ML19341D007
| ML19341D007 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 02/24/1981 |
| From: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 11345, NUDOCS 8103040692 | |
| Download: ML19341D007 (7) | |
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General offices: 1945 West Parnail Road, Jackson, MI 49201 * (517) 788 0453 7ebruary 24, 1981 V M,, g S N 9
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'2 Earold R Denton, Directer vJa Orrice of :Iuelear Reactor Regulation q
US :Iuclear Regulatcry Cc==ission
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'4ashington, DC 20555 4
M MIDLCID PROJECT
- IRR E7/1UATIO:t A'ID AFFRCVAL OF REACTOR FRESSURE VESSEL SUFFORT 'ICDIFICATIO:lS FILE: 35.h/3, 0505.803 UFI:
023005, 70*01*10*03 SERIAL:
ll3h5 Cur correspondence, Serial-9787 dated December 10, 1960 to the :ac's Region III Office, advised of our intent to meet with the Office of :iuclear Reacter Regula-tion (IIRR) to resolve any remaining Staff concerns on the reactor vessel support modificaticns.
~"ne intent of this meeting vculd have been to assist the IGR in the completion of its evaluation and approval of the design concept en a time-table which would suppert our construction schedule. On February 12, 1981, the Staff advised us of their remaining concerns and suggested that these could be dccumented and resolved without the necessity of a meeting.
The attached Enc 1csure represents a ecmplete response to the nine (9) questions verbally conveyed to us by the Staff.
It is our expectation that these responses will allow the :ER to ec=plete its evaluation and provide the :IRC's Region II Office with an approval of the design concept. If the :33 does find the modified support concept acceptable, we are prepared to implement this same modification en boti. Unit 1 and Unit 2.
"'his should relieve any GR concerns on the acceptability of t. - Unit 2 reacter vescel anchor belts.
In order to support our construction schedule, :mR's apnroval to the Region II!
Office vould be appreciated by March 1,1981.
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Enclosure Responses to !!RE Questions en Reactor 7essel Support Modification C0 Director of Office of Inspection fe Enforcement (US :30)
Director, Office of '< anagement (US liRC)
Cen=ander, :!aval Surf ace '4eapcns Center Reglenal Director, Office of Inspection
'e Enforcement (Region :::)
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2 LJAuge, ETEC w/a RLBaker,B!,W/AA,v/a JWCook, P-26-336B v/a RJCook, Midland Resident Inspector v/a LHCurtis, Bechtel v/a DFJudd, B'4W v/a GSKeeley, P-lh-ll3B v/a CBechhoefer, ASLB v/o GALinenberger, ASLB v/o FPCovan, ASLB v/o ^
AS'4L Appeal Panel v/o MMCherry, Esq v/o MSinclair, v/o CRStephens, US IIRC v/o WDPaton, Esq, US :TRC v/o FJKelly, Esq, Attorney General v/o GETaylor, Esq. Asst Attorney General v/o WEMarshall v/o f
REACTOR PRESSURE VESSEI, SUPPORT MODIFICATIO I MIDLAND PIXIT UNITS 1 AND 2 MIDLXID, MICHIGAN CCNSUMERS PC'a'ER COMPANY PESPONSES TO *:ER CUESTICNS FEBRUARY 27, 1981
Midland Units 1 & 2 RF7 Support Modiff Mtic o
RESPOWSES TO ?:33 CUES *IO:S 1.
What is the design margin in the reactor support system compared with the AISC Specification allowable stresses?
1.1. The Studs:
The studs' material, ASTM A354, is not covered in Section 1.4.1.1 of Based on the Teledyne Engineering Ser-vices study,the AISC., the allowable tensile stresses are given in the table below. In the same table the maximum design stresses from preliminary conservative design loads (see response to 5) are shown.
Preloadine Short-term Loading Allowable Design Allowable Maximum Design Cl}
43 ksi ( )
40.5 ksi 6 ksi 1.5 ksi
(=.046 fy)
(=.331 fy)
(1) At a temperature of 150*F.
At a temperature of 70' (when tensioned) the stresses in the studs are 4.5 ksi (min) to 5.5 ksi (max).
(2) But not more than one-half the lowest measured detensioning load on any stud which is considered to contribute to load carrying capability in the new design concept. Detensioning load may be increased above the prestress load required for nut rotation in order to determine the allowable sher t-term stress.
(3) This stress is only in one stud, stresses in other studs are lower as shown below in figure (1).
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- TR-3887-2, Rev 1 - Acceptability For Service of Midland RPV Anchor Studs, May 20, 1980.
Midland Units 1 & 2 RPV Support Modificatf'c Pais 2 of 6 1.2. Upper Lateral Supports.
The bracket material, ASTM A 516, is not directly covered as stated in Section 1.4.1.1 of the AISC. However, allowable stresses were calculated in a manner consistent with the AISC Specification. The allowable stresses are given in the table below. In the same table the maximum design stresses from preliminary conservative design loads (see response to 5), are shown.
