ML20236H972
| ML20236H972 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/31/1987 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| TAC-65726, NUDOCS 8708060005 | |
| Download: ML20236H972 (18) | |
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gg Pbetland General Electric Coiiipisiy Devid W. Cockfield Vice President, Nuclear July 31, 1987 j
I Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555
Dear Sir:
Pipe Support Design Verification In letters dated July 10, July 15, and July 27, 1987, PGE described the efforts undertaken to verify the design adequacy of piping supports at the Trojan Nuclear Plant. An Action Plan was developed and implemented for this purpose. Although not all activitics identified on the Action Plan are complete, PGE has concluded that it is safe to comence heatup and a retdrn to power operation of the Trojan Nuclear Plant for the following reasons:
1.
The design deficiency has been identified and corrected.
2.
The capacity depth of the rock bolts has been shown to be adequate.
3.
No-piping supports other than the ten main steamline supports and 16 safety-related pipe anchors were inadequately designed.
Our letter of July 27, 1987 responded to Nuclear Regulatory Commission (NRC) concerns identified during the NRC review of the support verification program in the offices of the Architect-Engineer (A-E) for the Trojan I
Nuclear Plant. Based on subsequent telephone conversations, the following additional information is provided to clarify the July 27 response on selected concerns as well as to provide Portland General Electric Company's (PCE's) assessment of the A-E's quality assuranco program.
Attachment A contains additional information pertaining to the adequacy of rock bolto.
A description of the method of weld verification is provided as Attachment B.
The assessment of the A-E's quality assurance program as it has been applied to the pipe support design verification effort is provided as Attachment C.
g OO B708060005 870731 r'
PDR ADOCK 05000344
- /
P PDR 121 S W Safrron Street, Portland, Oregon 97204
-. __.--_ - _ ~
Pordand General Ekx:fricCoiip:niy U.S. Nuclear Regulatory Commission July 31, 1987 Page 2 In PGE's letter on pipe support design verification dated July 10, 1987, PGE committed to a long-term program to verify designs for other safety-related supports. The scope of work and schedule are to be submitted by September 1, 1987. Attachment D contains a preliminary l
description of the scope and schedule for the long-term program.
In this letter and previous correspondence on this issun, we have referred to supports being " verified" as being " adequate, acceptable, or satis-factory". For the purposes of clarification, a support which is " verified" to be " adequate,' acceptable, or satisfactory" means the support meets the code allowables of the Trojan Final Safety Analysis Report (FSAR) under the required loads or load combinations. A proup of recently identified pipe supports is still being verified in this manner. Additionally, the actual l
loads for the 138 civil-designed safety-related pipe snubber and restraint anchorages (265 origin 611y) have not all been compared to the loads used for the support verification. These two ongoing activities will be completed before commencing with Plant heatup.
In conclusion, PCE is satisfied with the performance and results of the pipe support design verification program. The program has demonstrated that support deficiencies were isolated to those supports designed by the A-E civil group and that broad support deficiencies do not exist. The program results have provided assurance that Trojan is safe to operate.
Sincerely, Attachment c:
Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission Mr. David Kish, Director State f Oregon Department of Energy Mr. R. C. Barr NRC Resident Inspector Trojan Nuclear Plant bacribed and sworn to before me this 31st day of July 1987.
v-lc 1
u Notary Public of Oregon My Commission Expires:
August 9, 1987
l Trojan Nuc1 car Plant Document Control Desk Docket 50-344 July 31, 1987 License NPF-1 Attachment A Page 1 of 12 CAPACITY OF ROCK BOLTS The capacity of grouted rock bolts can be developed from the combination of three effects.
These effects are (1) setting the cone and shell, (2) grout effects in the immediate vicinity of the expanded shell, and (3) bond between the grout and the bolt. These effects are discussed below.
Setting Cone and Shell The first effect results from setting the cone and shell assembly at the embedmont depth. This is accomplished as follows:
insert bolt in drilled hole, and torque to set the malleable shell which is expanded into the side of the drilled hole by wedging action of the cone. At this state of the resistance development, the rock bolt would function in a manner similar to a simple expansion anchor. A technique suggested by ACI 349 to calculate the pullout capacity of simple expansion anchors is to assume a 45' shear cone emanating from the far end of the expanded cone [P = 44 (f'c)1/2 A gg).
e Grout Effects in Immediate Vicinity of the Expanded Shell The second effect results from grouting the bolt and considering the behavior of the expanded shell acting together with grout in the immedi-ate vicinity of the expanded shell. The grout fills the voids around the expanded shell and the smaller threaded cone which is forced into the far end of the expanded shell during the setting process. With the voids filled with grout, the load transmitted from the shell and cone assembly now has the additional benefit of wedging action against the grout which in turn transmits the load into the concrete.
