration bellows at Millstone 1 meets the current code rules for bellows design in the AS!!E B&PV,Section III, Subsection lie-3365.2 i.

5. EXTREME ENVIRONMENTAL SNOW LOADS Ques.t. ion The effects of extreme environmental loads due to snow on building roofs should be addressed; the snow load was defined as 115 psf in the staff's March'30, 1981, Safety Evalua tion on SEP Topic II-2. A.

Response

An evaluation of the roof of each safety related building at Millstone 1 has been performed. The evaluation consisted of (a) calculating the capacity of each roof, (b) subtracting the actual dead load, and (c) comparing the residual capacity with the required roof snow load.

Table 5 shows the results of the evaluation. The capacity was computed as either (a) yield for steel reinforcing in concrete sections or (b! yield of the extrene fiber in steel sections. This is conservative since the ultimate strength of Grade 40 reinforcing steel and A36 structural steel is substantially greater than the minimu yield used in design.

The required roof load was based on ANSI A58.1-1982 using a ground snow load of 115 pounds'per square foot. The capacity of each roof exceeds the extreme roof snow load.

Conclusion From the investigation of this topic, it may be concluded that the capacities of the roof systems of the critical structures at !!illstone 1 are adequate f or the extreme environmental snow load as defined in the HRC Staff's March 30, 1981 safety evaluation on SEP Topic II-2.A.

6. CONBINATION OF EARTHQUAKE AND PIPE BREAK Question The combination of safe shutdown earthquake and pipe break loads (e.g. LOCA) has not been examined except for the drywell. The staff requires that this combination should be examined on a sam,pling basis to confirm the margin of safety in the affected critical structures of the plant.

Resoonse This issue concerns requirements of the current Standard Review Plan which are not included in the design basis of Millstone 1. In order to provide the required information, a complete review of the buildings affected by .nipe break outside containment (P80C) has been performed. The following procedures was used, m The original PBOC report (8,9) was reviewed to determine the worst case systems and the location of postulated breaks.

m Arbitrary intermediate breaks and longitudinal breaks at terminal ends were eliminated, consistent with NRC Generic Letter 87-11.

m The buildings containing pipe break locations were reviewed to identify the critical structural components for seismic capability.

m Piping layout drawings were reviewed, supplemented by a walkdown, to determine whether the pipe breaks will affect critical structural components.

m SSE seismic loads on affected critical structural components were computed based on (7).

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E~  : Pressure, jet, and-pipe whip loads were a e calculated based'on-(9].

u Strosses in the affected' structural-components were conservatively calculated and: combined-absolutely.>

The' worst case systems identified in [9] are the condensate, feedwater, reactor cleanup,.and isolation condenser systems (it was-stated therein that the effects of the main steam system breaks were less severe than those_of the feedwater system breaks). The buildings containing these systems are 4

. the reactor'and~ turbine buildings.

• The seismically' critical structural-elements in these buildings are the condenser. shiele' aalls (surrounding the heater bays) in the turbine building, and the exterior walls-in the reactor building. The primary contfinment shield wall and the pipe tunnel walls have much higher capacity than the reactor building exterior walls, and are therefore not

- explicitly considered. The interior colurns in the reactor

- building do not contribute to the seismic resistance of the building; however, column failure could cause collapse of a floor section. Table-6A summarizes the break locations and affected structural components.

The affected critical structural components which needed evaluation are:

! a The east condenser shield wall due to a break i in the feedwater system, a The west wall of the r.eactor building due to a break in the reactor water cleanup system.

o e The reactor building column on the N and 8 I

lines due to a break in the isolation condenser j system.

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a The reactor building east wall due to a break in the~ isolation condenser system.

Tables 6B and 6C provide a summary of the evaluations of the

' structural components. Each one has adequate capacity to resist the combination of safe shutdown earthquake and pipe

-break loads. Since these are the bounding cases for the plant and the calculations are conservative, it may be concluded that there is an adequate margin of safety in the critical structures of the plant.

