ML20246N944
| ML20246N944 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/16/1989 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM TAC-65753, TAC-65754, NUDOCS 8903280074 | |
| Download: ML20246N944 (11) | |
Text
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Dunn POWER GOMPANY P.O. BOX 33189 CIIAnLOTTP, N.O. 28242 l
HALD. TUCKER Tetzenown i
vics ramminzwr (704) 373 4531 moos.zam emonnemum i
March 16, 1989
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Document Control Desk l
U. S. Nuclear Regulatory Commission I
Washington, D. C.
20555
Subject:
Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Performance Testing of Relief and Safety Valves Gentlemen:
Dr.
K.. N. Jabbour's letter dated October 18, 1988 transmitted a request for additional information regarding the performance testing of relief and safety valves (NUREG-0737, Item II.D.1). These questions were based on my submittals dated April 29, May 31, and June ~14, 1988. Please find attached responses to question Nos. 1, 3, and 5 from your request for additional information. As indicated per my February 24, 1989 letter to the Document Control Desk, Duke Power is in the process of obtaining vendor supplied reanalyses regarding the performance testing of Catawba's relief and safety valves. Responses to question Nos. 2 and 4 will be provided prior to July 31, 1989 after all reanalyses are completed and reviewed by Duke Power personnel.
Very truly yours, Ap
. /.,4c/?A L
H. B. Tucker JGT/2/II-D-11 0
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Mr. M. L. Ernst Acting Regional Administrator, Region'II U. S. Nuclear Regulatory Commission 1
101 Marietta St., NW, Suite 2900 Atlanta, Georgia 30323 Mr. W. T. Orders NRC Resident Inspector Catawba Nuclear Station h
8903280074 890316 l
l PDR ADOCK 05000413 j
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d Questi(m 1:
In Reference 1, Duke Power Co. (DpC) responded to question 1 of the NRC request for additional information by stating that the plant specific valve closing pressure rise was 9 psi.
The value of 9 psi appears to be too low even for the short inlet pipe used with the Catawba safety valves.
Verify this value includes both frictional and acoustic wave components of the. pressure rise calculation as provided in Reference 4.
For valve closing, if the plant specific pressure rise exceeds the test pressure rise, justify the plant valves will operate stably.
RESPONSE
The valve closing pressure rise including both frictional and acoustic effects was reverified in accordance with EPRI guidelines.
A calculation error was determined.
The corrected value for valve closing pressure rise for the worst case valve is 71 psi.
This value is slightly higher than the Dresser 31739A short inlet pipe pressure rise but well below the pressure rise of the long inlet configuration.
The Catawba pressure rise is bounded by the EPRI tests and valve performance is expected to be stable.
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Question 3:
Provide test data to support Duke Power Company's. assertion that 51 ft-lb provides adequate torque to close the block valve under full differential pressure conditions.
RESPONSE
Duke Power cannot provide test data supportingJ51,ft-lb as sufficient torque to operate the PORV block valve. While 51 ft-lb is considered adequate based on manufacturer's experience,Lwe will increase the torque settings of the installed actuator so that the torque output is the same as in the EPRI PWR Safety and Relief Valve Test.
The EPRI test report noted that shutoff was achieved at a setting as low as 1.5.
This corresponds to a torque of 79 ft-lb (see attached curve for spring pack 60-600-0048-1).
Conservatively assuming a. lubricated stem.(coefficient n#
friction equals 0.15), the block valve stem factor is.0074. ~ The test au,uator, then, was supplying a thrust as high as 79/.0074 or 10,676' pounds.
The block valve actuators will be set up (using MOV' diagnostic equipment) to provide a minimum of 10,676 pounds of thrust. The Unit 1 actuators are already set up to the above thrust. The Unit 2 actuators will be set.up during the Unit 2 end of cycle 2 refueling update which is scheduled to start in March 1939.
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v Quantion Efa)-
Provide the dates for the additions and any addendo of the ASME and ANSI Codes used in the structural analysis.
RESPONSE
The Catawba Codes of Record for piping analyses are the ASME Boiler and Pressure Vessel Code, Section til 1974 Edation including Summer 1974 Addenda, Summer 1978 Addenda for flange qualsfecat on, and the ANSI Code for Power Pipeng. 831 1 1973 Edetion including Summer 1974 Addenda.
Question 5(b):
Because Table 1.2 only discusses the ASME and AISC Codes, clarify the relationship between Tables 1.2 and 1.3 provided in DPC's response to question 12c.
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Response
The stress criteria shown in Table 1.2 is applicable to portions outside the ASME Section III Subsection NF boundary for Class 1 piping supports and the l
entire support on ANSI B31.1 piping. The load combination as defined in Table 1.2 is for all ASME Class piping and is also applicable to all overlapping ANSI B31.1 piping.
The codes and standards listed in Table 1.3 are often used in conjunction with one another for support designs.
