Text

B&W Owners Group Bolting Task Force Summary Rept of Nrc/B&W Owners Group 830506 Meeting on Reactor Vessel Internals Bolting
	ML20071N926
Person / Time
Site:	Oconee
Issue date:	05/31/1983
From:	; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML15238A809	List: ML20071N926;
References
	BAW-1784, TAC-51432, TAC-51556, NUDOCS 8306070286
	Download: ML20071N926 (124)
	v • d • e

.

,- -P* BAW 1784 May 1983 i

B&W OWNERS GROUP ,

BOLTING TASK FORCE l

SUMMARY

REPORT OF NUCLEAR REGULATORY COMMISSION /

BABCOCK & WILCOX OWNERS GROUP MEETING ON REACTOR VESSEL INTERNALS BOLTING MAY 6,1983 ,

Arkansas Power & Light l

Consumers Power Co.

Duke Power Co.

Florida Power Corp.

GPU Nuclear

Sacramento Municipal Utility District i Toledo Edison Co.

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SUMMARY

REPORT OF i

! NUCLEAR REGULATORY COPWISSION /

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BABCOCK & WILCOX OWNERS GROUP .

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MEETING ON

REACTOR VESSEL INTERNALS BOLTING 4

MAY 6, 1983 -

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1 TABLE OF CONTENTS l Section Page No.

1.0 Introduction 1 2.0 Technical Background 4 2.1 General Description 4 2.2 Problem Definition 14 2.3 Upper and Lower Core Barrel Bolts Design Basis 14 2.4 Ultrasonic Examinations 15 3.0 Inspections and Examinations 19 3.1 Site Inspections 19 3.2 Laboratory Examinations 19 3.2.1 Oconee 1 & 2 Examinations 19 3.2.2 Rancho Seco UCB Bolt Exam 26 3.2.3 Overall Conclusions 29 4.0 Safety Implications 30 4.1 UCB/LCB Design Margins & Joint 30 Failure Consequences 4.2 Justification for Continued Operation 36 4.2.1 Davis Besse 1 36 4.2.2 Oconee Nuclear Station 39 4.2.3 ANO-1 42 5.0 Future Actions 43 e

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APPENDICES A. NRC - B&W Owner Group Meeting Slides B. Site Ultrasonic Examination orocedure C. Reactor Yessel Internals Bolting Installation Torque vs Preload Test Data D. NRC Questions on Installation Torque and Prestress e

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LIST OF TABLES I

Page No.

2.1 Bolted Joints with A286 Material 11 2.2 Minimum Number of UCB Bolts to Maintain Joint Integrity 16 2.3 Minimum Number of LCB Bolts to Maintain Joint Integrity 17 3.1 UT Inspection Results (# Cracked /# Inspected) 20 3.2 Total UT Inspection Results 21 4.1 Likelihood of Core Drop Very Low Based on Structural 31 Margins and Results of UT Inspections 4.2 Upper Core Barrel Bolts Inspection Results 40 4.3 Lower Core Barrel Bolts Inspection Results 41 C-1 1-3/4" Diameter Upper and Lower Core Barrel Bolt Stress / C-2 Torque Test Data C-2 1" Diameter Lower Thermal Shield Bolt Stress / Torque Test C-3 Data C-3 1-1/2" Diameter Upper Thermal Shield Bolt Stress / Torque C-4

+ Test Data .

D-2 Preliminary Stress Level Comparison D-2 ,

D-3 Installation Torque - Ft-lbs for Upper and Lower Core D-3 Barrel Bolts iii

LIST OF FIGURES Page No.

1.0 B&W Owners Group Organization 2 2.1 177FA Plant Reactor Vessel Internals Assembly 5 2.2 177FA Plant R.V. Internals A286 Bolted Joints 6 2.3 Upper Core Barrel to Core Support Shield Bolted Joint Detail 7 2.4 Detail Upper Thermal Shield Restraint 8.

2.5 Lower Core Barrel to Lower Grid Joint Detail 9 2.6 Lower Grid to Flow Distributor Bolt 10 2.7 Lug and Guide Block Locations 13; 3.1 Upper Core Barrel Bolts with Abnormal Indications - Rancho Seco 22 3.2 Upper Core Barrel Bolts with Abnormal Indications - Crystal 23' River-3 3.3 Core Barrel Lower Grid Bolt Locations - Crystal River 3 24 3.4 Upper Core Barrel Bolts with Abnormal Indications - ANO-1 25 3.5 Sketch of Upper Core Barrel Bolt Showing Location of Fracture 27 3.6 Rancho Seco Upper Core Barrel Bolt #79 28 4.1 Reactor Vessel and Internals Evaluation Cross-Section 32 4.2 Upper Core Barrel with Joint Failure 33 4.3 Core Drop-Inches vs Additional By-Pass Flow 35 4.4 Reactor Vessel and Internals Cross-sectional Elevation 37 4.5 Core and Internals in Dropped Condition from Lower Core Barrel 38 Joint Failure 1

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1. INTRODUCTION As a result of failures discovered at Oconee in July of 1981 in the bolts fastening the lower portion of the reactor vessel internals thermal shield to the lower grid assembly, a program for reactor vessel internal bolt inspection and repair was initiated. This program resulted in the discovery in March 1983 at Rancho Seco similar failures of bolts in the upper core barrel to core support shield joint. In April of 1983, ultrasonic tests at Crystal River 3 showed abnormal indications in 4 lower core barrel bolts and a number of upper core barrel bolts. The discovery of the additional anomalies in core barrel joints led the B&W Owners Group (B&WOG) Steering Committee to direct the formation of a special Task Force on internals bolts (see Figure 1.0). This task force is made up of representatives of all B&W designed 177 FA plant Owners and is chaired by C.W. Hendrix, Jr. of Duke Power Company.

The Task Force mission is to provide a forum for sharing information on inspections, testing, and repairs related to the internals bol-ting problem and thereby maintaining cognizance of all activities pertaining to internals bolt failures. The Task Force will also maintain a consistent licensing posture for generic safety issues that may be raised. It will provide a means of monitoring generic issues and will provide a neans to expedite of the performance of necessary generic activities.

Thus, the B&W Owners Group has taken two immediate actions related to the internals bolting concern. One action was to provide an initial evaluation

that currently operating facilities provide no undue risk to the public health and safety. This is based on the assessments that there are large design margins in reactor vessel internals bolting systems, that current inspec'tions reveal bolt failure levels well below those required for joint failures and that even should joint failure occur no significant safety hazard will result.

The second action was to establish an organization among the B&W Owners (i.e.,

the Bolting Task Force) to continue to investigate and evaluate new information for resolution of problems found. The Bolting Task Force reports to the Steering Committee.

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ge The results of the initial evaluation were presented to the NRC on May 6, 1983. The handouts for that meeting are contained in Appendix A and the remainder of this report provides a summary of this initial evaluation and the additional information requested by the NRC after the May 6th meeting. The May 6,1983 presentation included specific statements by Toledo Edison and Duke Power concerning their justification for continued operation of the Davis Besse and Oconee Units. A statement was also made by representatives of Arkansas Power and Light regarding their restart plans for ANO-1.

The purpose of this document is to forwalize the information presented during the May 6, 1983, meeting.

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2. TECHNICAL BACKGROUND 2.1 General Description The reactor vessel internals, shown in Figure 2.1, are designed primarily to support and restrain the core while maintaining fuel assembly alignment and alignment between the fuel assemblies and the control rod assemblies.

Secondarily, the internals direct the flow of reactor coolant through the core, provide gamma and neutron shielding and guides for incore instrumen-tation. Lastly, the internals support internals vent valves and, in some plants, surveillance specimen holder tubes.

The internals are designed to withstand loading from normal operation and upset conditions as weil as design basis accidents. All reactor vessel internals have five bolted joints using A286 bolting material located as shown in Figure 2.2. These are: .

o upper core barrel .to core support shield, (Figure 2.3) o upper thennal shield to upper core barrel, (Figure 2.4) o lower thermal shield to lower grid assembly, (Figure 2.5) o lower core barrel to lower grid assembly, (Figure 2.5) o lower grid assembly to flow distributor. (Figure 2.6)

Additionally, four plants - Davis Besse 1, Crystal River 3, and Midland 1 and

) 2 - have surveillance specimen holder tubes bolted to the thennal shield.

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[ Of these joints, only the upper core barrel (UCB) and lower core barrel (LCB) bolting has core support significance. Should the bolts in either of these joints fail, the core and internals would drop onto guide lugs welded to the inside wall of the reactor vessel. These twelve (12) guide lugs are equally spaced around the internal circumference of the reactor vessel. The essen-tially L-shaped lugs are machined from Inconel ASME S8168 plate and are attached to the inner reactor vessel wall below the core and reactor vessel internals by full penetration welds. The lugs have a 3-1/4" thick 13-1/2" long horizontal leg and a 9" high vertical leg, tapering from 3" thick at the bottom to 2-5/8" at the top. (See Figure 2.7) As mentioned, these lugs 4

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provide support of the core and reactor vessel internals should either the upper core barrel to core support shield or the lower core barrel to lower guide assembly joint fails and the core drops. The guide lugs, together with guide block positioned on either side of each lug and fastened to the internals, provide restraints to horizontal or rotational motion of the core in either the normal or dropped : ore positions.

The bolting in the first five joints listed above is made of a nickel /chrone steel (A286) designated as SA 453 GR660, Conditi,on A material (solution treated at 1650F for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and age hardened at 1325F for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ). The quantity, location and size of these bolts are shown in Table 2.1. All bolts I

originally installed in these joints were hot headed. In addition, the thermal shield bolts were heavily cold worked before head forming or threading. All bolts in these five joints have welded locking clips to capture the bolt and prevent bolt rotation.

The sixth joint on the four plants 'with surveillance specimen holder tubes is also A286 material but is Condition B (solution treated at 1800F for 1 hr and age hardened at 1325F for 16 hrs) with the heads machined instead of hot headed.