Conclusion From the results of the investigation ot this topic, it may be concluded that the margin of safety of the critical structures at Millstone 1 is sufficient to withstand the simultaneous occurrence of pipe rupture and safe shutdown earthquake per the NRC Standard Review Plan, NUREG-0800.

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CONCLUSIONS-Northeast Nuclear Energy Company (NNECo)_has performed an investigation of each open issue relating to SEP Topic III-

-7.B for the Hillstone Unit 1 Nuclear Power Station [6].

These issues are:

1. Current code requirements for compression members (3.1.2).

2. Shear load capacity of coped beam connections (3.1.3).

3. Columns with spliced reinforcement subject to stress reversal (3.1.4),

4. Design of containment penetration bellows (3.1.5).

5. Effects of extreme environmental loads due to snow on building roofs (3.2.1).

6. Combination of safe shutdown earthquake loads and pipe break loads (3.2.2).

For each topic the data was reviewed and evaluated using conservative, bounding analyses. From the results of the evaluations it may be concluded that the structural design of Millstone Unit 1 provides sufficient capacity to satisfactorily resolve all open issues relating to SEP Tcpic III-7.B.

REFERENCES

1. Letter from'W.G. Council (NNECo) to D.G. Eisenhut (NRC), dated December 28, 1983.

2. Letter from W.G. Council to D.M. Crutchfield, dated January 11, 1984.

3. Letter from W.G. Council and R.W. Bishop to D.M.

Crutchfield dated February 2, 1984.

4. Letter fr0m U.G. Council to D.M. Crutchfield, dated August 11, 1982.

5. Tube Turns Report TT-1-66, "The Primary Containment Penetration Piping Expansion Joint Assemblies X-7A thru X-7D for Millstone Nuclear Station."

(References 1 to 5 are duplicated from Reference 6.)

6. Letter fror C.O. Thomas (NRC) to E.J. Mroczka (NNECo),

dated March 10, 1987.

7. Impell Report 02-0240-1094, "Generation of In-structure Response Spectra, Millstone Unit 1, Systematic Evaluation Plan," June 1982.

8. Letter from U.G. Council to D.M. Crutchfield, dated December 4, 1981.

9. NNEco Special Report "Effects of a High Energy Pipe Break Outside of Primary Containment." August 1973.

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. .V' TABLE 1 TEE BEAMS

.Se_ction ' Drawing Evaluation ST7UF47.5 51005, N-U This is a vertical hanger supporting a tank. As such, it.is always in tension. Thus the code provision is not applicable.

ST4B6.5 51014, B-B This is.a stiffener for a checkered plate grating.

The stem of the tee is welded to the plate. Thus it is supported, and the code provision does not apply.

ST3B6 51015, P-P Same as ST4B6.5, above.

STSUF10.5 51021, G-G This is a knee brace to distribute some of the diaphragm load in the roof steel to the bottom flange of the main girders. The d/t value is 20.6, so no reduction in allowable stress is required (the 1980 code requires a reduction in allowable stress if the d/t of a tee is greater than 21.16) .

ST6B7* 51029, !!-11 These are stiffeners for hatch covers. The stens ST6MF13.5* of the tees are welded to the botto. of the plates forming the hatch covers (see Figures 4 and 5).

Thus the stems are supported, and the code provisions do not apply.

ST4WFS.5 51030, D-D These are used in a bracing frame between colurns in the turbine building elevator machinery room.

The d/t value is 17. 4, so no reduction in allowable stress is required.