In this stage of the resistance development, the behavior is similar to a maxi-bolt.
The capacity of maxi-bolts obtained from a series of tests are shown in Tables 1 and 2.
The results in Table 1 show three anchors failed in the steel and two failed in the concrete. The average of the five failure loads would be a conservative estimate of failure controlled by concrete since only two of the anchors actually failed the concrete.
The embedmont for theco five is 7.5 inches, which is essentially the same as the embedmont of the orte-inch rock bolts used for pipe support anchorage.
These failure loads can be used to estimate the factors used in the application of the cone pullout formula x4 (f ')1/2 Aegg. Using e
these failure loads, the cone coefficient, x, can be calculated.
This value is tabulated in Table 1 for 4 = 0.85, f ' = 3500 psi and the e
embedment of 7.5 inches. The first three values tabulated are all below that for concrete failure since the steel portion failed.
o
. Trojan Nuclear Plant Document Control Desk Docket 50-344 July 31, 1967 License NPF-1 Attachment A Page 2 of 12
)
Considering the usual scatter (i 10 percent around the mean) for data concerning the same concrete' mix, an average cone coefficient'of approxi-
)
mately'3.6 could be expected.
Similar values are. tabulated in Table 2 for
]
4 = 0.85, f ' = 4000_ psi, and an embedmont of 7.5 inches. For the four.
./
e specimens, there were no concrete failures. Again, the average value for the cone coeffic.ient representing a_conceete failure would obviously be greater than the' average of these values but no real indication is available from the data. Considering normal data scatter in concrete, an average conc coefficient from these data representing concrete failure of approximately 3.5 could be expected.
Bond Between Grout and Bolt The third effect is the bond between the reinforcing steel used to make'the rock bolt and the surrounding grout. As the load is applied to the bolt, some of the load will transfer directly from the bolt through the grout to
{
the surrounding concrete.
Pullout' test data available on reinforcing steel attached to concrete by' grouting to fill a hole in the concrete are shown in Table 3.
This table reports load'at failure, the mechanism of failure and the cono coefficient.
These data show an average cone coefficient of approximately 5.5.
Capacity of GrouteG Rotk Bolts In reviewing the data, tce large difference between the values of the cope coefficients for the maxi-5olts and the grouted reinforcing steel can be explained by the load transt?r mechanisms. For the maxi-bolt, all'cf the load-is transferred at the far end of the enchor, resulting in considerable load concentration. 'For the grouted in reinforcing steel, the load transfer is distributed'over the length of the bar and not concentrated. Rationally, this implies the smoother the load transfer, the higher the cono coefficient.
The load transfer for the grouted rock bolt is smoother than for the maxi-bolt since sone of the load is transferred by bond along the length of the bolt.
In addition, the grout completely fills the voids around the shell and cone, thus providing a smoother load trancition in the vicinity of the
~
shell. This type of behavior implies the cone coefficient is greater than 3.5 rnd 3.6 obtained from the two sets of maxi-bolt data. The load transfer for the grouted rock bolt in likewise not as smooth as for the grouted reinforcing steel, implying the cone coefficient is less than 5.S.
Based oc this information, taking a cone coefficient of 4 with a 4 = 0.85 seems to be reasonable with some margin.
Capacity for One-Inch Grouted Rock Bolts For the one-inch grouted roct boat used in the pipe supports inside Containment, the above approach estimates the tension capacity to be:
Teap
- 4+Acff(fc') / = 4w(0.85)(7.375)2 (6500)1/2 = 46.8 kips l
I l
l Trojan Nucicar Plant Document Control Desk i
l Docket 50-344 July 31, 1987 l
License NPF-1 Attachment A j
Page 3 of 12 I
l This value is for an embedmont of 7-3/8 inch at.d a concrete strength of I
6500 psi. The embedment of 7-3/8 inch is the acto.a1 observed enbedment from several anchors which PGE has removed during support modifications by core drilling.
Concrete with an initial compressive strength ot 5000 psi is estimated to have a strength of 6500 psi af ter approximately 15 yeasy.
]
The allowabin lord used in the pipe support evaluation is 21 kips for normal 1 cad or werking stresn level with 26.25 kips and 31.4 kips as the allowable loads for the OBE and 'ISE, respectively. The corresponding ration of Lap; city to allowable loads are 2.23, 1.78, and 1.49, respectively. The corresponding allowable shear loads for the pipe suppcrt evaluation are il kips, 13.7S kips, and 16.5 kips respectively.
Comparison of Rock Bolt Load Demand __to A110wable Loads Table 4 provides a summary of bounding case (maxinrum) interaction ratios for rock bolts in safety-related pipe snubber supports, seismic restraints and anchors. The interaction ratio is fy35/3
/ T)S/3 (Tal
- (fi~
in which V and T are the rock bolt chear and tension demands, respec-tively, and Y and T are the rock bolt shear and tension allowables, a
a respectit/aly, appropriate for the governing load combination for the support rock bolts.