B&W-designed 177 FA plants that use A286 material for internals bolting are:

Plant NSS ASME Code Edition I o Oconee 1, 2 & 3 Summer 1967 l o ANO-1 Summer 1967 o Crystal River 3 Summer 1967 o Rancho Seco Summer 1967 o Davis Besse 1 Summer 1968

, o Midland 1 & 2 (under construction) Summer 1968 l

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TABLE 2.1 BOLTED JOINTS WITH A-286 MATERIAL UPPER CORE BARREL (120 - 1 3/4 DIA.)

LOWER CORE BARREL (108 - 1/ 3/4 DIA.)

FLOW DISTRIBUTOR TO LOWER GRID (96 - 1 DIA.)

, UPPER THERMAL SHIELD (60 - 1 1/2 DIA.)

LOWER THEINAL SHIELD (96 - 1 DIA.)

SURVEILLANCE HOLDER TUBE (12- 3/4 DIA. PER TUBE)

SA 453 GR 660 (A-286) IS AN AGE HARDEED, HIGH STRENGTH, CORROSION RESISTANT MATERIAL. BOLTS ARE CONDITION A/ UPSET HEADED EXCEPT THE SURVEILLANCE HOLDER TUBE BOLTS WHICH ARE CDNDITION B/ MACHINED. -

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examination, was made of upper and lower core barrel bolts for all three Oconee units. No abnomal indications were noted. Following these exarr.inations, Rancho Seco, in March of 1983 during their refueling outage, noted 19 of 120 upper core barrel bolt failures. In April of 1983 during the Crystal River 3 refueling outage, 51 of 120 upper core barrel bolts and 4 of 108 lower core barrel bolts showed abnonnal indications on U.T. examination.

These findings led ANO-1 to extend their refueling outage to U.T. examine their UCB bolts. Cnly 7 o' f 120 abnonnal indications were noted.

.TMI-1 is not included in this concern because the bolting for their reactor vessel is made from Inconel X750. Specifically, the material is ASTM A637-70, Grade 688, Type 2 (solution annealed at greater than 1800*F). General Public Utilities Nuclear had UT examinations made of 9'6 of 120 of their UCB bolts and no abrormal indications were found.

2.3 Upper and Lower Core Barrel Bolts Desian Basis l The upper and lower core barrel bolts were designed to withstand the loads l

generated during normal operation and upset conditions due to joint prestress, i

1 dead weight, thermal, flow induced vibration and operating basis earthquake.

The UCB and LCB bolts were also analyzed for faulted conditions considering the loads generated by a large break LOCA and core bounce, a safe shutdown

! earthquake and dead weight of the core and internals.

l Examination of these analyses indicates that large design margins exist for the UCB/LCB bolt rings. The minimum number of bolts need to maintain bolt integrity has been calculated for both normal operation and faulted conditions. The results are shown in Table 2.2 and 2.3. In the calculation 14

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of the minimum number of bolts for normal operation as shown in these tables, prestress was included in the loads considered. Thus, credit was taken for the compressive forces between the flanges of the joint produced by this prestress. This results in joints with higher prestress requiring fewer bolts to maintain joint integrity. This is not the case for faulted conditions.

Prestress was not included in the faulted condition calculation since the joint was assumed to be open with the loads being carried only by the bolts themselves.

Tables 2.2 and 2.3 show that a large number of either the UCB bolts or the LCB bolts can fail before joint integrity is lost. These large margins exist because the bolting rings were originally designed for low strength bolting.

Changes in design loadings for faulted conditions, and the original design requirement to maintain the joint closed, resulted in the selection of high strength bolting. With high strength bolting at . higher initial prestress, considerable additional margin for normal operating and faulted conditions was achieved.

2.4 Ultrasonic Examinations

Special procedures were developed from laboratory test work done at B&W's Lynchburg Research Center. These procedures and a brief description of the laboratory backup for these examinations are provided in Appendix C of this report.

l The technique used three calibration bolts, one bolt with no notch, one bolt with a 15% notch, and one bolt with a 50% notch. A 2.25 MHz transducer was used. The transducers were 1/2", 3/4", or 1" diameter, depending on bolt size, was used. Signal response from head to shank at 40% of full screen height was used to set sensitivity.

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TABLE 2.2 MINIMUM NUMBER OF UCB BOLTS TO MAINTAIN JOINT INTEGRITY 120 UCB BOLTS NORMAL (1) FAULTED (2)

PRESTRESS-PSI OPERATION CONDITIONS PLANT OPERATING PLANTS OCONEE 1 10,000 27 50 OCONEE 2 24,000' 12 45 0CONEE 3 36,500 8 43 DAVIS BESSE(3) 36,500 8 46 PLANTS RECENTLY INSPECTED RANCHO SECO 36,500 8 45 CRYSTAL RIVER 3 36,500 8 45 ANO-1 36,500 8 45 (1) NORMAL OPERATION LOADS CONSIDERED PRESTRESS, THERMAL, DEAD WEIGHT, A FIV (2) FAULTED CONDITION LOADS CONSIDERED LBLOCA, SSE, DEAD WEIGHT, AND CORE BOUNCE. CALCULATION CONSIDERED ACTUAL ULTIMATE STRENGTHS RATHER THAN CODE MINIMUMS.

(3) DB-1 HAS A RAISED REACTOR COOLANT LOOP RATHER THAN A LOWER COOLANT LOOP DESIGN USED IN ALL OTHER 177 FA UNITS.

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108 LCB BOLTS NORMAL I) FAULTED (2)

PRESTRESS-PSI OPERATIUN CONDITIONS PLANT OPERATING PLANTS -

OCONEE 1 10,000 24 21 OCONEE 2 19,000 13 18 OCONEE 3 28,000 -

9- 18 DAVIS BESSE(3) 28,000 9 16 PLANTS RECENTLY INSPECTED RANCHO SECO 19,000 13 18 CRYSTAL RIVER 3 28,000 9 18 ANO-1 19,000 13 19 (1) NORMAL OPERATION LOADS CONSIDERED PRESTRESS, THERMAL, DEAD WEIGHT FIV.

(2) FAULTED CONDITION LOADS CONSIDERED LBLOCA, SSE, DEAD WEIGHT, AND CORE

, BOUNCE. CALCULATION CONSIDERED ACTUAL. ULTIMATE STRENGTHS RATHER THAN i CODE MINIMUMS.

(3) DB-1 HAS A RAISED REACTOR COOLANT LOOP RATHER THAN A LOWERE COOLANT LOOP DESIGN USED IN ALL OTHER 177 FA UNITS. -

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' abnormal indications - Similarly, laboratory destructive analysis of two bolts dithabnormalindicationsandoneboltwithnoindicationstakenfromthe Rancho Seco UCB bolt ring corralated well with the field UT results. Bolt

(#79) from Rancho Seco which showed no indication by field UT did show a small anomaly in the shank during the laboratory examination. This anomaly was not the result of intergranular attack nor located in a part of the shank where such attack has been noted on failed bolts (see Section 3.2.2). In addition,

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. shank diameter set by the calibration.

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, 3. INSPECTIONS AND EXAMINATIONS Is 3.1 SiteInsdections To date, UT examinations have been made at the three Cconee units, Rancho Seco, Crystal River,3, and ANO-1 Mar bolts with A286 material and,at TMI-1 on bolts made from Inconel X750. Table 3.1 shows a sunnary of these tests by plant and joint location. Figure 3.1 shows the pattern of abnormal indications found in the UCB bolt ring at Rancho Seco. Figure 3.2 and 3.3.

show the pattern of UT_ indications at Crystal River 3 for tne UCB and LCB bolt h rings, respectively. Figure 3.4 shows the 7 abnormal indicatiora .in the UCS 3 bolt ring found at ANO-1. The inspection to date indicates a fairly evenly f distributed,and random pattern. There also appears to be no crientation that coincides with . reactor coolant flow patterns or component design geometry.

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Table 3.2-summarizes the results of field inspections performed to date on UC3; LCS, SSHT and flow distributor bolts.

3.2 Laboratory Examination 3.2.1 Oconee 1 & 2 Examinations In the late summer and fall of 1981, 21scmples were taken of lower thermal shield, (LTS) bolts from Oconee 1 and 19 LTS bolts were taken from Oconee 2 in January 1982. Visual examination, fluorescent penetrant tests, scenning.

i electron microscope (SEMs) and metallography examinations were made on each bolt from Oconee 1 and a similar set of tests were run on the 19 Oconee 2 bolts. . However, a UT examination was used instead of PT for the Oconee 2

) bolts. Test results showed intergranular attack at the head to shank tran's ition region each of the bolts that showed abnormal indications by site UT examination.

Three ' upper themal shield and tso upper / core barrel bolts from Oconee 1 were o

also examined. No cracking was noted by visual examination, fluorescent pene-trant test, or SEM surface examination of any cf these boits. This confirmed field UT examination results made during the Oconee 1 refueling outage.

Lastly,. one flow distributor bolt from Oconee 1 was visually examined and fluorescent penetrant tested and no cratting wainoted.

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TABLE 3.1 < ,

UT INSPECTION RESULTS (# CRACKED / # INSPECTED)

BOLTED JOINT OCONEE 1 OCONEE 2 OC_0 NEE 3 ANO-1 RANCil0 SECO CR-3 (9-81) (1-82) (6-82) (5-83) (3-83) (4-83) l 0/21 UPPER CORE BARREL 0/30 0/30 7/120 19/120 51/120 (120 3/4 DIA.)

LOWER CORE BARREL 0/16 0 / 2 11 0/24 ----

0/108 4/108 j (108 3/4 DIA.) .

FLOW DIST/ LOWER GRID 0/22 0/25 0/25 ----

0/93 0/96

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(96 - 1 DIA.)

