ML20092N962
| ML20092N962 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 06/19/1984 |
| From: | Reed C COMMONWEALTH EDISON CO. |
| To: | Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML20090G600 | List: |
| References | |
| 8818N, NUDOCS 8407050258 | |
| Download: ML20092N962 (57) | |
Text
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/ 'N Commonwealth Edison
/ ) one First Nttional Plaza. Chicago, Hienois i' ^ ? ^ Address Reply to: Post Othee Box 767 Chicago. tilinois 60690 ,. g June 19, 1984 Mr. R. C. DeYoung Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Byron Generating Station Units 1 and 2 Independent Design Inspection NRC Inspection Report Nos. 50-454/83-32 NRC Docket Nos. 50-454 and 50-455 References (a): December 30, 1983 letter from Cordell Reed to R. C. DeYoung. (b): March 23, 1984 letter from J. Nelson Grace to Cordell Reed. (c): May 14, 1984 letter from J. Nelson Grace to Cordell Reed.
Dear Mr._DeYoung:
This letter supplies additional information regarding Commonwealth Edison's responses to the findings, unresolved items, observations and general concerns which were identified during the
. Byron integrated design inspection.
Attachment A to this letter contains responses to the NRC questions contained in references (b) and (c) regarding issues not associated with pipe break analyses. The pipe break issues will be addressed in a separate letter ln the near future. Please address further questions regarding this matter to this office. i One signed original-and fifteen copies of this letter and the enclosure are provided for NRC review. Very truly yours, A Cordell Reed Vice President bs cc: J.G. Keppler - w/ Attachment 8818N t }S 8407050258 840619 PDR ADOCK 05000454 G PDR h
[h Commonwealth Edison
) One First NLtional Plaza. CNeigo. minois
( " T.? 7 Address Reply to: Post Office Box 767 I,Q Chicago Illinois 60690
. ,m .
): June 19, 1984 Mr. R. C. DeYoung Office of Inspection and Enforcement i
-U.S. Nuclear Regulatory Commission i
.W ashington, D.C. 20555 ,
i.
Subject:
Byron Generating Station Units 1 and 2 !' Independent Design Inspection NRC Inspection Report Nos. 50-454/83-32 ! NRC Docket Nos. 50-454 and 50-455
- i. -
References (a): December 30, 1983 letter from Cordell Reed to R. C. DeYoung. (b): March 23, 1984 letter from J. Nelson Grace to Cordell Reed. (c): May 14, 1984 letter from J. Nelson Grace , to Cordell Reed.
'l l
Dear.Mr. DeYoung:
G This letter supplies additional information regarding l Commonwealth Edison's responses to the findings, unresolved items, i' observations and general concerns which were identified during the Byron integrated design inspection. Attachment A to this letter contains responses to the NRC questions contained in references (b) and (c) regarding issues not associated with pipe break analyses. The pipe break issues will be ; addressed in a separate letter in the near future. Please address further questions regarding this matter to l this office. ; One signed original and fifteen copies of this letter and , the enclosure are provided for NRC review. 7 Very truly yours, i
/^= -
g ; i Cordell Reed Vice President ! bs r cc: J.G. Keppler - w/ Attachment > l 8818N I
,.. - , , . _ - . -- .-- . . . - . _ . - . - ~ . - - - . . - - _ . . - .-_ . . - . - _ . - - - - - - . .
s i Attachment A l lO i' BYRON IDI RESPONSES l Table of Contents Section f I -Response to NRC Letter Dated March 23, 1984 l II Response to NRC Letter Dated May 14, 1984 l ( "O l. 1-Y f. l l l O
i i f I t Section I ! l Response to NRC Letter Dated March 23, 1984
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l l , i I t t l I l e L 9 P 5 1 > f r t I b i t i h I, i t h [: I L e
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b BYRON-IDI O- General Item:
" Systematic review and corrective action program to assure that the necessary calculations in the mechanical systems discipline are identified, performed and updated as needed to support the current design.
Your response stated'that 3.11 safety-related calculations , in.the Project Management Division calculation books were j being reviewed in accordance with an approved instruction to determine.that they were technically adequate to support the current Byron /Braidwood design and to determine if the format conformed to the applicable version of Procedure GQ 3.08. You also stated that no hardware changes had resulted from these reviews, which were about 80% complete. You are requested to provide the following additional information about this review program when completed:
- 1. Describe the provisions in the review program to determine i that all necessary calculations in this discipline have been identified and performed. Indicate the number of new calculations, if any, that were performed.
I
- 2. Indicate the number of calculations, if any, where:
i () (a) Hardware changes were made (b) Reanalysis was performed - (c) Updated information was incorporated or documentation was. improved (d) Incomplete calculations had been approved (e) Additional action was taken to correct root causes or generic deficiencies In addition, with respect to the overall project, your i response noted that Commonwealth Edison quality assurance audits have included design calculations and that problems j identified in those audits were pursued to determine root causes and seek out generic deficiencies. You are requested to provide the following information with respect to previous Commonwealth Edison audits of Sargent & Lundy design calculations: I
- 1. The number of calculations audited
- 2. The number of calculations, if any, where:
Gl.0-1 i l
. . _ _ _ _ _ _ . , _ . _ . - _ _ _ . _ _ _ _ _ . . _ . , _ . . . _ _ _ . . _ _ _ _ . . _ _ . _ _ - _ . _ ~ _ _ _ _ _ _
BYRON-IDI l I (a) Hardware changes were made ; (b) Reanalysis was performed ~ f (c) Updated information was incorporated or documen- 3 ' tation was improved l
.(d) Incomplete calculations had been approved i i
Additional action was taken to correct root causes l
! (e' or generic deficiencies." l t
i RESPONSE ' The review of all safety-related calculations in the Project l
-Management Division (PMD) calculation books, performed by l i
the Byron project engineers, as described in the responses ' submitted with the December.30, 1983 letter, has been completed. This review was initiated to address the stated IDI concerns about the adequacy of the calculations previously performed [ j by Sargent & Lundy Byron project PMD engineers. The results of this review indicated that the current design was adequately f
,7- supported by these calculations. ,
i: As stated in this request for additional information, the f l objeccives of this review were to verify that the existing ! i calculations met the standards and instructions in effect I at the time the calculations were performed, and to verify i that these calculations were technically adequate. This review program did not include any specific provisions to determine that all necessary calculations by Byron PMD engineers had ! L been identified and performed. 'However, the following two ' i considerations should resolve this particular issue: l-
- l. The safety-related calculations in the Byron PMD calculation i books represent the calculations that were determined l to be the necessary and sufficient calculations required i
by the Project Manager and/or Mechanical Project Engineer, l as required by Sargent & Lundy Quality Assurance Procedure ' GQ-3.08. To address the concern, the Mechanical Project Management Engineers initiated a survey to confirm that the necessary PMD calculations have been performed. Two addi- ; tional calculations resulted from the survey. These calculations were performed to provide documentation for ! the size of the diesel oil day tank and the diesel oil j storage tank. These calculations represent the compilation { of the formally documented calculations performed by PMD ; engineers during the course of the project. As such, we believe that all necessary calculations were performed. 3