• These sections were identified in the original submittal as requiring a reduction in allowable stress if the sters are unsupported, llote: 1. Drawing prefix 25202 applies to all drawings

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TABLE 2

' COPED BEAM CollNECTIOllS

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Applied End Allowable Connection Load Shear Shear Drcwing Section __ Detail _ (1b/.f t) (kip) (kip) 51005,P-P 14B17.2 2-3/4"# bolts 660 7.7 20.8 OK 14WF30 2-3/4"e bolts 660 7.7 20.8 OK 51018,Y-Y 21UF62 Uelded Not Applicable OK 51021 12B16.5 2-3/4"e bolts 660 8.3. 22.8 OK 12WF27 2-3/4", bolts 660 8.3 23.8 OK 51027,AB-AB 24WF76 Welded Not Applicable OK AH-AH 16HF40 "

21HF55 "

AK-AK 21HF55 "

AR-AR 24UF68 "

AS-AS 14WF30 AT-AT 14WF30 AU-AU 30HF108 51034,DET B 14WF61 Welded Hot Applicable OK DET C 14WF61 "

DET E 14UF30 "

DET F 27UF145 " "

DET G 14WF61 Hotes: 1. Connections determined from visual inspection.

2. Applied load is from actual service condition.

3. Allowable shear is from AISC-1980, Section 1.5.1.2.2.

4. Drawing prefix 25202 applies to drawing nu=bers.

5. Connections are shown in the following figures:

25202-51005 Figure 1 25202-51018,51027 Figures 2-7 25202-51021 Figure 8

, 25202-51034 Figure 9 l

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1 TABLE 3A COLUM11 SPLICE DATA From To Column Colu=n~ Splice ACI349 Percent Eloy .Eley- -Lines. Size. Detail Rebar Length Length Developed 82.75 108.5 M7 - 8 , l'7 - 11 36x36 J. 12#10 30 37 81%

65.75 82.75 L7 36x36 K 20#10 30 37 81%

11 8 36x36 K 20#11 33 47 70%

M7-8,!!9-11 36x36 J 12#10 30 37 81%

42.50 65.75 K10,1110 58x50 P 40#11 33 47 70%

!!7 - 9 42x42 H 24#11 33 47 70%

1110-11 36x36 J 12#10 30 37 81%

14.50 42.50 L7,L11 48x36 A 14#10 30 37 81%

!!7 ,111 1 54x54 B 28#11 33 47 70%

'!!o t e s : 1. The reinforcement is spliced a minimum of 24 bar diameters per Drawing 25202-11087.

2. All splices satisfy the requirerent to develop 25%.of the tensile capacity of the bars.

n TABLE 3B COLUMil SPLICE TEllSIOli Dead Vert Horiz Concrete Steel Allowable From To Column _ Load SSE SSE Tension Tension Tension E13v Elev _ Size (kip) .( kip) .( kip) .( psi)_

_( ps.1)_ _ (psi)_.

82.75 108.5 36x36 -208 27 261 80 720 14600 65.75 82.75 36x36 -376 49 1523 196 1764 13300 42.50 65.75 42x42 -410 53 2368 11 99 13300 36x36 -558 73 172 -- --

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-14.50 42,50 54x54 -342 44 434 136 1224 13300 flotos: 1. Since the modulus of rupture for 3000 psi concrete is 410 psi, the concrete remains.uneracked and the steel stress is equal tc the concrete stress times Es/Ec.

2. The allowable tensile stress is 50% of the tensile capacity developed by the column splices.

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TABLE 4 CC.'ITAI11 melit PE!!ETRATIOli BELLOWS m Axial Lateral Equivalent.' Percent Movement Movement ~ Compression of Load _pase _ (in) _ _(in)i (per conv) Allowable LOCA: Drywell .970 .300 .1197 34.2 Thermal: Piping .440 --

.0367 10.5 Seismic: Drywell .131 .116 Piping --- ---

Self-wt .002 .002

.133 .118 .0263 7.5 Total 1.543 .418 .1827 52.2 floto: Per (5), movements are converted to an equivalent compression per convolution by dividing axial movements by 12 and lateral ecvements by 7.75. The allowable total equivalent compression is 0.35.