UPPER TilERMAL SilIELD 0/25 0/20 0/20 ----

0/60 0/60 (60 - Il DIA.)

. LOWER TilERMAL SHIELD 11/13 28/93 53/96 51/96 77/96 74/96 (96 - 1 DIA.) 918 IlEADS SEVERAL SEVERAL 48 75 HEADS 71 ilEADS I TWISTED llEADS HEADS HEADS TWISTED TWISTED OFF TWISTED TWISTED TWISTED OFF OFF OFF OFF 0FF i SURVEILLANCE Il0LDER j

TUBE (72 - 3/4 DIA.)* ---- ---- ---- ---- ----

25/72 DAVIS BESSE 1 PERFORMED VISUAL INSPECTIONS DURING THEIR 1982 REFUELING OUTAGE OF AND FOUND NO ABNORMAL INDICATIONS

DB-1 AND CR-3 ONLY !

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TABLE 3.2 TOTAL UT INSECTION ESULTS ,

( UCB, LCB, FLOW DISTRIBLTTOR)

TOTAL BOLTS TOTAL BOLTED JOINT INSNCTED INDICATIONS %

UPPER COREBAREL 441 77 17,5 LOWER COE BARREL 280 4 1,1f FLOW BIST/ LOWER GRID 261 0 -

0 SSHT (CR-3 ONLY) 72 25 34.7 21

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, " J-All failures noted in the LTS bolts from Oconee 1&2 were in the head to shank transition region and were attributed to intergranular stress assisted cracking. Also, no cracking was detected n the thread regions. It was also concluded that bolts that were acceptable by site UT were verified to have no flaws by laboratory examinations.

3.2.2 Rancho Seco UCB Bolt Examinations Two bolts (Bolt Nos. 41 & 116) that showed abnornal indications and one bolt that showed no indications (Bolt 79) from the upper core barrel ring were examined in the laboratory. All three bolts were visually and ultrasonically examined in the laboratorp. Metallographic examinations were made on one good bolt and one bolt with cracking. SEM/EDX fracture surface examination was made of bolt 116.

Bolt 79, the bolt showing no indications by field UT, was fluorescent penetrant tested and SEM/EDX examined. Small flaws containing localized deposits were ,noted in the upper shank region, see Figure 3.6. The surfaces between the flaws opened by ductile rupture when pulled apart. The flaw was not intergranular in nature and showed no secondary cracking. The flaw size was well below the minimum 15% site UT calibration standard and, therefore, would not be expected to be noted by the field UT examination.

The results of the examination of the two cracked bolts showed both bolts to have cracking initiated in the head to shank transition region similar to that found in the lower thermal shield bolts and that the cracking propagated through 95% of the shank area (see Figure 3.5) the cracking noted was typical of intergranular fracture and showed extensive secondary cracking.

l l These results led to the conclusion that Rancho Seco UCB failures are typical of previously noted LTS bolt failures. It was also judged that the flaw in the bolt that showed no indications by site UT was an anomaly not associated with the other bolt failures and is metallurgically inert showing no evidence of cracking or propagation and no significant loss of load carrying capa-bility.

26

% - _ . , -g. __--- % - -. - - , -w--

l SKETCH OF (FPER CORE BARREL BOLT . '.-

SHCMING LOCATION OF FRACTURE '

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3.2.3 Overall Conclusions The results of the laboratory examinations of internals bolting from Oconee 1 and 2 and Rancho Seco indicate that all bolt failures in the reactor vessel internals seen to date are due to intergranular stress assisted cracking located in the bolt head to shank transition region. In addition, the laboratory examinations have confirmed the findings of the field ultrasonic examinations.

P e

29

~~ * ***

- = = .

l l

4 SAFETY IMPLICATIONS 4.1 UCB/LCB Desian Margins & Joint Failure Consecuences The justification for the continued operation of the three Oconee plants and Davis Besse 1 and the restart of ANO-1, is based on several factors. Briefly stated, the present conditions are considered satisfactory. Large structural margins exist, severe degradation is detectable, and even if it were not detected, the consequences of core barrel joint failure do not constitute a significant reduction in public health and safety. Further discussion of these factors is presented below.

The upper core barrel and lower core barrel joints have large structural margins which reduce greatly the likelihood in the span time remaining for operation of these plants, that either the upper or lower core barrel joint would fail allowing the core to drop on to the guide lugs. The results of Rancho Seco, Crystal River 3, and ANO-1 inspections shown in Table 4.1, indicated that only 7% of the UCB bolts are needed for joint integrity during normal operation. This may be compared to the results of site UT examinations which showed the following percent of the UCB bolts to have no abnormal indications.

Rancho Seco 84%

CR3 58%

ANO-1 94%

Similarly for the lower core barrel joints during normal operation, only 11%'

of the bolts are needed for the LCB joint to remain intact at R/S while 100%

( of the bolts showed good UT examination, while at CR-3 only 8% of the bolts i are needed and 96% show to be good.

I l Despite the large structural margins should severe UCB or LCB bolt degradation occur it is very likely to be detected by periodic neutron noise monitoring.

{ If it were not detected and should the core drop, the dropped condition would

, be detected by loose parts monitoring backed by possible changes in self l powered neutron detector (SPND) and incore thermocouple readouts and reactor l coolant activity levels.

l 30

. ~ . ..

_ , _ , , _ . . .~. . - - - . -- - - -

.. ~

\

TABLE 4.1 :

LIKELIHOOD OF CORE DROP VERY LOW BASED ON STRUCTURAL PARGINS AND RESULTS OF UT INSPECTIONS UPPER CORE BARREL (120 BOLTS)

MIN # BOLTS RESULTS Oc UT INSPECTIONS REQUIRED FOR buuo Dtreu lvE PLANT NORML OPER BO_LR E RANCHO SECO .8 101 19 CR-3 8 69 -

51 ANO-1 8 113 7 LOWER CORE BARREL (108 BOLTS)

RANCHO SECO 13 108 0

! CR-3

. 9 104 4 31

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I REACTOR VESSEL AND INTERNALS 4

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ELEVATION CROSS-SECTION '*

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I The consequences of both upper and lower core barrel joint failure have been assessed. - Upper core barrel joint failure during normal operation will cause the core to drop .54" (the gap present at normal operating conditions) onto the reactor vessel guide lugs. Figure 4.1 shows a cross-section of the core and internals with the darkened portion representing components that would drop on UCB joint failure and Figure 4.2 shows the UCB bolt and joint detail in the failed condition.

After core drop, the bolt heads would be held in place by the locking clips, thus precluding becoming loose parts and the bolt shanks would remain captured I

within the core support shield bolt holes. The nominal radial clearance between the bolt shank and the bolt hole is 0.142". While there may be some damage to peripheral fuel assemblies by contact between the baffle plate and upper end fitting of the fuel assemblies, both lateral and rotation motion of 1

the upper core barrel and core would be restrained by the bolt shanks and

, coolable core geometry would be maintained. Likewise the guide lugs and guide j blocks with a 25 mil maximum (range 0.20-0.25 mils) clearance would provide similar restraint at the bottom of the core (See Fig. 2.7). In addition, the combination of bolt shanks at the upper end of the core barrel and guide lugs

and ' guide blocks at the lower end would prevent core tilt in the dropped condition.

With the core in the dropped condition, approximately 5% added bypass flow is s

predicted (see Figure 4.3) which is within existing thermal margins. The fuel assembly upper end fittings remain engaged in the upper grid assembly and since the control rods penetrate 6-1/2" into the fuel assembly guide tubes, control rod insertion would be possible.

A preliminary assessment of the ability of the core guide lugs to support the core in the unlikely event of a complete severance of the UCB or LCB bolted t

joint has been made and the effects o'f the drop on the guide lugs for normal operation, and accident conditions have been evaluated.

o Core Drop - The worst case lug load would occur if the UC8 bolt ring failed during hot standby. This was analyzed using a non-linear dynamic analysis of the lugs. The results indicate a maximum stress intensity of 36 ksi in the critical areas of the lug. This can be compared to a minimum ultimate material strength of 80 ksi.

34

' ' ~'

MAXIMUM 16 -

14 -

/ NOMINAL -

MININUM

/

12 -

e i

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. 8 -

8 / ,

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0 .20 .40 .60 .80 1.0 Core Drop, i n .-

CORE DROP VS ADDITIONAL BYPASS FLOW Fig. 4.3 35

1 Should the drop occur simultaneously with a LBLOCA, the additional load would increase the critical guide lug stress to 52 ksi. This is still well below the guide lug ultimate strength, o Normal Flow - In the dropped condition, the core is held up by flow forces which result in a net downward load on the lugs of only 3,400 pounds. The random turbulence flow excitation of the core would result in low cyclic loads on the guide lugs which could be tolerated indefinitely.

o Accident Condition - A combined LBLOCA and SSE can be resisted by a horizontal load on the end of the guide lugs, in the plane of the lug

, and by the shanks of the 120 severed 1-3/4" bolts. A safety factor of approximately 2.3 and 11.9 exists for the lugs and the bolts, respectively.

o Conclusions - Based on the above preliminary assessments, it is concluded that a dropped core can withstand normal and accident loads with the core remaining in a coolable geometry.

~

The conclusion that can be reached from these considerations is that while UCB bolt ring failure is likely to cause mechanical damage, a significant safety concern does not exist since the core can be shut down, restrained, and cooled without loss of reactor coolant pressure boundary integrity.

Similar conclusions can be reached for failure of the lower core barrel joint since the considerations are the same for this failure mode as for the UCB joint failure, with only two exceptions. Since the core barrel to core support shield joint remains intact no core bypass flow is expected and the load on the guide lugs is reduced by the weight of the core barrel. Figures 4.4 and 4.5 depict the lower core barrel joint failure and in the core dropped condition.

4.2 Justification for Continued Operation 4.2.1 Davis Besse 1 J

As discussed in the previous section, both the upper and lower core barrel bolting have large structural design margins before loss of joint integrity.

