ML20215B016

Full-Flow Filter Recovery & Equipment Assessment

ML20215B016
Person / Time
Site:	South Texas
Issue date:	06/01/1987
From:	Boyle D, Rex J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19292H390	List: ML20215B000 ML20215B016 NRC-87-3231, Forwards Nonproprietary WCAP-11508 & Proprietary WCAP-11506, Full-Flow Filter Recovery & Equipment Assessment. Proprietary Rept Withheld (Ref 10CFR2.790).Application for Withholding Rept & Affidavit Encl
References
WCAP-11508, NUDOCS 8706170186
Download: ML20215B016 (63)

Text

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WESTINGHOUSE CLASS '3 WCAP-11508 FULL-FLOW FILTER RECOVERY AND EQUIPMENT ASSESSMENT South Texas Project Unit 1 JUNE 1,1987 Compiled by: J. A. Rex Approved by .O D. E. Boyle, Managef Primary Component Engineering I

WESTINGHOUSE ELECTRIC CORPORATION Generation Technology Systems Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 l .

B706170186 870611 PDR ADOCK 05000498 A PDR 3

l TABLE OF CONTENTS l .

Section Title Page -

1

1.0 INTRODUCTION

1-1

2.0 BACKGROUND

H!'JTORY AND DESIGN OF THE FULL-FLOW FILTERS 2-1 3.0 IDENTIFICATION OF SYSTEMS AND AREAS FOR FILTER DEBRIS ,

INSPECTION 3-1 -

3.1 Primary Coolant System 3-1 3.2 Chemical and Volume Control System / Boron Thermal 3-2 g Regeneration System ,

3.3 Residual Heat Removal System 3-3 '

3.4 Liquid Waste Processing System 3-3 .*

4.0 GENERAL PLANT INSPECTIONS 4-1 4.1 Primary Coolant System 4-1 4.2 Chemical and Volume Control System / Boron Thermal 4-5 Regeneration System I 4.3 Residual Heat Removal System 4-6 4.4 Liquid Waste Processing System 4-6 5.0 EQUIPMENT EVALUATION DUE TO LOOSE FILTER DEBRIS 5-1 5.1 Reactor Vessel 5-1 5.2 Reactor Internals 5-1 5.2.1 Upper Internals 5-1 .

5.P.2 Lower Internals 5-1 r 5.3 Control Rod Drive Mechanisms 5-1 5.4 Steam Generators 5-1 5.5 Pressurizer 5-2 5.6 Reactor Coolant Pumps 5-2 5.7 Primary System Piping 5-2 5.8 Auxiliary Systems 5-2 2433s @ 60187.10 j

TABLE OF CONTENTS (Continued)

Section Title Page 6.0 DETAILED ASSESSMENT OF FILTER SCREEN MATERIAL 6-1 6.1 Inventory and Happing Record of Intact and Recovered 6-1 Filter Material -

6.1.1 Scope of Filter Degradation 6-1 6.1.2 Calculation of Unrecovered Filter Debris 6-3 6.2 Metallurgical Examination and Fabrication Investigation 6-7 of Filter Material 7.0 PLANT OPERATION WITH UNRECOVERED FILTER DEBRIS 7-1 7.1 Effect of Filter Debris on Nuclear Fuel 7-1 7.2 Effect of Filter Debris on Reactor Vessel and 7-3 Reactor Internals I 7.2.1 Reactor Vessel 7-3 7.2.2 Reactor Upper Internals 7-3 7.2.3 Reactor Lower Internals 7-5 7.3 Effect of Filter Debris on Control Rod Drive Mechanisms 7-5 7.4 Effect of Filter Debris on Steam Generators 7-7 7.5 Effect of Filter Debris on Pressurizer 7-8 7.6 Effect of Filter Debris on Reactor Coolant Pumps 7-8 7.7 Effect of Filter Debris on Primary System Piping 7-10 7.8 Effect of Filter Debris on Bottom Mounted 7-10 Instrumentation 7.9 Effect of Filter Debris on Auxiliary Systems 7-11 7.10 Effect of Filter Debris on RCS Materials 7-15 -

7.11 Conclusions 7-15 8.0 SAFETY EVALUATION 8-1  !

Appendix A SAFETY EVALUATION CONCERNING FILTCR DEBRIS FROM THE A-1 FULL-FLOW FILTERS 1

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1 1.0 INTpDUCTION i l

After completion of hot functional testing at South Texas Project Unit 1, which involved 28 days under full flow at hot conditions, inspection of the full-flow filters installed on the lower core support plate revealed that  !

degradation of 57 of the 192 filters had occurred. These filters are used during hot functional (as well as cold hydro) testing to help remove debris j

_from tr;e primary system. J 9

As a result, this filter debris appears to '

~ have been circulated througbout the Primary Coolant System and parts of the l auxiliary systems during the hot functional testing.