)
Gl.0-2 _ _ ~ _ . - . _ _ _ _ . _ _ _ _ . _ _ - _ . . .. _ . _ , . . - _ _ _ . . _ _ , . - _ _ _ _ . , , _ . , _ . . _ . _ , . . _ _ , , , . _ . . . . . _ , _ , _ _ - _
BYRON-IDI r (,
- 2. In order to provide additional assurance that Sargent &
Lundy has adequately addressed this issue, Commonwealth Edison Company (CECO) has retained the services of Bechtel 7 Power Corporation to perform an Independent Design Review (IDR) of three systems on the Byron /Braidwood plants. This systematic review will include an evaluation of the design adequacy and the design process 'on these systems, i and will ensure that the output documents (e.g. , calculations) meet the licensing commitments and safety-related design - requirements. The following tabulation summarizes the results of Sargent
& Lundy's review of the existing PMD calculations:
No. Of Category Calculations a) Hardware changes were made 0 b) Reanalysis was performed 0 c) Updated information was incorporated or documentation was improved 73 , d) Incomplete calculations had been approved 0 ('~) \ 2 e) Additional action was taken to correct root causes or generic deficiencies 0 f) No changes to original calculations 39 Total calculations Reviewed 112 The following clarifications to the categories presented above need to be made. Category (b) was defined as those instances where an existing calculation was found to be technically deficient or was not conservative relative to the existing design, and therefore, required a detailed analysis to verify the adequacy of the plant design. No calculations were deter-mined to be included in this category. Category (c) was defined as those instances where the existing calculation was technically adequate and supported the current design; however, revisions to the calculation format, list of references, updated infor-mation, or other related areas were made in order to improve the documentation aspects of the calculation. In no instance did these changes result in a design change or hardware change. Gl.0-3 h
+
t BYRON-IDI
- 3. In the above letter from the NRC, Commonwealth Edison I was requested to respond and supply information regarding the number of calculations for the Byron Project which ;
were examined during Commonwealth Edison Quality Assurance [ Department audits of Sargent & Lundy. Edison audits of i design calculations began in early 1979; and during the ! period February 1979 through January 1984, Edison Quality I Assurance conducted 22 audits of Sargent & Lundy which examined work pertaining to the Byron Project. During [ 12 of the 22 audits, approximately 325 calculations ; for the Byron /Braidwood project were reviewed. It should l be noted that, for the most part, calculations are applicable ! to both projects; however, some calculations were unique i to either Byron or Braidwood. All of the calculations were processed by the same team of Sargent & Lundy personnel , because the Byron /Braidwood stations are designed under [ a single project group. i e A summary of the corrective actions resulting from the ; above audits is as follows (using the same definitions as in Item 2 above): ;
/
a) No calculations were found to have problems which } resulted in hardware changes. () b) One calculation resulted in reanalysis. c) 16 calculations were noted where updated information ; was incorporated or documentation was improved.
-A breakdown of the 16 calculations is as follows:
14 required documented calculations to be originated to support the design, 1 resulted in a Design Criteria change, and lastly, 1 calculation was ' updated to show the correct load value. d) No cases were noted where incomplete calculations , had been approved. i e) As a result of the 17 calculational deficiencies ' referred to in 3(b) and 3(c) above, Sargent & L Lundy performed extensive reviews for two of the - deficiencies. The action taken to prevent recurrence t included revisions to procedures. ( b I Gl.0-4 l
BYRON-IDI i i i Finding 2-1: Diesel Engine Air Intake
' "You are requested to provide for our review a copy of i the documented walkdown which concluded that there are -
no additional non-safety-related components that will impair the function of the intake line." :
RESPONSE
A copy of the documented walkdown (dated 11-15-83) is enclosed. l i i i k- T / - l l i I i l s
?
l l l i l l t r
- O f F2.1-1 l
ATTACHMENT TO FINDING 2-1 UNIT 1 AUXILIARY FEELFATER DIESEL AIR INTAKE LOCATION REVIEW f) #
%/ LINE NO.:
1DOBlA14, Piping Subsystem' lD025 MATERIAL: A-106 GrB; 0.375 inch wall ROUTING: The air intake line runs a short distance in the Auxiliary L Building and. penetrates "L" line wall into the Turbine Building at EL 391'-10". The line turns upward immediately and pene- ', trates level 401. At level 401 the line extends upward for approximately 6 feet with two 900 elbows and a debris screen i welded onto the end of the pipe. Intake air for the Auxiliary fecdwater diesel is taken from the Turbine Building air volume. l The total length of intake pipe is approximately 30 feet. P PIPE SUPPORT: The pipe is seismically supported in the Auxiliary and Turbine Buildings by an anchor provided at "L" line wall. One pipe support located in the Turbine Building is provided for support of the j ') debris screen and doub'.e elbows. i) TURBINE BUILDING INTERACTION: ' l Below El 401: The intake line runs vertically from El 391'-10" to level 401 and h located 2'-10" from the edge of "L" line wall. l This area is free of larger components and no hazard to pipe integrity exists. The pipe is physically above most nearby components. i Above El 401: The intake line ends approximately 2'-10".from "L" line wall. In plan, the pipe is located West of the CO2 ! storage tank ~and North of the Turbine Building elevator shaft. All piping within eight feet of the intake line is 6" or less in diameter. The nearest larger diameter pipe is 10G21C24. This OG line r is located above the floor with centerline at El 413'-3". In plan the G line is located approximately eight feet East of the intake line. The elevator shaft structure would provide ample protection to shield the intake line from a swinging QG ! line. Two cable trays run vertically overhead but would not be capable of jeopardizing the integrity of the intake air line. Page 1 of 2
ATTACHMENT TO FINDING 2-1 , ( Cont ' d) CONCLUSION: t The Auxiliary feedwater pump diesel intake air line is a ['
'- reasonably short run of pipe. The air intake pipe is routed to provide a minimum exposure to non-safety related components capable of damaging it. The air intake location is acceptable. .
UNIT 2: The Unit 2 air intake location is similar to Unit 1, however the installation status is incomplete at this time. An , inspection of area indicated a similar piping and equipment I arrangement. ; REFERENCES,:
- 1. Piping analytical drawing 1D025, sheets 1 and 2, Rev. 1
- 2. Composite drawing M-330, Rev. T
- 3. Photographs taken 11-15-83. .
t -- Prepared //'/f'88 fm Reviewe ll-f5 Approved / /-/5'-83 i l 1 l l O Page 2 of 2
i I B/B 1
'/ Finding 2-3: Basis for Time Delay "You are requested to describe the basis for determining that the hydraulic transient, considering worst case conditions l such as minimum technical specifications condensate storage tank level, high condensate water temperature, simultaneous pump start and runout flows, does not result in pump trip or undesirable addition of relatively impure ESW system water to the steam generators."
