ML20237H188
| ML20237H188 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 08/18/1987 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20237H192 | List: |
| References | |
| TAC-65726, TAC-65736, NUDOCS 8708240389 | |
| Download: ML20237H188 (131) | |
Text
{{#Wiki_filter:__ I 1 M M I ge Portland General Electric Coiripemiy j C David W. Cockfield Vice President. Nuclear August 18, 1987 Trojan Nuclear Plant Docket 50-344
)
i License NPF-1 l l-1 l 1 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555
Dear Sir:
l Pipe Support Desir,n Verification f i In previous correspondence dated July 10 (clarified on July 15), July 27, and l July 31, 1997, portland General Electric Company (PGE) described and provided j the status of implementation of the Pipo Support Design Verification Program. ' The short-term program included 493 cupports. A description of the chronology and scope of the Pipe Support Design Verification Program is provided in Attachment A. These supports have been verified and the short-term program is ; complete. Discrepancies identified during the program have all been corroc- j ted. PCE has concluded that it is safe for Trojan to commence heatup and-return to power. Pursuant to a telephone conference call on July 31, following Nuclear Regu- l latory Commission (NRC) Staff review of the PGE July 31, 1987 letter, PGE l initiated steps to provide greater assurance that no safety problem existed i with the piping supports at the Trojan Nuclear Plant. Further analysis and justification for the support anchorages was performed and is provided in Attachment B. i a The NRC Staff had remaining questions concerning the process used to verify the adequacy of the support wolds. The process of verifying welds is des-cribed in Attachment C. The weld verification program was evaluated during ; the Quality Assurance Audit described below. The audit findings and the ! Architect-Engincor's (A-E) responso provide further assurance that the wold verification program was adequato. Additionally, because questions were generated concerning a potential break-down of the A-E Quality Assurance Program as it applied to the Pipe Support Design Verification Program, a consultant was hired to perform an Audit / Quality-Technical-Review of the A-E's work. The consultant's report is pro-vided as Attachment D. Included in this attachment is the A-E's response to the key findings and PGE's assessment of the A-E's responso. 8708240389 87081B PDR ADOCK 05000344 P PDR \ e sw samn sneet %rtana. Oregon 97204 1 r
Portland GeneralBechicCompany Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Page 2 PGE will conduct a long-term Pipe Support Design Verification Program to look at other safety-related pipe supports and the pipe whip restraints that uti-lize rock bolts. This program is described in Attachment E. PGE will provide the final detallr of the long-term program to the NRC by September 30, 1987, as described in Attachment E. The results of the long-term program will be reported to the NRC upon completion. Sincerely, Attachment c: Mr. John B. Martin i Regional Administrator, Region V U. S. Nuclear Regulatory Commission Mr. David Kish, Director State of Oregon Department of Energy Mr. R. C. Barr NRC Resident Inspector Trojan Nuclear Plant Subscribed and sworn to before me this 38th day of August, 1987. i l- d&k + 1), ~ Notary Public 6' My Commission Expires: 4 u v>uM- /[N/ r
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment A Page 1 of 7
SUMMARY
OF PIPE SUPPORT DESIGN VERIFICATION PROGRAM CHRONOLOGY AND SCOPE l' L ! On June 4, 1987, a Nonconformance Report (NCR 87-214) was written due to observed separation between the wall and grout, and between the grout and baseplate for main steam support EBB-1-1-SS-81 (SS-81) on the 'B' main steam line. The separation was observed during an inservice visual in-I spection of the snubber. A similar condition was observed on support SS-86 on the 'A' main steam line. l The Civil Engineering Branch of Portland General Electric Company's (PGE)
- Nuclear Plant Engineering Department evaluated the nonconformances.
l During the evaluation it was determined the supports were not designed for the required dynamic loads of 83 kips. The load requirements exceeded the capacity of the support anchorage (rock bolts) by a factor of 2-to-3. ! PGE developed an inspection plan for the remaining 62 safety-related Main Steam System supports. Additionally, PGE directed the Architect-Enginear. (A-E) for the Trojan Nuclear Plant to determine the root cause of the design deficiency and to develop a program to determine the extent of the deficiency. t The A-E determined the design error was made by their Civil Engineering Group and consisted of a failure to incorporate the dynamic system loads in the support design. The A-E proposed a program to verify all~ support designs performed by their Civil Engineering Group. Specifically, the scope of the program included 64 safety-related Main Steam System sup-ports (34 designed by a PGE subcontractor, and 30 designed partially by the subcontractor and partially by the A-E Civil Engineering Group), 33 Civil-designed safety-related pipe anchors, 265 civil-designed safety-related pipe snubber and restraint anchorages, and five Civil-designed-nonsafety-related pipe support structures providing anchorage for 41 supports. PGE expanded this program to include verification of all safety-related supports with dynamic loads (76). The supports in this latter category included 26 which were also in the pipe snubber and restraint anchorages and structural members category (50 of the 76 supports were designed by l the PGE subcontractor and 26 partially by the subcontractor and partially ! by the A-E Civil Engineering Group). The verification was to demonstrate each support met code allowable stresses under the required loads and/or load combinations described in the Trojan Final Safety Analysis Report (FSAR). The support anchorages, base plates, members and bolts were verified in this manner. (Note: If a support is referred to as " veri-fled", " adequate", " satisfactory'", or " acceptable", it means the support satisfies the FSAR requirements and code allowables.) i i
3 J l Trojan Nuclear Plant Document Cont.rol Desk I Docket 50-344 August 18, li87 License NPF-1 f Attachment A j page 2 of 7 -l On June 18-19, 1987, a PGE management team conducted a quality assurance surveillance of the A-E's implementation of the verification program. The team was headed by the Manager of the Nucicar Quality Assurance Department and included the Branch Manager of Mechanical Engineering and the Branch Manager of Nuc1:ar Ragulation. The objectives of the sur-veillance were to verify the A-E had adequately defined the problem, had identified and was taking appropriate corrective action, and that the j documentation for the corrective action program was in place and was being complied with. The team concluded the objectives were being men. The verification program identified 10 Main Steam System supports (SS-33, SS-86, SS-80, SS-81, SR-86, SS-87, SS-88, SR-90, SS-91 and SS-92) which were not adequately designed for the specified dynamic load. The cri-ginal dynamic loads for these supports were generated based on pressure transient data from a test on a turbine generator sbailar to Trojan's. These loads were confirmed to be conservative by subsequent turbine-trip testing at Trojan. As part of the verification program, new dynamic loads for the supports were calculated using the Trojan test data. The new loads were a factor of 1.6-to-3.1 times lower than the original dynamic loads. Utilizing the new loads, 3 of the 10 supports (SS-83, SS-80 and SS-87) were determined to be acceptable without modification. The remaining seven supports were modified. All 10 of the Main Steam System supports discovered to be inadequately designed were part of a group of 14 supports designed separately in late 1975 for dynamic load concerns. Sixteen of the supports in the Civil-designed pipe anchors category were inadequate for the original load in the A-E f'iles. The design of these 16 rupports did not meet the original loads which were intended to be preliminary pending transmittal of the actual loads for these supports by the NSSS vendor. The actual loads were never transmitted. Upon receipt of these loads during the current verification effort it was demonstrated the supports were adequate for the actual loads. Four of the 33 supports in the original verification program were on piping that was disconnected during an unrelated design modification completed during the 1987 , outage. The verification of these anchors was not required. On June 22 and 23, 1987, members of the Nuclear Regulatory Commission (NRC) staff performed a review of the Pipe Support Design Verification Program at the A-E offices. The reviewers identified 10 concerns. On July 9, 1987, the NRC issued a letter to PGE pursuant to Part 50.54(f) of j Title 10 of the Code of Federal Regulations [10 CFR 50.54(f)] requesting j l l _ _ ;-------------....__.?
1
)
i Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment A Page 3 of 7 3 PGE to demonstrate the design adequacy of safety-related pipe supports at Troj an. PGE responded to the NRC letter and the NRC concerns identified during the NRC review with letters on July 10 and July 15, 1987. On July 16, 1987 PGE made a presentation to the NRC on the P.ipe Support Design Verification Program in Bethesda, Maryland. On July 21-23, 1987, the NRC staff conducted another review of the veri-fication program in the A-E's office. Nineteen additional concerns were identified. Nine of these concerns were identified as being short-term to be resolved before startup while the remaining concerns could be resolved on a longer schedule. One of the short-term issues regarded the loads used for the verification of the 265 safety-related pipe snubber and restraint anchorages. The scope of this review had originally been to verify that the Civil Engineering Group had performed the design correctly for the original specified loads. The NRC expressed concerns that the support loads be verified to ensure the most current value was being used. The actual support loads were not on file with the A-E no they had to be requested from the Nuclear Steam Supply System (NSSS) vendor. Additionally, during the NRC review, questions were raised whether all civil-designed supports had, in fact, been verified. In response to this concern, the A-E conducted an additional review of drawings and field change notices to ensure all Civil-designed pipe supports were l identified. This review yicided an additional 40 safety and non-safety-related supports designed by the Civil Engineering Group. The verifi-cation program was expanded to include these additional supports. The loads used for the design verification were the latest loads calculated for the individual supports. The modified scope of the verification program is reflected in Tables Al and A2. On July 27, 1987, PGE responded to the nine short-term NRC concerns. As a result of subsequent discussions, some of the concerns were resolved and a final list of issues was developed. These issues were rock bolt capacity, adequacy of loads received from the NSSS vendor for verifi-cation of supports in the pipe snubber and restraint anchorage category, adequacy of welds in supports, completion of the verification program for the additional 40 supports, and the adequacy of the A-E's quality assur-ance program as it was applied to the verification effort. PGE's letter of July 31, 1987 provided additional information on these issues. Following further discussions with the NRC staff, PGE prepared a more conservative analysis for rock bolt capacity, retained a consultant to perform an independent quality assurance audit and technical review of Trojan's A-E, directed the A-E to complete the verification of the remaining 40 supports and the load verification and met with the NRC reviewers to better understand their concerns,
1 l Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 -) License NPF-1 Attachment A Page 4 of 7 During this period, four supports'(SS-80, SS-91, SS-83, SS-1210) were modified to increase the design margin for the rock bolts. In addition,- PGE performed a demonstration test of the installed rock bolts and verified the 265 pipe snubber and restraint anchorages were acceptable for the actual loads received from the NSSS vendor. The verification program has now been completed for all supports. A total of 36 support or pipe anchor designs were found to be inadequate. Ten of these supports were in the Main Steam System as described above. Sixteen safety-related pipe anchors were not adequately designed for the original preliminary loads. The remaining ten supports, identified during the verification of the 40 additional supports, were modified to meet FSAR design criteria. These results are summarized in Table A1. The results of the rock bolt testing, weld verification, and independent quality assurance audit are provided in Attachment B. All required modifications to supports with design deficiencies have been completed. Corrective actions for rock bolts are described in Attachment B. All supports included in the Pipe Support Design Verifi-cation Program meet all FSAR requirements and were verified using the latest pipe stress-based loads. I l SAB/mr 1949W i
l eor i one n rco tl nv f s f ml m o c p a o . s e s ao i i oop sncsd gsue i r i eur t mtt u t u t rtf u a s siih r a t a ec l nsd rsdti s c s wad o yd n3 oe u u . e s d aa hd s;q s ry s sev e o f c ree t t r t t hi R dl so nl ogr r so r rt e e t aah a o t o o c zwry uctt p ec p pne yeoc nt nuo p va p por . l npa a ecaon u if u u r p u ea s f s s sd so nd uq t r7 n i edd aes e xru8 o l lt l nsan i es d i oo9i l l a l eaoe t R u7 a Sf F1 t A A s A Tbl v i y f l i se t t d N rae A L roun epqg 0O 6 0 0 0 0 _ 6 l P b pei 1 1 1 _ 3 mud s N USA e i O N D I ft i T oo C N A N r ) ) O yo l l I bt a) a) f T c i y iy A rd a tl tl d C eer rl rl I b nt mgn au au t F pf pf a l I AR uio N sc ( ( ( ( i n EV E eb D u 04 33 O 0 0 60 25 0 0 4 i m L S 1 l BT AR TO l e P i ) ) r P y - vg l l U S bti n cCi i a i a w rd e r t t s E eet re r r r P b niee a a o I mgh en p p
- h P uicni ( (
- N srig *
- n F eAgn 0 3 5 1 6 0 9 O D nE 3 3 6 4 2 4 0 E 2 4 4 Y
R A l d M a
- M t d *
- t U f oe 4 3 5 1 9 0
- c S osT rn 6 3 6 4 2 4 6 t/ ea / / 2 / 2 / 4 rrddl 4 3 / 1 / 0 6 eoeiP b pws 6 3 5 6
4
- 4 /
3 o mpenn 2 6 9 s uuioi 7 4 i NS vC d e _ R _ d s s ee es t t r c gr n e ei n au i d ae l t m erbd) n i p pa s ec a nomnd _ o ae i ir r R u n gh eae i Mp P Pt e - r y i cm t h . t i s b yt D sn d a - p dP d d ed m t S eal el - i e e eR ne e h D at e r t m t t aM ft t t raR c ae a ad ar i l rul 4 s l t l l nsl S o W i ot ey e es e eaeagr - p vpcRt - D R ys R R np s i pu e t y St r
- s yr
- rau yert ou NS t
r C uryf St t a L f _ r _ teapmo t o t b oc os r s eS A _ o eh ebh u ee pd ee f - T p f ep f c f ucr vp pa h pd an O _ p at u an annt ii uo ti nS o T i u SA SSAS _ SSS FP SL OP a( N P S 1 2 3 4 5 6 - i' l ~ ll
~1 i
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment A l Page 6 of 7 TABLE A2 ARCHITECT-ENGINEER DESIGNED
- PIPE SUPPORTS FROM THE ORICINAL DESIGN BY SYSTEM Number of: (Verified / Total)
System Snubbers Restraints Hanners Anchors Total A. Safety-Related , Chemical and Volume 33/33 11/11 2/2 8/8 54/54 Centrol i Safety Injection 27/27 6/6 0 4/4 37/37 Pressurizer Spray 20/20 0 0 3/3 23/23 Pressurizer Relief 26/26 0 0 0 26/26 Component Cooling 6/6 12/12 0 6/6 24/24 Water Reactor Coolant Loop 8/8 0 0 4/4** 12/12 ; Drain Reactor Coolant 46/46 13/13 4/4 0 63/63 Instrumentation Residual Heat Removal 27/27 2/2 0 7/7 36/36 Reactor Coolant Pump 28/28 10/10 0 7/7 45/45 Seal injection Steam Generator 12/12 0 0 0 12/12 Blowdown j Main P,edwater 4/4 0 0 0 4/4 (Insida Containment) 1 Main Steam 8/8 22/22 0 0 30/30 SUBTOTAL A 245/245 76/76 6/6 39/39 366/365
- Known designs by the Civil Discipline. 1
- These anchors were disconnected during the 1987 outage. Therefore, I verification was not required. !
1 l l
1 I Trojan Nuclear Plant Document Control Desk l Docket 50-344 August 18, 1987 License NPF-1 Attachment A Pago 7 of 7 TABLE A2 (Conel.) ARCHITECT-ENGINEER DESIGNED
- PIPE SUPPORTS FROM THE ORIGINAL DESIGN BY SYSTEM Number of: (Verified / Total)
System Snubbers Restraints Hangers Anchors Total B. Non-Safety-Related l Moisture Separator 6/6 18/18 12/12 0 36/36 Reheater Relief Valve Discharge Turbine Step Valves 3/3 0 0 0 3/3 Main Feedwater Pump 0 0 1/1 0 1/1 Seal Drain Tank i Main Steam to Main 0 0 1/1 0 1/1 Feedwater Pump I Main Steam Bypass 0 2/2 0 0 2/2 SUBTOTAL B 9/9 20/20 14/14 0 43/43 GRAND TOTAL (A+B) 254/254 96/96 20/20 39/39 409/409
- Known designs by the Civil Discipline. ,
l I 1
)
l l l 1949W.887 1 l
I Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 ' Attachment B Page 1 of 17 SAFETY-RELATED PIPING SUPPORT ANCHORAGES A. Program Summary and Results 1.0 Program Summary During the course of the Pipe Support Design Verification Pro-gram, the following questions arose regarding the use of rock bolts for anchorage of some of the supports to the concrete:
- a. The failure mechanism anticipated for the rock bolts (ie, grout bonding, concrete shear-cene failure, or bolt tension failure).
- b. The significance of variance of the embedment depth from the nominal 8-inch depth.
- c. The capability of the as-installed rock bolts to provide sufficient anchorage for the safety-related pipe supports with a factor of safety of at least two.
To address these issues, Portland General Electric Company (PGE) has completed a test program, measurement of accessible rock bolt. embedment depths, verification of capacity versus demand (using s factor of safety of two) and a failure load analysis. Further ' more, during the load versus demand comparisons, several very conservative assumptions were used, viz:
)
- a. The shear-cone embedment depth was taken from the top of the i expansion shell rather than the bottom. Further reductions l were taken to allow for the accuracy of bolt embedment depth measurement,
- b. The presence of reinforcing steel was ignored.
- c. A linear interaction ratio was used to measure acceptability, l d. Seismic load demands were conservatively taken as 1.67 times {
the Operating Basis Earthquake (OBE) with only a 0.5 percent 9j dampint ratio for the piping. q 1 i
- e. Typically, for a given support, the bolt with the minimum capacity was compared to the maximum demand of any rock bolt ,, !
used in the support. 7
- f. Where rock bolts could not be measured due to inaccessi-bility, a minimum rock bolt depth (using 2.26 standard devi-ations from the mean measured depth) was assumed. >
e'
* ,1 .
l l Trojan Nuclear plant Document Control Desk i Docket 50-344 August 18, 1987 License NPF-1 Attachment B page 2 of 17 2.0 Results The results of the above program were:
- a. The rock bolts are all adequate to carry the maximum support
- load with at least'a facter of safety of two. Four supports '~
were modified 1to achieve this safety factor. f' b. From the tests, there was no evidence of a bond failure
! , mechanists occurring between a rock bolt and grout or between grout and the concrate. Failure is more properly represented .
by.'the shear-cone mechanism as described in American Concrete 3 Inrtit.ute (ACI) Standard 349.
- c. A afngle test-to-destruction illustrated the/sffect of the ;
aborg conservatism because a "real" factor-of-safety of 4.9
) , was r,bserved.
B. Backgrounds t' I i Several nethods are employed by the nuclear irgdustry for anchor-4tg piping supports to concrete walls or concrete slabs. These methods include the use ef' Expansion anchors (shell or wedge type), grouted bolts or thrp ded rods, cast-in-M ' acey nolts, through bolts, and rock bolts. f l Phillips Self-Drilling Anchor Bolts, Hilti Kwik Bolt 9, and phillips Wedge Anchors are three types of expansion anchors used for supports at Trojan. Figure B1 shows these three types of anchors. An expansion anchor develops ' bolt tension by compres-sive force between the expanded bolt or wedge ring and the sides of a. predrilled hole. Expansion anchors rely on contact with the concrete to develop their renistance to pullout, u 1 Grouted bolts or threaded rods are inserted into a predrilled f hole of predetermined depth. Grout is then placed in the hole j around the bolt. Their pullout capacity is based' either on the f concrete shear-cone development or the bolt or threaded rod mate- l rial properties, dependiny, upon embedment depth. l ) l Cast-in-place bolts are inserted into place and concrete is { [ poured around them. The pullhut capacity of cast-in-place bolts j j is developed in the same manner as for grouted bolts or threaded i rods. L, I s 1 M 1 _--_--_-___-L_____. - _ _ - _ _ _ - - _ - - J-----_- ._ _ _ _ _ - _ - _ _ - _ _ - _ .
I Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18,.1987 l License NPF-1 Attachment B I Page 3 of 17 4 Through-bolting is a direct means of attaching a support when there is access to both sides of a wall or slab. A hole is i drilled through the wall or concrete slab and a bolt is anchored I to the opposite side of the wall or slab and grouted in place. The full tensile capacity of the bolt material can usually be developed in.this case. A typical rock bolt is shown in Figure B2. Rock bolts are a com-bination of reinforcing steel rods with an anchor head wedged into the concrete and grouted in. The manufacturer's instruc- ] tions for installing the rock bolts are described in i Section C.2.0. l
?