36

i i

t T '

REACTOR VESSEL AND INTERNALS CROSS-SECTIONAL ELEVATION ,,

'w la -

x T P T_T._v v._7;--

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part of internal and the core dro onto guide lugs ' \ / /1 n

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_.- U Fig. 4.4 p-.

'37

CORE AND INTERNALS IN DROAPED CONDITION FROM LOWER CORE BARREL JOINT FAILURE i

CORE BARREL CYLINDER NOMINAL RADIAL --

(- -\

z MAXIMUM GAP = 0.045" CLEARANCE =0.225" -

N. i'g I

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L

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1 1/2" NOMINAL RADIAL CLEARANCE 1

} N.B.

/ LATERAL CLEARANCE BETWEEN VESSEL GUIDE

\ LUG AND GUIDE LUG VESSEL BLOCK = O25" MAX.

GUIDE LUG f -

L GUIDE LUG BLOCK 38 FIGURE 4.5

In addition, inspections to date at other plants show failures well below the maximum allowable.

Neutron noise monitoring analysis on May 5,1983, shows no change in the signature from that recorded at the beginning of the cycle. Consequences of joint failure for both normal operation and the very low probability large break LOCA event have been examined and found to be acceptable from a safety standpoint. Lastly, since the Davis Besse refueling outage is scheduled to begin July 29, 1983, there is very limited time for operation. These considerations lead to the judgement that continued operation of the Davis Besse plant is safe and justifiable.

4.2.2 Oconee Nuclear Station The justification for continued operation of the thrca Oconee units is as follows. As shown in Table 4.2 and 4.3, between 15 to 25% of the upper and lower core barrel joints were inspected in late 1981 and the first half of

, 1982 and found to have' no abnormal field UT indications. Backing this is confirming laboratory analysis of bolt samples taken from Oconee 1 and the verification of the UT technique by the bolt samples from Rancho Seco.

As was demonstrated in Section 4.1 and 4.2, there is a similar large structural margins for UCB/LCB bolt ring integrity at these three units as at Davis Besse. In addition, neutron noise analysis is presently being performed at regular intervals for all three units. Duke Power Co. considers that the beam mode vibration of the core barrel is present for all three units. An anomalous spectral indication has been noted at Oconee 1 but detailed analysis shows no significant vibrations in excess of those expected and the signals have not changed significantly stable since 200 EFPD.

In addition, there is an extremely low probability of occurrence of either a large break LOCA or a severe seismic event that would cause significant loading to be applied to the core support structure for the short term for operation of these units (i.e., shutdown for refueling is expected within the next 6 weeks at Oconee 1, within about 5 months at Oconee 2 and in about one year at Oconee 3).

39

'.l' Table 4.2 OCONEE UPPER CORE BARREL BOLTS INSPECTION RESULTS NUMBER NUMBER OF STATION QUANTITY INSPECTED 1 DEFECTS FOUND OCONEE 1 .120 21 (17.5%) 0 OCONEE 2 120 30 (25%) 0 0CONEE 3 120 30 (25%) 0 NOTE 1 - INSPECTED BY ULTRASONIC TESTING l

l 40

~

_ L __ .

Table 4.3 OCONEE' LOWER CORE BARREL BOLTS INSPECTION RESULTS NUMBER NUMBER OF STATION QUANTITY INSPECTED 1 DEFECTS FOUND OCONEE 1 108 16 (15%) 0 OCONEE 2 108 24 (22%) 0 OCONEE 3 108 24 (22%) 0 NOTE 1 - INSPECTED BY ULTRASONIC TESTING e

a O

41

-e ew ,

These considerations led to the conclusion that continued operation of the three Oconee units until their next refueling outage is justified. Strength-ening this judgement are the inspection plans being put in place for the Oconee station: 1) Neutron noise surveillance will continue on all three units at a three-week interval. 2) UT inspection of 100% of the Oconee 1 UCB bolts.during the refueling outage with additional examinations depending on the results of the UT of the UCB bolts. Additionally, the continued operation of the Oconee 2 and 3 units will be reassessed based on the results of the Oconee 1 inspection.

4.2.3 ANO-1 ANO-1 prolonged their present refueling outage to UT inspect their upper core barrel bolts. Only 7 UT indications were noted. These favorable results together with the large structural margins in the UCB/LCB bolt rings justify restart without repair or further inspections. In addition, Arkansas is installing neutron noise analysis equipment and will periodically monitor the signature to insure satisfactory behavior. Also supporting this judgement will be the increasing data base for understanding the bolting failures that will be obtained from near term inspections performed for Davis Besse and the Oconee units.

42 1

a

5. FUTURE ACTIONS Future actions by specific utilities and the Bolting Task Force may be summarized as follows:

o Crystal River and Rancho Seco are proceeding with repair plans and will keep NRC informed.

o AP&L plans to restart ANO-1 shortly, o Oconee I will be shutting down soon for refueling and inspection-to be followed shortly after by 08-1. UCB bolts will be inspected and results will be provided to the NRC.

o B&W Owners Group Task Force is developing their long range plans. A Task Force meeting dedicated to the development of such plans is presently scheduled for May 24. The plan, when formalized, will be provided to the

, NRC.

l i

l 43 L

APPENDIX A NRC - B8W OWNERS GROUP M ETING SLIDES REACTOR VESSEL INTERNALS BOLTING MAY 6, 1983 l

l 1

f

. 4 . -. .

NRC.- B&W OWNERS GROUP MEETING AGENDA REACTOR VESSEL ItHERNALS BOLTItG MAY 6, 1983 -

I. ItERODUCTION A. B&WCG ACTIONS TED MYERS-TED B. B&WOG TASK FORCE APPROACH SKIP HENDRIX - DUKE II. TECHNICAL BACKGROUND A. PROBLEM IDENTIFICATION D003 LEE - B&W AND INTERNALS JOINTS

DESIGN BASIS B. INSPECTION ETHOUS GARY ABELL - B&W III. INSPECTION AND EXAMINATION RESULTS A. SITE LARRY TITTLE - FPC B. LABORATORY BOB PIASCIK - BaW IV. SArtifIMPlJCATIONS DOUG LE - B&W V. PLANT STATUS LARRY YOUNG - TED

__ PAUL GUILL - DUKE DAN HOWARD - APa' VI. FUTUREACTIOBIS Slide 1

f r

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, 5' j /'

,. }

4.

TASK FORCE MISSION s ! '.

s j /

'r e

I, MAINTAIN COGNIZANCE OF ALL ACTIVITIES RELATED'TO INTERNAL BOLT FAILURES,

/ y

, - ; II, PROVIDE FORLM FOR SHARING OF INFORMATION.

III, MAINTAIN'dONSISTENT LICalSING POSTURE.

IV, CONSIDER ALL GENERIC ASPECTS ADDRESSING BOTH SAFETY AND ECONOMIC ISSUES, V, EXPEDITE PERFORMANCE OF NECESSARY ACTIVITIES, L

l l

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l e ,

I f n

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)

f

(; u Slide 2 i t. .

I

l B&W O M RS ACTIONS TO DATE o INITIAL EVALUATIGft f

,, THERE ARE LARCE DESIGN MARGINS IN 1BE REACTOR VESSEL BOLTING SYSTETtS ClRRENT INSPECTIONS STATUS REVEALS BOLT FAILURE LEVELS WELL BELOW THOSE REQUIRED FOR JOINT FAILURES EVEN IF JOINT FAILURE OCCURS THERE IS NO SIGNIFICANT SAFETY HAZARD

' INVOLVED THEREFORE CURRENTLY OPERATING FACILITIES PROVIDE NO UNDUE RISK TO PUBLIC HEALTH AND SAFETY o ESTABLISHED AN ORGANIZATION AMONG THE B&W OWNERS TO CONTINUE,TO IfNESTIGATE AND EVALUATE NEW

' ~

INFORMATION AND WILL CONTIt0E TO RESOLVE THE PROBLEMS WE HAVE FOUt0

/

t Slide 3

A 286 BOLT CHRONOLOGY DA4 o THERMAL SHIELD BOLT FAILURES JULY 15, 1981 AT OCONEE 1 o THERMAL SHIELD REPAIRS OCONEE 1 NOVDEER 1981 OCONEE 2 FEB. 22, 1982 OCONEE 3 JULY 7, 1982 AND-1 JAN,10,1983 RANCHO SECO UNDERWAY t

CR-3 UNDERWAY DB-1 1984 o UPPER CORE BARREL BOLT MARCH 25, 1983 DEFECTS AT RANCHO SECO o UPPER CORE BARREL BOLT APRIL 10, 1983 DEFECTS AT CR-3 Slide 4

~

+ -

PROBLEM DEFINITION o UT INDICATIONS WERE DETECTED AS EXPECTED IN TliE LOWER THERMAL SHIELD BOLTS IN THE PROCESS OF MAKING PLANNED REPAIRS.

o UNEXPECTED (K INDICATIONS WERE DETECTED IN THE FOLLOWING BOLTING RINGS:

- UPPER CORE BARREL

- LOWER CORE BARREL

- SURVEILLANCE HOLDER TUBE o INDICATIONS WERE FOUND IN SA453 GR660 (A-286) BOLTS o THE UPPER CORE BARREL AND LOWER C0"E BARREL BOLTING RINGS HAVE CORE SUPPORT STRUCTURAL SIGNIFICANCE o CONCERNS ARE APPLICABLE TO 177FA PLANTS EXCEPT TMI-1 i

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Slide 5

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LOWER CORE BARREL (108 - 1/ 3/4 DIA.)

FLOW DISTRIBIROR TO LOWER GRID (% - 1 DIA.)

UPPER THERMAL SHIELD (60 - 11/2 DIA.)

LOWER THERMAL SHIELD (96 - 1 DIA.)