The limiting criteria velue for the interaction ratio is 1.0 (ie, the value at which the combination of shear and teasion demands equals the combination of shear and tension allowables). As can be seen in the tables, the interactions trend toward ration much less than 1.0, which shows that the pipe support rock boJts have large inher-ent capacity margins in addition to those prtvided by the selection of the allowable loads.
Compaelson of Rock Bolt Load Demand to Capacity The ratio of capacity to actual demand can be obtained by combining the ration of capacity to allowable loads with the interaction ratios given in Table 4.
For example, for generic support No. II from the 265 group (from C-386, P,heet 2, Type II), the ratio of capacity to demand for the SSE load combination is 1.49/0.38 = 3.9.
Note the vast majority of the interaction ratios ere less than 0.5.
SAB/mr 1918W
Trojan Nucicar Plant Document Control Desk Docket 50-344 July-31, 1987 License NPF-1 Attachment A Pago 4 of 12 TABLE 1 MAXI-BOLT CA?ACITY TEST RESULTS FROM PITTSBURGH TE.4 TING LABORATORY Cone i
Coefficient Sample Lcad et Failure (Kips)
Failure Mechanism (x*)
1 30.05 Steel Anchor 3.38 2
29.86 Steel Anchor 3 36 3
28.88 Stoel Anchor 3.25 4
29.39 Concrete 3.01 5
30.88 Concrete 3.47 Iension Tests: Pittsburgh Testing Laboratory, June 11, 1980.
. Test samples:
Drillco maxi-bolt wedge anchor, 5/8-ir.ch-diameter with 7-1/2-inch cabedment.
Concrete:
Design 28-day compressive strength of f ' = 3500 psi.
e
- x = load et failure 4A rg (f ')l/2 e
e ll.
l 1
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SAB/mr 1918W
l
- s..,
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Trojan' Nuclear. Plant Document' Control Desk Docket 50-344
.Tuly 31, 1987 License NPF-1 Attachment'A' Page 5'of 12 TABLE 2 KAXI-BOLT CAFACITY TEST RESULTS FROM TENWESSEE VALLEY AUTHORITY Cone coefficient Sample Load at Failute'(Kips)
Failure Mechanism (x*)-
-1
'30,45 Steel' Anchor 3.21 2-129,33 Steel Anchor-3.09 3
.29.83 Steel Anchor 3.14 4.
30.45 Steel Anchor 3.21 Tension Tests: Tennessee Valley Authority,-CEB Report No. 80-64,
. October 24, 1980.:
Test ' san.ple s :
Drillco maxi-bolt weds,e anchor, S/8-inchWanet.er with 7-1/2-inc.h embedment.
Co,ncrete:
Design 28-day corprecnive strength of f ' = 4000 psi e
(interpreted feor_ tesi repcrt).
- x = load at failvare
$A rg (fe')llI~
e L
u
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fi l-SAB/;nr 1918W a
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Trojan Nuclear' P.lantL Document Control Desk Docket 50-344 July.31, 1987; i
'Licenwe WPF-1 Attachment A Page 6 of 12 TABLE 3 REINFORCING' STEEL GROUTED,IN CONCRETE HOLES Cone coefficient Sampl e.:
, Load at Failure * (Kips)
-(x*)
1.
27,510 5.22 2
28,000 5.32 3
28,250 5.37 4
27,500 5.22 1
'S 26,000 4.94 6
25,300 4,80 7
28,400 5.39-u 8
28,850
'5.48 j
9 28,600
'5.43-10
'25,300 4.80-11 29,750 S.65 12 32,200 6.12 e
- Steel controlled in all cases.
f ' = 3000 psi, e
embedment = 6 inches.
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"*i Trojan Nuclear plant Document Control Desk Docket 50-344 July 31, 1987 License NpF-1 Attachment B page 1 of 2 VERIFICATION OF SUPPORT WELDS During the pipe support design verification effort variouc aspects of the support designs were verified. One of these aspects was welds in structural members. These welds were verified for acceptability based on the design loads for the applicable support member. The welds were identified by referring to the drawings and procurement specifications.
The following describes how the welds were verified and how support welds are identified.
Weld Verification The welds in the supports were reviewod using the Trojan design documents listed below. The reviewer obtained the design drawing and the design calculation for the support being reviewed and considered the following three aspects of the weld:
1.
The magnitude of the load.
2.
How the load was' applied (chear, tension, moment).
3.
Weld type, size, and length.
With this information, an experienced engineer can determine whether a detailed evaluation of the weld is necessary.