Load Combination Allowable Stress l Max. Design Stress Or Interaction Value*
l
- 1) D.L. + L.L.
Fb = 20.1 ksi 1.0 ksi
- 2) D.L. + OBE
[a f;b = 1.25
.1
< 1.25 Fa Fb
- The interaction of axial compression and bending is according to the AISC Specification, Section 1.6.1.
D.L. =
Dead Load L.L. =
Live Load Safe shut-down earthquake SSE
=
LOCA =
Loss of Coolant accident loads, the worse accident from a guillotine break at the hot leg outlet from the RPV or the cold leg inlet to the RPV.
For LOCA the capacity is determined based on the allowable stresses given, however, yield strain may be exceeded.
Operating basis earthquake OBE
=
te Maximum design stresses in axial compression fa
=
Maximum design stresses in bending fb
=
Allowable stress in axial compression, including slender-Fa
=
ness and temperature ef fects,19.66 ksi (=.52 Fy).
Allowable stress in bending, 20.1 ksi, including tempera-Fb
=
ture ef fects, (=.53 Fy).
Minimum yield stress of the material as per Fy
=
ASTM specification.
2.
Provide the final analysis result when all analyses are complete.
_2 Maximum design stresses in the studs and the upper lateral supports calculated based on the final loads obtained af ter completion of the final analysis, will be provided to the NRC. These maximum design stresses will be compared with allowable stresses given in the response to question 1 above to demonstrate the margin.
forecast date for conpletion of this task is December 1982.
The current 3.
Schedule for final stress reports.
. The final stress reports will be sent to the NRC no later than 2 months prior to fuel load.
Midland Units 1 & 2 i
RP7 Support Modificatic Page 3 of o in the upper lateral support ?
4.
What is the source of bending moment Sources of bending moment in the upper lateral support are.
-18 k.fc 1.
D.L. + L.L.
M
=
max 692 k.ft 11.
Cavity pressure from LOCA M
=
2 k.f t M
=
111.
Eccentricity of axial load from SSE 38 k.f t M
Eccentricity of axial load from LOCA
=
iv.
max All moments above are at the interface between the bracket and are based on preliminary conservative and the embedment, design loads.
Positive moments produce tension at the bottom of the brackets NOTE:
and negative moments produce tension at the top.
Provide detailed explanation and breakdown of the LOCA and SSE 5.-
loads and how they were combined.
The breakdown of the SSE and LOCA loads used in the design of the studs and the upper lateral support brackets are as follows.
Axial load Studs (g):
(uplift)
Shear Moment SSE (During cold shut-down)* 233 kips 1551 kips 39,000 k.ft SSE (During operation) 350 kips 171 kips 810 k.ft 6521 kips 1505 kips 4,067 k.ft LOCA the vessel skirt interface not at the (1) The loads given above are at skirt-pedestal interface as reported in the reports previously sub-These loads were used in calculating the stresses in the mitted.
studs.
Moment (2)
Axial Compression Brackets:
I)
-10 k.ft SSE (During operation) 83 kips IN
- 22 k.ft 1,689 kips LOCA (2) The moments are from D.L., the eccentricity of the axial load, and the cavity pressure in case of LOCA.
(3) As a conservative design basis, the bracket loads previously reported have been arbitrarily increased by a factor of 1.5.
The preliminary loads from SSE (during operation) and LOCA were con-bined using the absolute sum.
These loads were derived assuming no upper lateral supports, which is the case during cold-shut down.
Midland Units 1 e e j
RP7 Support Modificatic Page h of 6 i
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i Confirm that the reactor vessel studs detensioned to a maximum of 6.
6 kai can withstand normal operating loads, especially vibratory loading during normal operation.
Babcock & Wilcox completed an assessment of the effects on the reactor vessel (RV) during normal operations of a lower pretension value for the RV holddown studs.
considered flow-induced vibrations In particular, this assessmentin varying degrees, during all stages of that could be present, normal operation.
The results of the assessment showed that a minimum pretension value of 1.5 ksi is adequate to ensure a stable reactor vessel orientation during normal operation.
7.
How was the gap criterion of 1/32-inch arrived at and what were the major considerations which went into this not gap; The gap size of 1/32 inch has been evaluated by B&W as the largest gap which, under the hypothesized accident conditions, would result in impact loadings on the upper lateral supports sufficiently small (when compared to LOCA and seismic loadings) that the calculations would remain in the linear regime.
The original plan for RV support modification and design (as reported in Preliminary Report NO. 1) called for a 1/32-inch gap by shimming in the hot condition. The primary purpose was to attempt to contain the loading analyses within the linear regime and, concurrently, to minimize possible thermal loadings on the upper lateral supports dur-ing normal reactor operation. The shimming criteria is currently under joint B&W and Bechtel evaluation to determine the optimum shimm-ing temperature.
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