. RESPONSE ,
i The auxiliary feedwater system preoperational test (2.3.10) included a section to determine whether a hydraulic transient on a simultaneous pump start would result in a sudden decrease in system pressure causing either a pump trip or the opening of the essential service' water supply valves. The test was conducted under the following conditions: l
- a. Condensate storage tank level was verified to be at i 197,500 gallons. (The test procedure requires the condensate storage tank level to be at the minimum e technical specification level of 200,000 gallons,
./) + 25,000 gallons.) ,
V
- b. Suction pressures initially recorded, ,
- c. AFW pumps started individually and suction pressures recorded, and
- d. AFW pumps started simultaneously and suction pressures recorded.
The maximum suction pressure transient recorded was a 10.2 psi drop in pressure for the motor-driven pump during the simultaneous start (initial suction pressure approximately i 20 psig); however, the suction pressure stabilized around 17 psig. The setpoint for switchover to essential service water ! is approximately 14.1 psia (1.22 in. Hg. Vac.). The ambient water temperature was not recorded since this parameter does not have any effect on the results as the difference in water ' vapor pressure between 40*F and 100*F is negligible. The pumps were not operated at a runout condition since runout , orifices are installed in ea,h auxiliary feedwater supply ' t I line to prevent this situati a from occurring. Based on the test results, it has been sucessfully verified l l l that a simultaneous pump start does not induce a hydraulic ' transient that will either open the essential service water s i supply valves or trip the pump due to a sudden loss of suction pressure. The test also verified that a time delay on the pump trip circuit is not required. F2.3-1 ,
y, ., , BYRdN- ID t
.(
l' .. ( j ' l Finding'2-13: Design Criteria Updating t "We. recognize.your statement that the design criteria are intended to guide design ef forts <in t 'ae initial phases I of design. However, we do not unde'. stand, bow development ; of a status list, by itself, will icovi'do an effective : sa,feguard .to assure that personnel 'perfor' ming safety-related l activities will not be mislead by obsolete information, l particular'ly in view of the' following: j
- 1. The design _ criteria appear to be important and useful documents with widespread distribution, including i availability to the plant staff and design engineers. ;
- 2. They are.c ntrolled design documents and this creates l a tendency to assume they are kept correct and current. }
- 3. Since they are generally not being updated, many will i contain obsolete and potentially misleading information. [
t Accordingly, you are requested to describe your plans j for additional measures, such as stamping all copies, to l assure that personnel performing ssfety-related activities 1 are not mislead by obsolete informatic .." l RESPONSE l Based on a review of the 30 Byron /Braidwood safety-related L design criteria documents, the March 23, 1984 status report : classified each design criteria under one of the following i l ! categories: Number of ; Design Criteria I i
. Design - The design criteria is correct, 18 l reflects the current engineering ;
l design, and can be used as a ; i design document. The design criteria is not 12 Information - 100 percent correct. It does, , however, provide a design basis ' for reference but cannot be used r as a design document. i The design criteria does not reflect 0 Obsolete - ' the current design and cannot be used as a design document. F2.13-1 !
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I i BYRON-IDI l P v While we still believe that the identification of the document status through a status list is adequate, those design criteria ' classified as "inforn.ation" (or if categorized as " obsolete" in the future) will be appropriately identified on each page and redistributed in accordance with the project distribution list. We believe we have addressed all of your concerns in , regard to this item. , i P
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I I F2.13-2 ,. l
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BYRON-IDI r
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Finding 2-18: Flooding Analysis ,
"For the RHR heat exchanger cubicles (Item B in your response),
you are requested to provide the load due to the maximum ! flooding level, the design live load for the floor and ' the ultimate capacity of the floor."
RESPONSE
In the RHR heat exchanger cubicles, the maximum flooding level is 101 inches, which results in a dead load of 0.525 KSF. The occupational live load used was 0.05 KSF and the total factored uniform design load for the SSE level load combination, of . I which the flooding is a part, is W = 4.91 KSF, The allow-able ultimate capacity of the floogSEbased on the strength l design method in ACI 318-71, is 6.65 KSF. Therefore, the ; floors in question have ample capacity for accommodating the flooding load.
?
(N r s_/ i t
)
l ! ! u l F2.18-1 l l
6 l i BYRON-IDI ! i O.- Finding 3-2: Functionality Criteria "Your response indicated that Class 1, 2, and 3 stresses - are being utilized to evaluate for functionality. However, , it is not clear how the analyst decides whether or not the i Class 1, 2, or 3 stress results from PIPSYS are acceptable i per the functionality requirements when certain stress i indices and stress intensification factors (specified ; in Tables 1 and 2 of the General Electric report) are ! higher than those listed in the Winter 1979 Addenda of l the ASME Code which is the basis for the PIPSYS calculations. l You are requested to explain further how these decisions ' are made." RESPONSE [ The following is a detailed discussion to illustrate that f the qualification methods used by Sargent & Lundy to satisfy , functional capability requirements are acceptable. This is i true regardless of which addenda of the code is utilized for f stress indices or stress intensification factors. I t I. DIFFERENCES IN CLASS 1 STRESS INDICES i-(f A comparison of the stress indices (Table F3-2.1 attached) illustrates that the stress indices specified in Winter ! 1979 Addenda of the code (basis for PIPSYS) are equal l to, or more conservati"e than, those specified in Table 1 of the General ElectJic report (NEDO-21985) except for the ! indices for branch connections and butt welding tees. Bi and B2 i L As stated in our original response of December 30, 1983, a portion of the Class 2 and 3 piping is evaluated for func- - tional capability using Class 1 analysis rules as delineated j in Section 2.2e of NEDO-21985. (Namely: Eq. 9 of NB-3600.) . For all cases where this approach was utilized, the allowable ! ! stress limit was considered to be 1.5 S The other two ! methods which can be used for evaluatind. Class 2 and 3 piping ; for functional capability are the scanning method and detailed } hand calculations per NEDO-21985. All three qualification l l methods are outlined in EMD TP-2, Rev. 4 (EMD-046032) , as , statod_in our original response. General Electric report (NEDO-21985) allows the use of ; 2.0 S as the allowable stress limit for the calculated ! stresEes for the branch connections and butt welding I tees using the higher values of By and B 2 indices. ; i The ratio of allowable stress limit recommended by the General Electric report (NEDO-21985) to the allowable ! _(Y
'~ 2.0 S y stress limit used by Sargent & Lundy is
- S equal to 1.33. . y. j I F3.2-1 l i
. _ ~ . . . _ , . . _ _ _ _ _ . _ . . _ -... _ _ _ _ _-_._ _ _ _ .. _ __ _ _ _ _ _ _ _ _ ._
BYRON-IDI By (S78) B 2 (S78) 4 The ratio of B y indices and B 2 ces B (W79) B (W79) v 1 2 for the worst case, is also equal to 1.33. Therefore, j the use of 1.5 S by the analyst, as the allowable stress limit will assure, functional capability, in accordance with our licensing commitment regardless of which addenda i of the code is utilized for the stress indices. II. DIFFERENCES IN CLASS 2 & 3 STRESS INTENSIFICATION FACTORS A comparison of the stress intensification factors (Table F3.2-2 attached) illustrates that the stress intensification factors specified in Winter 1979 Addenda of the code (Basis for l PIPSYS) are equal to or more conservative than those spe- , cified in Table 2 of the General Electric report (NEDO-21985) ! except for welding elbows or pipe bends and welding tees. As stated in our original response of December 30, 1983, , most of the Class 2 and 3 piping is evaluated for functional capability using the stress scanning method, which is a very conservative approach. This means that the Service Level C PIPSYS stresses are scanned to assure that they i do not exceed the Service Level B allowable stress limits. ; (Namely 1.2 Sh). The other two methods which can be used for evaluating Class 2 and 3 piping for functional capa-
~N bility are the detailed hand calculations per NEDO-21985 - (d or Class 1 analysis rules. These qualification methods are outlined in EMD TP-2, Rev. 4 (EMD-046032), as stated in our original response.