This experier.ce is gained by performing detailed calculations whero the demand of the welds is significant and by performing various bounding wold evaluations. For example, an engineer who has dono detailed wold evaluations recently would know that the capacity of a fillet weld in shear is approximately 900 pounds per inch per 1/16 inch of wold size. If a member was subjected to a tension load of 500 pounds, Which is being resisted by 6 inches of 1/4-inch fillet weld, a detailed evaluation would not be required. Another example encountered in the verification program was an angle, 1/4-inch thick, attached to other members with a 1/4-inch fillet weld all around. The wold was at least as strong as the member, therefore a detailed evaluation was not performed. The 1/4-inch angle, however, would be evaluated.
If the weld adequacy could not be easily jndged, calculations were performed and documented. As was the case for all verification work, the weld verification was subjected to an independent review.
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Notes.on drawings (e.g., Drawing C-408, Turbine Bu'_1 ding
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Miscellaneous Structures, Note'2).
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Weld syrabols on [tawings -(e.g.. Drawing C-386, Conti1rjnent 2
Internals.Suppressor and Rigid Restraint Steel Details,lSheet 2).
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ileference. to ' AISC code on drawings (e.g., Drawing ' C-120. T)T cal edetails, Structural Steel, Note 1).
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~k6quirements Specification for Furnishing. Detailing l t
Fabrication, and Delivery of Structural Steel", Sect 15n 13.21).
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Docket 50-344 July 31, 1987 License NPF-1 Attachment C Page 1 of 1 Portland General Electric Company's (FGE's) Assessment of the Architect-Engineer's (A-E's)
Quality Assurance (OA) Program and Performance I
PGE considers the overall performance and the QA Program of the Trojan A-E to be adequate. There have been deficiencies identified in the A-E's j
performance in the support verification program, however, these deficien-cies have been identified and corrected, and do not change the conclu-sions of the verification program. The deficiencies do not indicate a j
breakdown in the A-E's QA Program.
1 i
PGE has been closely involved with the A-E efforts to verify the adequacy of safety-related pipe supports at the Trojan Nuclear Plant. PGE direc-j ted the A-E to develop an Action Plan to encure that the issue was j
properly defined and bounded.
PGE reviewed and approved this Act ion Plan
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and conducted a quality assurance surveillance in the A-E's offices to
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assess the initial verification activities associated with the investi-I gation and resolution of the main steam piping supports. Tha team was headed by tho Manager of the Nuclear QA Department and included the Branch Manager of Mechanical Engineering and the Branch Manager of Nuclear Regulation. This assessment was conducted early in the investi-l gation of this issue to determine if the issue was accurately defined, if l
appropriate corrective actions were being taken to resolve the issue, and i
if the. Action Plan was being properly implemented. The PGE surveillance team concluded that the A-E was properly implementing the Action Plan.
l Subsequently, the Branch Manager of Civil Engineering for PGE has been in I
the A-E's offices on two occasions overseeing the activities and efforts in completing the Action Plan.
In this capacity, he has monitored the support verification program.
The Trojan A-E performs engineering in support of the Nuclear Plant Engineering (KPE) Department of PGE.
Their engineering performance is judged to be acceptable based on PCE's review of their work.
PGE has evaluated the impact of the individual deficiencies found during j
the support verification program to determine if they are indicative of a breakdown in the QA program or A-E performance. When viewed as a whole, it has been concluded that these deficiencies are not indicative of a breakdown in the QA program and are not significant so as to preclude safe operation of the Trojan Nuclear Plant.
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Trojan Nuclear plant Document Control Desk Docket 50-344 July 31, 1987 License NpF-1 Attachment D page 1 of 1 LONG-TERM VERIFICATION PROGR_AM The objective'of the long-term support verification program is to ensure that'the safety-related supports are designed in accordance with the Final Safety Analysis Report. This program will incli
ihe following major activities:
1.
A sample of all safety-related supports will be fully verified (i.e., confirm that the correct design loads are used in the support design and that the support design meets the FSAR code allowables under the' required load and Joad combinations). This sample will be sized such that following the completion of the short-and long-term pipe support verification programs, at least 25 percent of the safety-related supports will be verified.
2.
This cample of safety-related supports will be selected to ensure a portion of all major categories of supports are verified. A list of safety-related pipe supports by system and type will be developed.
3.
If support designs are found that are not ucceptable (ie, cannot be verified), the sample of supports will be expanded.
4.
A Quality Assurance Audit of A-E design activities.
This audit will include a review of the following:
1 a.
The procedures for performance of design calculations and fc_
the interface of design work between different design groups. Also, the procedures which ensure that new or revised loads are properly incorporated into the design of existing supports, b.
The procedures for documentation of design work.
In accordance with our letter of July 10, 1987, the scope and schedule of the long-term verification program description will be finalized by September 1 1987.
It is expected that the longiterm verification program will be substantially complete by about February 1,1988.
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