C. Analysis of Rock Bolt Capacities The original rock bolt design criteria, which were also used in the recent reevaluation of safety-related pipe support anchorages and dis- j cussed in the PGE to Nuclear Regulatory Commission (NRC) lettere of l July 27 and 31, 1987, were based on the appropriate standards avail-able at the time (ACI 318-63) and Williams Form Engineering Company (Williams) catalog and were conservative even though codes and stan-dards currently available are much more explicit. i Subsequent to discussions with NRC staff reviewers concerning thase submittals, a new, more conservative criteria for the 1-inch diameter rock bolts was developed and is described in this attachment. I l 1.0 Material Properties Affecting Rock Bolt Capacities 1.1 Rock Bolts The 1-inch diameter rock bolts are Williams hollow-core l robar rock bolts Type US-8-HC-SCS-158, which have a maxi-mum working load to elastic limit (yield strength) of 37.0 Kips and an ultimate strength of 50.0 Kips (Attachment B1). The rock bolt head assemblies are the ' Type A' short cone , and shell. .The shell measures 2.0 inches in length and 1-1/2 inches in diameter at the minimum of the shell j taper. The ' Type A' short cone and shell are appropriate , for the installations in which they were used (Attachment l B2). 1.2 Concrete Containment concrete in which the vast majority of rock l bolts for safety-related pipe supports are ancLored has a design compressive strength of 5,000 psi and is reinforced. l t i l
J o- j i Trojan Nuclear Plant Document Control Desk Docket 50-344' August 18, 1987 License NPF-1 Attachment B Page 4 of 17 Evaluation of data for Containment concrete'90-day cured-test cylinders (47 samples,' Attachment b3) shows the fol-lowing actual values: i Mean,= 6,876 psi I Standard deviation = 535 psi l-ACI 214 Criterion 3. states that for no greater than 1 percent of tests more than 500_ psi below ef ';- fe ' = 6,130 psi There are.16 safety-related pipe supports with 1-inch dia-meter rock bolts with nominal 8-inch embedment which are outsidi Containment in the Fuel Building. Concrete in which they are anchored has a design compressive strength. . of 3,000 psi and is reinforced. The 90-day cured test ' cylinder data (60 samples, Attachment B3) shows: Mean = 5,094 psi Standard deviation = 625 psi
'Again,'ACI 214 Criterion 3 for no greater than 1 percent of tests more than 500 psi below'fe ';
fe ' = 4,140 psi 1.3 Rock Bolt Grout The grout used for the' rock bolt installations was mixed I from Chem Comp cement, which was also supplied by Williams. From the Williams 1974-1975 catalog, the repre-sentative grout compressive strength is indicated to be as follows (Attachment B4): 1-day strength - 2,100 psi j 2-day strength .4,200 psi 1 3-day strength - 5,100 psi
-l The grout mixture is shown to be 4-1/2 gallons of water per 94-pound sack of cement. This low water-to-coment ratio (0.398 by weight) produces a high-strength grout, as the strength data demonstrates. The grout is'also a mix-ture which. expands as it sets up (ie, chemically compen-sated for shrinkage).
i
l Trojan Nuclear plant Document Control Desk 'I Docket 50-344 August 18, 1987 License NPF-1 Attachment B page 5 of 17 From inspection of the rock bolt samples core-drilled dur-ing the main steam line support modifications (12), the grout was seen to be of very good quality and adequately bonded with both the rock bolt and concrete (1-5/8-inch original drill hole for the rock bolt, 2-1/4-inch ID core drill, such that the samples also contained a layer of the i native concrete). 2.0 Fock Bolt Installations The rock bolts were installed in accordance with the Williams Form and Rock Bolt Engineering 1974-75 catalog. The procedure involves the following sequences for rock bolt installation:
- a. Drill the hole for the rock bolt. For the 1-inch diameter rock bolt of the type supplied (US-8-HC-SCS-158), the con- 1 crete drill hole diameter of 1-5/8 inch was specified by I the Williams catalog.
The drill hole diameter was verified by inspection of the samples removed by core drilling during the main steam line support modifications.
- b. Insert the rock bolt into the hole with the anchor head assembly (thrust ring, slip rings, expansion shell, and cone) positioned on the end of the bolt.
- c. Expand the shell by rotating the setting tool on the pro-truding end of the bolt with a wrench (preferably an impact wrench). From the Williams 1974-75 catalog, the torque to expand the shell is indicated to be in the range of 50-200 foot-pounds, depending on the strength of the medium in which the rock bolt is anchored.
For concrete with a design compressive strength of 5,000 psi, this setting torque would have been approxi-t mately 200 footpounds. From inspection of.the drilled-out l rock bolt samples, it was verified that the shells were properly expanded. The bottom of the cone was observed to l be plied nearly flush with the bottom of the shell, and in some cases, the cone was pulled completely inside the shell. l
- d. Grout the ar.nular space around the rock bolt, observing that a steady stream of grout is discharged through the de-air hole to assure that the entire hole is filled and that the entire area of the bolt is well grouted. Inspec-tion of the cored-out samples showed evidence of com-pletely grouted installations.
1 Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment B ! Page 6 of 17 3.0 Rock Bolt Anchorage Capacities 3.1 Modeling of Rock Bolt Pull-Out Resistance A completed rock bolt installation in concrete is a rein-forcing steel rod with an anchor head wedged into the con-crete and the annular space between the rod and concrete completely grouted to create a composite installation. Modes of resistance of a rock bolt installation can be demonstrated, and rock bolt capacities determined. Where ACI 318-63 code provisions are absent or nonexplicit for these applications, the provisions of ACI 318-63, Section 104, which allow for demonstration of adequacy based on documented test data and analysis, are considered applicable. In-place rock bolt tension tests, described further in Section C.3.3, were performed to demonstrate the ability of representative pipe support rock bolt anchorages to develop the relied-upon pullout capacities. These tests have con-l firmed that the capacity development in the rock bolt-grout-concrete system is not limited by conventional rein-forcing bar bond, but is rather more properly represented by the shear-cone mechanism described in ACI Standard 349, , which is addressed in the following Section. 1 3.2 Anchorage Capacity This section demonstrates the results of applying much more conservative rock bolt anchorage criteria than were refer- , enced in Attachment A to our July 31, 1987 letter. ! The ACI 349, Appendix B, empirical pullout strength rela-tionship and the current Williams rock bolt conservative, shear-cone effective area, A egg, representation are used. The pullout strength, Tu ' 18 Tu " "u Ae gg, where values used are: l Uu
- 4 + (f e')1/2 (per ACI 318-63, Section 1707, and ACI 349, Appendix B) l l 4 = 0.85 (per ACI 318-63, Section 1504) l
Trojan Nuclear plant Document Control Desk j Docket 50-344 August 18, 1987 License NPF-1 Attachment B l page 7 of 17 i For 5,000 psi concrete design strength, fe' = 6,130 psi. I' For 3,000 psi concrete design strength, f e' = 4,140 psi. (Refer to Section C.l.2 for values of f e '.) l From the Williams Catalog, Ae rt is based on a shear cone ! emanating from the top edges of the rock bolt expansion i shell and extending to the concrete surface at an angle of 45 degrees as shown in Figure B3. l The Aert becomes 1 (D + 2 L)2, in which l 4 D is the diameter of the expansion shell at the narrow end (D = 1.5 inch), and L is the distance from the con-crete surface to the top of the expansion shell. From measurements of the drilled-out rock bolt samples, a representative value is L = 5.5 inches, based on an 8-inch nominal embedment depth as shown in Figure B4. Thus, Ae rr = g (1.5 + 11.0)2 = 122.72 in.2 l The pullout strength becomes: l 5,000 psi design strength concrete; i Tu = 4(0.85)(6130)1/2(122.72) = 32,668 lb ' 3,000 psi design strength concrete; Tu = 4(0.85)(4140)t/a(122.72) = 26,847 lb The shear strength, Vu , of a grouted anchor is not sig-nificantly affected by concrete strength unless the anchors l are located near an edge (ACI 349, Appendix B). The i Bechtel design guide (Design Guide C-2.34) used for current l anchor installation provides limitations on anchor bolt l edge distances to eliminate the shear-out concern. This document was used for guidance to verify the acceptability of edge distances. The anchor bolt shear capacity is then conventionally based on the shear yield strength of the anchor bolt material. _ _ _ _ _ _ _ _ _ - _ _ _ _ _-_-_ - ________-______-___________-_____ ____. . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _____. . _ _ _ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ ~
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment B Page 8 of 17 I l Shear yield, fys, Per the American Institute of Steel i Construction (AISC) Manual (eighth edition, Commentary Section 1.5.1.2), is given in terms of tension yield, fyte l as: l l f = Est Y8 (3) The value of tension yield for the rock bolts is l fyt = 37.0 Kips (see Section C.1.1). Thus, 1 Vu "fs y -37.0 = 21.36 Kips (3) In summary, the rock bolt pull-out capacity and shear capacity derived from the criteria' described in this section are as follows: 5,000 psi design strength concrete: Tu = 32.67 Kips i Vu = 21.36 Kips 3,000 psi design strength concrete: j Tu = 26.85 Kips Vu = 21.36 Kips These pull-out capacity values are less than the rock bolt yield value of 37 Kips. To arrive at basic allowable loads for working stress de-sign, which was the design method used for the subject pipe support designs, a reasonable factor of safety of 2.0 was selected. Review of the typical loading combinations for working stress and ultimate strength design under similar loading conditions indicates that 2.0 is an appropriate and reasonable factor of safety between working stress and ultimate strength allowables. The factor of safety of 2.0 to ultimate is also consistent with the Williams recommendation for rock bolt designs (Attachment B1). The basic allowable loads for working stress design are then: l l
l l Trojan Nuclear plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment B page 9 of 17 5,000 psi design strength concrete: a= u = 32.67 Kips = 16.34 Kips 2.0 2.0 a= u = 21.36 Kips = 10.68 Kips l 2.0 2.0 3,000 psi design strength concrete: a= u_ = 26.85 Kips = 13.43 Kips 2.0 2.0 a= u = 21.36 Kips = 10.68 Kips 2.0 2.0 In accordance with the original Trojan design practice, basic allowable loads were increased by 1.25 for load combinations which include the OBE, and 1.5 for the load combinations which include the SSE. This criteria results in the allowable loads for the 1-inch diameter rock bolts with 8-inch nominal embedment as shown in Table Bl. All safety-related pipe supports which have 1-inch diameter rock bolts with a nominal 8-inch embedment were reviewed to determine the supports that have the highest load demands. The allowable loads were compared to these demands, and a linear shear-tension interaction relationship was applied. The pipe supports which originally had the largest interaction ratios are shown in Table B2. For four pipe supports which had inter-action ratios greater than 1.0, modifications to significantly increase the support anchorage capacities have been completed. l 3.3 Test Data l On August 11 and 12, 1987, PGE performed tension testing of nine 1-inch-diameter rock bolts with a nominal 8-inch embedment in the anchorages of safety-related pipe supports at the Trojan Nuclear plant. The rock bolts were installed in 5000 psi de-sign strength concrete. This testing was performed for the purpose of: (1) demonstrating that lower bound bond failure l performance would not devalop either at' the interf ace between the rock bolt and grout, or at the interface between the grout 1 l
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987' ] License NPF-1 Attachment B i Page 10 of 17 and concrete; and (2) to demonstrate that the relied-upon i pull-out capacities, as presented in Section C.3.2, could be developed. This testing was performed in accordance with PGE Temporary Plant Test Procedure 218 (TPT-218). The safety-related pipe supports from which the test sample rock bolts were selected is shown in Table B3, and the results of the testing are shown in Table B4 In summary, all of the pipe-support rock bolt anchorages tested 4 withstood the test load of 34.5 Kips without failure, and l there was no evidence of initiation of a bond failure j mechanism developing between rock bolt and grout or between i grout and concrete. The test load value of 34.5 Kips was based upon the calculated capacity for rock bolts in j 5000 psi design strength concrete calculated as described d in Section C.3.2. Following the test, the concrete around one of the rock bolts was sounded and exhibited looseness of the concrete, ; though the rock bolt held the load of 34.5 Kips. A retest I to failure on this rock bolt (SS-1106, Bolt No. 4, embedment based on measurements = 6-3/4 inches) resulted in a tension failure of the rock bolt steel at a load of 51.5 Kips, with no pull-out of the rock bolt anchor head. One of the preliminary steps in the TPT-218 procedure is to verify the embedment length of the rock bolt by ultrasonic measurement. Review of the initial ultrasonic measurements disclosed that some of the rock bolt embedment lengths were less than the nominal 8-inch embedment. In order to pro-vide assurance that capacities corresponding to rock bolt embedmonts less than the nominal 8-inches are adequate to resist the load demands with a factor of safety of 2.0, all accessible rock bolts with 8-inch design embedments on pipe supports and restraints have been ultrasonically measured. The accessible rock bolts are those that are not in high radiation areas, that do not involve excessive man-rem exposures for scaffolding, and that do not involve unacceptable personnel safety risks. This population includes 23 of the 29 most heavily loaded supports. l For the category of accessible rock bolts, pull-out capa-cities were calculated for the measured rock bolt embedment depths using the criteria in Section C.3.2, and these capa-cities were compared to the maximum rock bolt load demands i for individual safety-related pipe supports. For the ente-gory of inaccessible rock bolts, a high confidence level-minimum expected rock bolt embedmont depth was determined based on a statistical evaluation of the embedmont depth
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment B Page 11 of 17 data from the accessible rock bolt category. A pullout capacity was calculated using the minimum expected rock bolt embedment depth. This pull-out capacity was then used for comparison with the rock bolt load demands for all of the inaccessible safety-related pipe supports. Results of the rock bolt capacity to demand comparisons are shown in Attachment B5. As can be seen, the shear-tension interaction ratios are less than 1.0 in all cases, and the pull-out capacity of the rock bolts provides a f actor of safety of at least 2.0 for the tension values resulting from load combinations which include the SSE, in all cases. 4.0 Conservatism in Criteria 4.1 Concrete Shear Cone A very significant conservatism is the use of an embedmant depth, L, for shear-cone effective area determination. The L distance has been taken from the top of the expansion shell, rather than the bottoa. This is shown in Figure B3. An L distance to the bottom centerline of the shell would be permitted by ACI-349, Appendix B, interpretations, and is also considered appropriate by Williams representatives (see Attachment B2). If an L distance to the bottom centerline of the shell were used, Ae rf = g (2 x 7.5)2 = 176.71 in.2 as opposed to the 4 Ae rr = 122.72 in.2 used in the formulation in Section C.3.2, and the capacity would be increased by a factor of 1.44. 4.2 Presence of Reinforcing Steel Although unquantifiable, the presence of the reinforcing steel results in some inherent contribution to an increase
- in the pullout capacity. Rock bolts with nominal 8-inch embedment are sufficiently deep to anchor beyond the first ;
layers of reinforcing steel. The Containment internals concrete, where most of the pipe supports with rock bolts are anchored, is particularly heavily reinforced. I
Trojan. Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment B page 12 of 17 4.3 Interaction Radios Use of the linear shear-tension interaction ratio of j T + V = <'1.0 l Ta Va as opposed to the nonlinear interaction ratio of
/T I
, 5/3 + ' y ) 5/3 , g 1,o
\ T,; yve ; l also represents a significant conservatism. Test data for concrete-embedded anchors, of various types, show that the j nonlinear shear-tension interaction ratio best represents i the capacity envelope for combined shear-tension loading. )
Use of the nonlinear shear-tension interaction ratio with I an exponent of 5/3 is endorsed by Electric Power Research ; Institute (EPRI) Report NP-5228 " Seismic Verificat:lon of ' Nuclear Plant Equipment Anchorage", which was developed j with input from NRC staff reviewers. l 4.4 Load Demands 1 The derivation of seismic loads also represents a conser- ' vatism. For Trojan, the Safe Shutdown Earthquake (SSE) loads were typically obtained by multiplying the OBE loads by 1,67, regardless of the piping-system's response fre-quency [ Final Safety Analysis Report (FSAR) Section 3.9.3.4.l(3)]. This ratio is applicable only in the high frequency range of the spectra, but was used throughout. Also, a damping ratio for piping of 0.5 per-cent was used for both the OBE and SSE (FSAR Section 3.7.3.5.3.). Realistic damping ratios are typically higher than this by significant amounts, which would reduce piping response loads and reactions on the pipe supports. D. Conclusions The technical aspects of rock bolt installations at Trojan have been summarized with focus on 1-inch diameter rock bolts with a nominal embedment of 8-inch. Overall conclusions are: The materials involved in the rock bolt capacity development are of high quality and strength, and proper for the applications.
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment B Page 13 of 17 ; 1 The rock bolts were installed properly. A much more conservative criterion than was originally used was developed and applied to the rock bolt installations. It was shown that safety-related pipe supports with 1-inch diameter rock . bolts having the minimum nominal 8-inch embedment can meet these l more conservative acceptance criteria, or have been modified to meet these criteria. I The safety-related pipe supports with rock bolt anchorages are ade-quate to resist the design loads and have sufficient capacity margins. 1 s
.J l
l l l I i I I i i I l l TEB/mc 1963W
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 l License NPF-1 Attachment B j Page 14 of 17 { I i TABLE B1 i CRITERIA 2 ALLOWABLE LOADS l FOR l-INCH DIAMETER ROCK BOLTS - WITH NOMINAL 8-INCH EMBEDMENT ' Basic Allowable OBE Load SSE Load l Load Combinations Combinations (kips) (kips) (kips) Concrete Design Strength = 5,000 psi: Tension (Ta ) 16.34 1.25 (16.34) = 20.43 1.50 (16.34) = 24.51 Shear (Va ) 10.68 1.25 (10.68) = 13.35 1.50 (10.68) = 16.02 Concrete Design Strength = 3,000 psi: Tension (Ta) 13.43 1.25 (13.43) = 16.79 1.50 (13.43) = 20.15 Shear (Va ) 10.68 1.25 (10.68) = 13.35 1.50 (10.68) = 16.02 I A rock bolt shear-tension linear interaction ratio of: Shear Load Demand + Tension Load Demand I 1.0 Shear Load Allowable Tension Load Allo'.*1ble E+ I- T f 1.0 will be used Va a J l l l i TEB/mc 1961W.887 , I i l I _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ __ _. ._ . _ _ ____m. __ ___ _ _ _ . . _ _ _ _ . _ J
I Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 g . License NPF-1 Attachment B l Page 15 of 17 TABLE B2 SAFETY-RELATED PIPE SUPPORTS EVALUATION OF HIGHEST LOADED 1-INCH DIAMETER ROCK BOLTS WITH NOMINAL 8-INCH EMBEDMENT Load Allowable (kips) Load Interaction Governing (Criteria per Demand Ratio Load Section 3.2) (kips) T +V Support Combination Ta Va T V Ta Va i 5.000-psi Concrete C-386 II SSE 24.51 16.02 13.4 5.0 0.86 SS-1135 SSE 24.51 16.02 14.4(1) 1.7 0.69 SS-1210 SSE 24.51 16.02 21.4 8.9 1.43(2) SS-1133 SSE 24.51 16.02 9.2 5.9 0.74 SS-80 SSE 24.51 16.02 0 16.0 '1.00(2) SR-82 OBE 20.43 13.35 0 6.5 0.49 SS-83 SSE 24.51 16.02 0 16.0 1.00(2) SR-85 OBE 20.43 13.35 0 6.5 0.49 SS-91 SSE 24.51 16.02 19.0 6.0 1.15(2) 3.000-psi concrete SK-248 OBE 16.79 13.35 0 9.3 0.70 SR-803 OBE 16.79 13.35 4.8 2.0 0.44 SR-300 OBE 16.79 13.35 4.8 2.0 0.44
- SR-815 OBE 16,79 13.35 6.3 5.2 0.76 l
l l l l (1) It is noted that ratio of relied upon capacity to rock bolt tension demand for the load combination which includes the SSE is 32.67 = 2.27. 14.4 (2) Supports have been modified to increase anchorage capacity. 1961W16.887
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment B Page 16 of 17 TABLE B3 SUPPORTS WITH 1-INCH ROCK ANCHORS TESTED l Item Location Drawing Test Load Number of No. System Support Number (feet) Reference (kips) Bolts Tested l
- 1. SI* SI-601R-1-1 SS1218 Cont. 45 C-386 34.5 1 Sheet 4A I
- 2. SI* SI-2501R-1-64 SS1106 Cont. 45 C-386 34.5 2 j SI-2501R-1-65 SR1118 Sheet 4A l
- 3. SI* SI-2501R-1-65 SR1122 cont. 45 C-386 34.5 1 I l Sheet 4A l
- 4. SI* SI-2501R-2-4 SS1157 Cont. 45 C-386 34.5 3 Sheet 4A I S. SI* SI-2501R-2-3 SR1155 Cont. 45 C-386 34.5 1 i Sheet 4A
- 6. MS** EBE-6-652 SS2102 Cont. 77 C-386 34.5 1 Sheet 6 9
- Safety injection System
** Main Steam System l
l i TEB/mr 1961W.887 i
l l l l Trojan Nuclear Plant Document Control Desk l Docket 50-344 August 18, 1987 ) License NPF-1 Attachment B Page 17 of 17 TkBLE B4 ROCK BOLT TEST RESULTS I Depth of Support Bolt Embedment Test Loa Number Number (inches) Cracks (kips Remarks SS-1106 1 6-11/16 Yes(2) 34.5 Hairline radial crack approximately 3 inches long each side of bolt after test. SS-1106 4 6-3/4 No(3) 34.5 Observed hairline radial crack before test. SS-1157 1 8-1/8 No(4) 34.5 Small area indication. SS-1157 4 8-1/4 No 34.5 SS-1157 6 7-7/16 No 34.5 SR-1122 3 7-11/16 No 34.5 SS-1218 5 8-1/16 No 34.5 SR-1155 1 8 No 34.5 SS-2102 4 7-1/16 No 34.5 NOTES: l (1) Test requirements (for 5000 psi concretc): 33 (-0, +2) kips. (2) Bolt No. I was core-drilled out for evaluation of the crack indication. j A horizontal crack plane about 2-inches below the concrete surface was observed. (3) Following the test, the area around bolts 1 and 4 was sounded. For a l distance from the bolt of about 10-1/2-inches in one direction and about 16-inches in another direction, describing an oblong shape, the sounding j I gave an indication of looseness of the concrete (see attached sketch). Bolt No. 4 was planned to be retested up to the capacity of the testing equipment i to determine if there was evidence of initiation of a concrete pull-out cone. The maximum load achieved during the rotest was 51.5 Kips, at which the rock bolt steel section failed. A shallow, approximately horizontal i crack plane through the concrete about 2-inches from the surface, similar to I that seen for Bolt No. 1, was also observed at Bolt No. 4.. The anchor head l of the rock bolt, however, did not pull out of the concrete. ; (4) Local sounding indication, but no visible crack. Inspections and further ; sounding showed no evidence of significant structural distress. This local ! area was subsequently chipped out and repaired with grout. ARA /mr/1961W18.887 i
R O i g H y,.y pg -
- e. ," ,-
,- SC PN L A I
3 l
+
r I LE I HG i sy' PD
./
$f .F E W
. - p
- G NLT B I
- -, e .?p SLO I
. g L
p PLB E I R R I iN 5
,gE1 L.R
'l L
l _D O U G
~
HF H
~ .