SURVEILLANCE HOLDER TUBE (73 - 3/4 DIA.)

SA 433 GR 660 (A-286) IS AN AGE HARDENED, HIGH STRENGTH, CORROSION RESISTANT MATERIAL. BOLTS ARE CONDITION A/ UPSET HEADED EXCEPT THE ,

SURVEILLANCE HOLDER TUBE BOLTS WHICH ARE CONDITION B/ MACHINED.

DB-1, CR-3 AND MIDLAND 1 & 2 DEL 5/4/83 Slide 8

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DESIGN BASIS DESIGNED TO WITHSTAND LOADINGS RESULTING FROM NORMAL AND UPSET CONDITIONS:

8 PRESTRESS I DEAD WEIGHT I THERMAL t OBE 0 VIBRATION DESIGNED TO WITHSTAND LOADING RESULTING FROM DESIGN BASIS ACCIDENT (FAULTED) CONDITIONS.

O SSE 8 LOCA .

I DEAD WElGHT 8 CORE B0UNCE '

ALLOWABLE STRESS CRITERIA FOR NORMAL AND UPSET CONDITIONS WAS SECTION 111, SUBSECTION NB.

ALLOWABLE STRESS CRITERIA FOR FAULTED CONDITIONS WAS AS PRESENTED IN FSAR.

JOINT DESIGN CRITERIA WAS NO SEPARATION DURING FAULTED CONDITION LOADINGS.

l SIGNIFICANT MARGINS ARE INCLUDED IN THE DESIGN (ALLOWABLE STRESS AND NUMBER OF BOLTS REQUIRED TO CARRY LOADINGS).

DEL 5/4/83 Slide 13

MINIf1JM NifBER OF BOLTS REQURIED FOR NORMAL OPERATION UPPER CORE BARREL 120 BOLTS PER RING (3 BOLTS RECJIRED TO SUPPORT CORE)

NLFBER PLANT OF BOLTS RANCHO SECO 8 CR-3 8 ANO-1 8 OCONEE 1 27 OCONEE 2 12 OCONEE 3 8

~

DAVIS-BESSE 1 8 MIDLAND 1 & 2 8 DEPENDS ON INSTALLATION TOR 0lJE Slide 14

MINIME NIMER OF BOLTS REQUIRED FOR NOFFAL OPERATION LOWER CORE BARREL 108 BOLTS PER RING NIMER PLANT OF BOLTS RANCHO SECO 13 CR-3 9-AND-1 13 OCONEE 1 24 0CONEE 2 13 OCONEE 3 9 DAVIS-EF$ E 1 9 MIDLAND 1 & 2 9 l

DEPENDS ON INSTALLATION TORQUE Slide 15

[ .___ . _ _ _ _ . -

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Crack Site I

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Slide 16 l ..

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ULTRASONIC CALIBRATION TECHN100E

_ 0 POSITION TRANSDUCER ON GOOD BOLT STANDARD AND SET SIGNAL RESPONSE FROM HEAD TO SHANK AT 40% FULL SCREEN HEIGHT (FSH).

RECORD DB REQUIRED FOR THIS SENSITIVITY.

O REPEAT WITH TRANSDUCER ON 15% NOTCH STANDARD.

O REPEAT WITH TRANSDUCER ON 50% NOTCH STANDARD.

Slide 17

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^I TRAN300CER CRACK CRACKED BOLT S1ide 18

~

CALIBRATION STANDARDS O BOLTS OF SAME NOMINAL MATERIAL AND SIZE.

O THREE BOLT STANDARDS - ONE CLEAN; ONE 15% NOTCH; ONE 50% NOTCH.

o NOTCHES CUT IN SHANK JUST BELOW HEAD. .

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RESULTS TO DATE o ALL FIELD OBSERVED IfFORMATION i

CORRELATES WITH UT RESULTS (BROKEN HEADS) o ALL LABORATORY DESTPJJCTI\E ANALYSIS OF BOLTS RLLED FROM SITES CORPEArts WIll! UT RESULTS l

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Slide 20

5-4-83 llT INSPECTION RESULTS (# CRACKED / # INSPECTED)

BOLTED JOINT OCONEE 1 OCONEE 2 OCONEE 3 ANO-1

' RANCil0 SECO CR-3 (9-81) (1-82) (6-82) (5-83) (3-83) (4-83)

UPPER CORE BARREL 0/21 0/30 0/30 7/120 19/120 51/120 (120 3/11 DIA.)

LOWER CORE BARREL 0/16 0/24 0 / 2 11 0/108 4/108 (108 3/4 DIA.)

FLOW DIST/ LOWER GRID 0/22 0/25 0/25 ----

0/93 0/96 ,

(96 - 1 DIA.)

UPPER TilERMAL SillELD 0/25 0/20 0/20 ----

0/60 0/60 ;

(60 - 11 DIA.)

LOWER TilERMAL SillELD 11/13 28/93 53/96 51/96 77/96 73/96 (96 - 1 DIA.) 914 HEADS SEVERAL SEVERAL 48 75 IIEADS 69 elEADS TWISTED HEADS llEADS HEADS TWISTED i

TWISTED OFF TWISTED TWISTED TWISTED OFF OFF OFF 0FF - '

0FF SURVEILLANCE il0LDER TUBE (72 - 3/li DIA.)* ---- ---- ---- ---- ----

25/72

'DB-1 AND CR-3 ONLY Slide 21

TOTAL (IT INSPECTION RESULTS

( UCB, LCB, FLOW DISTRIBUTOR)

TOTAL BOLTS TOTAL BOLTED J0ItIT INSPECED INDICATIONS %

UPPER CORE BARREL 441 77 17.5 LOWER CORE BARREL 280 4 1,4 FLOW DIST/ LOWER GRID 261 0 0 Slide 22

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/04 3ECT/ONZ Z tos Slide 26

OCONEE I EXAMINATIONS

1. _ LOWER THERMAL SHIELD BOLT (8)

- VISUAL, PT, SEM, METALL0 GRAPHY RESULTS - INTERGRANULAR FRACTURE IN THE BOLT HFAD TO SHANK TRANSITION REGION.

2. UPPER THERMAL SHIELD RESTRAINT BOLT (3)

- VISUAL, PT, SEM SURFACE EXAM RESULTS - NO CRACKING.

3. UPPER CORE BARREL BOLT (2) .

- VISUAL, PT, SEM SURFACE EXAM RESULTS - NO CRACKING.

4. FLOW DISTRIBUTOR BOLT (1)

- VISUAL, PT RESULTS - NO CRACKING l ~'

Slide 27

SKETCH OF LOWER THERNAL SHIELD BOLT ..

SHOWING LOCATION OF FRACTURE -

l , .

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LOCATION OF PRINCIPAL .

FRACTURE a w i

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Slide 28

E OCONEE II EXAMINATIONS LOWER THERMAL SHIELD BOLTS (19)

VISUAL, UT, SEM,.METALLOGRAPHY RESULTS - INTERGRANULAR CRACKING IN THE BOLT HEAD TO SHANK TRANSITION REGION, .

l Slide 29 l

1 - - -

r . .

4 f

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CONCLUS!0NS (0CONEE EXAMS)

1. ALL FAILURE DUE TO INTERGRAAULAR STRESS ASSISTED CRACKING.

d 2, ALL FAILURES LOCATED IN THE HEAD TO SHANK y-1 .,

TRANSITION REGION.

.s

.- 3. ' BOLTS WHICH CONTAINED NO SITE U/T If0ICATIONS

~

WEREVERIFIEDTOHAVENOFLAWS(CRACKS [

_. USING DETAILED LAB 9RATORY EXAMINATIONS.

l 4

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4 Slide 30

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, . _ , . , . - - __ . - . . L : :_ - . __ . L , _ _ . _ . .

RANCHO SECO UPPER CORE BARREL BOLT MATERIAL - ALLOY A-286 CONDITION A PROCESSING - HOT HEADED

- SOLUTION TREATED

- 0 1650 F FOR 2 HOURS

/

- AGE' HARDENED

~

0 1325 F FOR 16 HOURS

- THREADS . ROLLED

'c 1

51ide 31 '

RANCHO SECO EXAMINATIONS BOLT TYPE: UPPER CORE BARREL' BOLT QUINTITY: 2 BOLTS WITH SITE U/T INDICATIONS (41, 116) 1 BOLT WITH NO SITE U/T INDICATIONS (79)

EXAMINATIONS: - VISUAL (ALL)

- U/T (ALL) '

.PT (BOLT 79)

- SEM/EDX FRACTURE SURFACE (BOLT 116)

- SEM/EDX OF SMALL DEFECT AREA OPEN BY TENSILE LOADING (BOLT 79)

- METALL0 GRAPHY (BOLT 41 & 79)

Slide 32

~

, RANCHO SECO EXAMINATION RESULTS BOLT #41, #116 (CONTAINING SITE U/T INDICATIONS)

1. INTERGRANULAR FRACTURE WITH EXTENSIVE SECONDARY CRACKING (BRANCHING).

2. CRACKING INITIATED IN THE HEAD TO SHANK TRANSITION REGION.

3. CRACK PROPAGATION SIMILAR TO THAT FOUND IN THE LOWER THERMAL SHIELD BOLT, I.E., FOLLOWED HAZ CURVATURE.

4. THE HEADS OF.BOTH BOLTS WERE ATTACHED BY SMALL LIGAMENTS FORMED BY THE SECONDARY CRACKING.

(CRACKING PROPAGATED THROUGH ~ 95% OF BOLT SHANK.) -

Slide 33

SKETCH OF UPPER CORE BARREL BOLT -

SHOMING LOCATION OF FRACTURE

/h/b/b/h.,/h/h.,y.