The normal allowable i
stresses are based on the AISC allowable values as defined in Part 5 of the AISC manual. AWS D1.1-73 is used as the structural welding code for the a
Non-NF portion of a support.
For Non-NF applications, MSS SP-69,.1976 serves
'l as a guide for selecting appropriate catalog components, manufactured to MSS
'i SP-58, 1975 standards, and structural members that meet the criteria of Table 1.2.
MSS-SP-69 is used as a supplement to Subsection NF for selection and i
application of pipe supports.
I Quantion Efeit For the load combinations provided in Table 1.0 (question 12c),
i clarify whether the allowable used for aquation 9 (faulted) also
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considered the constraint thet the stress be not greater than 2.o Sy.
If not, lustify not using this constrasnt prov ded in the ASME Code.
Also, equalson 13 Should have included the OBE seismic anchof j
movements.
Justify not includang thss load or redo the analys6s including it i
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RE12.ONAE:
The Code of Record for Catawba (from question Sal does not require a 1imst of 2.0 Sy when appiy6ng the requirements of NB-3652'Eq.(9).
Eq(9).for faulted is satisfied using a. stress llimit of 3.0 Sm.
In the Code of Record for Catawba, Eq(13).does not' require.OBE seismic L
anchor movements to be included.
Eq(12) includes.thermat plus l
thermal anchor movements, where Eq(13) from NB-3652 specifically excludes seismic anchor movements.
Question 5(d):
For the load combinations provided in Table 1.2 (ouestion'12c),.
(a) clarify.why the faulted load combination allowable was 1.5 times ~the AISC normal allowable stress when AISC code only allows a normal allowable increase of 1.33.for any seismic load; (b) also provide the allowables used for the Class 1 piping supports that were stated in note 1 of Table 1.2 to be based on Subsection NF of the ASME Code.
Response
(a) The portion of a pipe support not within the. Subsection NF boundary is defined as being part of the building steel structure.
NUREG-75/087
" Standard Review Plan" (SRP) Sections 3.8.3 and 3.8.4 defines the allowable stress limits for internal steel. structures. Using the elastic working stress design methods of Part 1 AISC specification, the SRP..
maximum allowable stress for normal and upset load combinations is the AISC normal allowable values.
For faulted load combinations, the SRP maximum stress is (1.6) ( AISC normal allowable values). The faulted:
I allowable stress was reduced from (1.6) to (1.5) (AISC normal values) to add a safety margin and be equivalent in most cases to-the 0.9Fy maximum stress value. 'These sections do not permit a 33% increase in allowable stress for seismic loadings.
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(b) For structural. steel portions within the NF jurisdictional' boundaries, the allowables are as follows: (2)
Condition (1)
Criteria Normal:
AISC Normal Allowable Stress Upset:
AISC Normal Allowable Stress Faulted:
1.33 AISC Normal Allowable Stress with 0.9 Fy maximum NOTES:
(1) The load combinations are the same as defined on Table 1.2 (2) Manufactured component standard supports used within the NF jurisdictional boundaries use the allowables specified in Subsection NF of the ASME BPVC,Section III, Division 1.
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p-1 Quantion Etalt in several of the taDies that compared the maximum calculated stresses to the allowable stresses (question'12d), there were poents-where the equation 10 calculated stresses exceeded the allowable, stress. 'This mas stated to be acceptable because. equation'12 and 13 or equation 11 were satisfled.
However, the tables did no' provsde a comparison of the calculated stresses and equation 12 and 13.or equat.lon 11-allowables for these poinis.
Provide this comparison for review.
RESPONSE
Tables 2 '. 0 and 2.2 I(attached) provide a comparison of the calculated-stresses and equation 12 and 13.
Question Effic in response to question 12b. DPC stated the time step used in the structural analysis was 0.002 s and the-lumped mass spacing was based on.a frequency of 30 Hz.
A time step of 0.002 s ss able to accurately-calculate frequency responses up to about 62 Hz.
Based on f
EG&G Idaho experience, a' frequency of at least 100 Hz needs to De
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considered in the structural analysis because the forcing functions from the valve discharge could excite the higher frequencies in the
- piping, if less than 100 Hz was used, significant dynamic responses in the system could be missed in the analyses and the piping stresses underpredicted.
DPC must justify that the analyses with a lumped l
mass point spacing based on 30 Hz and 0.002 s time step l
conservatively caIculated the psping stresses or redo the analysis accounting for the hegher frequencies
RESPONSE
The Time History Analyses was reanalyzed using a lumped mass spacing based on a frequency of 100 Hz.
These result 5 when compared with the original analysis using a lumped mass spacing based on 30 Hz do not show a significant change in the results Therefore, 30'Hz.Is a reasonable basis for a mass point spacing.
For further verification, the time history analysis for 100 Hz was reanalyzed with.a tsme stop l
of 0.001 seconds and compared with the results from the 30 Hz analysis with a time step of 0.002 seconds.