The General Electric report (NEDO-21985) allows the use of 1.5 S as the allowable stress limit for calculated stresses forY all piping components. ; The most conservative ratio of allowable stress limit recommended by the General Electric report (NEDO-2198 5) , to the allowable stress limit used by Sargent & Lundy l 1.5 S y
- is equal to 2.0. This ratio addresses the
,1.2 Sh.
most conservative assumptions of material properties. The ratio of the stress intensification factors for welding ; 1.3 elbows and pipe bends 2/3 (S78) is 1.93 0.75 x 0.9 ! (W79) _ h 2/3 , ,
*S h is the allowable stress limit at 557' F. S is the yield stress value at 557' F. The ratio was determiEed for the J
worst case. F3.2 :
BYRON-IDI (} and the ratio of the stress intensification factors for welding tees .llf is 1.28. fh 8 Therefore, the use of the stress scanning method will assure functional capability in accordance with our licensing com-mitment regardless of which addenda of the code is utilized for the stress indices. We believe that we have addressed all of your concerns in regard to this item. a [ i i i 5 I s i b l P F3.2-3 i i l _. . . ~ . _ _ . ~ . _ - . . _ . - - - _ . _ . . . _ . . . . - - , _ _ _ . - _ - . - . _ _ _ _ _ _ . _ - _ . -
3YRON-IDI TABLE F3.2-1
.y }
() . COMPARISON OF C1. ASS 1 STRESS INDICES BETWEEN THE 1979 WINTER ADDENDUM OF THE 1977 ASME CODE AND THE 1978 $UMMER ADDENDUM OF THE 1977 ASME CODE 4 1979 WINTER ADDENDUM 1978 SUMMER ADDENDUM OF 1977 ASME CODE OF 1977 ASME CODE FITTING (BASIS FOR PIPSYS) (BASIS FOR NEDO-21985) Straight Pipe 5 = 0.5 3 B = 0.5 3 B = 1.0 B = 1.0 2 2 Curved Pipe or Butt 5 = 0.5 3 B g = 0.5 max U21 ding Elbows ! I*
- 1.46 max - > 1.0 2" h 2/3 B B 2" h 2/3
.Pranch Connections B g_ a 0.5 B g = 0.5 unless B 2r rB 2b "
then By = 0.67 B = 0.75C B 2r 2r 3 1.0 h = 0.75C2r * ; A r')
-" il 1.0 (R[ f e' C
2r
=
0.8 ([)} rj m) C 2r = 0.8l 7 )l (r l (m [l 1.0 !
~ ,A .B 2b
= 0.5 C 2b 21.0_ B2b=. 0.5 C 2b l "
.t )
! ! fT' fR fr' fT (R fr') * *
--" j ~> 1.5 l(3)lfe')'r) (
3l[*)I l(m)"Y {[d')llfr')
= C l C
r/ l r) ('p ) = 3r)l[")l( m) l
. Butt Welding Tees B y
= 0.5 Bg = 0.5 unten B2r or B 2b " I then By = 0.67 fA,) fR ;
B 2b
= 0.4 l- 1 1.0 B 2b =0.4{ lh IR*) !
fR)
.B = 1 1.0 B 2r = 0.5l T,j}
l b3 ,. 2r 0.75lT,jl ( ( Butt Welding Reducers B = 1.0 B g = 1.0 3 B = 1.0 B = 1.0 2 2 Girth Fillet Weld to B = 0.75 B = 0.5 + y 3 *
.S:cket Weld Fittings, src. 5 =
2 1.5 B = 1.0 2 i l
~ .f] .
.{ \
-- (._/ l
)
F3.2-4 [ i
.x EYRON-IDI t-*
.y i '
l
, l TABLE F3.2-2~
i l
/ -
COMPARISON OF CLASS 2 AND 3 STRESS INTENSIFICATION FACTORS
\. ; ,
EETWEEN TR' 1978 WINTER ADDEMDUM OF THE 1977. ASME CODE AND TRE 1978 SUlstER armyunimi 0F THE 1977 ASME CODE l
. . E l l
1978 WINTER N 1978 SUMMER ADDENDUM OF 1977 ASME CODE OF 1977 ASME CODE 7ITitNG (BASIS FOR PIPSYS)* (BASIS FOR NEDo-21985)* i l i~ Straight Pipe Bett Weld 1.0 1.0 1.3 ' 0.9 Welding Elbow or Pipe Send 2/3 2/3 ! D.75 ( h h ! EFT 0.751 0.75i t
.UFT 0.75i 0.751 l WDT 0.75i 0.90i ,
Fillet Weld Joint, Brazed 0.75(2.1) = 1.58 1.0 l Joint, Etc.