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Slide 34

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RANCHO SECO EXAMINATION PESULTS BOLT #79 (CONTAINING NO SITE U/T INDICATIONS)

1. FLAW CONTAINS LOCALIZED DEPOSITS i

2. FLAW IS NOT INTERGRANULAR IN NATURE.

3. NO SECONDARY INTERGRANULAR CRACKING WAS OBSERVED.

4. SURFACES BETWEEN FLAWS OPENED BY DUCTILE RUPTURE WHEN PULLED APART.

5. TOTAL FLAW SIZE MUCH LESS THAN 15% SITE U/T CALIBRATION STANDARD.

1 Slide 35

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RANCHO SECO EXAMINATION CONCLUSION 1.

RANCHO SECO UPPER CORE BARREL BOLT FAILURES ARE TYPICAL OF PREVIOUS. BOLT FAILURES.

2.

BOLT #79 FLAW IS AN ANOMALLY AND NOT ASSOCIATED WITH BOLT FAILURES. THE FLAW IS METALLURGICALL/ INERT SHOWING NO EVfDENCE '0F PROPAGATION. .

Slide 37

OVERALL CONCLUSIONS

1. ALL BOLT FAILURES ARE DUE TO INTERGRANULAR STRESS ASSISTED CRACKING LOCATED IN THE BOLT HEAD TO SHANK TRANSITION REGION.

2. DETAILED LABORATORY EXAMINATIONS CONFIRM SITE UT FINDINGS. .

, Slide 38

~

LIKELIHOOD OF CORE DROP VERY LOW BASED ON STRUCTURAL MARGINS AND RESULTS OF UT INSECTIONS I

l UPPER CORE BARREL (120 BOLTS)

MIN # BOLTS RESULTS OF UT INSPECTIONS REQUIRED FOR buuu it tu 1vt PLANT NORMAL OPER BOLIS. B_OLH RANCHO SEC0 8 101 19 CR-3 8 69 51 AND-1 8 113 7 LOWER COPE BARREL (108 BOLTS) .

RANCHO SECO 13 108 0-CR-3 9 104 4 Slide 39

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onto guide lugs j1 I

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l Slide 40

- - __ __ =__________ _

UPPER CORE BARREL BOLTING RING FAILURE CONSEQUENCES o BOLTS FAIL UNDER THE HEAD AS DETERMINED FROM LABORATORY EXAMINATIONS o . CORE BARREL ASSEMBLY AND FUEL DROPS .54" AND IS SUPPORTED BY GUIDE LUGS o BOLT HEAD REMAINS CAPTURED BY LOCKING DEVICE PRECLUDING LOOSE PARTS o BOLT SHANKS REMAIN ENGAGED (2h") IN CORE SUPPORT SHIELD THUS RESTRAINING CORE BARREL ASSEMBLY. RADIAL CLEARANCE BETWEEN SHANK AND HOLE IS .?42".

-o NOMINAL RADIAL GAP BETWEEN CORE BARREL CYLINDER AND CORE SUPPORT SHIELD IS .170". FLOW BYPASS AROUND COPE IS 5%. ADEQUATE THERMAL MARGIN EXISTS.

o FUEL ASSEMBLY UPPER END FITTING REMAINS EFEAGED IN UPPER GRID o CONTROL RODS CAN BE INSERTED SINCE IN FULLY WITHDRAWN POSITION, THEY PBETRATE INTO FUEL ASSEMBLY GUIDE TUBES 6-i INCHES.

o CORE BARREL BAFFLE PLATE LIKELY TO CONTACT PERIPHERAL FUEL ASSEMBLIES AT THE UPPER END FITTING WHICH APE CAPABLE OF CARRYING HIGH LOADS IN SHEAR.

o LATERAL LOADS (INCLUDING ASSYETRIC LOCA LOADS) ARE RESISTED BY:

THE BOLT SHANKS AT THE UPPER SEVERED JOINT THE GUIDE LUGS AT THE BOTTOM Slide 41

UPPER CORE BARREL BOLTING RING FAILURE CONSEQUENCES, CONT'D, 1

o VERTICAL LOADS (INCLUDING COPE BOUNCE) ARE RESTRAINED IN: ;

THE UPWARD DIRECTION BY THE INTACT CORE SUPPORT SHIELD AND UPPER GRID STRUCTlJRE THE DOWtMARD DIRECTION BY THE GUIDE LUGS WITH A 30% MARGIN TO CODE ALLOWABLES o FATIGUE EVALUATION OF GUIDE LUGS RESULTS IN LOW CYCLE LOADS (3400 LBS) WHICH COULD BE TOLERATED It0EFINITELY CONCLUSION: UPPERCOREBARP5.BOLTINGRINGFAILUREISLIKELY TO CAUSE ECHANICAL DAMAGE BUT IT IS t0T A SAFETY CONCERN BECAUSE:

CORE CAN BE SHUIDOWN CORE WILL BE RESTRAINED AND COOLED REACTOR C00LAfff PRESSURE BOUNDARY WILL BE MAINTAINED Slide 43

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Slide 42

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Slide 44 I

LOWER CORE BARREL BOLTING RING FAILURE CONSEQUENCES o BOLTS FAIL UNDER THE HEADS o FUEL AND LOWER GRID ASSEMBLY DROPS .54" AND IS SUPPORTED BY GUIDE LUGS o BOLT HEADS REMAIN CAPTURED BY LOCKING DEVICES o BOLT SHANKS REMAIN ENGAGED IN LOWER GRID ASSEMBLY o FUEL ASSEMBLY ENGAGEFENT AND C0iVfROL ROD INSERTION CAPABILITIES IS MAINTAINED .

o LATERAL MDTION OF THE LOWER GRID ASSEMBLY RESTRAINED BY THE GUIDE BLOCKS AND GUIDE LUGS ( 020 To .040 IflCH CLEARANCE) o VERTICAL MDTION RESTRAINED BY THE CORE BARREL AND GUIDE LUGS.

o LOADINGS ON GUIDE LUGS ARE LESS SEVERE THAN FOR UPPER CORE BARREL BOLTING RING FAILURE CONCLUSION: SAE AS FOR UPPER CORE BARREL BOLTING RING FAILURE Slide 45

BASIS FOR CONTINUED DAVIS-BESSE OPERATION o LARGE STRUCTURAL MARGINS o OTHER PLANT INSPECTION RESULTS o SATISFACTORY PRESEfff CONDITION NElfiRON NOISE (VERIFIED ON MAY 4,1983)

LOOSE PARTS COOLANT ACTIVITY o JOINT LOOSENING IS DETECTABLE o ACCEPTABLE SAFETY CONSEQUENCES DUE TO CORE DROP DURING NORMAL OPERATION o VERY LOW PROBABILITY OF CTBINED CORE DROP PLUS HIGH LOAD (LBLOCA) ACCIDENT o ACCEPTABLE SAFETY CONSEQUENCES DUE TO CORE DROP PLUS HIGH LOAD ACCIDBff o LIMITED PLANT OPERATION UNTIL REFUELING (JULY 29, 1983)

Slide 46

JUSTIFICATION FOR CONTINUED OPERATION FOR OCONEE NUCLEAR STATION e PAST UT EXAMINATIONS e LAB EXAM 0F BOLT SAMPLES e SIGNIFICANT STRUCTURAL MARGIN ,

e NNA RESULTS e ABILITY TO DETECT CORE DROP e CONSEQUENCES OF CORE DROP e LOW PROBABILITY OF DBA e FUTURE INSPECTION PLANS l

l Slide 47 l -- . . . -,

UPPER CORE BARREL BOLTS INSPECTION RESULTS NUMBER NUMBER OF STATION QUANTITY INSPECTED 1 DEFECTS FOUND OCONEE 1 120 21 (17.5%) 0 OCONEE 2 120 30 (25%) 0 OCONEE 3 120 30 (25%) 0 NOTE 1 - INSPECTED BY ULTRAS 0hlG TESTING f

, Slide 48

l l

l LOWER CORE BARREL BOLTS INSPECTION RESULTS NUMBER NUMBER OF-STATION QUANTITY INSPECTEDI DEFECTS FOUND OCONEE 1 108 16 (15%) 0 OCONEE 2 108 24 (22%) 0 OCONEE 3 108 24 (22%) 0 NOTE 1-INSPECTEDBYULTRASONICTESTING Slide 49

LAB EXAMINATION OF BOLT SAMPLES O REMOVED 2 UCB BOLTS FROM 0-1 DURING 1981 REFUELING OUTAGE (#1, #60) 0 FIELD lif SHOWED BOTH BOLTS L7E K i i N ~T4 C. "T O BOLT #60; HAD NO INDICATIONS OF FAILURE PER THE FOLLOWING TESTS: ,

- FLUORESCEtR DYE PENETRANT

- SCANNING ELECTRON MICROSCOPE O BOLT #1: HAD NO INDICATIONS OF FAILURE PER THE FOLLOWING TESTS:

- FLUORESCENT DYE PENETRANT l - ENHANCED LAB ULTRASONIC TECHNIQUE l

\

l t

i Slide 50

NNA RESULTS e NO ANOMALOUS SPECTRAL BEHAVIOR HAS BEEN OBSERVED FOR UNITS 2, 3.

e ANOMALOUS SPECTRAL BEHAVIOR HAS BEEN OBSERVED FOR UNIT 1.

- NOT INDICATIVE OF A CORE DROP

- ANOMALY HAS STABILIZED FOR 200 EFPD e BEAM MODE VIBRATION OF THE CORE BARREL IS

~

PRESENT FOR ALL THREE UNITS.