- rY*
P L C I
- M.?
/'
h E N F S-A
~
s T
?rj L
% 4 I O g T L-B Rtf HIK I
^
. g K
W h'4
/ -
r ! EiB l TUBE thhE ']~ [ [ GROUT ft ! r ( u i [s r GROUT CANAL z u I _ THRUST SLIP m RING l RINGS _
-EXPANSION SHELL CONE- =l h. 1 FIGURE B2 WI _ _l AMS HO__OW CORE ROCK BOLTS
1
= D+ZL - i O
bh '\NI k 'l
/ . .
s..
~
'.~
~. '
\ ~, ,
Y, . N / e
, hQ ' *J,o h
\ .
.- .g
/ ,
l N .p. . C' :. ., /
. . - N -
/
oJ ,. x - I ..
/
- b? N .
/ ,
\ ' '* V . ' *
/ $$
, Y, .
/ "
. .\ .
/ .
! ]
'y ;
's
's i / O,"
. \
v i j e
/ i
. . n 6 MEA.E C. CME 60RFAC.C h. 'j
' FOR. t>E. TEEM \ MAT. \OM ll ll! ' i JA CF NECTED AEA I lI! I A. L , ,o J. _ _ . _ Lt 0 0
- 90 o= ik ; Toe or satt_.t.
l . i ciAmt1ma
~
Aeg=bo+zti comcemz eacm cme cuzcnue A E" A , A eg;. FWORE-53
e 1 1, DAMETE.R. I I R O C K S O t._T DE- AIR:. HOLE Folt cWOUTI146 ,
=.L:
w m.:=-% .-, ll 4 gg AtcMolt
,, PLATE !
, mw .i m w \ , j h.. . . : .. . ;. .: . . lI .....
/f . . . . .;)
ll
*:~ .:. *
* . . c : '. . : m- , l
. i o l
.r e -
3, .
- 8. J ll .
.\
1.9
. j ll
. CiROUT .. )
.p. !
e,. . J. ll .D: j ll ll -
.' i
,Q
~
Il o
. . l ~4 to !
yge.;a 1 - rj. o ..
,. . li, v
<i j ,
/.. *T I T
N ~ d EXPA Stow l!\ ./ ' SM2kG
". . O Ol -
LL v
=TC l [ [
t i pp t M* * *
... = 1%= '80 .
l-tu. otevcTEe R.ocx. Sot.:r wrra K C M N.A.L. 8'N. EM6E.DMEMT f l F % U E E.- 64
Tro an Nuclear Plant Document Control Desk' . Doc et 50-344 August 18, 1987 License NPF-1 Attachment B1 j Page 1 of 3
?
.,,..... 1
. v :; ;; . ....
.+ l ,. ,' , '.* : $7, . [
. u ...
- n dy ame%. *r s
-.. w :. : s .. .
. .s s u t .. .. . g f gl c.
RG lt j d . n.y.w
~.ij.}b; i,\j:{:p ghg ..,
i Mr ..
..c.
- a sg%., ..
u
, ,i'f.h, ~} ('.
..;^ .
- ng ,' ,g.: ' ., ;.
.p. < , e e,;,, ;;
;y,.% 4 ,~Y .g.a- .u . 4 . ..
t
; ; a . .. b.
. _ ." . . .r ; ~, .
.t 'fr ,, ['f') , ,. ' ' '
RA i
*.-'g.a $ 6 ..Ihre,:g, .ed
{;i;7A%C s.,. Nk[l g .7 s N pqp d % # Y @gN .k
;g 3 , s.
m , ye - pg<g+.g%
.. g..ws, . l,-
{y / b.e 4 / i 4.- . f' %W ~ ' g'
~
5
,m
. %, - %s.wuih% - . .y~.w.
. .)..
3 Y$ - gg ' ?c., b"' ,q' ,.u... v,.
.#i $,.,fJ ,4 ,, y 0 h $ Nh4
~7 <g.gy.;g if ,
i- . ,y
--.,y .<tpfgg.,rgg,z w .
+
co A 4 t
. . g 4 h u -
2
* '. 3 Trojan Nuclear Plant (Docket 50-344 !"';"" ""'* m n WILLIAMS
- PRESTRESSABLE HOLLOW CORE Document Control Desk EXPANstoN SLIP August 10, 1987
( sHELL RINGS THRUST Attachment B1
, \ [ RING Page 2 of 3 0
1 i'
. v; , /jg s.ren 6
HOLLOW BAR MEETS ASTM A 615 , i
$ PRESTRESSABLE
- POSITIVE GROUTING e PERMANENT 1
1 Through years of development, Williams This grouting procedure is a critical point i has produced and patented the only Pre. in creating a permanent rock bolt, since , 1 stressable, Hollow-Core Groutable, an ungrouted bolt is subject to rock relax- l Rebar, Spin lock Rock Bolt System. The ation and deterioration. In the Williams ! f
- " hollow core" allows the bolt to always hollow core system, as the grout rises it L
be grouted from the lowest gravitational spreads through the many fissures and t point in an up bolting situation, the grout voids to solidify the rock and create is pumped in through the plastic grout monolithic section including rock, grout [ tube and begins to fill the drill hole from and bolt. The bolt is completely engulfed j p J the plate. The grout rises until the entire in grout and is protected from corrosion. j' I hole is filled and the grout returns This is extremely important and only ob-j through the hollow bar. In down grouting tained with Williams hollow core grouting situations, the grout is pumped through systems since epoxy and pre measured
$ the hollow bar and starts at the bottom of cement grouting systems do not allow for I the hole. It rises and returns through the the grout which spreads through the rock.
[M de-air tube when the hole is filled. ADVANTAGES OF Wil.UAMS+ HOLLOW CORE over two tube or resin systems The deformed, high bond rod of the US. RESIN OR PRE MEASURED GROUT SYSTEMS 1 Fissures, voids. or rock fractures cannot be pf t measured to ( determine the proper amount of grout required. I! 2 Water or other cebris is left in the hole and ts mixed with grout. N M l
- 3. Bolt is poorly grouted and left exposed to corrosion and failure.
1 ( _ l ) TWO TUBE SYSTEM "WlO,'aM" i
- 1. Groutmg tube is very susceptible to damage curing installation. ,
- 2. Damage to the tube may result in premature grout return .
p r
~ ' ~ ~ ~ ~ ~
- 3. Poorly grouted bolt is left susceptible to corrosion and failure WILLIAMS' / ..m.,,,,,,
HOLLOW CORE SYSTEM i " """
- 1. Grout is pumped through the bolt forcing out water and debris and replating with grout.
- 2. All fissures, voids. and fractures are pressure grouted starting : d f ple es in pe t on
- 3. Bolt is ;>ermanent. Safe from corrosion or fai!ure. i \ [ j
Docket 50-344 i g License NPF-1 d i GROUTABLE RE BAR " SPIN LOCK"* ROCK BOLTS
- xevsotE $
Document Control Desk PLATE ) August 18, 1987 . 9 Attachment B1 g Page 3 of 3 i M} i . MU 1_ -- - ss 3 HEX N) j u. NUT
. 3 HC series, is manufactured by a special GROUT process from a high grade of steel {j ',"tof TUBE q substantially exceeding the strength of j common re bar stock. The " SPIN-LOCK' "
head assemblies provide 300* perimeter ! 'l , 4 h !
)dfdj m
s fi expansion anchorage and are designed -
,il y to develop the full strength of the hollow- j h)h
~
$ q core rod. {[ j j l [
Because the " SPIN LOCK"' head as- 4o d5 [ sembly develops the full strength of the Ji1' yE J;l rod, the Hollow-Core rock bolt may be pre-stressed to the desired design load and -
%dg:
j p{e *
~ a r' r t croutina.
SO K p 3 % F. S . - :. 2.o 1g i 7sK nyy_g -- - - . WILUAMS* HOLLOW CORE RESAR ROCK B01.TSg,
! s t
MAXIMUM DIA-RECOMMENDED DESIGN LOAD WORKINC LOAD TO ULTIMATE ROCK TYPE DRILL HOLE TYPE HEAD TO PART REFEPENCE h
.4 AT APPROX. 2:1 ELASTIC STRENGTH D:!L [1) ASS'Y EXPAND ON NUMBER NUMBE R L SAFETYFACTOR LIMIT SHELL(2) NUT [
HARD & MEDIUM 15tr 41 rpm A 13 200 3JO H1 Hob A 13 USS NC FCS 158 j l HARD & MEDIUM 134' 44 mm A 14 i 250 3tr R1H 06 A 14 USS HC SCS 175 HARD & MEDIUV 17 B" 48 mm A tr, 250 300 R1 Hot A 15 US+ HC E CS 178 q 13 25.0tc ins 37 000 lbs % 000 IM MEDiuv & WEAK 1 $$ ' - 41 mm D 13 250 300 R1Ho6 B 13 USS HC l CS 1M
}g 2t mm 11250 [f kg 16 bJ0 kg 22 700 kg MEDIUM & WE AK 13,4' - 44 mm B 14 250 3(0 R1H06D 14 US S HC LCS-175 MEDlUM & Wf At 17 B' 40 mm D 15 2% 300 R1 hob D 15 US b HC LCS 176 1 g WE AK ROCK & CONCRETE 134 - 44 mrt C 14 250 300 R1HU6 C 14 USS HC LCSF 17E WE AK ROCK h CONCRETE 134" 44 mm D 14 250 3J0 R1H OB D 1.8 US 8 HC.LCSi r 175 HAHD & VEDIUM 2 51 mm B 16 250 NOTE 0; R1H 11016 US 11 HC.L C5 2'X) ,
MEDIUf/ 6 WEM 2 144 ' 57 mm H 16 $50 NOT E D) P1H 11 D 1b US 11 HC LCS 225 MEDIUM & WE At 21l2" 63 mm C 2v $50 NOTE Di R1H 11 B 20 US 11 HC-L CS ?$0 ) 13tr- e 50.000 lbs 74.000 bs 100 00rgs EAK ROCF & CONCRETE 21.4 ~ 57 mrn C 10 750 NOTEi3i < R1H 11 C 18 US i1 HC LOSF 225 35 mm 22 675 wg 33150 t g 45250
- g WEAK ROCK & CONCRETE 21!T 6J mm C 20 750 NOTL O,' R1H 11C 20 US 11-HC LCSF 250 WE AK ROCK & CONGRETE 21/4" . 5? rnm D tB 750 NOTF 0 R1H 11 D 1e US 11 HC L CLLF 225 WE AK ROCh & CONCRETE 217 63 mm D 20 750 NOT E (3, RtH 11 D 20 US 11 HC LCSLF 250 l 2' . 6 100 000 lbs. 146 (no los, 200 000 lt s HOCK a CONCRETE ,! 76 mm D 24 1000 NO 7 E (3! R1H 16 E 24 US 10 HL i CSHf .EXs 5G 45350 kg 67.150 eg 60 73u >c WEAK ROCK & CONCAETE 317 89 mm D 26 1000 NOTEci R1H 16-D 26 US 16 HC LCSLF 35( ~l l
NOTES. l m Ca'e thould te ta**:n to prevent d'ilung ovew.md ncdn (2) A funtifran Of t,tf dta Strength Moft torque may tie requif8'lt uf) !O"Q bo'?L O' m Special ruth COndef aunk COnhalf yOgr Wilhams Reprnen! alive for more speck deidelb Qi htfen 10 fleb6 fed it!h5fle lOnd using a h01DA f am hydfdJl lC jdCh COnbJll iDur Nill 43NL beitr8MnldDvt
-,.S,-._...,-,~.,~,-,-,_.,
;4 r
., i
~ * '
zu, sammru unw a
4 Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18 1987
. , . '-- , . . License NPF-1 Attachment B2
?', .y Page 1 of 4
,1 .,
.t p .
, . . g
' WILLI AMS FORM ENGIN EERING CORPORA rlON
' q.
J- 1501 MAoisoN AVt., s.E o ORAND RA#tDs. McHIGAN 49607. M16) 452 3107 e TlLfX No. 22 M16 f_.7. .. . -
-) "INTERNAT10NALLY KNOWN" w e n !J, August 11, 1967
~
l Bechtel-Western Power Co. l '50 Beale Street ' l San Fransisco, CA 94105 !' Attn Dr. W.H. White l ( Dear Mr. Whites i I In addition to the Williams letter of August 5, 1987 ve include the following information: A strength reduction factor ( ) of .65 was used (per ACI code 349) by Williams Form Engineering to calcu ate the concrete embedment depths of the various Williams rock and concrete anchor bolts. The type "A" anchor assembly (short cone and shell) is acceptable for use in concrete, especially when used in higher compressive strength , concretes. The acceptability of the performance of this mechanical l anchorage would certainly have been noted at the time of installation when the bolts were prestressed (thus pretested) prior to grouting. When embedment depths are to be adjusted for depths that are not shown in the Williams Catalog, a designer can consult ACI 349, section B-7 for expansion anchors in concrete. Variables such as strength reduction
- factors, compressive strengths, etc., may reduce embedment depths shown in the Williams catalog.
We hope this information is useful for your needs. , Yours truly, s , ! We 7 p 5 1 J., Miles
. h li
- Engineering Manager WILLIAliS FORM ENGINEERING CORPORATION d...
*e#
_ _ - -oo - - , Docket 50-344 August 18, 1987 I 7 ,- - . License NPF-1 Attachment B2 l
/ . Page 2 of 4 1
= c l
-~
WILLIAMS FORM CNQiNEERING CORPORATION l g
'*v '
1901 uADitCN Avt 8 L
- 04AND RAPIDE, MICWlGAN 48607. slt) 4444107 e TELEX NO 22 9418 "lNTERNAtl0NALLY KNOWM" 7- w
, August 5, 1987 I
Mr. Bill White Bechtel Power 50 Beale street San Francisco, CA 94105 I Re Williams 1" Hollow Core l Rock Bolts at Trojan Plant ! l
Dear Bill:
We are concerned to hear of your problems at the Trojan Plant. We hope the following information will help. l The minimum embedmont chart shown in our current catalog was added l in 1960 and, therefore, was not there when this project was l designed. The AC: 349-79 Section B.7 for expansion anchors is the normal and the accepted method of calculating the shear cone strength of mechanical anchors in concrete. With the various sizes and lengths of expansion anchors that W:,111ams produces, the problem of how much catalog room to devote to wnich anchor was settled by advertising the figure that represented the largest and longest anchor assembly for each ! diameter bolt produced. Also, unlike ACI, we calculated the shear cone strength from the top of the shall to further insure a conservative depth (Which reflects the total contact area of the shell). It is the general accepted prt.ctice to calculate the shear cene from the centerline of the bottom of the expansion shell as indicated in AC:. This also may be semowhat conservative by AC1 since it neglects the bottom of the shell area in contact with the l cencrete. g ..,- i fi$Il j] k 4 une u i.. .....:.....;
-Auguct D, 1937 Document Co11 trol Desk
. Trojan Nuclear Plant August 18 1987 Docket 50-344 Attachment B2 { License NPF-1 Page 3 of 4 4 4A any ancnor assembly was to be used that was not the largest and/or longest for its diameter,.it would be-advantageous for the designer to calculate for this condition. If for instance, a short cone and shell assembly is installed to the same depth as the advertised embedmont depth, it might indleate a much higher concrete shear cone strength than would be necessary, r These figures all represent a conservative approach'for general applications to insure that the mode of failure is the ultimate strength of the bolt and not the concrete. All Williams calculations were made on non-reinforced concrete. If a concentration of rebar is within the projected shear cone area, the resulting actual additional strength could be many times the calculated values used for non-reinforced concrete. I cur hollow core rock belt has been designed with triple redundancy to help prevent helt pull-out. First, the spin-lock cone and shell anchor is mechanically locked in place. Secondly, the ancher mass acts as a " dead man" anchor against the grout' column. Thirdly, the bolt deformations are bended into the grout column. As I under-stand our Wil-X-Coment was used with this installation, I have attached a physical testing report on expected strengths of the grout. , If you have any additional questions or comments, pisase contact us. Also if you would like someone from our office to meet with you or to assist with any testing, pleaso let me know. Best regards, j 7
.om A.,5pama ;
Vice P#esd nt of ; Marketity'and Sales TAS/sds Enclosures - Physical Testing Repcrt of Wil-X-Cament Grout
#668 Anchor Bolting System Catalog
u- - - - - - - - - - -
. Vgepre esawo Document Control Desk i Docket 50-344 I License NPF-1
.'"l'C" "O August 18, 1987 Williams W19..X.coman y c1t Attachment B2 egegne 0*e .
Page 4 of 4 Warehouse ASTtt Neat Cement C ogle sies sow W/C Matie V or64**tti _ _ _5mple __ c . i n e Hodifid 'M 8'T" T
" " " ' "- - . wil.x per 556 st. we u. i i ue_. m utn .eamwe- i a n C w 11,,A , ,9'_ "
4 l ;
.e.w.1 cre s.o r we ,i i
#f 08 Sit towc iMal_ stW _
wet at fit twoor ewen4L _ l
*.t t uce t r ewat St sti.Patti wt'a00 ,
i
.eewayag (t - - - -
wt.. dos. 4ut eet ive p
- 4~ c- __ _
q u* s aamut ut _
,e a 71.0 ._
i i s te i -> 82 i 821 l l , J, aest.a.i .cu. ll 1 1 s e e ccut i a t :*.e i 8,6 > 98tibivt l?8fwCTw i.e.m 50.0
" -- l I l W/C l
l t o . . ., e.. ev i I .I 1773 ; 4192 ~f
" H 3032- 6792 e 4 i *D ' 9025 2*s*9' _8 4367 e o* n .*9
- I i 5392 1 12075 1 I e o i gaagettsg=ce t a t me n 'w*, tu t ,
! u l l
sa s e -se ,, , l l I l i e..,. a u n n , .
,,'.., 6_
l at ttwtisaticw.wes. ~. ... ...,, . . . , . _ I
.w ,
-l ;
7/24 I - PNVilC AL finflit E. BurMS , L. M A I'J41 Caff,Allt Mtp lewice CMtansLT NI SAftCC*PLfftD_
Trojan Nuclear Plant Document Control Dssk , l Docket 50-344 August 18, 1987 NDP 200-4, Rsv. O I License NPF-1 Attachment B3 Attachmint..C 5 pages , NUCLEAR DIVISION CALCULATION COVER SHEET i Title DM/#d8 0AS t (NMVY o ; ; e 1 Projeet Tra , e Calculation No. Id" b/ O Y I,
- Supersedes Calculation No.
Structure System Component References (RDC/SPEER/MR, etc) ///A Calculation Ob_iective/Su : nary Description h / $$M$3tMoe &T $d $3VM 6+< k
$Q ld(4 {${C Afix et w / C Z -
l Computer Code Dated Program Version /SM6 GI l/ 8.d/ Operating Syste'* /! Hardwatb Configuration i 4 Checked Approved . Rev Date by Date j No. Revision Orirsinator Date SW f
~
A'kW?/ El7/27 / 1 67 /68 di{f/JJ (/
)
1 l I i j a f SGG/mr 1 0974G l 1
,'y' wu/uc owvu U!