^
Reducer 0.751 0.75i V i
*All Values Must be 1 1.0 )
I t [ l i k 1 l i i i ( i, i r i I r F3.2-5 l
~ _. _... . _..__,.__.,-.___.___ _ __. _.._.. _ ,, _ .__ __ _ _.,..___,__.. __,_.....____, .__ _ .-.- _ _ , _ _ ,
BYRON-IDI If' ( Finding 3-6: Pipe Support Added Mass ,
"Your response stated that one example cited in the report !
did not, in fact, violate your criteria with respect to l added mass and the other example exceeded the criteria by an insignificant amount (53 lb vs. 52 lb criterion). ; However, your responte did not address the overall concern. . You are requested to confirm on a systematic basis that ! your procedures for added mass are being uniformly followed or, if not, there is no significant effect on the analysis ! results." 1 RESPONSE i A criteria for inclusion of the support added masses in piping analysis does exist as stated in our previous response. Devi-ations from the existing criteria will not adversely affect . the validity of the analysis results based on the inherent ! conservatism in the total design process. This has been demonstrated by conducting a technical evaluation ; of previously completed piping analyses. The sample for this ; evaluation was determined utilizing the military standard . statistical sampling scheme (MIL-STD-105D). The evaluation [~)
\- has been completed and was documented-on June 15, 1984.
[
- Furthermore, to ensure uniform application in the use of support added masses in the future, a retraining program is being i 4
- established. Detailed classroom and hands-on instruction are being conducted with emphasis placed in this area. !
f i I i
?
t i t i I i i I i l- ! l i F3.6-1 I i r
BYRON-IDI i,
;f'% t
- (_) Finding 3-7: Overlap of Plate "You are requested to provide a copy of the following docu-ments for our review:
- 1. FCR F-9079
- 2. The backup calculation
- 3. Revision C to Support Drawing lAF03009R"
RESPONSE
A copy of each of the requested documents is contained in the attachments, and each of these is discussed below: L
- l. Attachment A, FCR F-9079: ,
^ f Byron FCR F-9079 (dated 4-16-81) was written against Revi- i sion B of support drawing 1AF03009R to have a 1/4-inch overlap on its embedded plate. This FCR was picked up on Revision C of the drawing and issued on 12-23-81. Subsequently, this FCR was closed out on 1-9-82 as shown on Part D of the first j - page of the FCR.
' 2. Attachment B, Structural Division Calculation:
A copy of Calculation No. 13.1.2, Pages 521, 522, and 523, is enclosed. Page 521 is the approval page of the calculation for Drawing 1AF03009R, Revision C, which incorporates FCR F-9079. This page contains the preparer's and reviewer's signatures and dates, and also shows approver's signature and date (top right side of page). i Page 522 shows that design load has been changed per Revision C of the drawing from 1753 pounds to 1056 pounds and the review : method has been identified (see " Remarks" column and notation for
" Remarks" column).
Page 523 is for backup calculations for the expansion anchor plates and bolt as indicated on Page 522. The supporting calculation for Revision C is shown on lower portion of this ca'.culation sheet with Revision C indication in a box. The calculations shown on upper portions of this sheet are for previous revisions.
- 3. Attachment C, Drawing lAF03009R, Revision C: l
[(,,,) Drawing 1AF03009R was revised per FCR F-9079 and to reflect the formal analysis loads. l F3.7-1 I _ _ _ _ . . ._ _ _ _ _ _ , _ ... _ __._ _ - _ _ _____._J
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SD&DD REVIEW RECORD FOR h MECH. COMPONENT SUPPORT DRAWING # M- /A F o3 oat e . l t l . DWG.REV. STATUS PREPARED BY DATE REVIEWED BY I DATE ; _. 4 H f& , /2.l//ff'l ,
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BYRON-IDI Finding 3-16: U-bolt Analysis "You are requested to describe the criteria employed for U-bolt design and confirm that they have been uniformly applied."
RESPONSE
The U-bolt review procedure is defined in the Westinghouse , Byron Pipe Support Design Reference Manual, Byron 1 and 2 (Revision 0, dated 11-22-63), and is summarized below. The criteria for evaluating b-bolts, for pipe sizes up to 2 inches, is based on results of tests on U-bolt samples. The vendor allowables, as published in applicable load capacity data sheets, have been shown to be conservative by results of failure tests for sizes of 1/2 inch to 2 inches for tension and side loads. Westinghouse has reviewed these test results and established allowable U-bolt loads by applying factors of safety to the test results. Safety factors of 4.0 (normal / upset) and 2.0 (faulted) are used to define test-based allow-ables. This test data was collected for ITT Grinnell Figure gs 137H U-bolts. Since the Byron Project uses both the ITT Grinnell ( ,) J-bolts and Elcen Figure 68A U-bolts, a comparison was made of U-bolts for various pipe sizes and shows the Elcen U-bolt to be identical in both material and bolt dimensions. The test data is therefore equally applicable to the Elcen U-bolt. For U-bolts for pipe sizes greater than 2 inches, the manu-facturer's load capacity data sheets are the basis for the acceptance criteria. The acceptance criteria for U-bolts for piping 2 inches and smaller, is based on test data. It is Westinghouse policy for the Byron Project to select U-bolts in the design phase using the vendor supplied load capacity data sheets. Allowables based on the test results discussed above may be used in the , as-built reconciliation phase. Westinghouse reviewed pipe support calculations to verify , that the criteria was correctly applied. This review covers analyses that apply to 62% of the 820 U-bolts in the Unit 1 containment and applicable scope in the auxiliary building. In all cases, the reviewed U-bolts inat the specified criteria.
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l BYRON-IDI : j. Finding 4-1: Transverse Wall Load Criteria '
"For the original Byron design, you are requested to describe ;
L how the design of peripheral walls, perpendicular to the !
- transverse load, actually considered the loads coming !
on to the outside of the walls from wind and tornado (wind l and differential pressure). [ l IFor a given wall loaded with a transverse wind load, dracribe j how the transverse load at the cente'r of a wall betwren t two supporting orthogonal walls (shear walls) is carried i out through the wall to the orthogonal walls which carry the shear force as in-plane shear in a normal shear wall : concept. Describe how Sargent & Lundy determined the load i t- capacity of the peripheral wall loaded transversely and ! converted these forces to stresses and then to reinforcing j i . areas, if required. Describe how these same loads are ; addressed in the final load check. The use of diagrams l to illustrate the details is recommended." j l t
RESPONSE
t As noted in Subsection 11.6.1.3 of the Structural Project
# l
- Design Criteria DC-ST-03 B/D (IDI Reference 4.31) , the tornado .
loads in Safety Category I structures were obtained by static analysis utilizing the DYNAS lumped mass model. Various positions ' i of the tornado were investigated to determine the maximum
. shear wall forcesi and, for each position, the location and magnitude of the resultant tornado surface pressure was deter- ,
mined and transferred at the mass center at each elevation. ' The maximum shear wall forces thus obtained were compared [ to those forces due to seismic loading and the walls were ! designed for the most critical loading combination. Seismic o load governed over wind and tornado in the shear wall design. j For a wall loaded with a given transverse load, the wall was : analyzed using a " strip method" (i.e., the most c,ritically loaded strip of the wall was isolated and treated as a " wall strip" which , j spans either vertically or norizontally and is supported at the
; intermediate floor slabs or adjacent walls as applicable, see
! Figures F4.1-1 and F4.1-2). In those cases where a vertical i strip was used, the transverse load was applied to the wall ! strip and transferred to the supporting intermediate slabs. ! The load is then transferred through the slabs to the orthogonal ! shear walls and down to the foundation. In those cases where a horizontal strip was used, the transverse load applied to l the shear wall was transferred directly to the supporting i I orthogonal walls.