This report dis usses the background history and the design of the Full Flow Filters used at South 'iexan Unit 1, the results of inspections performed on the equipment contained in the primary loop and certain auxiliary systems, results of metallurgical examinations performed on the filters and evaluations of the effects of the unrecovered filter debris on the equipment contained in the primary system and certain auxiliary systems.

D Additionally, contained in Appendix A of this report is a copy of the Safety Evaluation concerning operation of the plant with unrecovered filter debris.

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2.0 BACKGROUND

HIS70RY AND DESIGN OF THE FULL FLOW FILTERS Westinghouse has previously used full flow filters on plants which utilize 12 foot cores. This filter is designed to attach to a lower core plate which is approximately 3 inches thick and is not suited to the South Texas Unit 1 q plant, an XL plant, which has a lower core plate approximately 17.5 inches

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Figure 1 shows the overall design of the full flow filter as installed on the lower core support plate. A total of 192 filter blocks are used at South Texas Unit 1 during cold hydro and hot functional 3

testing, after which they are removed prior to the first core load. These 1 filters are attached by filter blocks bolted to the lower core support piate -

by bolts that clamp onto the top and bottom of the sursport plate. The semicircular arc at the top indicates the active filter screen element. The left-hand figure depicts the 134 filter blocks which are used at the fuel -

element locations that do not have bottom mounted instrumentation (BMI), while the right-hand figure depicts filter blocks used at the remaining 58 locations having BMI, for South Texas Unit 1. The filter screens are the same for both types of filter block and the main difference is the attachment to the underside of the plate.

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Vy 3.0 IDENTIFICATION OF SYSTEMS AND AREAS FOR FILTER DEBRIS INSPECTION ,

In addition to equipment contained in the Primary Coolant System, an l evaluation was performed on five auxiliary systems, Chemical Volume and

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Control / Boron Thermal Regeneration Systems, Baron Recycle System, Residual Heat Removal Systein, Liquid Waste Processing System, and Sampling System, to determine inspections to be performed for Filter debris. Of these five ,

auxiliary systems, three were deemed necessary for inspection, Chemical Volume and Control / Boron Thermal Regeneration Systems, Residual Heat Removal System, - O and Liquid Waste Processing System.

I Areas that were inspected and cleaned in the Primary Coolant System and the , ,,

three auxiliary systems are identified in the following paragraphs.

3.1 Primary Coolant System Upper Internals Packaga Core exit thermocouples - Inspect for straightness and electrical continuity.

Control Rod Guide Tubes - Hydrolaze and inspect with boroscope (as ,

many times as necessary)

Lower Internals (core barrel)- Hydrolaze and visually inspect the e following areas Core barrel ledge P

Core barrel spray nozzles Specimen baskets' Baffle /former area Incore instrumentation columns Behind neutron pads Reactor Ve.ssel Closure Head Control R>d Drive Mechanisms - Hydrolaze and visually inspect (using boroscope)

Remaining head penetrations - Hydrolaze and visually inspect (using boroscope) 2433s/052987 10 3g

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i Reactor Vessel Bottom (Lower Plenum) - Visually inspect Thermowells - Visually inspect gap between thermowell and boss  ;

Hot Leg - Visually inspect l Cold Leg - Visually inspect Cross Over Leg - Visually inspect Steam Generator j Hot leg channel head - Visually inspect Cold leg channel head - Visually inspect Divider piste drain holes - Visually inspect Tubes - Visually inspect near bottom of tubes, using plug blow  :

through method, remove any material lodged in tubes, 4

visually inspect using boroscope tubes that contained filter debris.

Pressurizer Spray nozzles - Boroscope inspect and hydrolaze Outlet screens - Visually inspect Loop Seals - Visually inspect (during PORV change-out) 3.2 Chemical Volume Control System / Boron Thermal Regeneration System Regen Heat Exchanger - Flush shell side of heat exchanger Letdown Orifices and Trim Valves - Visually inspect Letdown Filter (Pre-filter) - Visually inspect Reactor Coolant Filter - Visually inspect Seal Water Return Filter - Visually inspect Inspections were deemed unnecessary for the following equipment.

Mixed Beds and Cation Beds - downstream of letdown filters BTRS Demineralizers - downstream of letdown filters  !

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VCT Spray Nozzles - downstream of letdown and RCS filters Excess Letdown Heat Exchanger - discharge goes to seal water return filter Charging Pump Svetion Strainers - downstream of RCS and seal water return filters l

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j PD Pump Suction Stabilizers - downstream of RCS and seal water return filters l .