[ Calci" ' ; 8-40 207 OJob ~ GENERAL COMPUTATION SHEET O File l i-Pr:iect/ Job r/Cf/M) I (- Subiect de/J4.2E7"4 %E'SZ69t) 4172 CA)h y-t/ q # //
^\
l By b) P Mr2E & o v' Date 3 /9 / 9 ')_ chk. ay ct,p(// }((1fp _D ste 6/2 /b) Orig. GYRev. D . f / / 1 j ; TTFTT I! !
, QlT{~ [~l l l ,
Il l { I I l, 1 1iii! i f I! I
! I I i li iI 4
-.._ 8 AC M. h . N L t d/ T 1 ,T h fr UL5 ! h* CN1kr'eh5 l I \ I gr
. 1 j An[d3[ystlWhad.c.~....'l'l{
> i ! l ! l lt lll O+._ l l
1 i . : '...a<.,,.. L
-n--
! i
' ! 5 I I !I I I I
!.76FE/&dA/6 6:5.* I i_ b._. _. t l t i i l-4e wM*
e - - . e- .e .- +d.= h e.geg .
< : i .
.L_-.- .i~. ; :
- ! 4
- !I I i
*e **'
~.
iiii>
, . . i 1 : : ,
' ?
f.. . . f . ' . ' . . . . .. .. ! __ l e.J . . .
,_fon#x t?? 7-Ser! ._- ..-- _ = ... . .. .[.._.l _j . {7 \ . -
T! i I \ ;
!.~]&,;rI_ /fcx zi 4 is -
+
; , . : i i i
; ; 1_.I fhIt5 hb blCS. _ $$t c 4 <'</ %. :tY.-Ah.6..[_.--...
. i : i , ,l ;j g
l
~~~ ~ '
a g*p/$c/rpy / l
, ', i I
. - . i.a - .._ . ,
[ k,. _ ,, / 8T b [. .. b
,1
, i i i : 3 1 ,_ ; i I 4
.._r a .
...L... [ ... _.(._.
. /r $o/1 tW e of tM CM Y~ Efdt?) d ff <
; . 5l 1
, , . 4 i' i (
. - i i .
i I.___II , .__ ,
/, h *;_. 38R- . / 287 [ Of J:~.j,_ .!' 8_/_ / _i i f,;2 -
i i ; i i : : i : !! ! 1 i i , s.___ i
.- g ., _ , !_ .:
u Y//f2' l
! I i ! ; Il! I.
Ii !
' i ! . _! ._ ,} ,l I ', ' l I I .,I _l ! ! I f l l i l l l l Ii 4
_4 ' I.. 1 ,,,[ [ 8( ' ' l
.[,, . ! ! } Ct! l kl ll l l.
' l I I h-. ',' I . I _.O. -
. c _c_
' : i l i .:+ . - I .
l i i i !
- _l .
- . - -. - ~ . .. .. - ~ . . ,
I, I.m..gI
- e a. ..me.
.e4 .t l I
. .i _.i._ .g o. '
e } - oI.. i ...__.i ! I i t i. M ! f _. Cd b
.- . _ d C.. .._ A
( - ejyL a__ i _st{&. . . /'6.[ . M ! /_k_/0C. !/$1/ rA.c, Lt }o 4_.WQ l
- 4. _._ L I
. L_w_
i I. i i i i i i i . .i . ! __, l l 4 i L. q.._ . 3. _4. . t I h-. a s Mr .if '
#d i
- I . i i il l l l
_a! l - _1. . . . ._ _ , . _ . . . - 1I
' . l
! l . .
l l ! l l l l
. .. 3... ..
u~u vo ~w~
%g
~
2Gde: ' O jib 8 40 207
- GENERAL COMPUTATION SHEET O Film.
Pr: Ject / Job 7;4(1//9 / / r~ (- - Subject Cos)Cf2t"75 'D 6er.Z GrA} d ff2 EA) 61 T f4
.A, 1 -
Date 2>/8/}t/) ._ Orig. WRev. D BY h3. OF/2EG-rd'f Date 8,/9,/B9. Chk. BY Q'ff f / fir}u - jf l l I I l l ? l
' _.,l t l l ml , -
i lj i i- l i ! I i i : !iiii! I i
! i I
\
! l ! i' ' ! Oc h ! dB.1&_. LA.3dD !D.S'= i! 61TA-.ipsli !
; ; ii ; i...._g.. i i i i i. . : -
i
.. _. i __.: _i i
, i : .ii , i
=.e b g
.I e
ii.
.: .,il ,! i i !
i<: i i i i l i i . &. w.
,. $ . a.
' f '
' ' ' _=
[ ,. ' '
' .l l j Ir ! '.
! % l 3 ' i * .
. l. ! . _ L . . -'$
I
- ff/kdy/ OTL . . . S , b. ' ' '
W ,..no .M-We . h kt. _l..E (00.J Rh fk $fS.E*5 4kE
.. A : .ro o ja m . A i v aco;.. q i a a . L _i i !, ' !
! i n! a _ ., __ : -
- : i i !!
I i I
- hg .
/ M00. - L2 ~E 3 f l f& AS Acr }d.(QLZ) i i i. i -
1 .
". . ~ U Yff h b ff b
, ! l }
4
~
Yl , S B1G r .S~op - 7. 97 x .C.B f-.#. 4/2 9 p ,.N'; i\
,. i
.....a--
- e. 3 .. .
': I i I
$. ./.-._u rf .h l-i .<l I _i 1m _I 1 .
___.x,.q_ff y.
! ! ' ! I I ' !
I 2.dj s OH.rrod z_.t:3f.h.hW = 4 l AB-y.sE !
. . i t i t i
. , , . .,i i I
$.% _h '
$.C 1
' i
; i - i i i :
t
....._.}...._.;_ ..
.domu c ce d,', ._ _ ,_
f
- u. i i _ .. L._ ~ . 4 i i ..
! I I
- i i i.
..e__
{II '
- .__!.__3asoc/' (M I dE.//A !-fd_S.C._. h d-fa.I I iM L _i_t ! i IJ.i i I
L_ - -- ,_. L 4i 1i i i 4 i i i i i i
' s.
i .,a i i St
.. , . -.4 f 8 .{U O b _Y l l$l .. . . ,
a - , ,. -
- 1
- ~fL4 2[4IG.pri.L&. LMi[~"ILJea' E2-
, ~
-3 7.,
.+ --_ .
ll
! ! ! f i
a 1I - i
. . V -
\
i O. -....
. I. f T : i i
!r_ l
\ , _ . . , . .
.cs
% 3
Y d - 4 *f 0 CONTAINMENTINTERNALSCONCRETECYLINDERTESTS T/T[f7 cubit' TEST 1 TEST 2 AVERAGE b ITEM TEST DATE POUR CYL. Mll NO. NO. NO. YAh)5 SHEET D-2 27 6720 6670 -6695 1 70 20-Dec-71 2001 Cl 714 6240~ 6430 D-2 134 6620 2 91 10-Mar-72 2004 Cl 914 6860 6755 D-2 134 6650 3 92 10-Mar-72 2004 Cl 91B 6100 6800 D-2 134 6900 4 92 10-Mar-72 2004 Cl 919 6740 6775 0-2 134 6810 5 92 10-Mar-72 2004 Cl 920 6920 6895 937 D-2 406 6B70 6 94 16-Mar-72 20 M Cl 7150 7165 D-2 406 7220 7 94 16-Mar-72 2004 Cl 939 7220 '7245 D-2 31 7270 B 94 16-Mar-72 2004 Cl 945 6830 6B65 D-2 31 6900 9 95 16-Mar-72 2004 Cl 947 6400 6510 6435 D-2 31 10 95 16-Mar-72 2004 Cl 946 6260 6170 6215 03-Apr-72 2006 Cl 1913 D-2 25
!! 101 22 7490 7320 7405 ;
12 103 11-Apr-72 2003 C1- 1043 D-2 1 D-2 63 6B60 7060 6960 13 108 21-Apr-72 2050 Cl 1099 40 6580 6700 6640 14 309 25-Apr-72 2006 C1 !!06 D-2 7770 04-May-72 2011 C1 1138 D-2 3 7903 7640 15 111 6330 5930 6130 l 09-May-72 2052 C1 1157 D-2 50 16 114 6330. 6760 6545 l 23-May-72 2013 Cl 1222 D-2 10 17 120 8000 7760 7880 I 24-May-72 2014 Cl 1227 D-2 80 IB 120 41 5990 6070 6025 J 19 121 26-May-72 2062 Cl 1241 D-2 D-2 62 7360 6900 7130 l 20 121 26-May-72 2033 CI 1242 D-2 22 6210 6400 6305 21 122 31-May-72 2061 Cl 1248 D-2 26 6870 6740 6805 22 123 02-Jun-72 2014 C1 1261 7020- 6940 1277 D-2 13 6860 23 125 07-Jun-72 2011 Cl 6420 D-2 37 6370 6470 24 124 12-Jun-72 2039 Cl 1269 6620 .6645 D-2 15 6670 25 127 12-Jun-72 2060 Cl 1295 5840 5775 D-2 162 5710 26 127 13-Jun-72 2040 Cl 1300 6500 6510 D-2 162 6440 27 127 13-Jun-72 2040 C1 1301 6730 D-2 140 6630 6830 2B 127 13-Jun-72 2056 Cl 1302 7320 D-2 87 7430 7210 29 133 29-Jun-72 2034 Cl 1368 D-2 100 5910 6140 6025 30 140 19-Jul-72 2035 Cl 1432 6635 ) 1433 D-2 100 6760 6510 31 140 19-Jul-72 2035 C1 7170 17-Aug-72 2036 Cl 1502 D-2 55 7390 6950 32 146 7730 7765 30-Aug-72 2068 Cl 1521 D-2 145 7800 33 14B 7450 7325 34 148 30-Aug 72 206B C1 1524 Il-Sep- 72 2056 Cl 1541 D-2 D-2 !!7 8 7200 6770 7110 6940 - 35 151 7725 } 11-Sep-72 2056 C1 1543 D-2 40 7610 7840 36 151 7460 7285 s 14-Sep-72 2022 Cl 1546 D-2 366 7110 37 151 6390 6310 14 Sep-72 2022 Cl 1548 D-2 48 6230 38 131 7070 14-Sep-72 2022 Cl 1551 D-2 7 6910 7230 39 151 7075 14-Sep-72 2022 Cl 1552 D-2 7 7160 6990 40 151 6340 D-2 9 6400 6280 41 157 13-Oct-722023Cl 1605 6350 D-2 360 6000 6620 42 157 16-Oct-72 2023 Cl 1604 7100 1599 D-2 360 6740 7460 43 156 16-Oct-72 2023 Cl 1600 0-2 360 6690 6760 6725 44 156 16-Oct-72 2023 Cl 12'5 1601 D-2 360 7360 7230 45 156 16-Oct-72 2023 Cl 7550 1667 D-2 36 7460 7640 46 164 13-Nov-72 2025 Cl D-2 76 B190 8290 8240 47 165 15-Nov-722071Cl 1677 6876.06 AVERAEE = 6876.0638 STD. DEV. = 534.70630 i l
\ Td Vio M11 B1 AND B2 CONCRE1E CYLINDER RECORDS Cil. CYL. AVERAGE h: WLf ITEM DATE POUR CYLIN. NO. NO. Mll ND. CUBIC. YARD 5 1 2 M g, gl 1 04-May-72 9524ACW 1136 B1 15 6140 5780 5960 2 05-May-72 B061 CB !!40 B1 60 5540 5380 5460 3 05-May-72 8062 CB 1142 B1 B6 5160 5290 5225
'4 09-May-72 B060 CB 1153 B1 70 4560 5270 5065 5 09-Hay-72 6550 V6 1156 ' B1 - 62 5410 5100 5295 6 10-May-72 B063 CB 1158 B1 80 4950 5060 5005 7 10-May-72 B071 CB 1159 B1. 23 6280 6170 6225 B ll-May-72 7037ATB 1164 B1 4 6000 5700 5B50 9 12-May-72 8070 CB 1168 B1 53 5030 4920 4975 10 16-May-72 4-455 1 1184' B1 10 ' 4990 $220 -5105 B1' 16 5550 5480 5515 72 4-455 1 1192 11 17-May 72 6-655 1195 12 18-May- B1 15 5460 5240 5350 13 19-May-72 6-655 1 1206 B1 14 5340 5240 5290 14 23-May-72 4-4551F 1217 B1 9 4670 4530 4600 69 4943 4780 4860 15 24-May-72 402 FAB 1225 B1 -
61 52 5030 4850 4940 16 01-Jun-72 9046 TB 1254. 17 06-Jun-72 7116 TB 1272 B1 52 6470 6260- 6365 14 4090 4250 4170 1B 07-Jun-72 5-461C1 1276 B1 20 4480 4210 4345 19 OB-Jun-72 7120 TB 1200 B1 14 '5320 5090 5205 20 09-Jun-72 5-461CI 1292 B1 77 5410 5570 5490 21 12-Jun-72 7117 7B 1293 B1 66 5050 5200 5125 22 13-Jun-72 7117 TB 1296 B1 1312 B1 35 4170 4090 4130 23 15-Jun-72 401 FAB 20 5060 5010 5035 24 15-Jun-72 7171 TB 1314 B1 20 4970 4950 4960 2516-Jun-72 4020 FAB 1315 B1 15 4670 4920 4795 26 19-Jun-72 4020 FAB 1319 B1 10 5410 5240 5325 27 21-Jun-72 666 FAB 1331 B1 93 4740 4600 4670 28 22-Jun-72 30205WB 1335 B1 20 5270 5410 5340 29 26-Jun-72 6554 VB 1349 B1 50 4830 4700 4765 30 27-Jun-72 3021SWB 1350 B1 B2 185 4550 4420 4485 31 28-Jun-72 5114 15 1357 63 4390 4510 4450 32 05-Jul-72 6557 Y6 1374 B2 33 05-Jul-72 BBE FAB 1376 B2 111 5060 5090 5075 B2 70 3930 4330 4130 34 11 Jul-72 BBA FAB 1396 140 4440 4640 4540 35 11-Jul-72 BBF FAB 1397 B2 B2 140 4660 4740 4700 36 11-Jul-72 BBF FAB 137B 68 4690' 4920 4805 3711-Jul-72 BBB FAB 1399 B2 15 6000 5590 5795 38 12-Jul-72 4000AA6 1404 B2 64 5570~ 5450 5510 39 13 Jul-72 6559 Y6 1409 B2 B2 90 4300 4480 4390 40 17-Jul-72 BBA FAB 1419
; 60 6120 6010 6065 41 17-Jul-72 B7B FAB 1423 F2
'21 4550 4620 4585 4216-Jul-72 6550 YB 1425 B2 43 20-Jul-72 6602 Y6 1437 B2 27 4350 4350 4350 B2 120 4050 3960 4005 44 21-Jul-72 B2 FAB 1440 45 21-Jul-72 B2 FAB 1442 B2 44 3910. 4020 3965 153 4940 5160 5050 46 28-Jul-72 511515 1454 B2 153 4490 4250 4370 47 26-Jul-72 5115 IS 1455 B2 B2 47 4700 4560 4630 48 31 Jul-72 6550AYB 1457 49 07-Aug-72 891 FAB 1477 B2 140 5570 5630 5600 50 07-Aug-72 691 FAB 1478 B2 140 5320 5500 5410 51 11-Aug-72 895 FAB 1487 B2 143 4920 5150 5035 52 11-Aug-72 B96 FAB 14BB B2 143 6010 5940 5975 53 14-Aug-72 3772 FAB 1494- B2 10 6990 6650 6820 54 15-Aug-72 89E FAB 1496 B2 19 5200 5090 5145 55 17-Aug-72 B9E FAB 1499 B2 40 4370 4620 4495 56 10-Aug-72 5156 IS 1504 B2 28 5500 5590 5545 72 4002ABB 1507 B2 47 5480 5340 5410 57 21-Aug 72 6702 YB1512 B2 64 6400 6210 6305 50 24-Aug 72 6710 TB1520 59 29-Aug- B2 27 5B20 5890 5B55 60 30-Aug-72 515B 15 1525' B2 33 4640 4790 4715 AVERASE CYLINDER TEST STREN6TH 5094.17 STANDARC DEVIATION 624.61
Trojan Nuclear Plant i-[ .U 3 Document Control Desk
'.Ni i lTlf,*',( ',,;3 Docket 50-344 h[ ' , s./.,.f ,3,5N'...E ~$t U,yX fi tkSf ']' Ni(2'((h[*
- jh,'
- . August 18, 1987 f6 1."'**W 4'4 W . p ; Attachment B4 License NPF-1 MM 'a
'02 - .T
,P
'- : , '. t ( w.:.h..:,$'. M Page 1 of 2 c.
y 1
- a. j. .j,,es, ,
}," l , .,
' ' ~
i '$ . . r ..: I N' > s._ u : y,p:5.:.y q.:- -
, :..'3;. ' j y w.:
l ma/.:@m 8e u :.E. i ,;z g:;I, I !;',;yA,_T 16 9 m r 3 Lxnmm@w.E!ItiNtG
- 6TiG!m wwwv.;agmyp%m:qawx W.n w .n . :m. ... s Yk .
[ f_' 'g'??5pfgg W.fgjJ.jVr*i
^
g a$na , 3..ki i
; >. ' v4 ;,l /&.
- f. f., ..Q-QQ-
. .yty . .
Ky
, uake,; a. . w .v e , s. 9.;g "
I.. $ 9 ) CM$f.NU-c 5" ,, y r .i .eji
'f
,J. , e ,i - i 'j ,
F ,,l}. , (~g ~
- E
*i1 iga 1
- m
Trojan Nuclear Plant . m Docket 50-344 . . Document Control Desk d License NPF-1 ., August 18, 1987 Attachment B4 l
< q y. . p. , <
"' " " " m M -
Page 2 of 2
~, .
" " ?-
of ,,: ,
,f .
-0
("[ "w X IOutmg SySl6m
- Q _ J _ _ JL _ _ a C0mp'I6tB:
'q %
1 2 WIL-KWIK-MT '" -
' WIL-X-CEMENT-GR A fast setting sement with on Inittel est t!ms from 3 to 8 minutes. WILX Cenem orout is e ehemloolly sorwermeted sw pte primary use is to held the deelt tube in piece and essi off the soment recommended for grouting Wtillems sh',s type sustdo entrance no the delli hole around the Williams Hollow Corv ,, Hollow Core Grouteble Rock Seite.
Acck Boel At the some time e levefine pod is built to emure . . . I proper beoring e<ee of the p(ete vuoihor. A mixture of t , s ! I. WS.X le chemicalfy compensated for shrinkees. It has a y r, part WJ4pek $st to 1 part ester (by vielght) will ,, Ngh bond Wlue end is erects reelstent for permanent
. produce especatmetely the following P.S.I. m- * , instettations and more durebf a grout. Secause it le a poemre etseng'hs ertth ett and mixtarre et 70 . . m v somem trout, it is norHgioelve and has a long Lemer ausgeretures evill lower comortesive ,,
- ~
ehelf life ashen beqrt dry., ,
, prong'.hvaskistfy for periods shown: l rc h
., 't 16 Muu s,600 1 Hr. 7,600 Averose senditione'reculk approximately 3 sr 20 Mwn; 4.800 A Hrs. 7.200
- 416 gallone of gueter per 94 lb. sack of 46 tan. . 300 5 Hrs. 8,000 l . tyll-X. TMs mirture indil produas ep-WIL*LCEM E NT O R OUT e GROUT PROVIDES p the W ng .8 .
$ des, WB KMk r g g g g ;,,4,e,com. , ,g g g, ,
_ ADDITIONAL STAINOTH ANo r mote CTION W h re m h> p cond pr. tie bo6es of seiking 4*ound PLA$ TIC OROUT i peratures esill lower normroedve p%3*e in 8leusenry or concreta vuell yygg yiT$ THRU e ./ A0 AW8Y CORMm, strengths in% for pwlods shown-o and Aoors ^9 PALLS HotE To REcloN ggygogg , goy ,y y ,gggpg-
.h ,'.bearpoweran.
On Ives se wu.navik4et should uesd egliers heeW machinery or projects
,ay,,,such ,,,,,, s A:n or ANcMon PROTECTING COMPLETE .2 esy storgeh - 4200 P.E.1. ,. '. fhalecerters are required end in 4 net ALL AT40N 8"O**8*'I' * **' 'ANosiNosANcHon 3 doy stroneth - 8100 P.E.I.
onouTPeo
.t -_
THMUTust AND mock TOGRTHER. AT MTC y >
, ao p.3,3, W1L KWIK et7 Y
?. .