)
F4.1-1 i l i _ - - - - _ - a
i BYRON-IDI ; i [ For the case of transverse wind and tornado loading on walls, i the controlling condition was tornado pressure plus a tornado-generated missile. The original wall design included an analysis which demonstrated that any peripheral wall was adequate to absorb this load and transfer it via a controlling strip to : the lateral load resisting system. The transverse flexural steel area required for each wall , strip was calculated using the flexural strength provisions of Chapter 10 of ACI 318-71. The transverse shear friction ' i steel area required was calculated using'the shear-friction provisions of Chapter 11 of ACI 318-71. Transverse loads are addressed in a similar manner in the final load check. t L I i I-i F4.1-2 ;
i
.. s BYRON-IDI ,
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E1JON-IDI
i Finding 4-2: Shear Friction Method (Gi "Our understanding of your response is as follows:
In the design of walls, the transverse shear stresses were computed from the transverse loads such as dynamic soils and water pressure, wind loads, tornado effects and horizontal seismic and compared to a value of V such as determined by Section 11.4 of ACI 318-71. If thE actual shear stress was less than or equal to the allowable, no ties were added. If the actual shear stress was greater than the allowable stress, ties were added to carry the stress exceeding that. carried by the concrete such as required in Section 11.6 of ACI 318-71. In either case, when the vertical reinforcement was sized, the value of the actual transverse shear was combined with the in-plane shear and the resultant used as the total shear to be carried. The area of steel was computed from the resultant enear value by using the shear-friction concept. The result is that in all cases, there is a margin in the vertical reinforcement relative to carrying in-plane shear loads. i ) You are requested to verify that this understanding is correct or, if it is not, describe how the actual design was executed. You are also requested to provide the details of how the transverse shear was combined with in-plane shear. The use of diagtams to illustrate details is recommended."
RESPONSE
4 Your understanding of our response is correct relative to the treatment of transverse shear stress, but differs somewhat relative to the method used for sizing. vertical reinforcement. In the design of the walls, the transverse shear stresses were computed from the transverse loads and compared to a value of v c determined using chapter 11 of ACI 318-71. If the actual shear stress was less than or equal to the allowable, no ties were added. If the actual shear stress was greater than the allowable stress, ties were added to carry the stress exceeding that carried by the concrete. In the design of the vertical reinforcement, however, the steel area required for transverse shear loads was added directly to the steel area required for in-plane shear loads. Both the transverse shear steel area and in-plane shear steel fT area were determined using the shear friction concept. V F4.2-1
i l BYRON-IDI .
)
Finding 6-3: Bases for Setpoints : Finding 6-7: Setpoint Accuracy Requirements ! Findire 6-8: Basis tor Reset Value "IEEE 279-1971, which is invoked by 10 CFR 50.55a(h) requires t
'A specific protection system design basis shall be provided l
.... The design basis shall document as a minimum .... l
, (2) the' generating station variables .... required to t t be monitored in order to provide protective actions .... ; . (4) prudent operational limits for each variable .... l (5) the margin .... between each operational limit and [ the level considered to mark the onset of unsafe conditions i
.... (6) the levels that, when reached, require protective ;
actions, (7) the range of transient and steady state conditions [
.... throughout which the system must perform, (8) the i malfunctions .... for which provisions must be incorporated to retain necessary protective actions and (9) minimum performance requirements including .... response times :
.... accuracies .... ranges .... of the magnitude and I
rate of change of sensed variables ....' Our. understanding of the intent of your responses is as follows: !
- 1. For all safety-related instruments in the Sargent
() & Lundy scope, you will assure that documented bases have been provided as required by IEEE 279. l t
- 2. For cases judged to be complex, you will assure that [
calculations have been provided to support the selection i of setpoints. ; i You are requested to indicate whether or not this under- I standing of your intent is correct. If it is not, please ; explain what is different in your intent." ! [
RESPONSE
The instrument data sheets for safety-related instruments ! I. ' document the bulk of the design basis information and, in conjunction with other information denoted below, comply ; i with our interpretation of IEEE-279 as follows (item numbers ; p correspond to referenced sections above): j (2) The generating station variables that are required i 4 to be monitored in order to provide protective actions , are determined during the design and review process , of a carticular system as documented on the system's I Piping and Instrument Diagrams (P&ID), Control and i Instrument Diagrams (CID) and the Logic Diagrams (LD). [ q The documentation for this activity is thus contained [ on the P&ID's, CID's, and LD's as well as the instrument index and the instrument data sheets. (See attached [ i i ! F6.3-1 ;
~ . _ _ ..-. _ __ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ . - . _
BYRON-IDI , i example of an instrument data sheet, refer to encircled Item 2 in boldface type). ; (4) The operational limits for each variable are documented on the Instrument Data Sheet. (See attached instrument data sheet, refer to encircled Item 4). i (5) The margin, with appropriate interpretive information, between each operational limit and the level considered to mark the onset of. unsafe conditions are determined from existing calculations, design drawings and/or vendor supplied component design data. The margin on the data sheet is the relationship between the instru-ment's range and the setpoint. (See attached instrument data sheet, refer to encircled Item 5). (6) The levels that, when reached, will require protective actions are determined using the design input discussed 1 in Item (2) above and documented on the data sheet. (See attached data sheet, refer to encircled Item 6). 1 (7) The range of transient and steady state conditions of the power supply and the environment during normal, (' abnormal, and accident circumstances throughout which . ( the system must perform are contained in the procurement specification which is referenced on the data sheet. In addition, the data sheet calls for seismic and/or Class lE qualifications, all of which are documented , in the EQ binders. (See attached instrument data ; sheet, refer to encircled Item 7). (8) Malfunctions, accidents, or other unusual events (for , example, fire, explosion, missiles, lightning, flood, ! earthquake, wind, etc.) which could physically damage protection system components or could cause environmental changes leading to functional degradation of system performance, and for which provisions must be incor-porated to retain necessary protective action, are i reviewed at the time a particular condition is identified (see also Item (7) above and encircled Item 7 on For instru-ment data sheet, for environmental conditions). example, cubicle flooding was reviewed and the review of these flood levels (as related to instrumentation) was incorporated into the flooding calculation. , (9) The required instrument accuracy of an instrument 1 is determined from an engineering assessment of infor-mation contained in the system calculations, design r- drawings and/or vendor supplied component design data. ( )f The setpoint accuracy required is then used in the review of vendor catalog information to establish F6.3-2
- . _ , , - . . .._.,,m.., ,..,,.__,,._,_.._.,m_,...,_,,,,_ , _ , _ . _ _ . , _ - , , . , , - _ _ ~ _ , -
t I i BYRON-IDI i Ii i jh !