TABLE 5 EXTREME EllVIRO11MEliTAL S!!OW LOADS

' Roof Critical Roof Roof Snow Bui.1 ding Elevation Element. Capacity Weigh.t Capacity Reactor 151 Steel Girders 154 21 133 OK-54 Concrete Slab 1155 150 1005 OK Turbine .105 Steel Beams 195 10 185 OK 77 Steel Joists 115 10 105 OK 55 Composite Beams 320 55 265 OK Control 55 Concrete Slab 1465 300 1165 OK Redwaste 14 Concrete Slab 1015 300 715 OK llotes: 1. All capacities are in pounds per. square foot"of roof area.

2. The roof snow load is 97 psf for a ground load of 115 psf.

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TABLE 6A PIPE BREAK EFFECTS ON STRUCTURES System. "

Break Effects C@ndensate Condensate A guillotine break will cause a jet from the pipe Booster Pumps to hit the outside wall on the 14 line. Also, the discharge pipe will whip against the 14 line wall.

Since the distance is very small, danage is local.

The crane columns and steel framing holding up the concrete floors will not be affected.

A jet load could hit the condenser shield wall on the 13 line. The diameter of the pipe is 10" at the pump connection and the distance to the wall is about 16 feet. In this area the shield wall has a vertical span of about 20 feet, is 3'-0" thick, and is reinforced with #8@l2". There are masonry blockouts. The concrete is 4'-0" wide and the masonry is 6'-6", alternating. There is also a parallel 3'-0" thick wall running along the 12 line, so local failure on the 13 line is not critical. If a jet hits the masonry, the 13 line wall will not be affected. If it hits the concrete, only one panel will be lost. The wall will stay up because the concrete has longitudinal reinforcing steel which can bridge over the failed section. Therefore, this break does not affect a critical structural component.

L.P. Heater Jets from a guillotina break will go up or down in Nozzles a plane parallel to the shield wall and will not hit it. Likewise, pipe whip will not inpact the shield wall. Therefore, this break does not affect a critical structural component.

L.I.P and Guillotine breaks at M2-2a or M2-2b inlets will I.P. Heater cause jets at 45 degrees, either up or down. The Nozzles downward jets will hit the floor. The upward jets will be partly blocked by the heaters and will hit the shield wall near the bottom and at an angle, so no damage will occur. Pipe impact will be on the floor at 11'-6". Therefore, this break does not affect a critical structural compon nt.

Jets from breaks at M2-3a or M2-3b outlets will be either straight up or down, missing the shield wall. Pipe whip will be upward,' hitting the operating floor. Therefore, this break does not affect a critical structural corponent.

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Reactor Feed A guillotine break at the pump suction nozzle will.

Pump Inlets result in vertical jets, missing the shield wall.

Pipe whip will be vertical also. Therefore, this break does not affect a critical structural component.

FGedwnter Reactor Feed A guillotine break here is the same as for the Pump Outlets inlet side. Therefore, this break does not affect a critical structural component.

H.I.P. and The ordentation of the breaks here are the same as H.P. Eeater for the L.I.P. and I.P. nozzles. Therefore, this Nozzles break does not affect a critical structural component.

Containment This is in the pipe tunnel which is not a Penetration seistically critical structural element.

Reactor Water Penetration Jets or pipe whips here cannot hit the secondary Cleon-up X-14 containment wall because of the shield wall on the K line. Therefore, this break does not affect a critical structural component.

Elbow breaks A guillotine break at stress points 55 or 571 on 8"-CUU-1 would cause a jet to hit the J line wall near the 8 line at Elevation 59.00. A guillotine break at 571 could also cause a pipe whip impacting the J line wall. This break affects a seismically critical structural element.

The line size is 8", pressure is approxinately

^ rsi, the jet length is 8 feet, and the pipe

'istance is 4 feet. The wall is supported at ion 42.50 and is 21 feet high (to the bottor c..; next floor slab).

Clean-up Aux. A guillotine break will cause vertical jets and Pump Mozzles pipe whip, which will miss the secondary containment wall. Therefore, this break does not affect a critical structural component.