~
PAE34vmg AT
/ OTEEL
,d ((sEsn OPTIONAL HAND PUMP
& l r. .c n'. FOR 5 MALL QUANTITY
%s N ,,,y,g INSTALLATION Y
- W AS M t PS t- opcmien: Head pe ered. e nr.ke ,.shi n
,? ' pump.
b 40 PAI. overage,30 PAI.mesknuns cc, on ,1 wie, se,ne ei, ps 6,ed er. ,i -en,. J, , q / Osaning: Piush samver throuch hopper and pump. , 1 Appros. $ise. f' O High, t't" oride, t 0" hande, Weight: 30 les Dry suetebt names Ia Other Features: seessmmended for use in 5 to 10 esf.
! containers Ino erese6.no 2Cf' He shtl. Omry he64
. cienning et interned parts Requires no tutirication.
, tMil t i Ana$ Approx. 8 M. noneth of grout hoss. A.eastWe en
" ~*
AIR POWERED
.GROUTPUMP -
g hetion: Air Drhen, terow feed Grout Pump. Cloening: Plush geneross cuantify of esotor through happer L. <$. Cepecity: 10 Gel. Hopper i
- f. Epose Outtet: 3 to 4 gel. por minute et 0 to 130 P.R.I.
and pump.
. . Approx. Blas: l' 3" long,2'4" high, l'.10" wise
- ftmas IIIsosWoments NeIght! IllIbe. Ofy WoI9ht
.b , i ,
er Akteuser Inlet: 90 to 150 P.8.I. er 130 eu, ft. of ele per min. to Other Poetures: Rectroutetor and hopper Apsetor for keeping mia .
,4 V eevelop 3 H.P. In solution.
w A!r sne futer and efter for air motor habrimetion
/A'.'y1/ " .Ae7 EO e*.*ATo";M. PNa Aa." ' ' " '"*th
*.wovon No. oD e
o, e+emwore . mht - m
En - '
- I Trojan Nuclear plant- Document Control Desk.
' Docket 50-344 l August' 18,'1987' LLicense NpF-1 Attachment B5 page 1 of 11.
ROCK BOLT TABLE DEVELOPMENT.
- 1. Intro (uetion Table B5-1 displays'the capacity, allowable loads, linear inter-action ratios and safety factors for 1-inch diameter,'8-inch nominal embedment depth rock bolts used to anchor: safety-related pipe _sup-ports'to' concrete. This attachment describes the_information in Table B5-1.
- 2. Maximum Load Identification <
l The data. presented in Table BS-1 portains.to the most' highly loaded l bolt in each support. Calculations were made to' determine the demand tension and shear loads for'the maximum bolt load in each support. Where the rock bolts for each of the supports have different embed-ment depths, the minimum embedment depth was assumed. For example, in a support confi6uration using four rock bolts with three at - 8.6-inch and one at 6.2-inch embedmont depths,1the one embedded f 6.2 inches was used for the worst-case loads. The procedure for ; selecting the maximum _ bolt loads is as follows:
+ For pipe support baseplates subject to direct tension only, all of the bolts are equally loaded. 1
+ For pipe support baseplates subject to direct shear only, all of the bolts are equally loaded, l
j
+ For baseplates subject to moments (as.well as tension and shear),
the maximum bolt load occurs on the bolt located furthest from the neutral axis on the tension side.
+ The conditions described above led to identification of the maxi-l mum load for nearly all of the supports. However, in cases where j the support contained numerous baseplates, each of which.had.dif-l ferent loads, an enveloping approach was used to determine the
! maximum bolt loads. This enveloping approach conservatively I assumed that the bolt with the maximum tension load is the same bolt in Which the maxinum shear occurs. . l I
- 3. Rock Bolt Embedment Lengths :
1 For all accessible pipe supports included under the present verifi-cation program, embedment measurements have been obtained for-rock bolts anchoring these supports. The total lengths were determined j for 609 rock bolts using ultrasonic methods. The embedment lengths I were calculated'by deducting the projections of. the bolts outside the ! concrete surface from the total lengths. k i
Trojan Nucicar plant Document Control Desk Docket 50-344 August 18, 1987 License NpF-1 Attachment B5 page 2 of 11 In order to allow for measurement error, embedmont lengths for measured bolts were determined by deducting an estimated measurement error for the embedment length from the calculated embedment lengths. The estimated error was determined by comparing the differences in measurements on 41 rock bolts. These rock bolts were measured inde-pendently by two separate crews using two separate ultrasonic length measuring instruments. The mean value of the measurement differences was found to be 0.16 inch and this was used as the estimated measure-ment error. The embedment lengths with the subtracted measurement error were used to determine the tension capacities of the embedment. In order to provide a basis for determining the embedment lengths of , inaccessibic rock bolts, a statistical evaluation was performed using the rock-bolt embedment length data measured from accessible supports. The measured embedment lengths for the 609 rock bolts tested were plotted in the histogram in Figure B5-1. The rarmal probability distribution which fits these data has a mean corrected embedment length (x) of 7.79 inches with a standard deviation (s) of 0.70 inch. A Chi-squared test was performed on the data and resulted in a value of 10.19. Since this value does not exceed the Chi-squared value at the 0.05 level of significance of 15.507 for this data distribution, the nonmal distribution provides a good fit. With the above statistical model, the embedment lengths can be estimated for inaccessible rock bolts. There is a 95.5 percent probability that the rock bolts in the inaccessible category have embedment lengths greater than i -2s, which corresponds to an embedment length of 6.39 inches. This was conservatively rounded down to 6.1 inches which is less than all the measured lengths in Table B5-1. This corresponds to 2.3 standard deviations from the mean. Thus there is a 98.3 percent probability that the embedmont i length of the rock bolts in the inaccessible category are greater j than 6.1 inches. I
- 4. Table Legend l The columns of Table B5-1 are explained as follows.
Column A " Support Number" This column lists the support designations of safety-related pipe supports which have 1-inch diameter rock bolts with the minimum nom-inal 8-inch embooment. The table includes only those supports with rock bolts. The remainder of the supports included in the verifi-cation program used other types of anchorages or longer rock bolts.
l I Trojan Nuclear plant Document Control Desk Docket 50-344 August 18, 1987 License NpF-1 Attachment B5 page 3 of 11 l Column B " Isometric Number" 3 l l l This column lists the isometric drawing number associated with the ! support number mentioned in Column A. Column C " Number of Rock Bolts in Support" l This column lists the total number of rock bolts utilized to anchor I l the pipe support mentioned in Column A. Table B5-1 presents data for j the maximum rock bolt load for each support { J Column D "Embedment Depth" I This column lists the minimum embedmont depth of rock bolts from the embedmont depths determined for all rock bolts listed in Column C for 1 all accessible supports. For inaccessible supports, the embedmont f ! length is listed in Column D and is assumed to be 6.1 inches as discussed in paragraph 3 above. Column E " Allowable Tension Ta (kips)" Allowable tension values are calculated based on the embedment depth of Column D and the methodology discussed in Section C.3.2 of Attachment B. Column F "C/C Distance" l This tabic provides the minimum spacing of the rock bolts for the l support listed under Column A. j Column G " Tension Reduction Factor" This table accounts for the possible effect of overlapping pullout cones based on the spacing of Column F and the effective embedment calculated to determine the allowable tension Ta for Column E. i l Column H " Allowable Shear" (Va) i I The allowable shear values (Va) are independent of embedment lengths ) and are in accordance with Section C.3.2. ,
)
i
i f ' Trojan Nuclear Plant Document Control Desk l J Docket 50-344 August 18, 1987 License NPF-1 Attachment B5 Page 4 of 11 < OBE LOAD CASE
~
Columns I and J " Tension Demand and Shear Demand" 1 Columns I and J contain the maximum tension and shear values required to j be resisted by any rock bolt from the total number of rock bolts listed in l Column C. These values are for load combinations that contain OBE as one j 4 of the loads. Column K " Interaction Ratio" j Tension and shear interactions are combined linearly to calculate the l interaction values of this column. SSE LOAD CASE Columns L and M " Tension Demand and Shear Demand" ; Columns L and M contain the maximum tension and shear values required to be resisted by any rock bolt from the total number of rock bolts listed in Column C. These values are for load combinations that contain SSE as one , J of the loads. Column N " Interaction Ratio" Tension and shear interactions are combined linearly to calculate the interaction values of this column. 2 Column O " Tension Factor of Safety for SSE Loads" Tension factor of safety for SSE is calculated as follows: Column 0 = Ultimate Capacity in Tension SSE Tension Demand { l i i ! I i 1 I t l 1987W 1 l l
Trojan Nuclear Plant Domment Connol Desk
- 4. Docket 50-344 August.18, 1987 License NPF-1 Attachment B5 Page 5 of 11 ED E ***
~
1*5 *19 9* *** oo9 11 on llHg -a-
~:=-e ggna s"gg ens. =xx: -n. ess e9o 911
=a as-o=
a-AA
^5 1 : CR 888 885 888 888 588 08 23; RO: as BR g$0g bbb b b bbb bbb bbb bbb bb bbb bbb bb bb t
l l ^
- SS. qqq ggg ooo gqq qqq oyg qq gyy gge b, y yg i --- Noo ooo ooo coo oo Mom ooo oo oN kg l
i I V V fE3 lgg, w So* ooo gee 11o odd ooo
*"o ooo
""1 ooo 1 *. 5 coo 1*
no
**o coo
**5 enn "1
-o 1"
c.
^5 .. ... ... ... ... ... .. .. ..
l Ih $38 ooo Ek! ooo kkE ooo kEE ooo EEk ooo N$$ ooo 5k' oo ooo
$95 ooo
$k oo oc Ek l- _
e S $ hag? ^ 5 ' t. o-e 1"* noo
- o coo
*d" oco 1*1 ooo 9'*
ooo "a oc "99 new oo9 ooo
*1 ao
. 15 on ,gg3 I-gh 335: oS 5- *** oao 5 W ~i; *
- 5 45 E n t EA-v -we 11o noo o11 ooo oco 111 ooo 1 *.
- coo 19 no coo
.ou a nn 9o a~
-o "1
c. W. , r l E 2 -s D e a W@b __vv _____________________________________________________ 3Ea,g"Df C u - 3 e5 $8 E E gg *. a. t **9 a. a.
- 91*
- a. *. .99 ** t o. t . o.*. o. a.
- EA6
* ?
-ash.P15]
s" a a a saa saa ses ses ass as saa s" e sa"s"as 9 A
! 88$ 882 8E8 888 888 88S F: $88 872 8F $8 s N $ Abb bab Abd AAA AAA AAb bb bAA Abb Ab bA 2 C e' 561
" stg .
SSR 888 888 888 888 888 88 888 R88 88 88 3 { S,gh:.. ... ... ... ..: ..- .. g3 :.. :- . gg]g ___________________c____________________________________ -g b, g a 2: 2 5, , Essy ; a, --s se1 en u- ete e- es sea n~- e te Igg - C o== ong ngo men >>= >>e e. Ogg === on oo f64 ' ~ 5 I Eje *": *s8 sRo ass as, asR *a ::S UES se as Ide* 0 Ja; Jis isi 444 iis Jia is sii sai is di 1l$ 3 EwS* 5 5 y g' i l i b Sg b b.
.o. ... ... .. .~. ~., og ... .-- .. .. ,; ;g$
nl5-l ,g =-Tm 3y ;. ;
-{
;;; y t
n 99 7t- s t fless t g "22 220o qs2 633 SqAa -- [6 sg le llll a s**ss"s o BBEIglB. oo t 6 t} 58 o, AA
~~
117 888 888
???
--- gg gs 64 66 23* as8 sa= ==a =sg gg I 6:666 t
- bss ::: ::: ::
666 &&& AAA &&&
- - ~~
- ~8 sin s"ii 9~ i II I"I
L Trojan ~ Nuclear P5at[t Document Control Desk l Docket 50-344 August 18, 1987 l
)
License NPF-1 Attachment B5 8 Page 6 of 11 )
- a,a a,a $55 9"* ". a 9 *,a Ma* *95 t o.
- 99* *19 19e E
N $h m3 age *u 3*u Rau-883 23# -n sSC 032 30" 222 "*' I
^5 o 588 GBR 8MN BEu 352 88s $58 388 888 383 880 3#u lh:[h
^
das did did dds did did did did ddd did did dds
$ $^ 1*5 *"5 1 *. *. *** *** ""9 9*9 "9" *99 "*9 * * ""
dI *** *** *** *** ** *** ** *** ** ** ** ""* l I i l8_ a 7$ oo o,,
- 9,,
non o.- oo-s-, oo-n,, ooo
,,n ooo ooo ooo ooo
,,9 o--
,,a Sn. .j vv w
}
, J
^5 .. ..
1 aas saa aan a8s aaa saa a8 ans I Iifeansesas-=aa8 4 Ew ooo ooo oco ooo ooo ooo Coo ooo ooo oco ooo ooo
^
g ! 5^ n 2
$ u? 1*"
~
ooo 91a ooo 151 co-
**o ooo 99.
ooo 19" ooo co* ooo 1 ooo
- ooo
*9 "*9' ooo *o*
o o e-o ]-
.* 21
.I l
I 235: e2 ) s_ aaa' aa, "", s s :y
$a r
gg 1 " La l$7v 915 coo 15, non 991 on-S t e. ooo ate ooo Maa 191 ooo ooo ooo ooo ooo 11* o-- 95"
-n.
N y *i g vvF_ J- -ga; y a
, g;
^
o O MND hbh Mbb PMb bhb WPM MMP bhb MbM PhD PPM hhh v sss dds sss dds sss sss sds sss dds sss sss sss g g j 3] "j
$$$ ,a o
_-u _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ . - - - _ _ - . _ - _ _ _ _ . _ _ _ _ _ - - - _ _ _ - _ _ - _ _ _ _ _
$s$ ,
g 888 88E FRR 888 288 888 888 S88 888 888 888 898 2d I b bbb b Y Y$ N bN Y Y b Y ,
. a 57t M
E g 888 88s 888 888 s88 888 888 888 888 888 888 888 1532. g Y MN MN Mh eMM M MN NMM M h MM ' I ____-______________________________________________-_-_ gj g 5L
.5
*** *,* *** *** *"* *** *** *** **" ,*n 1
~ubwv Mbb
**1 Mbo -vo dem bbM Meo.
>>m gen
. 9**
>>M mbb MMC o-o B
h, E4 ob - - --. I_ Q~ ese s e e@ aae ese ses see see ase ses ses see eae _-__~____________--_--_-_______________________________b s_eh g-s Ih.
$l9 .~. .n. n.. n.~ nn~ n.. m.m ... ... ... ... ...
]gge.
In er
-- d 97Y ?TT ? I 17f 6II 6FF T
6 f I YY I?? f?? f77 Y M .. E Et. ' 200 una u?C 22* 2A2 An2 22 200 2%C 022 m?? ??? j gg) j E t= $ s&s sss sjs sss sss sss sss sss sss hna as hss ssa sss sss sas sjj s jjj hsann habu s}s
~ n nun un h8s un nnn nnn kasnn n 8 l ksk ins sai 666 666 666 666 6kk sks sik ka}3ska]
su cem neu ex: muu ::: aga sus ss: 88 unh un:u nun au=u ac una anu uun uun una nu una unu
.l- 9 AAA 666 Ass 666 A&& A&& AAA 666 A&& A6A &&& 666 i
1
Docicet5bMA[ License NPF-1 Augu E 18 8[ Attachment B5 Page 7 of 11 B
,, es, de, n-~ -e1 1s- ,-1 en, s,e e-e ese -e- sn, l 8 g
g s0 c'" *:N R* =2' ='M *** 22- **E 222 :$ R2* AA AA
=_=_-___________-_______-____
^5 888 :2 228 080 5:2 2RS 8 282 2:8 Bs8 088 8:2 llo:$
A L add i add add did add did add add add add add ada . k 931 ooo $99 $99 ghyk ooo ooo SOS ooo ooo 1 . o, ooo g o, q o-- owq co-
, g o, g
-oo qqq ooo e, v, g ooo qqq ooo ,
{ 1
$_5 1 1 9 o*9 5*"
- 9 *. "'" 1"" 1"* 999 * * *C 5 ". " o*
I j g$ oo -en n-o -on o-a -no onn -on n-o ooo 1 *.1 "--n
".1 I vv _ -oo B
l ^I2$994@$9994$95E868944E5885699 8 bf ooo ooo ooo ooo coo ooo ooo coo ooo ooo @8E 8@9 ooo ooo ( I c Q f_ 5 ^ **o 6 5 $ 1 g)_?ooo o. o 1 Soa act ott ooo 1** a*9 oa* oco ooo ooo ooo ooo ooo tot ooo 119 ooo ana ooo een ]4 ooo ,g g
=h E a 5:goa
-E ego 5'55r
- 1 Ea coo 759 on- 99" *o9 a. a q 9t= "95 99* -55 511 999 ooo Egyg t
vvp_lE7v --o co- oon cuo o-- oo- --o ooo 9*1 o-- to9 ; g _ 2 57sB 3-8 ____________________________________________________ gg Eh I 1R'g{of , hh __S
,5- ih
*. a. a.
222
- a. a.
222 222 222 222 222 222
. .. t *. 9 9 t *.
- a. t aqq qat 22= 222 222
- a. q *.
22 22
- a. *
-8323
*6 B 2 9 2 5ItI-
__ A _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ h 338 $88 R*S 888 S$8 888 888 888 888 88E
$ adA daA did aAA daA AAA AdA Add AAA 888 828 e4 e lE AAd AAA did 5 E . ,s l .
~
5 I l I 888 888 888 888 888 888 888 888 888 888 5La ggji. l v lb o_ hvc sii ssi sii sie sig iis sii sai sai is; si; iss 888 8ss 64 w 5E
- 5yl8 m g
.5 o e-- ene -se ene -ee e .e ete een see ens e .e -se -as1 ggn ego ggg onn ggs >>n -go og- >>o ang son gge 04 jg3 EIp 2:2 328 222 232 222 022 2R lEj BCE ses ses ses ss; ess ss; ses sss 222 22; sse sse ss; s$s 222 es E
.G]obgi
_____~______.___________.____..________________.____________ , b s p;E M . nne een ene wee weo eon ag a goe nee -nn ewe wnn wee y .$5l 6 85 -6_7 8 4 7, It - MM b b MN MM -
==- =44 II: : n e 5 s
=== ::=
y e :: 22: e v
%$s I6 Ass ses Ass sas nahh hhh nnu s:e s
hhna sse== saa ss nnu Ass un ses Ass saa hns hhna nhnaa sls e a x2, asi hsa as un ils+ un sua
=- =s ==m===
ass s === s:s 2: See === ses ::: s*: sag I === === === === === === =a == s :- s i l3=666A&6&&&AAA a 466 &&& A66 66A &&A his 6&& dsA
Docket 50-344 August 18, 1987 License NPF-1 Attachment B5 Page 8 of 11 BD _ ,, ,,, 9, ,,, ,9, ,,9 ,,, ,,, ,9, ,,, -,, 99-E ama E R.v 2;$ 02" 0"* "8* 'AN 28% -
*2R =3" ** 024 d
A A O 238 GNS 554 33M 382 288 BES ESR $53 DES R:: 583 hkh das id6 did did did did add sad did ddd did did A:
$ ' $^ ""o ooo ooo 9""
ooo ooo "1" ooo 99 ooo
"*5 ooo no-o""
ooo o", ,ao oo- ooo odo ooo igf een on -no ..n .n. own -ne ene een me om. -me lN@f vv$g* - h Add GAi . s i n' Aid isA AdA AdA JAs Add g Add JAA idd , l
's5 ... ...
I & _fhkkk ooo
$$$ sk$
ooo oop $$$ ooo kk$ ooo
$8k ooo kkN ooo kkk ooo 8kk ooo
- 8. k k ooo kkk ooo kkk oco
^
b
% 5^ "19 *"* *** **" $9" 599 a * '9" #11 91 *. ""9 919 g $gg l7_7 ooo ooo. ooo ooo ooo ooo ooo -oo ooo ooo ooo ooo l * $, -
E. E Bg$A A
~
5:l l
-.o2 .
I - I" **1 1** 99" ***. ""* **" ""* "** "9" =aa x=* 1aa 5 * :5 - gg, coo con g lg7v n-m on- no- no- -o- non -oo oo. -oo ooo - vv w,c g _______________________________________________________ - g t 3 g[ PEM . Ew E gg Og
- a. t
. **t
*** . t*t t o. a. **9 .. 9 *.
- the t o. a. **t St* 6. a t
. fe5ye, A 222 2*2 222 222 222 222 222 2
- 222 222 222 222 - *sgh G g 26sb A
SE8 888 858 888 SE8 BEE K$2 888 $88 88$ $88 883 3 8 9 AsA AdA AdA AAA AdA Add add AAA &AA AAd iAA Aad c 2 QFh a, E 57 5*24 O lcj&v .k.88 . .k.!.8 .k.!.k .k.k.k .k.kk 8. k ! 35S5
- .k .k .8.k .8
. k . .k8k . .k.k.k .k.k.8 .k.8k ,
se : __"___~_____________________________________..____________ 5g'l A b. g BE *** **" " ""* *"* * * *** "** Esse 2_0 . *1* e ."o * .gr og
**1 M** **9 jfgm n=e oe. .
*n ==g ... o* **g -er nog 3E 282 :$R 280 SRS RED 235 $38 0:2 822 222 22 228 5 I2hfe
- did JJJ isi did did iss sii did did iis sii Jii E
gggg h$g wS 85 8 .~ ... ... ... ... ~~~ ... g., ... ,,. ... ..n }$5F M h 8r .-E QB **iu b*2. %" i B
@ @@@ @@ --- - --N NMM MM.
