ML20136D427

Forwards Rev 3 to SER for Core Stratification Sample Acquisition in Response to NRC Comments. Increasing Depth of Core Probe Activity Down to & Through Flow Distribution Plate Proposed

ML20136D427
Person / Time
Site:	Three Mile Island
Issue date:	12/31/1985
From:	Standerfer F GENERAL PUBLIC UTILITIES CORP.
To:	Travers W Office of Nuclear Reactor Regulation
Shared Package
ML20136D432	List: ML20136D427 ML20136E018
References
0373A, 373A, 4410-85-L-0248, 4410-85-L-248, NUDOCS 8601060166
Download: ML20136D427 (17)

Text

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GPU Nuclear Corporation Nuclear  :::ome:r8o s Middletown, Pennsylvania 17057-0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:

(717) 948-8461 4410-85-L-0248 Document ID 0373A December 31, 1985 TMI-2 Cleanup Project Directorate .

Attn: Dr. W. D. Travers L.

Director r-US Nuclear Regulatory Commission  :- 1, 9, c/o Three Mile Island Nuclear Station y @

Middletown, PA 17057 y

.M

~~

Dear Dr. Travers:

C0 Three Mlle Island Nuclear Station, Unit 2 (TMI-2) u Operating License No. DPR-73 Docket No. 50-320 Safety Evaluation Report for Core Stratification Sample Acquisition, Revision 3, Response to NRC Comments on Safety Evaluation Report for Core Stratification Sample Acquisition, Revision 1 Attached for your review and approval is Revision 3 to the Safety Evaluation (SER) for Core Stratification Sample Acquisition activities. This revision includes responses to portions of the NRC comments on Revision 1 of the SER and proposes increasing the depth of the core bore activity down to and through the flow distribution plate. A discussion of reactor vessel integrity and revised man-rem estimates has been included.

Also attached are CPU Nuclear's responses to the NRC comments on Revision 1 of the SER. NRC comments were previously provided by NRC Letter NRC/TMI-85-095, W. D. Travers to F. R. Standerfer, dated November 22, 1985.

1 A

Sincer7 / C

,. R. Standerfer i\g Vice President / Director, TMI-2 i FRS/RBS/eml Attachments I 960106016603})$$:

PDR ADUCK ppg 42 0 GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

ATTACHENT (4410-85-L-0248)

Response to NRC Comments on Core Stratification Sample Acquisition SER, Revision 1

1. COFNENT:

Describe your control program that assures that a length of drill string long enough to reach the lower vessel head is not available. This should include the case where a drill hole is abandoned due to a broken bit.

RESPONSE

The length of drill string inside containment at any time is administratively limited to no more than required for a single core bore of a depth not exceeding the maximum depth limit (i.e., the fully extended drill string will be no less than 6 inches from the reactor vessel inner wall). Additional sections of drill pipe will be allowed into the reactor building only after previously used sections have been disposed of properly. This disposal may include placement in a defueling canister or direct removal from the reactor building. In addition, the drill rod and casing pipe storage racks have been designed to limit the amount of pipe that may be stored at the drilling platform, so that extra pipe is not available.

In the event that a drill bit breaks, the hole will not be " abandoned" until all sections of the drill string, including broken pieces, are removed from the hole. Consequently, a broken drill string will not impact the drill string length control program. If for some unforeseen reason, a portion of the drill string cannot be removed, the next core bore will not begin until the drill string storage rack at the drilling platform has been filled and all other lengths of drill string have been disposed of and/or accounted for.

2. COMENT:

What load restrictions (i.e., torsional, horizontal and vertical force limits) will be placed on the core bore equipment to ensure that incore instrument nozzles will not be degraded. What is their bases?

RESPONSE

No load restrictions will be placed on the core bore equipment specifically for protection of incore instrument nozzles. The basis for this is provided below.

The core bore operation will exert a downward force on the core region debris bed and on the core support assembly. This downward force is automatically controlled and will not exceed 10,000 pounds based on the operational limitations of the core bore equipment. This downward force cannot be imparted on the lower reactor vessel head incore instrument nozzles unless there is a direct, solid link between the drill bit and I

ATTACHW NT (4410-85-L-0248) nozzle. Since none of the drill locations will be directly over an incore nozzle, this link can only be created by debris. To ensure that l incore nozzle loading will be precluded, the depth of core bore will be limited such that the bit will not pass into the rubble bed in the lower vessel head regi on. The determination of maximum drill depth will be based on TV camera viewing of the lower head region immediately below the i

drill path prior to the start of drilling at a particular location.