The change-in results i
are not significant Therefore, using a time step of 0.002 seconds and a lumped mass spacing based on 30 Hz.is adequete to account for the higher frequencies for the structural analysis 1
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1 TABLE 1.2 Load Combinations and Stress criteria for Supports, Restraints, and Anchors LOAD COMBINATION CRITERIA 1.
Norinal Thermal Displacement AISC Normal Weight Allowable Stress 2.
Upset Thermal Displacement AISC Normal' Weight Allowable Stress 0BE-(Inertia)
OBE Seismic Anchor Movements Relief Valve Transient i
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Faulted Thermal Displacement (1) 1.5 x AISC. Normal Weight Allowable Stress SSE (Inertia) with a 0.9 FI SSE Seismic Anchor Movements.
maximum Relief' Valve Transient l
NOTES:
1)
Criteria shown is for non-NF.
For portions within the NF jurisdictional boundaries, the allowables specified in Subsection NF of the ASME BPVC,Section III, Division 1 are used.
2)
Relief Valve Transient = Maximum absolute value load from PORV discharge transient and Safety Relief Valve discharge transient.
3)
Loads on supports are combined in the'following manner:
Normal (+),,= Weight + Maximum positive thermal Normal (-) = Weight + Maximum negative thermal j
Upset (+) = Normal (+)+[ absolute sumation of CBE Inertia, OBE Seismic Anchor Movements, Relief Valve Transients]
Upset (-)= Normal (-)-(hbsolutesummationofCBEInertia,OBE I
Seismic Anchor Movements, Relief Valve Transients) l
, Faulted (+) = Normal (-) + [ Absolute summation of SSE Inertia,
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SSE Seismic Anchor Movements, Relief Valve i
Faulted (-) = Normal (-) - [ Absolute sumation of SSE Inertia, SSE Seismic Anchor Movements, Relief Valve Transients) s
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J TABLE 1.3 CODES AND STANDARDS FOR PIPE SUPPORT DESIGN 1.
ASME Boiler and Pressure Vessel Code.Section III
- Division 1. Rules for Construction of Nuclear Power Plant Components. 1974 Edition including all addenda
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through the Summer 1975 Addenda.
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2.
MSS SP-58. 1975. Pipe Hangers and Supports - Materials.
Design and Manufacture, j
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AWS D1.1-73. Structural Welding Code - Steel.
1 4.
AISC Manual of Steel Construction. Seventh Edition with Specification for the Design. Fabrication T Erection of Structural Steel for Buildings, together with Supplements
- 1. #2. and #3.
5.
MSS SP-69. 1976. Pipe Hanners and Supports - Selection and Application.
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l TABLE 20 (REVISED)
UNIT 1 i
[
UPSTREAM PIPING (CLASS 1)
MAXIMUM STRESSES 1
i COMPONENT JOINT STRESS ALLOWABLE CONDITION DESCRIPTION NAME RESULTS STRESS AATIO 1
i Eq. 9 (Desi0n) 6" Elbow 31 19431 24120 0.806 Eq. 9 (Faulted) 6" Elbow 31 38773 48240 0.804 i
Eq.
10 6"x6"x3" 96 92387 48480 1.906*
j TEE (See i
. Note 1) 1 Eq.
12 AWTT 127 36837 48216 0.764 i
TEE 6"x6"x3" 96 35999 48480
.0 743
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4 TEE j
I Eq.
13 AWTT 122 45520 48180 0.945 VALVE l
I 6"x6"x3" 96 42161 48480 0.870 TEE l
l Fatigue Usage 6"x6"x3" 96 1.0 0.039 Factor TEE i
I Notes:
1)
Acceptable since equations 12 & 13 are satisfeed.
2)
AWTT = As-Welded Tapered Transition Joent L_-_______-.
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TABLE 2.2 (REV1 SED)
UNIT 2 UPSTREAM PtPING (CLASS 1)
MAXfMUM STRESSES COMPONENT JOINT STRESS ALLOWABLE CONDITf0N DESCRIPTION NAME RESULTS STRESS RATlO Eq. 9 (Design) 6" Elbow 31 22418 24120 0.929 Eq. 9 (Faulted) 6" Elbow 31 31900 48240 0.661 Eq.
10 6"x3" 98 75370 48480 1.555*
Reducer (See Note'1)
Eq.
12 6"x6"x3" 96 32401 48240 0.672 4
TEE q
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l 6"x3" 98 17829 48480 0.368 i
Reducer i
Eq. 13 6"x3" 99 47387 48480 0.977 j
Reducer i
6"x3" 98 46958 48480 0.969 l
Reducer l
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Fatigue Usage 6"x6"x3" 96 u :.042 1.0 0.042 I
Factor TEE I
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Notes:
1)
Acceptable since equations 12 & 13 are satisfied.
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