YYY YYY Y@tf fff --n ff? ??? "4,? ?, AAA A
--- AAA --
AAA AAA
*6 *5 h $$$ $$$ $$$ $$$ $$$ hhk 555 $$$ $d8 $$k $$$ $$$
??? ((S SS
? S[{
SSR "R$ S h8 SSS
??. [SS S
[?? SS@ .?? Q.R.S ?.. ((2 ..? $.88
??
EGE EGE EEE GEE GEE k""h" k hkh kka k k" EEE GEE MEE gas 388 s:: 2:2 en: *2 m2: ::s *:x me: :s ::: l$ :: :: ___ ___!!!_@@s sss sss $$$ ses ses m
- ::= :::
i 30: :: ::: Ass
- := ::: :::
$ss $66 666 666, '
Document Control Desk
-Trojan' Nuclear Plant Docket ~50-344' . August 18' 1987 License.NPF-1: Attacimient B5 Page 9 of 11 i 5 9 8
~
$g $*21 N2" 11 *. "** 1"* "** * * * *1"
. 9"5 *9" **q *"* 9 *. * ***
N "R0- 2*OR- ESC -ERR 0 30 Egg 322 nM, car -Rga I a
^5 3: 885 SSE 235 ss s ESS E88 588 did did add add did did did add sss Ess sss 530 l=h A
[ did add did did ll 2g S
~
, "oo o
Oi 009 ooo
""O
- 1. o* -oo n
91* oo o "S* coo 115 ooo 91* ooo 1* ooo
"$1 coo "9*
ooo 10 ooo 1 l t l gg I *os e-e onn ena a99 ec-151 coo 1* 1 oo o 19* coo 191 "a9 co- ""1 51* 111 155 ooo ooo ooo ooo ooe V V _^_5 l g b 802 898
'5
-f _59 ooo EE~ EE* EEE EE8 988 EE8 8 8 E8 888 EEE fB.5I l 5
^
ooo ooo ooo ooo ooo coo ooo ooo ooo ooo ooo , a g: ox gb {
^ 5
- 35
- 2) "OT 099 Nwo *11 910 f 090 """ 1"1 000 g ooo Goo ooo ooo ooo ooo ooo 100 ooo " O.1 "O9 C"O #
Doo ooo ooo j ggf g E e5
. *Ej j
- .o, --n no- en- -n- ne. -,. ~~e k lYga iAA das siA Add dda did did did did add did &&A --- --n -oo -n, sst ygf ]
- b. = vvE___________________________________________________
g ]g ____fg l t e a. *tt *99 **t a. t
- t o. a. t o. a. 999 the t e a, t e a. 999 g a 222 222 222 222 222 22 222 222 222 222 222 222 ij l
i 9 ____9___________________________________________________ .c at a E
$83 888 SE8 388 988 888 883 888 E88 BSE E88 888 5 ?l g ! g$ dad AAd adA aAA isA AAA Add AAA &AA T*eg '
g did &AA AAA e g 3333 l 5 I ag:l
]sA 1 888 88S 888 888 S88 888 888 888 888 888 888 888 bfg ,
fii sii sii iff idi sii sii das iis sii dif fia hffj 1*O *** *** 24 *1*- 9 * *. *1*- -- **" *** 1** Maa **a *** 155 j -v -e .-- ' og -- - g=g g --~ E i i' 5 i9'~0 3 999 @@E $89 E99 988 $$$ 985 $E9 882 82E
-.- ... .. -.. .. ... .-- ... -.. -. .E.@8992 ... sIjl~ ,
_______________________________________________________ gW% ea . a 5 b n .n. ... ..n ... -.. ... ... ... .nu ... nn. .., }3 $ct$5 i V aa S E.E u acts) YYY Y-- yyy e 6 gjj 22 rev 6"6"" 4 "nnf7x22 x-A sll
-nu
- ~:~:~ !!! !! 82 5
*t fjj 666 sse Ass lli lil lil &&s ses iss
}}} }}
MMM MM ESS nn hREun bbb FRl n8hnS nEbb Enh nu khh bbh bhh hkh kkk hhh hhh h$s hNb ~h$$ 8 8 nh8 y huh 8 SSS~n t ggo -nn -.N n e - . gum I g co8 333 --- EES --- EE8 --- -EBISE $B5
- $ne o88 ~~~
Ee$ 8 BII ---
$~8
~ n i
} . - - - - _ - _ - - _ &&& _&&6A66 AAAAAA AA6 AAA AAA Ash isi 666,664
i Trojan Nuclear Plant Document Control Desk l Docket 50-344 August 18, 1987 { License NPF-1 Attachment B5 i Page 10 of 11 )
- *** *** 9 ' *. *a1 Te1 1** ea* 99* a** 9** Ca1 5p 2*" ""9 l A 22" "S* 0:3 - %2R -s2m - $33 : 4 sas -- 2#2 N2 A
l ______________________________________..________________ 1
^5 2 s Q $22 883 R82 353 828 528 888 *22 558 588 822 l
1g~l$ did add add did did did did did did did did did , 1 i
^
,_ .on no oon .-o ... ooo o-o oo. ... oo. ooo o.o eg$)
I lE7 o.,
"1" wo-
*1*
ooo 99" 15"
--o one ooo n--
1"" ooo 1"9 ooo 9'o o-- wvM_ i l Ie5
- ... ... ... .. 1
@ ~K_ ooo coo oco Coo ooo ooo ooo ooo ooo ooo ooo e 5 oo.
h
^
_ *2E
- --- --- co- von 335: oS glgagg0idA i
-o- no- o-- -n- ...
Add daA sad ooo o-o con p t= .s b add ida did did ddd add add did - s
.qy*j 1
\
-$ v g, 3 g"m 5_ *Ej%t!
$g R E e s ,-
o Un-
,s ,- ~, --, s~e eco
~ne ooo see 19, ,~n 999 e,~ neo :-s a2 y vv 7v o-o o-o coo -oo ooo ooo ooo gg .
_gy : . l
$p @
M ____=_________________________________________________ w _ 25 e , E h ".".". ". ". 9 9 ". ". ""9 *"" 999 h 222 222 222 222 222 222 9ese 9 ". U ". D 999 999 " 9 ". "" 8 e . ... 3# 222 222 222 222 2 2 ,"
^ ,
! KK8 *** R28 E38 885 SS8 SER ERS RRR 838 888 888 24 3 )
9 dia- did dia dia aad dia add did das E as aad aaa .~ )? i Se re ir m
^ 5*og i
l$ 2g I 888 sss 888 888 888 888 888 888 888 888 888 888 II2j v dds ses 4de 444 844 444 dde 444 444 444 444 444 v 2 j'<gE
- 7 gg
_ ] 3E l gjjg g 2c
~* :,:*, e9, e, ese ees see ese --9 -~, ene *ee ene 22* 22" ""2 ::" "N2 22" *22 "2" """ "*"
2 5 l s]_- ess seg ea; ses ess es; sde ses ses ess ;;s ;;s M - -M _ s]i gg3g e esg egg one e,n ,,, nne -,, .e egg nnn ne. en. ySSE y-
,laIq e
- _ri! .
. b h
- --- --- -- sss --y t 4
AAA AAA e6 AAA AA-2-- Yf Yf. f )
--- f .f? --
5 0 ? ? ??? T?? 77$ $$$ $$$ $$$ $$$ $$$ $ $$8 $$b
= o= nn: n= nasn a0s ~~
sa nn sas
~~~ saa bhsu nun Iss$
un ss nbu hunas M - gg2a :n=
.. 3, --, .,~ -
8gg 22 2 - 22: 4 222 222 223 22- 222 2== 22 222
- i4 kii 6 lii ils! 666 &&& 666 666 &&& Ass sah 66
______..___________________________________________-- 6_-
j mssere6r mrsmets nenam yn3ns3 I8' I684- ^ ^ ^ - l dooMr3 50-C71 yaa'ecynteua gg i Iposusa Nd3-I de3a II oJ II i f 5 01 1 S / .C I 52.O/ 0 0/ x? c a 579 S .O ,', l 09 31 l 1 l 03 / 52 9 Q i 9 / .9 r I I
/ 09 !
4.G I i 9 57,g y I5/ S .g 94 8 1 2 '7,l
/ 52 8 R G AI a
I. / / 08 iC [ p7,7 e y 3 '#/ 's 57 7 N [ n 67 k 9 N s 52 7 g c ,y 07 6.II ~ 9 5 's ' 9 57 6 1 N 03% %. ni \ % S.G 99 6 6./ _s
\ 52 6 1 70
- 1. O .G i
1 57 5 i I I I I I I I I l l i l i I I I I I I I O O 9 b N
)eV( YCNEUQERF l
7
]
Trojan Nucicar Plant Document Control Desk j Docket 50-344 August.18, 1987 License NPF-1 Attachment C Page 1 of 29 l Design Verification of Support Welds During the pipe support design verification program, welds joining struc-tural elements of the supports were verified for acceptability based on i current pipe stress loads for the applicable support structure. The welds were identified from drawings, procurement specifications and in a few instances were obtained from field walkdowns where weld details were not apparent on the drawing. The following describes how the welds were verified and how the supports and their welds were identified. Safety-Related Pipe Snubbers and Restraints A calculation assessing the adequacy of 265 cafety-related pipe snubbers and restraint anchorages, including the welded connection evaluations, was generated. This calculation, J-307, contains the following:
+ Cover sheet indicating the purpose for the calculation, source of design loads and the referenced design criteria and drawings.
An introductory section providing, among other things, the evalu-ation methods. The calculation indicates the typical supports, Types I through V, were evaluated by numeric calculations. Sup-ports that were similar to the generic ones, but differ in some ways were evaluated-by indicating the differences and determining their adequacy by either calculation or judgment depending on the extent of the variation. Supports that were not similar to the generic types were evaluated by engineering judgment if the con-figuration was simple and loads were small, typically less than 1 kip, and well below the capacity of the support. For complex configurations or substantial loads, calculations were performed for the structural elements, including wolds, if the stress levels were judged to be significant compared to the capacity of the structural element.
- Detailed numerical analysis of each case for Types I, II and III supports including calculations for the structural members, welds, base plate and anchor bolts.
I
- Reference to a separate calculation, H-19, for the evaluation of Type IV supports.
l Rationale for concluding that Type V supports were acceptable by l comparison to Typo IV supports. l
f I Trojan Nuclear plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C page 2 of 29 l i
+ Specific evaluations for each non-typical support. The extent of the evaluation varies as stated above depending on the magnitude of l
the load, the complexity of the configuration and the similarity to one of the typical support types. In this manner welded connections for safety-related pipe snubber and ( restraints supports were evaluated to confirm their adequacy. Other Support Types The remaining types of supports that were part of the verification program include safety-related main steam supports, safety-related pipe anchors, non-safety-related pipe support structures and other miscellaneous sup-ports designed by the A-E's Civil Group. Because of the relatively few number of each of these support types and their varying configuration, no generic evaluations were made. In the verification of each support, the weld capacity was evaluated. This was done by engineering judgment if the wold capacity was clearly greater than that of the member (such as full penetration welds) or if the load was much smaller than the capacity furnished by the specified weld. The welds for the remaining supports were analyzed to demonstrate their adequacy. The analysis included determination of the loads applied to the welds and a comparison of these j loads to the weld capacity. In this manner welded connections for the safety-related main steam supports, safety-related pipe anchors, non-safety-related pipe support structures and other miscellaneous supports were evaluated to confirm l their adequacy. Support and Weld Identification Process The evaluating engineer located the support detail and determined the wold configuration on the civil drawings in one of the following ways:
+ Schedules were provided on the pipe snubber restraint and anchor support drawings, which identified the support number and referred to a specific detail. The welding configuration was either called out on the drawing details, described by notes on the drawings, or shown by reference to another drawing. In a few cases, field measurements were necessary to confirm the as-built weld configuration.
+ Where schedules were not provided, the support identification was shown directly on the detail. These details either provided the weld configuration or the drawing notes stated the welding require-ments. In a few cases, field measurements were necessary to con-firm the as-built weld configuration.
l t
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C Page 3 of 29 In this manner the evaluaters were able to determine the. support details, including the weld configuration for each support being verified. Documentation For each support, weld verification has been documented either in a cal-culational format or by summarizing the engineering judgment. This docu- i mentaticn is included in the support calculation file. Table 1 shows how the welds for each support were evaluated. i I l 1 1 i SAD!mr f
)
1940W.0887 l l 1 _ _ . _ _ _ . _ - _ _ _ _ _ - - - _ - - _ - - _-- -_- _ a
Trojan Nuclear Plant Docum:nt Control Desk Dockch 50-344 August 18, 1987 License NPF-1 Attachment C ! TABLE Cl Pago 4 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation Support Full No. SYSTEM l Penetration Calculation l Judgment j Other l Wold SS-80 Main Steam X SS-83 Main Steam X SR-82 Main Steam X SR-85 Main Steam X SR-86 Main Steam X SR-90 Main Steam X SS-87 Main Steam X SS-91 Main Steam X SR-89 Main Steam X SR-93 Main Steam X
)
l SS-81 Main Steam X by PGE l SS-88 Main Steam X by PGE SS-92 Main Steam X by PGE SS-86 Main Steam I by PGE SR-57 Main Steam X SR-58 Main Steam X SR-60 Main Steam X SR-61 Main Steam X SR-63 Main Steam X SR-64 Main Steam X SR-66 Main Steam X i SR-67 Main Steam X SR-300 Main Steam X
)
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C TABLE C1 Page .'s of 29 SAFETY-RELATED PIPE SUPPORT WELD. DESIGN VERIFICATION Method of Critical Weld Evaluation ) Support Full ] No. SYSTEM lPenetratica Calculation l Judgment ' Other l I Weld SR-301 Main Steam X q l SR-302 Main Steam I SR-303 Main Steam X < SR-304 Main Steam X SR-305 Main Steam X SR-305 Main Steam X 5 SR-307 Main Steam X l l l l l _ - _ _ _ _ _ _ _ _ _ - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - _ _ - _ _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ . __ _ _ _ _ _ - _ __ a
Trojan Nucicar Plant Document Control Dssk Docket 50-344 August 28, 1987 License NPF-1 Attachment C TABLE Cl- Page 6 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION l Method of Critical Weld Evaluation Support Full No. SYSTEM Penetration Calculation Judgment Other l Weld H9 Main Steam X H10 Main Steam X H11 Main Steam H j H12 Main Steam X l SR-33 Main Steam X SR-34 Main Steam X SR-36 Main Steam X SR-37 Main Steam X SR-82 Main Steam X l j SR-85 Main Steam X l SS-35 Main Steam X SS-38 Main Steam X SS-80 Main Steam X SS-81 Main Steam X j SS-83 Main Steam X l l SS-86 Main Steam X H13 Main Steam X H14 Main Steam X H15 Main Steam X H16 Main Steam X H17 Main Steam X H18 Main Steam X SR-39 Main Steam X i i w _ - ---__.__________2__ _ _ _ _ . _
1 i 1 Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C TABLE C1 Page 7 of 29 f 1 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION I Method of Critical Weld Evaluation Support Full 1 No. SYSTEM Penetration . Calculation Judgment l Other Weld SR-40 Main Steam X SR-86 Main Steam X SR-89 Main Steam X SR-90 Main Steam X l SR-93 Main Steam X l SS-87 Main Steam X l SS-88 Main Steam X SS-91 Main Steam X SS-92 Main Steam X SR-304 Main Steam X SR-305 Main Steam X SR-306 Main Steam X SR-307 Main Steam X l H2 Main Steam X l l H8 Main Steam X SR-57 Main Steam X I l SR-58 Main Steam X l l l l SR-66 Main Steam X i I SR-67 Main Steam X SR-68 Main Steam X SR-69 Main Steam X SR-74 Main Steam X i SR-75 Main Steam X j l l l l 1 l 1
- -- .-_ ___ _ _ _ - - - - - - - _ - - _ - - - . - _ _ _ _ _ - - - . _ _ _ -. _ __ _ _ ___.---- _ _ __ _____d
l - Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 i License NPF-1 Attachment C 1 TABLE C1 Page 8 of 29 j SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation i Support Full , l No. SYSTEM Penetration Calculation ' Judgment Other ! I Weld ! SS-76 Main Steam X l SS-77 Main Steam X M4 Main Steam X i M6 Main Steam X SR-60 Main Steam X SR-61 Main Steam X SR-63 Main Steam X SR-64 Main Steam X l SR-70 Main Steam X SR-71 Main Steam X SR-72 Main Steam X l SR-73 Main Steam X SS-78 Main Steam X SS-79 Main Steam X SR-300 Main Steam X SR-301 Main Steam X SR-302 Main Steam X SR-303 Main Steam X l i 1 i , I
Trojen Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 Licenso NPF-1 Attachment C TABLE C1 Page 9 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation Support Full No. SYSTEM Penetration Calculation Judgment Other Weld SR-247 Component Cooling Water X SR-248 Component Cooling Water X SR-249 Component Cooling Water X i SR-803 Component Cooling Water X l SR-804 Component Cooling Water X SR-805 Component Cooling Water X SR-810 Component Cooling Water X SR-811 Component Cooling Water X SR-812 Component Cooling Water X SR-813 Component Cooling Water X SR-814 Component Cooling Water X SR-815 Component Cooling Water I SS-1002 CVCS Normal Letdown I SS-1003 CVCS Normal Letdown X i l
-Trojan Nuclear Plant Document Control Desk -Docket 50-344 August 18, 1987 License NPF-1 - Attachment C.
TABLE C1 Page 10 of 29 ) i SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION j 1 l Method of Critical Weld Evaluation -I Support Full No. SYSTEM Penetration Calculation Judgment Other l Weld SS-1006 CVCS Normal Letdown X
- SS-1007 CVCS Normal Letdown X SS-1008 CVCS Normal Letdown X l SS-1009 CVCS Normal Letdown X I
SS-1010 CVCS Normal Letdown X l SS-1011 CVCS Normal Letdown X l l SS-1012 CVCS Normal Letdown X l l SS-1013 Reactor Cool System Drein X l SS-1014 Reactor Cool System Drain X j SS-1015 Reactor Cool System Drain X l l SS-1016 Reector Cool System Drain X ! SS-1017 Reactor Cool System Drain X l SS-1018 Reactor Cool Systcm Drain X SS-1019 Reactor Cool System Drain X SS-1020 Reactor Cool System Drain I SS-1021 Reactor Cool System Drain X l SS-1022 Reactor Cool System Drain X l l SS-1023 Reactor Cool System Drain X SS-1024 Boron Injection X l SS-1025 Boron Injection X i SS-1026 Boron Injection X . 1 l l 4 d.
Trojcn Nuclear Plant Document Control Dask Docket-50-344 August. 18, 1987 License NPF-1 Attachment C TABLE C1 Page 11 of 29
, f SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION i
Method of Critical Weld Evaluation i I Support Full No. SYSTEM Penetration Calculation Judgment l Other I Wold i SS-1027 Boron Injection X j SS-1028 Boron Injection X SS-1029 Boron Injection I SS-1030 Boron Injection X SS-1031 Boron Injection X SS-1032 Boron Injection X SS-1033 Boron Injection X SS-1034 Boron Injection X SS-1035 Boron Injection I SS-1036 Safety Injection X SS-1037 Pressurizer Spray X SS-1038 Pressurizer Safety Disch X SS-1039 Pressurizer Safety Disch X SS-1040 Pressurizer Safety Disch X SS-1041 Pressurizer Safety Disch X i ! SS-1042 Pressurizer Safety Dicch X SS-1043 Pressurizer Safety Disch X SS-1047 Pressurizer Safety Disch X f l SS-1048 Pressurizer Safety Disch X SS-1049 Pressurizer Safety Disch X SS-1052 Safety Injection X ( _ _ _ _ _ _ _ _ _ _________________.__________.m_ _ _ _ _ _ _ _ _
Trojen Nuclear Plant Docum:nt Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C TABLE C1 Page 12 of 29
, SAFETY-RELATED PIPE SUPPORT WELD DESICN VERIFICATION d
Method of Critical Weld Evaluation Support Full No. SYSTEM Penetration Calculation Judgment other l Weld SS-1053 Safety Injection X SS-1054 Reactor Cool Pump Seal Inj X SS-1055 Reactor Cool Pump Seal Inj X SS-1056 Reactor Cool Pump Seal Inj X SS-1057 Reactor Cool Pump Seal Inj X SS-1058 Reactor Cool Pump Seal Inj X SS-1059 Reactor Cool Pump Seal Inj X SR-1060 Reactor Cool Pump Seal Inj X SR-1062 Reactor Cool Pump Seal Inj X SR-1063 Reactor Cool Pump Seal Inj X SR-IO64 Reactor Cool Pump Seal Inj X SR-1066 Reactor Cool Pump Seal Inj X SR-1068 Reactor Cool Pump Seal Inj X S3-1070 Pressurizer Spray X SS-1072 Pressurizer Spray X SS-1074 Pressurizer Spray X SS-1075 Pressurizer Spray X SS-1076 Pressurizer Spray X SS-1077 Pressurizer Spray X J SS-1078 Pressurizer Spray X SS-1079 Pressurizer Spray X , SS-1080 Pressurizer Spray X SS-1082- Pressurizer Spray X h .