\~ the instrument selection. The selected instruments !
are documented on the instrument data sheet by manufac- ' turer cad model number using vendor standard designs 9- which envelop the system operating requirements. (see ; attached instrument data sheet, refer to encircled Item ! 9). Past experience has shown that response times and '
- ranges of the magnitudes and rates of change of sensed variables have had little effect on the instrument ,
selection. Therefore, these parameters are not reviewed i unless a specific application is needed. I II. Complex setpoints for safety-related Sargent & Lundy instru-ments have been identified via a documented memo and calcu-lations exist for these instruments. t
?
f l I E e s i F6.3-3 ! i
. - _ . . . - - . ~ . . - . - . . _ - - - . - . . . - . . _ .. . .
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() t Unresolved' Item 2-1: Diesel Engine Exhaust Pipe i "You are requested to describe the basis for determining that tornado misbiles will not crimp the auxiliary feedwater ; pump diesel engine exhaust stack completely closed. Include ' a discussion of the potential for damaging the hinged cap in such a way as to incapacitate the pump <"
RESPONSE
Sargent & Lundy performed an analysis to determine the effect I of tornado missiles impacting the auxiliary feedwater pump l diesel exhaust stack. Tornado missiles defined as Spectrum j II missiles in Section 3.5.1.4 of NUREG-0800 were postulated. i The calculations demonstrate that crimping the exhaust pipe i due to a tornado missile impact results in a maximum 60% l reduction in flow area at the roof interface. Missile impact i on the 1/8-inch thick aluminum weather cap will not affect the flow area since the cap will either be destroyed or blown out of the stack by the exhaust pressure. The 60% reduction in flow area of the exhaust stack at the roof interface wil () not incapacitate the auxiliary feedwater pump diesel.
- i. 1 1
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P BYRON-IDI
~/ i Unresolved Item 4-2: Top Reinforcing for Slabs "The design procedure outlined in the response, if applied for all slab designs on the Byron project, would yield conservative steel areas (bottom steel) for midspan positive moments. You are requested to indicate whether this concept Indicate if the procedure was used throughout the plant.
described in the original answer for supplying negative steel (top steel) at each slab boundary was used throughout the project and if so what portion of the maximum moment for a simply supported case was provided in negative moment capacity at the boundaries. Indicate, by providing detailed references to written documents, how these project-wide concepts (if used) were provided to individual designers in the way of instructions or procedures. If no project-wide concept was applied, indicate what technique was used in providing slab reinforcing based on varying boundary conditions."
RESPONSE
As noted in our original response, negative moment steel equal to that at the continuous support was provided at the junction
-[']
(_,- to the wall of slab 4AB53. This is a conse'rvative design approach which was used for all slabs supported by walls. This typical detailing is shown as Slab Support Types 5 and 6 (see Figures U4.2-1 and U4.2-2) on Structural Drawing S-473 which was included in project Specification F/L 2722 and initially released to the appropriate contractors on August 7, 1974. These standard details are specified for construction at all slab-to-wall junctions and, because they are standard details, no other reinforcing arrangements could have been used. Thus, their use is a project-wide concept and did not depend on the judgment of the individual designer involved. V i l U4.2-1
i BYRON-IDI : r r FIGUPE U4.2-l* [ l I i [ i I i DowgL TO M ATCH i n SLA6 RE.INF:. [ SKEY ! MIN. [ Ld. . Lc.t . j
~
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't OO'NE L TO MATCH SL A6 REINF. ,
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I L BYRON-IDI ! ('~) (_/. l Unresolved Item 6-2: Pressure Switch Qualification l r
"Wher, review of the pressure switch qualification data I is complete, you are requested to provide a description i of the basis for acceptance. If qualification by similarity l
-with the tested switch is used, describe the rationale '
for using similarity." t i
RESPONSE
The qualification program for pressure switches 1PSL-AF051 ( i and 1PSL-AF055 is described in the following discussion. l, The original pressure switch specified for this application i was United Electric Model J-302-S156, which is a metal bellows { type sensor. Later, due to operating requirements, this switch 4 was changed to Model J-302-552, which is a teflon diaphragm i type sensor. Since the test program for Model J-302-S156 I was in progress, it was decided to continue the test and evaluate ! the acceptability of the report upon receipt. ; i Since the time of the IDI, the report has been received, reviewed, , and found to be unacceptable for qualification of Model J-302-552. ! O' Due to internal mechanism differences between the two switch models, seismic testing of Model J-302-552 is required and in progress. 1 Since the switches are located in a mild environment, the ! environmental qualification will be by a combination of sim- l ilarity between the tested and supplied switch models for ! parts that are identical, and a material analysis for parts that are difterent. , t i l l 1 I i I I I i f I U6.2 ' 1 i l l I _ _ _ ~ __ _ _____. .
...__._._..,.,.J
i I i I i i Section II ! t Response to NRC Letter Dated May 14, 1984 : I i, r P I e I I l I r I i j 4
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i t i i f F I r l l r I r I I i L l i I l I l i I L _ . - . . - . _ - - . ~ - - . . - - . . - . - . _ . . - . . . . . . _ - , . _ _ ,_ _ - _ _ . - , . _ _ . - - - . - -. --
M BYRON-IDI
~
Finding 2-1: Diesel Engine Air Intake "Our March 23, 1984 letter requested a copy of the documented walkdown which concluded that there are no additional non-safety-related components that will impair the function of the intake line. Please indicate how the turbine building crane was assessed relative to potential failure during a seismic event and consequential damage to the diesel intake line, unless this is covered in the documented walkdown."