Consequently, the drill bit downward force can only be exerted in the rubble above the flow distribution plate where the force will be distributed to the core support assembly and is unlikely to impart a 1 ,t on the nozzles.

If the drill string / bit were capable of " catching" an incore instrument string and wrapping the string around the drill bit as it rotates, a stress could be imparted to an instrument nozzle or to the instrument tube below the vessel lower head. This type of event is not considered credible for the following reasons:

a. The drill bit / string configuration is such that there is no feature which could grab and hold an instrument string.

b. Each core bore will be centered over a fuel assembly which has no instrument string. If an adjacent instrument fuel assembly collasped into the path of a core bore, the bit would drill through the assembly and sever the string. The only other drill bit contact with an instrument string would have to be with a " loose" string from an adjacent fuel assembly location. The instrument strings in an intact core are contained within an instrument tube in the center of a fuel assembly. It is not considered feasible that the surrounding fuel assembly and instrument tube could disintegrate or melt, thus exposing the instrument string, without the destruction of the instrument string.

3. COWENT:

Please provide details of drillin0 platform and actions including:

4

a. Load and load distribution of platform

b. Location of supports and contact points

c. Dynamic effect of drilling action

d. Total we1 0 ht of drill r10

RESPONSE

t The Drill Indexing Platform Structure Assembly is comprised of three major subassemblies identified as the Wing As',embly (consisting of 4"x4"xl/4" structural tubing), the Upper Lev 9l Assembly (consisting of 1

ATTACHMENT (4410-85-L-0248) 6"x6"xl/4" structural tubing), and the Lower Level Assembly (consisting of 4"x4"xl/4" and 6"x6"xl/4" structural tubing). Drawing 419932 in Volume II of EGG-TMI-6824, "TMI-2 Core Stratification Sample Project System Design Description Drawings" shows these assemblies. The structure assembly was evaluated for structural integrity using four different load cases with each load case assuming three different indexing positions for a total of twelve load combinations. Table 1 shows the loads in common to the twelve load combinations. This total load is 22, 400 pounds. Table 2 shows the additional loads caused by drilling actions as described in the table. Table 3 gives a summary of the twelve load combinations. Figure 1 graphically presents the locations of the applied loads. The maximum primary stresses to the main structural members are shown in Table 4 and are shown to be much less than the AISC allowables.

TABLE 1 WEIGilT LOADS FOR ALL LOAD COMBINATIONS Location of l Label Description Application

• Magnitude i

- Drill Indexing Platfom Structure (Distributed) 6" tube:40 lb/ft (3400 lb total) 4" tube:15 lb/ft W1 El. 337.833 1650 lb

Drill (1250) Indexing plus 2 operators Platfom (200 Left each Wing)

W2 Drill Indexing Platfom Right Wing El. 337.833 -1250 lb (1250)

W3 Drilling equipment center of gravity El. 337.833 11.050 lb Drill Unit (3535) 24 inches Tilt platfom (2200) from 'P' underwaterstructure(2650)

Drill roller platfom (760) .

Casing (300)

Drill tube, core barrel, bit (550)

Middle and top clamp (350 & 275)

Bottom clamp (130)

Hydraulic control panels (300)

W4 Cask (4700) and cask roller platfom El. 333.886 5050 lb (350) below 'P'

• NOTE: Weight W3 and W4 have three alternate locations relative to point 'P' as specified by dimension "X." (Sco Tablo 3 and Figure 1) .

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TABLE 2 ADDITIONAL LOADS FROM CORE DRILLING (REACTION LOADS ON BIT)

Location of Case Label Description Application

• Magnitude

1. Initiation of drilling #1 El. 305 F1 Anticipated force on bit, x-dir.& vert, below 'P' 1118 lb M1 Anticipated moment on bit 100 ft-lb

2. Initiation of drilling #2 F2 Anticipated force on bit, z-dir.& vert. 1118 lb M1 Anticipated moment on bit 100 ft-lb

3. Bit jams while pushing F3 Maximum force on bit 17.500 lb M2 Maximum moment on bit 5200 ft-lb

4. Bit jams while pulling F4 Maximum force on bit 10.000 lb M2 Maximum moment on bit U 5200 ft-lb (NOTE: All reactions have three alternate locations below point 'P' as specified by dimension "X." (See Table 3 and Figure 1)