Regenerative Guillotine breaks will cause vertical jets and Heat Exch. pipe whips. Therefore, this break does not affect Mozz.les a critical. structural component.

Non-legen. Guillotine breaks will cause vertical jets and Heat Exch. pipe whips. Therefore, this break does not affect Nozz1ts a critical structural component.

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Anchor at A guillotine break would cause jet loads and pipe  ;

Elevation impact on this wall or the K'line wall. These are 57.17 shield walls and not part of the teactor building o primary structure. Therefore, this break does not affect a critical structural component.

Penetration This is where the relief line dumps into the X-212 torus. A break here would cause a jet down onto the torus; however, a jet would not come out of the penetration because there is no pressure reservoir. The pipe whip would be upwards, striking the underside of the Elevation 14.50 slab rather than the secondary containment wall.

Therefore, this break does not affect a critical structural component.

Isolation Isolation Protective devices have been installed on the Condenser Condenser inlet and outlet piping to mitigate the effects of Nozzles pipe rupture. Therefore, this break does not affect a critical structural component.

Penetration A guillotine break here would result in a jet X-10A coming out of the penetration which would impact Column N-8, 21 feet from the break, or the secondary containment, 50 feet from the break, at a 30 degree angle. Failure of the column could cause collapse of the floor above, and failure of the wall could cause loss of secondary containment. Therefore, this break affects critical structural elements.

The jet from the pipe i= pacts the drywell shield wall, which is not evaluated. The pipe pivots about the penetration through the Elevation 82.75 slab, and will be forced down by the weight of the motor operated valve. The pipe impact will be on the Elevation 65.75 slab rather than one of the columns.

Penetration The effects of this break are similar te the X-10A X-11B break, except that the penetration is not in line with a column and the pipe size is 10" rather than 14". Therefore, X-10A is the controlling break.

~

TABLE 6B SEIStiIC AllD PIPE BREAK LOADS

• - Seismic Pressure Jet Pipe-Force Force Force I= pact.

Brc.ck. System Structure .( g ). (psi). (kip) (ft-kip)

1. Feedwater Condenser Shield Wall .47 1.3  !!A 17 A -.

2. RWCU Reactor Bldg. J Line .47 1.5 68 271

3. Iso. Cond. Reactor Bldg. Col. !!-8 .50  !!A 30  !!A

4. Iso. Cond. ' Reactor Bldg. P Line .50 1.2 268  !!A

!!ote: -All leads are normal to the wall. Vertical loads are very small.

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TABLE 6C COMBINED SEISMIC AND PIPE BREAK LOADS BENDING MOMENTS IN CRITICAL STRUCTURAL ELEMENTS Elastic Pipe Whip Seismic Pressure Jet Total Capacity Ductility Breck fin-kip) .in-kip)

( .in-kip).

( Jin-kip). Jin-kip)_ Used

1. 8233 7267 --

15500 16630 --

OK

2. 2518 2572 8543 13633 16630 8.3 OK

3. 185 --

2430 2615 10432 --

OK

4. 239 441 5103 5783 6622 --

OK Notes: 1. Loads are considered to act on a vertical strip whose width is six times the wall thickness. This is conservati"e since there is horizontal as well as vertical reinforcement. The P line wall uses the full panel width since the jet load is spread over a large area.

2. A dynamic load factor of 2.0 is used with the jet force.

3. The elastic moment capacity is computed as .9Mu. No credit is taken for dynamic intensification factors.

3. The pipe whip response is computed (a) neglecting resistance of piping and restraints, (b) assuming a fully plastic impact (all momentur transferred to the wall), (c) assuming all other loads (seismic, pressure, and jet) are static and at the their maximum value at the time of the impact, and (d) neglecting the remaining elastic energy absorption. Thus, the ductility used is extre=ely conservative.

4. The J line wall (break 2) has a large ductility because there is both tension and compression steel; however, ACI 349-76 limits the allowable ductility of walls in bending to 10.

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