RESPONSE
The turbine building crane was not included in the subject walkdown since only the non-safety-related equipment in the immediate vicinity of the auxiliary feedwater diesel air intake line was reviewed. The diesel air intake line is located at grade elevation 401 feet while the turbine building crane is located above the main operating floor (at approximately elevation 500 feet). The turbine building crane rail girders are designed to withstand SSL loads. The bridge is normally parked at the south end of the turbine building during power operation, corresponding to a horizontal distance of nearly 300 feet from the diesel air intake line. In the unlikely - ~ event that the bridge fails during an SSE, the auxiliary feed-water diesel air intake line will not be affected. i i i i 1 O F2.1-1
BYRON-IDI I; O N,) Finding 2-4: Time Delay on Logic Diagram
"(1) What system ensures that logic diagrams will be revised :
when the associated schematic diagram is revised? (2) Please indicate the systems associated with each drawing referenced in FCR Nc. F21,265." l RESPONSE g (1) Project Instruction PI-BB-58, " Incorporation of Electrical ' Schematic Control Diagram Changes Into Control Logic Diagrams," has been written to formalize the engineering [ practice described in the previous response to this finding, l t This instruction requires that changes to schematics are reviewed against the logic diagrams and, if applicable, requires that logic diagrams are revised. In addition, ' Sargent & Lundy is conducting a review of the logic diagrams ; against the schematics. This review is scheduled to be completed in mid-July 1984. [ (2) The following drawings are referenced in FCR-21265 with the respective systems requiring revisions: (S
\
1-4030 OG01 OG (Off-Gas) 1-4062B WO (Chilled Water) 1-4062C WO (Chilled Water) ' 3-4062E Bill of Material l-4062G WO (Chilled Water) 1-4062H WO (Chilled Water) 1-4600E FW (Main Feedwater) 1-4030 SX01 SX (Essential Service Water) 1-4611B AP (Auxiliary Power 480 V and above) 2-4045B EH and TG (Turbine EHC and Turbine Generator Auxiliaries) l b (-)
%J F2.4-1 ,
l BYRON-1DI i (~ L N')
,g FiEding 2-8: Missing Calculation For Containment Spary m
We believe that the FSAR statements are design bases and '
..a licensing commitments. Our letter dated March 23, 1984 (page 1 of enclosure) requested you to describe the provisions in your review porgram (of Project Management '
Division's calculations) to determine that all necessary calculations have been identified and performed. Please indicate how you ensured that necessary calculations were ; identified and performed relative to FSAR statements. i
RESPONSE
We concur with the IDI Team that the FSAR statements represent , licensing commitments. The Independent Design Review will address the issue concerning calculations, a's described in i our response to the General Item of NRC letter dated March 23, ! 1984. (The response to the General Item is contained in i Section I of this attachment.) - l t l t i I t r l i l
?
l l F2.8-1 i i
I < BYRON-IDI b o Finding 6-12: Equipment Status Display Criteria "Please inform us of the date that we can review the final design of the Equipment Status Display System." i RESPONSE 4 The " Documentation Package for Equipment Status Display System; Byron and Braidwood Nuclear Power Stations - Units 1 and 2," is now complete and available for review. This document contains final design information for the ESD system. O O F6.12-1
s BYRON-IDI [ i
<~
N ,%) l Unresolved Item 3-1: Rod Hangers and Pipe Rest Supports "The following outline is provided to clarify the team's intent:
- 1. Use of infinite support stiffness met the licensing i commitment in the sense that there was no specific commitment to use realistic stiffness in piping analyses.
- 2. Our sample problem indicated that calculated piping stresses varied somewhat when realistic stiffnesses were employed, but not enough to matter with respect to the piping stress.
- 3. Our sample problem indicated that calculated seismic support loads varied when realistic stiffnesses were !
employed. The maximum increase in a support load was 70 percent. This result is shown in Table 2 of the EG&G report at Sargent & Lundy Node 98A: ; 609 lb - EG&G calculated SSE load using reasonable stiffness l 358 lb - S&L calculated SSE load using infinite stiffness : f~) i N/ 251 lb - 70 percent increase over the S&L calculated load (a) In the sample problem, this type of variation was not concidered to matter with respect to ! support strength in view of the large margins typically provided. (b) However, we were concerned about your up-lift t check procedures for non-linear supports such' as pipe rests and rod hangers. When the seismic loads exceeded the dead weight and thermal loads further checking was performed to assure that unloading did not cause problems, e.g., checking I ! of pounding action and of increased loads on I adjacent supports. Our concern was as follows: [ (i) If reasonable support stiffnesses were , f used, the predicted seismic loads would ! be substantially greater in some cases. (ii) Some non-linear supports which were not , originally predicted to unload and thus i were not checked would be expected to l unload. I l l l O(~% l U3.1-1 i
--_-.__..__._.__,___.-___--.J
.. .- . - .-. - - - . _ _ . .- . _ - - -_ . . ~ .
l 1; ? BYRON-IDI [
) [
(iii) We, therefore, intended to suggest that t you check additional non-linear supports for unloading - for example, those where seismic loads exceed about half of deadweight and thermal loads. g You are requested to describe your plans to assure that ! seismic unloading of non-linear supports, where that can ( 1 be expected, will not cause overstress due to pounding ( i or increased loads on adjacent supports." 1
\
i j RESPONSE e
-Piping analysis is a design tool for providing a basis for selection of support hardware and evaluation of piping' stresses.
c The-EGEG analysis method considers pipe support stiffness values whereas the Sargent & Lundy analysis method considers pipe support to be infinitely rigid. 1 Botn EGEG and Sargent & Lundy methodologies are acceptable j means for analyzing a piping system. Large margins and con-siderable conservatisms do exist in both approaches. These are demonstrated and discussed in detail in technical literature 4 () such as NUREG/CR-3526, " Impact of Changes in Damping and Spectrum Peak Broadening on the Seismic Response of Piping Systems.' In addition, it is not reasonable to take extreme differences r resulting from the.two methodologies and review the results of one analysis method (EG&G) against the other (S&L) and to suggest that the uplift limits should be increased for rod hangers and rest type supports. Sargent & Lundy's support modeling practice was discussed in detail with the staff of the NRC Mechanical Engineering i Branch on August 19, 1983 and it was found to be acceptable. l
. Attached for your reference are two menos by the NRC Mechanical
!- Engineering Branch and the Division of Licensing which state that the approach used by Sargent & Lundy is acceptable. I Reference 1, " Memorandum from D. G. Eisenhut" states:
" Based on our review of the Sargent & Lundy design practices, the staff concludes that the method used by Sargent &
l-i Lundy for the modelling of the pipe supports in the piping L design analyses together with the engineering rationale presented in some detail in the attachment provides an adequate bas!.s for the calculation of piping stresses and support loads." O U3.1-2
BYRON-IDI O Reference 2, " Memorandum from D. Terao" states:
"It is the staff's belief that S&L's design practice of modelling supports as infinitely rigid is acceptable when used with sound engineering judgement."
We believe that we have addressed your concerns in regard to this item. O I i l 9 U3.1-3
. _ _ _ _ . _ . . _ _ . _ . . _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _ - _ _ - ~ _ _ _ _ . . - _,
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BYRON-IDI 4 . ., [ . t 1 i k i
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ATTACHMENT IN RESPONSE TO { UNRESOLVED ITEM 3-1 . t I i l l rs
- 1. NRC Memo from D. G. Eisenhut to R. L. Bangart, dated October 3 1~,. 1983. t
)
.2. NRC Memo from D. Terao to R. J. Bosnak, dated September 19, i 1983. f i
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