Trojen Nuclear Iisnt 3 Document Control Desk Docket 50-344 ' August 18, 1987 License NPF-1 Attachment C TABLE C1 Page 13 of 29
\ SAFETY-RELATED P1PE SUPPORT WELD DESIGN VERIFICATION
+ s.,
3 %s , Method of Critical Weld Evaluation Support ' Full No. SYSTEM l Penetration l Calculation ; Judgment Other Wold SS-1083 PressurizerSpe)y , I SS-1084 Pressurizer Spray X S 3 ES-1085 Pressurizer Spiay ,, } X Pres.surizer Spray SS.1086 X a SS-1087 Boron Injectior- s X L i SS-1088 Boron ,Inj ection I SR-1089 Itoron Injection X SF-1090 Boron Injection X I SS-1091 ' Boron Injection X l SS-1092 Byron Injection X
\
SS-1093 Boron injection X SS-1096 Safety Injection X m SS-1G98 Safety Injection X
' ttS-1099 Safety Injection .
X f 4SS-1A00 Safety Injection X
,1 SS-11(1 Safety Injection $ X SSa1102 Safety Injection I SP- 1103 Safety Injection X SS-1104 $bafetyInjection '
X s
'3S-1105 S$fety InjecU on X SS-1106 Safety Injection X SS- 1,107 Safety Injection X
'l SS!1108 Safety Injection X
\: s 9
i 1 l Trojen Nuclear Plant Document Control Desk i Docket 50-344 August 18, 1981 License NPF-1 Attachment C # TABLE C1 Page 14 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critica:. Weld Evs;1untion , Support l Full .j No. SYSTEM ' Penetration Calculation , Judgment ' Other 1 Weld SS-1109 Safety Injection X l SS-1111 Safety Injection I j 1 SS-111,? Safety Injection X SS-1113 Safety Injection X SS-1114 Safety Injection X SS-1115 Safety Injection X i SR-1116 . Safety Injection X SS-1117 Safety Injection X SR-1118 Safety Injection X SS-1119 Safety Injection X SR-1120 Safety Injection X SR-1122 Safety Injection X l SR-1124 Safety Injection X ) SS-1126 Pressurizer Safety Disch X SS-1127 Pressurizer Safety Disch X SS-1128 Pressurizer Safety Disch X SS-1129 Pressurizer Safety Disch I l SS-1130 Pressurizer Safety Disch X SS-1131 Pressurizer Safety Disch X SS-1132 Pressurizer Safety Disch X l SS-1133 RHR Suction X SS-1134 RHR Suction X l S _. ______.____m___w
l Trojan Nuclear Plant - Document Control Desk l Docket 50-344 August 18, 1987 { l Licence NPF-1 Attachment C TABLE C1 Page 15 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation ) Support Full No. SYSTEM Penetration : Calculation Judgment Other
- l. Weld SS-1135 RHR Suction X SS-1136 RHR Suction X SS-1137 Reactor Cool System Drain X !
SS-1138 Reactor Cool System Drain X SS-1139 Safety Injection X SS-1140 Safety Injection X SS-1141 Safety Injection X f SS-1142 Safety Injection X SS-1143 Safety Injection X SS-1144 Safety Injection X SS-1145 Safety Injection X SS-1146 Safety Injection X SS-1147 Safety Injection X SS-1148 Safety Injection X SS-1150 Safety Injection X SS-1151 Safety Injection X SS-1153 Safety Injection X - SS-1154 Safety Injection X
, SR-1155 Safety Injection X SS-1156 Safety Injection X SS-1157 Safety Injection X SS-1158 Safety Injection X SS-1159 Safety Injection X
\ l
Trojen Nuclear Plant Document Control Desk j Docket 50-344 August 18, 1987 q License NPF-1 Attachment C l i TABLE C1 Page 16 of 29 [ i SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of critien1 Weld Evaluation ___ j Support Full ) No. SYSTEM Penetration Calculation Judgment Other Wold . SS-1160 Safety Injection I SS-1162 Reactor Cool Pump Seal Inj X q l SS-1164 Reactor Cool Pump Seal Inj X l SS-1168 CVCS Letdown X S5-1169 CVCS Letdown X 4 SS-1170 CVCS Letdown X j SS-1171 CVCS Letdown X l SS-1172 Reactor Cool Instrument I l t SS-1173 - Reactor Cool Instrument X SS-1174 Reactor Cool Instrument X SS-1175 Reactor Cool Instrument X SS-1176 Reactor Cool Instrument X SS-1177 Reactor Cool Instrument X SS-1178 Reactor Cool Instrument X SS-1179 Reactor Cool Instrument X l SS-1180 Reactor Cool Instrument X ) SS-1181 Reactor Cool Instrument I l SS-1182 Reactor Cool Instrument I SS-1183 Reactor Cool Instrument X i l SS-1184 Eeactor Cool Instrument X f l I SS-1185 Reactor Cool Instrument X j SS-1186 Reactor Cool Instrument X SS-1187 Reactor Cool Instrument X ! l l l I l l i
--__ _a
I i Trojen Nuclear Plant Document Control Desk l- August 18, 1987 Docket 50-344 Licence NPF-1 Attachment C TABLE C1 Page 17 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation Support Full No. SYSTEM Penetration Calculation Judgment Other Weld SS-1188 Reactor Cool Instrument X l SS-1189 Reactor Cool Instrument X l SS-1190 Reactor Cool Instrument X j i SS-1191 Reactor Cool Instrument X SS-1192 Reactor Cool Instrument X SS-1193 Reactor Cool Instrument X SS-1194 Reactor Cool Instrument X < SS-1195 Reactor Cool Instrument X ! SS-1196 CVCS Letdown X SS-1197 CVCS Letdown X SS-1198 Reactor Cool Pump Seal Inj X SS-1199 Reactor Cool Pump Seal Inj X j SS-1200 Reactor Cool Pump Seal Inj X SS-1201 Reactor Cool Pump Seal Inj X SS-1202 Pressurizer Spray X SS-1203 Pressurizer Spray X SS-1204 Pressurizer Spray X SS-1205 Pressurizer Spray I SS~1206 Reactor Cool Instrument X SS-1207 Reactor Cool Instrument X 1 SS-1208 Reactor Cool Instrument X SS-1209 Reactor Cool Instrument I SS-1210 RHR Suction X
i i Trojan Nucisar Plant Document Control Desk ] Docket 50-3G4 August 18. 1987 License NPF-1 Attachment C l TABLE Cl Page 18 of 29 i SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION l Method of Critical Weld Evaluation Support Full l No. SYSTEM l Penetration Calculation Judgment ! Other l ! Weld SS-1211 Safety Injection X SS-1212 Safety Injection X SS-1213 Safety Injection X SR-1214 Safety Injection I SS-1215 Safety Injection X I l SS-1216 Safety Injection SS-1217 Safety Injection X SS-1218 ; Safety Injection X SS-1219 Reactor Cool Instrument X l SS-1220 Reactor Cool Pump Seal Inj X SS-1221 Reactor Cool Pump Seal Inj X SR-1222 Reactor Cool Pump Seal Inj X SS-1223 Reactor Cool Pump Seal Inj X SR-1224 Reactor Cool Pump Seal Inj X l SR-1225 Reactor Cool Pump Seal Inj X l SR-1226 Reactor Cool Pump Seal Inj . I j SS-1227 Reactor Cool Pump Seal Inj X l ' l SR-1228 Reactor Cool Pump Seal Inj X SS-1229 Reactor Cool Pump Seal Inj X SS-1230 Reactor Cool Pump Seal Inj X SS-1231 Reactor Cool Pump Seal Inj X SS-1232 Reactor Cool Pump Seal Inj X l SS-1233 Reactor Cool Pump Seal Inj X l l l l l I l _j
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C TABLE C1 Page 19 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation Support Full ! No. SYSTEM Penetration Calculation Judgment Other l Wold SS-1234 Reactor Cool Pump Seal Inj X SS-1235 Reactor Cool Pump Seal Inj X SS-1236 Reactor Cool Pump Seal Inj X l SS-1237 Reactor Cool Pump Seal Inj X l
1238 - Reactor Cool Pump Seal Inj X SS 4239 Pressurizer Spray X SS-1240 Reactor Cool Instrument X l
SR-1242 Reactor Cool Instrument X SR-1243 Reactor Cool Instrument X i SR-1244 Reactor Cool Instrument X l l l SR-1245 Reactor Cool Instrument X ) l SR-1246 Reactor Cool Instrument X SR-1247 Reactor Cool Instrument X SR-1248 Reactor Cool Instrument X l SR-1249 Reactor Cool Instrument X SR-1250 Reactor Cool Instrument X SR-1251 Reactor Cool Instrument X SR-1252 Reactor Cool Instrument X SR-1253 Reactor Cool Instrument X SR-1254 Reactor Cool Instrument X SS-1257 Pressurizer Safety Disch X l
Trojen Nuclear Plant. Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C TABLE C1 Page 20 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation Support Full No. SYSTEM Penetration l Calculation Judgment Other Weld SS-1258 Pressurizer Safety Disch X SS-1259 Pressurizer Safety Disch X l SS-1260 Pressurizer Safety Disch X SS-1262 Pressurizer Safety Disch X i SS-1263 Pressurizer Safety Disch X SS-1264 Pressurizer Safety Disch X SS-1266 Pressurizer Safety Disch X SS-1267 Pressurizer Safety Disch X SS-1268 Pressurizer Safety Disch X SS-2101 Steam Generator Blowdown X SS-2102 Steam Generator Blowdown X SS-2104 Steam Generator Blowdown X SS-2105 Steam Generator Blowdown X SS-2106 Steam Generator Blowdown X SS-2109 Steam Generator Blowdown X SS-2110 Steam Generator Blowdown X l SS-2111 Component Cooling Water X SS-2112 Reactor Cool Pump Seal Inj X SS-2113 Component Cooling Water X SS-2114 Component Cooling Water X SS-2115 Steam Generator Blowdown X SS-2116 Steam Generator Blowdown X l 1 l ? l l l l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _______.__________-._a
Trojan Nuclear Plant Document Control Desk l Docket 50-344 August 18, 1987 License NPF-1 Attachment C i TABLE C1 Page 21 of 29 t SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION 1 Method of Critical Weld Evaluation No. SYSTEM l Penetration Calculation Judgment Other l Weld SS-2117 React Cool Pmp Seal Wtr Inj X SS-2118 React Cool Pmp Seal Wtr Inj X SS-2119 Steam Generator Blowdown X SS-3001 Steam Generator Blowdown X SS-3003 Steam Generator Blowdown X l SS-3004 Steam Generator Blowdown X 1 l l l l l l I I l w_----_--------------
Trojan Nuclear Plant Document Control Desk i i Docket 50-344 August 18, 1987 License NPF-1 Attachment C TABLE C1 Page 22 of 29 ! SAFETY-RELATED PIPE SUPPORT WELD DESICN VERIFICATION l 1 l c Method of Critical Wald Evaluation l Support l Full No. SYSTEM Penetration (Calculation Judgment Other l Weld l H-115 Reactor Coolant Instrument X ! l H-116 Reactor Coolant Instrument X H-117 Reactor Coolant Instrument X J H-118 Reactor Coolant Instrument X SS-143 Reactor Coolant Instrument X SS-144 Reactor Coolant Instrument X I SS-145 Reactor Coolant Instrument X SS-146 Reactor Coolant Instrument X I SS-147 Reactor Coolant Instrument X SS-148 Reactor Coolant Instrument X SS-149 Reactor Coolant Instrument X l SS-150 Reactor Coolant Instrument X SS-10 Main Feed X SS-11 Main Feed X SS-13 Main Feed X SS-14 Main Feed X SR-65 Chemical & Volume Control X SR-66 Chemical & Volume Control X SR-78 Chemical & Volume Control X SR-79 Chemical & Volume Control X SR-97 Chemical & Volume Control X SR-98 Chemical & Volume Control X l SR-82 Chemical & Volume Control X ! l l 1
l 1 Trojan Nucicar Plant Document Control Desk I Docket 50-344 August 18, 1987 License NPF-1 Attachment C TABLE C1 Page 23 of 29 i SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION . 1 Method of Critical Weld Evaluation l No. SYSTEM penetration Calculation l Judgment l Other l Wold H-59 Chemical & Volume Control X H-57 Chemical & Volume Control X l SA-53 Component Cool Water (CCW) X SA-250 Chem & Vol control (CVCS) X SR-807 CCW X 1 SR-809 CCW X SR-816 CCW X l 1 SS-3002 CCW X SR-1001 CVCS X 1 ! SR-1004 CCW X l 1 I SR-1005 CCW X SS-1050 Reactor Coolant Instrument X SS-1051 Reactor Coolant Instrument X SS-1241 RCP Seal Injection X SS-1255 Safety Injection X I l l
Trojan Nuclotr Plant Document Control Dssk j Docket 50-344 August 18, 1987 License NPF-1 Attachment C , TABLE C1 Page 24 of 29 l SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation j Support Full l No. SYSTEM Penetration Calculation l Judgment l Other ( Weld SR-149 Main Steam X SR-16 Main Steam X t SR-8 Main Steam X 1 SR-9 Main Steam X ! SR-10 Main Steam X l SR-14 Main Steam X l SR-15 Main Steam X SR-17 Main Steam X 1 SR-18 Main Steam X 1 l SR-19 Main Steam X SR-26 Main Steam
- X SR-27 Main Steam I
! SR-168 Main Steam X SR-301 Main Steam X l SR-302 Main Stee.m X SR-303 Main Steam I SA-310 Pressurizer Blowdown System X SA-315 Pressurizer Blowdown System X SS-18 Pressurizer Blowdown System X SS-23 Pressurizer Blowdown System X SS-29 Pressurizer Blowdown System X SS-47 Pressurizer Blowdown System X SS-300 Pressurizer Blowdonm System X l
l l l Trojen Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C j TABLE C1 Page 25 of 29 j l SAFETY-RELATED PIPE SUPPORT WELD DESICN VERIFICATION Method of Critical Weld Evaluation Support Full i No. SYSTEM l Penetration Calculation Judgment l Other Weld SS-301 Pressurizer Blowdown System X SS-302 Pressurizer Blowdown System X SS-303 Pressurizer Blowdown System X SS-306 Pressurizer Blowdown System X SS-307 Pressurizer Blowdown System X SS-308 Pressurizer Blowdown System X SS-311 Pressurizer Blowdown System X SS-316 Pressurizer Blowdown System X SS-1038 Pressurizer Blowdown System X SS-1039 Pressurizer Blowdown System X
's SS-1040 Pressurizer Blowdown System X SS-1041 Pressurizer Blowdown System X SS-1042 Pressurizer Blowdown System X SS-1043 Pressurizer Blowdown System X j i
SS-1257 Pressurizer Blowdown System X l SS-1258 Pressurizer Blowdown System X SS-1259 Pressurizer Blowdown System X SS-1260 Pressurizer Blowdown System X l l SS-1262 Pressurizer Blowdown System X l SS-1263 Pressurizer Blowdown System X I ! SS-1264 Pressurizer Blowdown System X SS-1266 Pressurizer Blowdown System X SS-1267 Pressurizer Blowdown System X SS-1268 Pressurizer Blowdown System X l ________-_________a
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment C TABLE C1 Page 26 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation Support Full No. SYSTEM l Penetration Calculation Judgment l Other l Weld SR-50 Pressurizer Blowdown System X SR-51 Pressurizer Blowdown System X SR-304 Pressurizer Blowdown System X SR-305 Pressurizer Blowdown System X SR-309 Pressurizer Blowdown System X SR-312 Pressurizer Blowdown System X SR-313 Pressurizer Blowdown System X 1 SR-314 Pressurizer Blowdown System X l SS-39 Pressurizer Blowdown System X SS-40 Pressurizer Blowdown System X SS-41 Pressurizer Blowdown System X I SS-42 Pressurizer Blowdown System X SS-43 Pressurizer Blowdown System X SS-44 Pressurizer Blowdown System X l SS-45 Pressurizer Blowdown System X SS-330 Pressurizer Blowdown System X SS-1126 Pressurizer Blowdown System X SS-1127 Pressurizer Blowdown System X SS-1128 Pressurizer Blowdown System X SS-1129 Pressurizer Blowdown System X SS-1130 Pressurizer Blowdown System X SS-1131 Pressurizer Blowdown System X SS '1132 Pressurizer Blowdown System X SR-15 Pressurizer Blowdown System X l
1 Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 i License NPF-1 Attachment C l TABLE C1 Page 27 of 29 ! SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critica: Wold Evaluation Support l Full No. SYSTEM Penetration Calculation Judgment Other l Weld SR-24 Pressurizer Blowdown System X I SR-30 Pressurizer Blowdown System X l SS-1047 Pressurizer Blowdown System X SS-1048 Pressurizer Blowdown System X SS-1049 Pressurizer Blowdown System X 1 l l l l i d
Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachmer.t C i TABLE C1 Page 28 of 29 SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION Method of Critical Weld Evaluation Support l Full No. SYSTEM l Penetration Calculation l Judgment l Other l Weld 1 Acu. Safety Injection X 2 Acu. Safety Injection X 3 Acu. Safety Injection X 4 Acu. Safety Injection X 5 RHR X 6 CVCS X 7 Pressurizer Spray X 8 Pressurizer Spray X 9 RHR X , 1 10 CVCS X 11 RHR X 12 RHR X 13 Pressurizer Spray X 14 RHR X 15 RHR X 16 RHR X 17 CVCS X l l 18 CVCS X l 19 CVCS X l l 20 CVCS X l 21 RCP Seal Water X 22 RCP Seal Water X l L__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
- t. .
l' \ Trojan Nuclear Plent Docum nt Control Desk ) Docket 50-344 August 18, 1987 1 License NPF-1 Attachment C i TABLE C1 Page 29 of 29 l SAFETY-RELATED PIPE SUPPORT WELD DESIGN VERIFICATION l
- j. Method of Critical Weld Evaluation _ '
i Support Full No. SYSTEM Penetration Calculation Judgment Other ; Weld ! l f 25 R.C. Drain & Vent X 26 R.C. Drain & Vent X 27 R.C. Drain & Vent X 28 RCP Seal Water X i 29 CVCS X 30 CVCS No Wold 31 RC Drain and Vent X 32 RCP Seal Water X l 33 RCP Seal Water X 34 RCP Seal Water X 35 RCP Seal Water X i l l l l 1986W
l Trojan Nuclear Plant Document Control Desk Docket 50-344 August 18, 1987 License NPF-1 Attachment D1 MD43RANDUM 6 Pages l l l August 11, 1987 RCJ-326-87 d 1 ! To: D. W. Cockfield , 0~" From: R. C. Jarman ! C. P. Yundt . [ l
Subject:
PGE QA and Technical Functions Assessment of the Indepen: lent QA i Audit and 7bchnical Quality Review (7QR) of the Trojan Nuclear Plant i Pipe Support Desictn Verification F tcrtcua j 1 DuriryJ the period of August 1 through 10, 1987 an independent audit and ) technical quality review of Bechtel's Pipe Support Design Verification Pwgccua for Trojan was planned and conducted. The purpose of the audit and technical quality review was to determine if there was a propcuunatic breakdown in ) l Bechtel's Quality Assurance Progcun during the Pipe Support Design Verification l l Program. IMPELL Corporation was selected to perform the audit based on their l experience and engineering expertise. The INPELL team consisted of a qualified audit team leader, a technical manager, two supervising engineers, three lead ! ( senior engineers, two senior engineers and a technical specialist. The l performance of the audit was monitored by PGE civil, mechanical, and Quality I Assurance Engineers. FGE and IMPELL developed the audit plan, the QA audit checklist, and the l technical quality review criteria checklists. The audit scope included a i thorough review of the verification process for a sample of supports and addressed specific concerns / weaknesses identified during the performance of the verification program. The deficiencies and renmmandations identified as a result of this audit / review were transmitted to Bechtel by IGE for evaluation and corrective action (s). ME requested that Bechtel respond immediately to two deficiencies and two renmmndations which required prampt evaluation (items 3, 4, 5, and
- 7. f) . Dechtel was requested to respond to all remaining items by August 21, 1987. IGE QA will conduct a follow-up audit on all identified deficiencies and renmmndations as part of FGE's long-term Pipe Support Design Verification Program.