TABLE 3

SUMMARY

OF LOAD COMBINATIONS Indexing Position of Point 'P' Load case (all cases (Location specified by dimension "x")

include weight loads) x = 1.01 f t x = 2.02 ft x = 4.05 ft

1. lA 18 1C

2. 2A 2B 2C

3. 3A 3B 3C

4. 4A 48 4C

FIGURE 1 i

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TABLE 4 MAXIMUM PRIMARY STRESSES AND THE ALLOWABLE STRESSES FOR THE MAIN STRUCTURAL MEMBERS 6"x6"xl/4" TUBING 4"x4"xl/4" TUBING MAXIMUM ALLOWABLE MAXIMUM ALLOWABLE SHEAR (psi) 1548 14,400^ 277 14,400" 23,760 c b d BENDING (psi) 5122 (B) 23,760 2409 (4) 656 (y) 23,760 c 1967d (x) 23,760 c COMPRESSION (psi) 130 21,255* 2490 19,723" f f STRESS LEVEL 0.249 4l 0.310 dL I NOTES:

a - 0.4 x Fy,Fy = 36,000 psi for A36 Steel, AISC, " Manual of Steel Construction", Eighth Edition 1980, Appendix, Table 2, Pg. 5-73 b - Bending Moment About B-Axis and Y-Axis c - 0.66 x Fy d - Bending Moment About B-Axis and X-Axis e - AISC, " Manual of Steel Construction", Eighth Edition, 1980, Table C1.8.1; Eq. (2.a.1-1) & Eq. (1.5-1) f - Ibid, eg. (1.6-2) l l

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ATTACHMENT (4410-85-L-0248)

4. COMMEtT:

Please provide analyses and calculations to substantiate that:

a. Existing structure can take the platform load and operating forces;

b. Reactor vessel can take the operating forces; and

c. Accidents in operation will not damage the system.

RESPONSE

a. The Defueling Work Platform (DWP) is positioned over the Reactor Vessel and is supported from the refueling canal floor by the Shielded Support Structure (SSS). The DNP consists of a circular beam approximately 17 feet in diameter with interior cross beams as shown in Figure 2. The platform circular and cross beams are welded girders made of 304 stainless steel. The platform is supported by 22 wheel assemblies that ride on a rail mounted on the SSS shown in Figure 3. The SSS consists of frame work made of ASTM A36 carbon steel resting on four columns as shown in Figure 4.

Structural analyses have been performed to evaluate the structural integrity of the DWP and the SSS. The analyses found that the load case shown in Figure 5 with the shield collar at Location 'A' (Denoted as " Case 2") to be the most limiting for the defueling equipment cases. All cases considered the live load of 100 p.s.f.

(total of 24,300 lbs.) and the platform shielding weight (total of 81,101 lbs.). Two cases were considered for the core drilling equipment. The core drilling cases considered the totaA live load and the total platform shielding weight along with the loads denoted as "A", "D", and "E" shown in Figure 5. Each core drilling case also includes the loads shown in Table 1 of the response to Comment 3 applied at the Reactor Vessel centerline. The two core drilling cases differ by the loads imposed on the platform during core drilling operations. Core drilling Case 1 assumes a 14,000 ft-lb moment across the working slot at the Reactor Vessel centerline to evaluate the reaction loads on the bit described as Cases 1 and 2 in Table 2 of the response to Comment 3. Core drilling Case 2 assumes a 10,000 pounds downward force applied at the Reactor Vessel centerline to evaluate the reaction load on the bit described as Case 4 in Table 2 of the response to Comment 3.

Table 5 shows the results of the three loading cases described above. All results are within the design limits for 304 stainless steel as taken from the ASPE Boiler and Pressure Vessel Code Section III, Division 1 - Appendix XVII. The nodes identified in Table 5 are depicted in Figure 6. Table 6 shows the structural design criteria used for the DWP and the SSS. The allowable stress for all the components given in Table 5 (with the exception of the filter canister support beam) is 20,000 psi. The allowable stress of the filter canister support beam is 18,000 psi.