Slide 51

ABILITY TO DETECT CORE DROP e NEUTRON NOISE SURVEILLANCE

-e LOOSE. PARTS MONITOR e SPNDs, INCORE T/C

- POSSIBLE DETECTABLE GLOBAL CHANGE IN

.THESE INSTRUMENTS l

Slide 52

_ PROBABILITY OF DBA o ACCIDENTS WITH POTENTIAL FOR SIGNIFICANT LOADING ON COPE SUPPORT STRUCTURE ARE LBLOCA AND SEVERE SEISMIC EVENTS, o THE PROBABILIT( OF OCCURRENCE OF THESE RARE EVENTS FOR THE LIMITED PERIOD OF INTERIM OPERATION IS ACCEPTABLY Si%LL, o- PERIOD OF INTERIM OPERATION:

- 2 WEEKS FOR UNIT 1

- 5 t0NTHS FOR l' NIT 2

- 1 YEAR FOR UNIT 3 Slide 53 t

ANTICIPATED SCHEDULE FOR OCONEE NUCLEAR STATION UNIT 1 SHUTDOWN FOR RERJELING ArTER COMPLETION OF UNIT 2's MAINTENANCE OUTAGE.

UNIT 2 MAINTENANCE OUTAGE TO EEGIN THIS hEEK90.

SHIRDOWN FOR REFUELING EARLY OCTOBER.

UNIT 3 SHUTDOWN FOR REFUELING APRIL 1984 4

1 4

Slide 54 L_

INSPECTION PLANS FOR OCONEE NUCLEAR STATION o CONTINUE NNA SURVEILLAN E

- SURVEILLANE FREQUENCY OF 3 WEEKS o PERFORM 100% UT OF UCB DURING 01 REFUELING OUTAGE o PERFORM ADDITIONAL EXAMINATIONS

. BASED ON THE ABOVE RESULTS J'

i o RE-EVALUATE OPERATION OF 02, 03

.( BASED ON THE AB0VE RESULTS F

l' l

.I Slide 55

StPMARY o UT's SHOW NO BOLT FAILURE o NNA RESULTS INDICATE BEAM MDDE VIBRATION IS PRESENT FOR ALL THREE UNITS o SUBSTRUCTURAL UPPER JOINT FAILURE IS NOT EXPECTED FOR INTERIM PERIOD OF OPERATION o HEALTH AND SAFETY OF C{NERAL PLBLIC IS NOT ENDANGERED BY A CORE DROP SCENARIO Slide 56

{

ANO-1

.o PLANT STATUS SHlHDOWN FOR REFUELING 11/82 STARTUP DELAYED ON 4/25/83 TO PERFORM INSPECTION OF UCB BOLTS INSPECTION COMPLETED 5/3/83 PREPARATIONS FOR RESTART IN PROGRESS

o. INSPECTION RESULTS 7 0F 129 UCB BOLTS EXHIBITED UT INDICATIONS o BASIS FOR RESTART

~

FAVORABLE INSPECTION RESULTS AVAILABLE MARGIN -

EVALUATION OF CONSEQUENCES OF-POTENTIAL FAILURE ~ .

o FIRURE ACTIONS CONTINUEDP4RTICIPATIONINTASKFORCE ACTION FUTURE INSPECTIONS /REPAIPS TO BE EVALUATED PEUTRON NOISE ANALYSIS EQUIPMENT BEING INSTALLED FOR USE THIS CYCLE Slide 57

FifrURE ACTIONS i

f o- THE UTILITIES PLAN TO FORMALLY DOClM NT B E INFORMATION PRESENTED TODAY (WITH 2 WEEKS) o CRYSTAL RIVER AND RANCHO SECO ARE PROCEEDING WIE PEPAIR PLANS AND WILL KEEP SYD MINER INFORED 1

o ; APaL PLANS TO PESTART AND-1 SHORTLY o OCONEE 1 WILL BE SHUTTING DOWN SOON FOR REFUELING AND INSPECTION; BOLTS WILL BE INSPECTED AND l, _

RESULTS WILL BE PROVIDED TO THE NRC

o. B&W OWNERS GROUP TASK FORCE IS DEVELOPING THEIR LONG RANGE PLANS WHICH WILL BE C0011UNICATED TO THE^NRC l

l l

l .

L

, Slide 58 m

8 4 t

I

?

APPENDIX B SITE ULTRASONIC EXAMINATION PROCEDURE 4

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Development of Upper Core Barrel Bolt U.T. Procedure The procedure for the ultrasonic examination of upper core barrel bolts was developed as a special case of a gen 2ral procedure for examining bolts used on reactor internals. The initial objectives of the development were positive identification of bolts with no cracks or cracks extending 50% or more through the shank. A probable identification of bolts with cracks from 15% to 50%

through the shank was also desired.

Because of physical constraints imposed on the examination by radiation considerations, the examination was to be conducted with the reactor internals immersed in at least 20 ft. of water. This implied that the transducer must be positioned remotely and that detailed scanning of the transducer was not practical.

Several transducer frequencies and sizes were considered, the selected parameters were chosen so that with the transducer centered on the bolt head, .

a small amount of ultrasound would impinge on the shank to head fillet region, where failures were apparently initiating. A small amount of ultrasound is required to provide sensitivity to small defects, but the amount must be small or the sensitivity would be low.

The selected technique was based on measuring the change in gain required to bring the shank-to-head fillet signal to 40% screen height while monitoring the reflection from the threaded end of the bolt. Cracks would cause a j reduction in gain.

l Calibration was performed using bolts with sawcuts made into the shank immediately under the head. The depths were approximately 0,15% and 50% of

the shank diameter.

l B-1

BABCOCK & WILCOX INSERVICE INSPECTION PROCEDURE EUEJECT: ULTRASONIC EXAMINATION OF UPPER RESTRAINT BLOCK BOLTS, FLOW DISTRIBUTOR BOLTS, CORE ISI-16 5, Rev . 1 BARREL BOLTS & LOWER THERMAL SHIELD BOLTS

1. SCOPE: This procedure shall govern the ultrasonic method of detecting and evalvating bolt defects in the head to shank region of 1 upper restraint block bolts, flow distributor bolts, core barrel bolts, and lower thermal shield bolts.

2. EXAMINER QUALIFICATIONS:

2.1 Ex aminer: The examiner performing the examination shall be qualified to Level II in accordance with the Babcock & Wilcox Company Administrative Procedure ISI-21. The Level II shall be responsible for and shall accept the results of the examination. The examiner shall have a minimum of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 training in the technique described in this procedure.

2.2 Assistant

The assistant shall be qualified to at least Trainee or Level I in accordance with the Babcock & Wilcox Company Administrative Procedure ISI-21. The assistant shall not independently evaluate or accept the results of the examination.

The assistant shall perform the examination in accordance with this procedure under the guidance of an examiner defined in 2.1. %

When the examination is performed by a Trainee, the Level II or Level III shall observe the performance of the examination to ensure that the requirements of this procedure are met.

3. EQUIPMENT: The equipment required to perform the measurements shall include at least the following:

3.1 UT Scope: A pulse-echo type ultrasonic flaw detection instrument shall be used.

3.2 Cables

Coaxial or microdot cables or a combination thereof may be used.

1 3.3 Search Units: Nominal 2.25 MHz 1/2" or 1" round transducer applicable to the bolt size and mounted in a spring loaded t

ixture (Figure 1) for remote operation.

3.4 Couplant

A suitable liquid couplant medium such as borated, demineralized, or distilled water shall be used. Water which meets the station specifications for compatability with the reactor vessel internal surfaces shall be considered adequate.

The couplant temperature for calibration shall be within 25*F (14*C) of the couplant temperature used for the examination.

A NIS PATIVE APPROVAL TEJC NKAL APPROVAL j ( LEV. III) QA APPROVAL

/2 -242-- mEgbAw A//A ISSUEQ/ REVISED BY ISSUE DATE REVISION DATE PAGE NO.

GAT 8-6-81 11-29-82 1 OF 5 B-2

_ _ _ _ . - = - _ _ __ ___ .- _ - ...

BABCOCK & WILCOX INSERVICE INSPECTION PROCEDURE l S UBJ ECT: ULTRA 5ONIC EXAMINATION OF UPPER RESTRAINT BLOCK BOLTS, FLOW DISTRIBUTOR BOLTS, CORE ISI-165, Re v . 1 :

BARREL BOLTS & LOWER THERMAL SHIELD BOLTS ___

l

, 4. CALIBRATION BLOCK:

4.1 Material

The calibration block or blocks shall be of the same

nominal composition as the component to be examined.

4.2 Size: The diameter of the calibration block or blocks shall be of the same nominal diameter as the component to be examined.

The minimum length of the calibration blocks shall be the length of the bolt being examined.

4.3 Reflectors

Three blocks (Figure 2) shall be used for each bolt examination performed. One bolt shall have no flaws (Figure 2A). One bolt shall have a transverse notch cut to a depth of approximately 15% of the bolt's shank diameter (Figure 2B). One bolt shall have a transverse notch cut to a depth of approximately 50% of the bolt's shank diameter (Figure 2C).

j 5. SYSTEM CALIBRATION:

5.1 Range

The sweep range shall be established by using either an IIW block, a step wedge, the actual calibration block, or a combination of these.

5.2 Sensitivity

The sensitivity level for the examination shall be established by positioning the search unit on the bolt head of the good bolt. The gain should be adjusted to provide a 40%

(+ 54) FSH response from the head to shank radius of the bolt.

Record the calibration settings on the calibration / data sheet 4 (Figure 3).

The calibration sensitivity shall be demonstrated capable of detecting cracks in the head to shank region of the bolt by positioning .the search unit on the bolt heads of the standards

! which are notched at 154 and 50%. Record the gain setting 1 L! required to set the notches at 40% (+ 54) FSH on the calibration / data sheet. -

6. EXAMINATION REQUIREMENTS:

Position the search unit housing fixture securely over the head of i i

' ' the bolt to be examined. Adjust the gain .to provide a 40% (+ 54)

FSH response from the head to shank radius of the bolc. A cracked i bolt will generally provide a defect signal at or near the signal lr from the head to shank radius and will display an amplitude greater

( '

l than 40% FSH. The amplitude of the back reflection will generally decrease in proportion to the size of the crack. This, in addition i to the results of the sensitivity demonstration on the 154 and 50%

l{ notches, should be the basis for determining the condition of the i bolt. The bolt number and the results of the examination should be recorded on the calibration / data sheet.

f PAGE NO.