The following is a FCE QA and Technical Functions assessmnt of the deficien-cies and recommendations for which an immediate response was requested.
- 3. OA/ TOR - Technical Procedures Renmondation: In the current support verification program, support evaluations are performed by the Bechtel Plant Design Group and Civil Group. Each Group uses its own design criteria procedures. The Plant Design Group uses
l l RCI-326-87 August 11, 1987 Pace 2 1 DC-11760-P-003, " Pipe Support Design Criteria for Trojan l- Nuclear Plant", Rev. 0,. dated 6/15/87. W e Civil Group l uses Standard 11760-C1, " Civil and Structural Design i l Criteria for the Trojan Nuclear Plant:, Rev. 5, dated l '10/6/86. Although reviewers did not compare the technical l contents of the Plant Design Group and Civil Group ( criteria on a point by point basis, it appears that the criteria developed by the Plant Design Group is a more complete document for pipe support qualification. Example: h e criteria developed by the Plant Design Group states that the material yield strength should be taken at the operating temperatures. E is is nct stated in the criteria developed by the Civil Group. B is is reflected' in Bechtel Calculation File J-301 for the following supports for which the temperature is not defined to q adjust the yield strength: SS-80, SR-82, SS-83, SR-85, ' SR-86, SS-87, SR-89, SR-90, SS-91, and SR-93. From the above sample supports reviewed during this audit, j there are no qualification issues directly related to the j l existence of two criteria. The adequacy of the supports reviewed were not affected by the differences in criteria. It is rcmmmanded that Bechtel review the technical ' contents of the two criteria and make certain that there l L is no inconsistency between the two criteria for pipe j , support qualification. l l Bechtel Response to the Item 3 Recommendation Both criteria are applicable and acceptable for design in the respective , l group. A comparison of the two criteria has revealed no technical differences i that would affect the validity of the pipe support verification program.
% e Civil discipline perforned design of pipe supports during Trojan's original l
design. m is activity was performed under the guidance of Criterion C1, and hence it was proper to use this criterion for the review effort for consistency with the design basis. Although the criterion of record for the Civil disci-pline's review was C1, the group was aware of the considerations contained in i Criterion P003. lasues such as temperature effects (explicitly contained in l P003, but not in C1) were addressed by the civil discipline. For example, such l temperature effects were documented in the pipe anchor calculations. In the l example in the recommendation, temperature had no effect and in this case was not documented. These documentation concerns will be addressed in the long-term program. ICE Assessment of the Item 3 Recommendation and Response Although the Bechtel Civil Engineering Group and Plant Design Group have utilized different criteria documents for the support verification, PGE's assessment of the finding is that the results of the verification effort are
RCJ-326-87 August 11, 1987 pace 3 unaffected. The specific tempernt.ure concerns identified have been determined to be insignificant due to the negligible change in yield strength from the reference temperature to the raninal temperature of the supports during operation. RIE considers the use of two different design criteria by two separate design groups unacceptable in the lorg-term program. Therefore, PGE will require that all design groups use consistent design criteria for Trojan.
- 4. TOR - Exoansion Anchors l Deficiency: Several support drawings indicated that baseplate bolts can be either Rock bolts or expansion anchor bolts.
! There is no documentation showing the type of bolt used in the support evaluation. The use of expansion anchor bolts may impact the design verification since different design and analysis xmmntions are req 1 tired to evaluate the expansion bolts. Bechtel Response to the Item 4 Deficiency Note 3 on Drawing C-386, Sheet 2, requires the use of Rock bolts except 1-in. diameter expansion anchors (by Phillips or USM Corporation) are acceptable substitutes in the following cases:
- a. Type I support using a W4x13 with snubber of less that 3 kips.
- b. Type I support using a W6x20 with snubber of less that 1 kip.
- c. Type II support with a snubber of less than 3 kips,
- d. Type IV support with a snubber of less than 3 kips.
- e. Special support with a snubber of less than 10 kips.
The different analysis assumptions required for evaluating expansion anchors are significant only when the applied loads are large relative to the expansion anchor allowables. For the conditions where cupansion anchors nre allowed for cases authorized, Note 3 limits the applied loads to the low values noted. The resultant loads in the expansion anchors are also low relative to their allowables. The factors of safety, between applied loads and anchor capacities exceed 10 in all cases. For the "special supports" (Case e), expansion anchors are used when specifi-cally called for on the support details. Again, the resultant loads in the expansion anchors are low relative to their allowables. Expansion anchors are specified for supports SR-1060, SR-1066, SS-1077, SS-1168, SR-1228, and SR-1244. The faulted (SSE) loads applied to these supports are 0.38k, 0.12k, 2.42k, 0.12k, 0.18k, and 0.40k, respectively. The factors of safety exceed 5 in all cases.
I RCJ-326-87 August 11,:1987 Pace 4 Since the loads applied to the expansion anchors are low relative to their l- allowables, detailed analysis in the verification calculations of the expansion anchors were not deemed necessary. l 1 PGE Assessment of the Item 4 Deficiency ard Response PGE has evaluated the IMPELL finding and concurs with the Bechtel statement that different design ard analysis are r==vy for expansion anchors only when the applied loads are large relative to the expansion anchor allowables. IGE feels that it is remry for the verification documentation to ' clearly irdicate the type of anchor used and the rationale for its acceptability regardless of the size of the loads. ICE does not believe this issue affects the conclusion of the verification program, but intends in the larg-term to have the documentation completed.
- 5. TOR - Weld Evaluations Deficiency: There was insufficient input data to define the details for the qualification of welds in the evaluation of some supports. Examples are:
SS-87 Sketch showed one weld symbol for ALL welds for the support. However, there was I a note requesting additional weld information and there was no evidence that this additional information was received. i All welds were then assumed to be 3/16" ; fillet all around for the evaluation. l SS-1135 Member weld connection was checked for SS-1268 adequacy; however, weld pattern, size and type were not identified. For SS-1135, welds were qualified based on a typical weld pattern and size shown on Generic , Support Type I which is not a typical " connection. For SS-1268, weld between member on baseplate is qualified by comparison to Type I generic support which calls for a 1/4" all around fillet. Bechtel Response to Item 5 Deficiency ! During the Support Verification Ptwjscun, input data on welds was obtained from Civil pipe support details, procurement specifications, and field walkdowns. In the cited examples, additional documentation should have been provided to give traceability to information obtained by field walkdowns. For SS-87, field information was obtained and used but not documented. This lack of documentation does not invalidate the pipe support verification program. For supports SS-1135 and SS-1268, the calculations and review were based upon correct assumptions and 1he deficiency is incorrect.
RC7-325-87 August 11, 1987 Eaoe 5
'Ihe evaluatirg engineer located the support detail and determined the weld configuration on the civil drawings in one of the following ways:
. Schedules were provided on the pipe snubber restraint and anchor support drawings, which identified the support number and referred to a specific drawing detail. 'Ihe details either called out the welding configuration or notes on the drawings, or referred to drawings which stated the welding requirements.
. Where schedules were not provided, the support identification was shown directly on the drawing detail. 'Ihese details either provided the weld configuration or the drawing notes stated the welding requirements.
Durirg the long-term program, increased emphasis will be placed on the need for i added documentation such that the input data and bases for engineering ! judgements will be traceable. IE Assessment of the Item 5 Deficiency and Response Initially during the support design verification program, support welds were evaluated, in many cases by engineering judgement which was not adequately documented to provide a clear basis for the judgements made. Although the lack of documentation has not affected the conclusions of the verification program, IGE considers the lack of documentation unacceptable. IE directed Bechtel to review the support weld verifications to ensure they were thoroughly completed and documented. At this time, the weld verification has been completed and the documentation has been updated.
- 7. Recommendations Related to documentation Rom e ndation f: While there is no technical impact on the results, due to the criteria implementation, the consideration of Rock Bolts spacing and edge distance requirements is not consistently documented.
Resolution: The application of Rock Bolts criteria should be consistently documented in the calculations during the long-term program. Bechtel Renonse to the Item 7.f Recommendation For the Support Verification Program where the support is judged sufficiently close to a concrete edge, edge distance has been considered. For example, edge distance was considered for SS-80 and SS-83 because they are located on top of a wall and the bolts are 6-in. away from the edge. (Calculation J-301, Rev. 1, dated August 3, 1987 evaluates and documents edge distance for the aforemen-tioned supports.)
RCT-326-87 August 11, 1987 Pace 6
'Ibe Support Verification Pmgam also addressed Rock Bolt spacing and the effect of overlapping cones where it was judged that the load demands and bolt spacing may be such as to require such calculations. For example, Calculation J-307 addresses the effect of overlapping shear cones.
Where, by judgement, it was noted that neither edge distances nor spacing were of concern, a conscious decision was made that edge distance and spacing did not apply. 'Ihis was not documented. For the long-term program more detailed documentation of these judgements will be provided. ! PGE Assessment of the Item 7.f Recommendation and Response l ! 'Ihis re-Mation points out the failure to document adequately the use of engineering judgement during the verification process. The re - ndation specifically indicates that the lack of documentation does not affect the conclusion of the verification program. In ortler to close this issue, Bechtel will be directed to address the documentation of Rock Bolt spacing and edge distance criteria as part of the long-term program. Summary Bechtel committed to the following additional actions in their letter forwarding the responses discussed above:
- 1. Add senior staff to the conpletion of the program.
- 2. Prepare a tabulation to document the sources ard magnitudes (original and current) of support loads.
- 3. Prepare a tabulation of how welds were reviewed on each support.
it is evident fram the findings and rm mrdations of the OA audit that significant documentation deficiencies exist in the support verification program. These deficiencies, however, have not affected the conclusion that the supports, except those previously identified as deficient, are acceptably designed. The documentation deficiencies do, in our opinion, indicate a partial breakdown of Bechtel's OA Pi v am. Nuclear Quality Assurance Department will track, via Nonconforming Activity Reports, the implementation of PGE accepted corrective action to all of the deficiencies and rer-ndations identified by IMPELL and verify the results by audit. RCJ/CPY:dal c: 'INOB Audit Distribution R. W. Griebe, OAC F. H. Lamoureaux T. D. Walt D. W. Cockfield R. C. Jarman C. A. Olmstead C. P. Yundt P. A. McMillan NQAD Files: Pipe Support Verification Audit of Bechtel QARID No. 7.3
('
~ Document Control Desk Trojan Nuclear Plant August 18, 1987 l Docket 50-344 l . Attachment D2 )
License NPF-1 7 pages 1 Bechtel Western Pcwer Corporation Engpneers-Consoucaws ] l Fifty Saale Strost , San Franc *c, CaWortua ', !' aw Ameess: 70 em seetL sen Fwensa s>,mr ) August 11. 193*1 BF-13044 DC# T028862 j i Mr. P., C. Jarman, Hanager
!Tuclear Quality Assurance Depcrt=ent For land General Electric Company 121 S. W. Salmon Street !
Portland, Oregon 97204 . i
Subject:
Fortland General Electric Cc=pany Trojan Nuclear Plant - Job ".1760 Trojen Nuclear Plant Pire Surcert Veri fic a tion _Procrayn
Reference:
PGE letter P3-11831 dated August 11. 1987 : I Pear Mr. Jarran: t We are in receipt of the referanced letter which covers the sarje : ) l cid;; eenducted ry IMPELL in cur effices during the peri:d Aagast i 4-5. HB~ cn tne Trojan Nuclear Plant Pipe Support Ver.fication , l Prograr. l j Je concur with the report's overall stater.ents that o ". . it is concluded that the prograr.matic aspect of the Support Verification Program is adequate to demonstrate the integrity of the structural supports included in the Orejan prograr." i e " Based upon docu=ents reviewed durin; this audit / review and extensive discussions with Bechtel personnel, the review tear concisded that the support design verifi:stier. pr:g a. has properly identified all supports originally desi;nel 'ey Sechtel Civil Group for Tr an nuclear Plant." e "Otr review of the QA prograr implen.entation on this s_p;crt c design verification pr:grar indicates that the prograr is in compliance with the established Svehtel Q4 progran requirements." l The ::ncerns hav= been rescived durin; tne audit and present ne significant GA er te:hni:a1 issues Je'.ative to s_pper-structural adequacy. Severa*. centerns relative te documentation issues are to be rescived during FGE's long-ters program." Tne sneitsure addresses the deficiencies and rect ==encati .s ra*: :7 IMFILL in tne repert required to rest the goals of the suppcrt verification prograr (!!cs . 3. 4, 5). In additien we have 1 12su tenet
Bechtsl Westem Power Corporation 1*r. R. C. Jarman Fage 2 1 provided a response to recommendation 7.f as requested by ?GE. Responses to the remaining items will be covered in a future submittal, by August 21, 1987. We believe that the completion of some of the ongoing actions listed below will serve to improve the level cf documentation of the program and thereby address several of the audit recommendations. To that end we have: ! l
- 1. Added senior staff to the completion of the program. l l
- 2. Prepared a tabulation to document the sources and .?.agnitudes (eriginal and current) of suppert loads.
- 2. Fre;ared a :?.bulatien ci how velis ven revie"ed en -a:h sup; rt.
e.er : possib'e. for esen deficiene or recommends.:icn. Ye neve 1 addressed the rect c at.se and ccrrective actions c.. an ice.:.-b;t-i .e: hasis. Mcwever, we believe some of the issues,.e.fter additie.a1 evaluation.by curselves, have been determined not te he defi:ivncies cased en clarification provided in ene enclosure. If there are questions on this caterial, please do not hesitate te call me. i 1 Very truly yours, L,ch$? i< J . cA.. P. , W. Fosse Project Engineer
.o.b . .: . .u.
Inc le s'_r e i 12615 (646)
D , i L 3. QA/TQR:- Tochnical Procedures Recommendation: In the current support verification program, support- j
.Civil evaluations-are Group.
performed by Bechtel' Plant Design Group and- ! , procedures. .The. Each-Group Plant Design: uses its own design Group-uses criteria DC-11760-P-003. L
" Pipe Support Design Criteria for Trojan Nuclear Plant", ;
Rev. O, dated 6/15/87.- The Civil Group uses. Standard. i
-11760-C1, " Civil and Structural Design Criteria for the Trojan Nuclear' Plant",-Rev. 5 dated 10/6/86.. Although reviewers did not compare: the technical concents of the
' Plant Design Group and Civil Group criteria on a' point by L point basis., it appears that the criteria developed by the Plant Design Group is a more complete document for Pipe ,
L support qualification. Example: The criteria developed by' , 1 the Plant Design Group states that the material yield-
-strength.should be taken at the operating tem;erstures.
This is not stated in the criteria developed by._the Civil Group. 1 This is reflected in Bechtel Calculation File J-301' I for.the fellowing suppcrts for which the' temperature is not defined to adjust the yield strength: SS-80, SR-83,"SS-S2, SR-85, S R - 8 6 ,. S S - 8 7 ,- SR-89, SR-90, SS-91, and SR-93. From the above sample supports reviewed during.this audit, there are no qualification issues directly related to the existence of two criteria. The adequacy of.the' supports .! reviewed was not affected by the differences in criteria. 1 It is recommended that Bechtel review the technical contents-of the two criteria and make certain that there is no i inconsistency qualification. between the two criteria for pipe support l- Response: t l i 30th criteria are applicable and acceptable for design ia the respective group. A comparison of the two criteria has i revealed no technical differences that would affect the validity of the pipe support verification program. l The Civil discipline perforked design of pipe supperts during Trojan's original design. .This activity;was
~ performed under the guidance of Criterion C1, and nence it was proper to use this criterion'for the' review effort for consistency with the design basis. .Although the. criterion of record fer.the Civi] discipline's review was C1, the -
group was aware of the considerations contained in Criterion-P003. Issues such as temperature effects (explicitly. contained in P003, but not in C1) were addressed by~the ' civil discipline. For' example, such temperature e.ffects were documented in the pipe ancher calediations. In the example in the recommendation, temperature had no effect and- ' in this case was not documented. These documentation concerns will be addressed in the long-term pregram. 1
._1___ ._ - -
4 '. TOR l- Expansion Anchors 3
.e Deficiency: ' r Several~ support drawings indicated that baseplate: bolts can g \%
r b be either Rock bolts or expansion anchor bolts. There is no-
.evaluation.-
documentation showing the type of bolt used in.the support-The use_of expansion anchor bolts may impact *
.the design verification since different design and analysiss assumptions are required to evaluate the expansion bolts.
Response
(i i Note 3 on Drawing C-386, Sheet 2, requires the use of Rock-bolts except 1-in. diameter expansion anctors (by Phillips er USM Corporation) are acceptable substitutes in the following cases: 1
- a. Type I support using a W4x*3 with snubber cf ".ess than 3 kips.
b. Type I support using a W6x20 with snubber of.less than 1 kip. Type II support with.a snubber of less than 3 kips. if d.
- e. Type IV support with a snubber of less than 3 kips. 3 Special support with a snubber of less than 10 kips!'
The different analysis assumptions required for evaluating ( expansion anchors are significant only when the applied loads are large relative to the expansion anchor-allowables. , i l For the conditions where expansion anchors are allowed for-cases.noted. values authorized, Note 3 limits the applied loads to the low i The resultant loads in the expansion anchors are also safety, low relative to their allowables. The factors,of' 10 in allbetween cases. applied loads and anchor capacities exceeds. For the "special supports" (Case e), expansion anchors ark used when specifically called for on the support details. , Again, the resultant loads in the expansion anchors are its relative to their allowables. Expansion anchors are specified for supports SR-1228, and SR-1244. SR-1060, SR-1066. SS-1077, SS-1168, 1 The faulted (SSE) loads applied to these supports are 0.38k, 0.12k, 2.42k, 0.12k, 0.18k. and 0.40k, respectively. 'The factors of safety exceed-5 in all cases. l . l Since the loads applied to the expansion anchers are low relative te their allowables detailed analyses in the. verification calculations of the expansion anchors were not-deemed'necessary. f j r
' f[
j' .
i
)
l
- 5. T2R . Weld Evaluations 1 Deficiency:
There was insufficient input data to define the details for the qualification of welds in the evaluation of some ! i supports. Examples are: - ' 55-87: Sketch showed one weld symbol for all welds for the
', support. However, there was a note requesting additional weld information and there was no l
1 evidence that this additional information was provided. ! All welds are then assumed to be 3/16-in. i fillet all around. SS-1135 & Member weld connection was checked for adequacy. SS-1268: However, the weld pattern, size, and type were not i identified. Welds were-quali.fied based en a :ypi:11 l weld pattern'and size shown fer Generic Suppert Type i ! which is not a typical connection. For SS-1262, ,j the weld between SS-1268 member on baseplate is qualified by comparison to Type I generic support ! l which calls for a 1/4-in. all around fillet. l Response: l During the support verification program, input data on welds 1 was cbtained from civil pipe support details, procurement specifications, and field walkdowns. In the cited examples. , l additional documentation should have been provided to give-traceability to information obtained by field welkdowns . 1 ! For SS-87, field information was obtained and used but not dccumented. This lack of documentation does not invalidate j ! the pipe support verification program. For supports SS-1135 ! l and SS-1268, the calculations and review were based upon j correct assumptions and the deficiency is incorrect. The evaluating engineer located the support detail and determined the weld configuration en the. civil drawings in ene of the following ways: ' o Schedules were provided on the pipe snubber restraint and anchor support drawings, which identified the support number and referred to a specific drawing detail. The details'either called out the welding configuration or notes on the drawings, or referred to drawings which stated the welding requirements. o Where schedules were not previded, the support identification was shewn directly on the drawing detail. These details either provided the weld configuration or the drawing-notes stated the welding requirements, i-s L _se-
s.. n
.m se, ') %
.)
During the long-term. program, increased emphasis will be
. placed on the need for added documentatiorf such that the of input' data and bases for engineering judgements will be traceable.
4
~\
l'
'\
p ., 2 k l
\
s s I s. ( I
7.f. { Recommendation: ! While there is no technical impact on the results due to the criteria implementation, the consideration of Rock bolts i spacing and edge distance requirements is not consistently documented. -
Response
For the support verification program where the support is judged sufficiently close to a concrete edge, edge distance has been considered. ..For example, edge distance was considered for SS-80 and 5S-83 because they are located on top of a wall and the bolts are 6-in. away from the edge. (Calculation J-301, Rev. 1, dated August 3. 1987 evaluates and documents edge distance fer the afererentiened supports.) The support verification program also addressed Roc.t. belt spacing and the effect of overlapping cones where it was judged that the load demands and bolt spacing r.ay be such as to require such calculations. For example, Calculatio- ) J-307 addresses the effect of overlapping shear cones, , i Where, by judgement, it was noted that neither edge distances nor spacing were of concern, a conscious decision ! i was made that edge distance and spacing did not apply. This judgement was not documented. For the long-term program more detailed documentati0n of these judgements will be provided.}}