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A. Canister Positioner (24000 lb) + Vertical Shielding (9006 lb)

B. Tool Positioner (7500 lb)

C. Tool Positioner Reaction (5000 ft ib)

D. Filter Canisters (7000 lb)

E. Vertical Shielding (10066 lb total)

F. Tool Rock (2500 lb)

G. Jib Crane Weight (750 lb)

H. Jib Crane Moment (8000 ft Ib)

1. Shield Collar (10000 lb) - (Can be at Location A or D)

FIGURE 5 WORK PLATFORM LOADING CONDITIONS UllDER NORMAL OPERATION (LOAD CASES 1 and 2)

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ATTACHMENT (4410-85-L-0248)

TABLE 5 MAXIMUM STRESS RESULTS FOR THE WORK PLATFORM Max. Stress (oby)

Defueling Core Drilling Component Nodes Equipment (Case 2) Case 1 Case 2 A. Main Cross 92-93 14,886 14,593 14,220 Beam #1 91-92 14,201 14,989 14,520 (Canister 90-91 13,353 15,456 14,869 Positioner) 89-90 12,435 15,851 15,147 88-89 11,014 15,857 15,153 B. Main Cross 106-107 8,665 6,579 6,669 Beam #2 107-108 8,138 6,740 6,801 (Tool 108-109 9,090 6,866 6,895 Positioner 109-110 9,063 6,916 6,914 110-111 8,960 6,916 6,914 C. Tool Positioner 131-132 1,081 605 609 132-133 1,139 716 722 133-134 1,123 754 762 D. Filter Canister 726-727 1,127 1,139 1,154 Support Beam 727-728 1,122 1,139 1,133 728-88 920 939 934 E. Jib Crane 820-821 10,936 12,242 11,725 Supt. Beams 718-83 9,935 11,040 10,571 717-718 9,746 10,821 10,361 822-823 11,937 12,223 11,987 F. Circular Beam 2-3 2,146 2,076 2,042 17-810 2,608 2,582 2,521

TABLE 6 STRUCTURAL DESIGN CRITERIA

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WORK PLATFORM SUPPORT STRUCTURE 304 SST A36 STEEL TENSION .Fg = 0.6 Fy Ft = 0.6 Fy SHEAR Fy = 0.4 Fy Fy = 0.4 Fy BENDING Fb = 0.66 Fy Fb = 0.66 Fy (compact sect.)

COMPRESSION Fa = 0.6 Fy Fa = 0.6 F (consider buckling) (consfderbuckling)

BEARING Fp = 1. 5 Ful t 7p = 1. 5 Fult (forbolts) (forbolts)

BOLTS ASTM A325 ASTM A325

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-tension F bt =Fult/2 Fbt = 44 ksi

-shear .Fby = 0.62 F ult /3 Fby = 21 ksi 304 SST F = 30 ksi F(It= 0 ksi A36 STL Fy = 36 ksi Fult = 58 ksi A325 BOLTS y = 81 ksi

.F F

ult = 105 ksi

ATTACHMENT (4410-85-L-0248)

b. Since the core bore equipment is supported by the DWP, which is in turn supported from the floor of the fuel transfer canal, the static equipment loads associated with the core bore operation are not imparted to the reactor vessel. The only significant operating force imparted to the vessel is the downward force exerted by the drill bit face. The core bore equipment is designed such that even with a single control system failure, the maximum force to the vessel will not exceed 10,000 pounds. Since the vessel has with stood pressures of up to 2300 psig following the TMI-2 accident, a force of 10,000 pounds transmitted to the vessel will not cause failure of the vessel. Damage to the reactor incore instrument nozzles is unlikely as discussed in the response to NRC comment Number 2.

c. Accidents which could cause damage to the reactor coolant system resulting from the operation of the core bore equipment are addressed in response to NRC comments 1, 2, Ab, and 5.

5. COMMENT:

Assuming failure of an incore instrument nozzle, what is the maximum leak rate? (Include comparison to makeup rate and recirculation rate.)

RESPONSE

As discussed in the response to NRC comment Number 2, a failure of an incore instrument nozzle as a consequence of the core bore activities is considered unlikely. If, however, a nozzle fails, the resulting leak rate from the vessel will be approximately 0.4 gpm as discussed in the i

SER for Heavy Load Handling Over the TMI-2 Reactor Vessel (the nozzle '

failure mechanism as a result of loads imparted by the core bore operation will be similar to the failure mechanism which results from loads imparted by a load drop accident). This leak rate is negligible compared to the makeup and recirculation capability as described in Technical Specification Change Request No. 46, Recovery Operations Plan Change Request No. 46 approved by NRC letter dated August 8, 1985.

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