[ B-3 2 OF 5

_ , - - _ _ _ . _ _ _ _ . . . . _ _ . , , . _ . _ _ _- . . . _ . _ , _ . . . _ . , _ _ . . , - ~ , . . , .- , _ . _ . ~ , . _ _ , _ _ , . . _ _ - _ _ _ - - - .

BABCOCK & WILCOX INSERVICE INSPECTION PROCEDURE 5 UBJEC T : ULTRA 5ONIC EXAMINATION OF UPPER RE S T R A I N T BLOCK BOLTS, FLOW DISTRIBUTOR BOLTS, CORE ISI-165, Rev. 1 BARREL BOLTS & LOWER THERMAL SHIELD BOLTS RANSOUCER .

,1 .

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CM c x ~

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SPRING _

,., CABLE TO INSTRUMENT I

Fi gure 1. Springloaded Fixture PAGE NO.

3 0F 5-3-4

n .

BABCOCK & WILCOX INSERVICE INSPECTION PROCEDURE

SUBJECT:

ULTRA 50NIC EXAMINATION OF UPPE R RE S TRAIN T BLOCK BOLTS, FLOW DISTRIBUTOR BOLTS, CORE 151-165, Rev. 1 BARREL BOLTS & LOWER THERMAL SHIELD BOLTS CALIBRATION BLOCK FIGURES

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Note 1 - For flow distributor bolts and thermal shield bol ts, the notch is cut approximately 1/16" below the flange.

Note 2 - For the upper restraint bolts and the core barrel bol ts, the notch is cut approximately 1/8" 5elow the flange.

Figure 2. Calibration Blocks.

PAGE NO.

4 0F 5 E-5

BABCOCX & WILCOX INSERVICE INSPECTION PROCEDURE TUB'JLcI:

UL TR A5 0NIG E XAMIN Ai10N Or UPPLR kt5iKAINI BLOCX BOLTS, FLOW DISTRIBUTOR BOLTS, CORE ISI-165, Rev. 1 BARREL BOLTS & LOWER THERMAL SHIELD BOLTS O m, G I

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PAGE NO.

i 1 5 0F 5 i

3-6

.'. J

. APPENDIX C REACTOR VESSEL INTERNALS BOLTING INSTALLATION TORQUE VS PRELOAD TEST DATA

.. s Torque Test Data Testing was conducted in 1972 and 1973 and later in 1981 in which prototypes of the 1" (lower thennal shield),1-1/2" (upper thermal shield restraint), and 1-3/4" diameter (upper and lower core barrel) bolts were strain gaged to determine bolt load vs. applied torque. The main purpose of these tests was to establish the required torque to obtain the prescribed design bolt preload stress. Strain gages were applied to the smooth portion of the bolt shanks away from the areas of stress concentration. In all tests, neolube was applied to the threads and under-head bearing surface of the bolts as was c-i e.s d-3 done on bolts installed in the reactor internals. Tables I, 4 and 4 present the pertinent data cbtained for the 1-3/4" diameter,1-1/2" diameter, and 1" diameter bolts, respectively.

A column in the attached tables labeled " Torque Sequence" describes the steps taken in installing b given bolt. Where two lines of values are shown, the step wife sequence was to torque from hand tight to a given value shown in the first line, then detension and retorque to the final value shown in the second line.

c-1

Table C-1 1-3/4" Diameter Upper and Lower Core Barrel Bolt Stress / Torque Test Data o Smooth portion of four bolt shanks were instrumented with strain gages.

o Neolube was applied uniformally to threads and bearing surface of head, o Bolts were torqued into the Rancho Seco reactor internals.

Stress in KSI

, Torque Torque Bolt Bolt Bolt Bolt Ft-lbs Seouence 1 2 3 4 Comments 0-2100 1000 0-1750 12.2 10.9 8.5 7.0 0-2100 Bolts 3&4 were relu-1000 0-1750 . - - --

11.4 8.2 bricated and retor-qued in same hole.

0-2100 1750 0-1750 19.1 18.9 16.9 13.8 0-2100 1750 0-1750 -- --

20.1 15.2 2000 0-3000 -- --

22.7 26.1 Under heads of bolts 384 w?re refinished.

2000 0-3000 -- -- --

24.7 Bolt 4 was relubri-cated & retorqued in different hole.

2400 0-3000 -- --

25.7 32.7 3000 0-3000 -- --

31.7 40.8 c-2

e- . -

i. -

Table C-2 1" Diameter Lower Thermal Shield Bolt Stress / Torque Test Data o Smooth portion of bolt shanks were instrumented with strain gaps.

o Neolube was applied uniformally to the threads and bearing surface, o Two bolts (Bolt Nos.1&2) were torqued into the Rancho Seco reactor internals and two bolts were tested on a mock-up in the Lab.

Bolt Stress in KSI Torque Torque Bolt Bolt Bolt Bolt Ft-lbs Sequence 1 2 3 4 Comments 0-450 350 0-350 30.8 35.8 37.9 37.6 0-450 Bolts 3&4 were relubri- '

350 0-350 -- --

36.4 37.7 cated and retorqued in same hole.

4 C-3

I l

l Table C-3 1-1/2" Diameter Upper Thermal Shield Restraint Bolt Stress / Torque Tests Data o Smooth portion of bolt shanks were instrumented with strain gages.

o Neolube was applied uniformally to threads and bearing surface of head, o Three bolts were torqued into a special fixture that simulated the reactor internals.

Torque Bolt Stress in KSI Torque Sequence Bolt I Bolt 2 Bolt 3 0-1100 700 0-1200 16.1 15.5 16.4 0-1100 800 0-1200 18.7 17.1 19.4 -

1100 0-1100 24.2 25.3 25.4 c-4

Appendix 0 NRC Question on Installation Torque and Prestress e

4

~

.o During the NRC meeting the staff requested that copies of slides on prestress and installation torque be provided for the UCB and LCB bolts at the affected 177 FA plants. These slides were not included in the meeting handouts. Table 0-1 and 0-2 provide the requested information. Table D-2 was developed to evaluate the peak normal operating stress in the region under the head. The peak stress is determined from total normal operating and prestress loads and is obtained by multiplying the total stress by the exhibited proper stress concentration factor to account for the diameter change at the head to shank transition. In addition, the question was raised why the three Oconee plants used different installation torques and, therefore, have different prestress values for the UCB and LCB bolts. Examination of records that. indicate after the preparation of specification and procedure for Oconee 1 bolt installation, increased attention was placed on the design margins for joint integrity under faulted condition loads. The increased torque was apparently specified for Oconee 2 and 3 to increase those margins to ensure the joint was maintained closed during faulted condition loadings.

The NRC also asked what steps will be taken for replacement bolts to insure that the torque actually applied in the field provides the prestress set by Engineering specification.

Torque vs load tests will be provided on several production line bolts to

correlate prestress with applied torque. UT calibration will then be made to l provide a method of field determination of actual prestress.

Installation in,the field will be by calibrated hydraulic torque wrench with the installation torcue values recorded. This data will then be verified by

! UT examination after torquing. Incorrectly torqued bolts will be detensioned I

and retorqued to the proper value then rechecked by UT.

l Information provided in Tables D-1 and D-2 is based on specific documentation l of field procedures or installation records where available. B&W and the i listed utilities are continuing to review records to ensure that the information is accurate. Until that work is complete the data reported in these two tables is subject to change. However, the changes are expected to l be within the ranges shown and should not effect conclusions drawn from this data.

D-1

, .q.

Table D-1 ~

~

PRELIMINARY STRESS LEVEL COMPARIS0N NOMINAL NOMINAL PRELOAD OPERATING TOTAL PEAK STRESS STRESS STRESS STRESS BOLTING RING PLANT (KSI) (KSI) (KSI)

(KSI)

UPPER CORE BARREL OCONEE 1 10 5 15 34.5 4

T R PLANTS 3 .5 5 4 j LOWER CORE BARREL OCONEE 1 10 8 i

18 41.5 -

OCONEE 2 27 62 i RANCHO SECO 19 8 27 62 ANO-1 - -

~

OC'0 NEE 3 28 ' 8 36 83

CR-3 i DB-1

MIDLAND 1 & 2 l UPPER THERMAL SHIELD OCONEE 1 17 3 20 40 CR-3 20 3 23 46 '

OTHER PLANTS 31.5 3 34.5 69 FLOW DISTRIBUTOR ALL PLANTS 33 2 35 70 i

LOWER THERMAL SHIELD ,ALL PLANTS 33 32 65 137 (IN!TIAL INSTALLAT10N) ,

f i

' TUBE SURVEILLANCE HOLDER CR-3 DB-1

]/ 45 3 48 101

{ YlELD STRESS OF A-286 VARIES FROM 100 TO 134 KSI. PRELOAD STRESS VALUES BASED ON NE0 LUBE.

D-2

. . +' E Table 0-2

. Installation Torque for Upper and Lower Core Barrel Bolts Torque-Ft-lbs Plant UCB Bolts LCS Bolts Oconee 1 950 950 Oconee 2 2000-3000* 1750 Oconee 3 3000 2400 ANO-1 3000 1750 Crystal River 3 3000 2400 Rancho Seco 3000 1750 Davis Besse 1 3000 2400 Midland 1 & 2 3000 2400

Documentation as to correct value presently under investigation.

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ML20071N926

Contents

Text

SUMMARY

SUMMARY

= >= C.D - >= >= O be >= E - J D == M w M O OM K w M 'CQ WM w S w >= c M C Z M G w o M = M Q w >= >

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2.2 Assistant

3.2 Cables

3.4 Couplant

4.1 Material

4.3 Reflectors

5.1 Range

5.2 Sensitivity

SUBJECT:

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