BTR Heat Exchangers - downstream of letdown filters Recycle Evaporator Feed Filter and Domineralizer - downstream of RCS, letdown and seal water r; turn filters Recycle Holdup Tank - downstream of RCS, letdown and seal water return filters 3.3 Residual Heat Removal System Heat Exchangers - Visually inspect, blow plugs through tubes Suction Isolation Valves - Visually inspect lnspection of the RHR Pump Suction Strainer was deemed unnecessary since the strainers were not installed during hot functional testing and any filter debris would have collected in the channel head of the RHR Heat Exchanger.

3.4 Liquid Waste Processing System Reactor Coolant Drain Tank - Visually inspect I

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4.0 GENERAL PLANT INSPECTIONS This section provides the results of the inspections described in Section 3 of this report. .

4.1 Primary Coolant System Upper Internals Package

- Core exit thermocouples - Inspected for straightness and no damage wan found. Electrical continuity checks were performed. Three thermoccuples were replaced due to reasons other than filter debris.

- Ccitrol Rod Guide Tubes - Hydrolazed several times until inspections indicated no presence of filter debris. No physical damage was found. Essentially all large pieces of filter debris were caught in 1 the upper core plate holes at the bottom of the support columns or in the lower flange of the guide tubes. Figures 2 and 3 show photographs of these regions.

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- Core barrel ledge - No filter debris or physicsl damage found.

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- Core barril spray nozzles - No filter debris or physical damage found.

- Spec.imen baskets - No filter debris or physical damage found.  ?

- Baffle /former area - No filter debris or physical damage found.-

- Incore instrumentation columns - No filter debris or physical damage found.

- Behind neutron pads - No filter debris e physical damage found.

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FIGURE 2 CAPTURED SCREEN MATERIAL ON SUPPORT COLUMN BASES 2381s/052987 10 4-2 l

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Reactor Vessel Closure Head

- Control Rod Drive Mechanisms - No filter debris or physical damage found, i

- Remaining head penetrations - No filter debris or physical damage found.

Reactor Vessel Bottom Head (Lower Plenum) - A number of small pieces of filter debris were found.  !

No physical damage was found.

Thermoweils - Filter debris was found (one wire piece). No physical damage due to filter debris found. 4 Hot Leg - No filter debris or physical damage found.

i Cold Leg - No filter debris or physical damage found.

Cross Over Leg - One small piece of filter debris found. No physical damage found.

Steam Generator Hot leg channel head - No filter debris or physical damage found.

Cold leg channel head - No filtse debris or physical damage found.

Divider plate drain holes - Some filter debris found. No physical damage found.

Tubes - Prior to the plug blow-through inspection performed in-May, inspections of the- steam generators revealed pieces of filter debris lodged in the drain hole and tube entrances. Steam generator "A" had the least volume, all found in the drain hole.

All filter debris found in steam generator."B" was at the tube entrances. Steam generator "C" had the largest volume of filter debris, one piece lodged in the drain hole and all other pieces lodged in the tube entrances. Steam generator "0" had the second largest volume of filter debris all of which were lodged in the i tube entrances. One piece resembled a small ball made of wires.

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The plug blow through inspection revealed filter debris in four tubes in steam generator "A", seven tubes in steam generator "B". .

sixteen tubes in steam generator "C", and eight. tubes in steam generator "D". Using a fiber optics video camera, 'the interior of the tubes where filter debris was found were inspected for damage. The inspection revealed some superficial buffing marks on the interior surface of the tube from the hot leg side of the tube entrance to a maximum height of 21 inches. From visual inspection of these tubes it was seen that the filter debris was lodged in the tube at the transition point between the expanded and un-expanded portions of the tube. The superficial marks were determined to have such insignificant depth that they could not be measured by eddy current testing.

The steam generator bowl, nozzle cover ring bolt holes, the manway drain holes, and the steam generator bowl drain holes were inspected visually or by fiber optic camera. Minor. particles'of filter debris were recovered during this inspection. No damage to these areas was found. Moreover, the hot, cold, and crossover '

legs were inspected and no filter debris was found.

The total material recovered from the steam generators was 55.78 grams (1.97 oz.). <

PressurHer Spray Nozzles - No filter debris or physical damage found.

Outlet Screen - Fifteen grams of filter debris found. No physical-damage found.

Loop Seals - Will be performed during PORV changeout.

4.2 Chemical Volume Control System /Boren Thermal Regeneration System Regen Heat Exchanger - No filter debris or physical damage found.

l Letdown Orifices and Trim Valves - No filter debris or physical damage found.

Letdown. Filter (Pre-filter) - Filters were changed 15 times during HFT. No filter debris found on or in l the last set of filters.

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Reactor Coolant Filter - Filters were changed numerous times during HFT with no inspection. No filter debris found on or in the last set of filters.

Seal Water Return Filter - No filter debris found.

i 4.3 Residual Heat Removal System Heat Exchangers - Small amount of filter debris (total of 3.2 grams) found in the channel heads of the heat exchangers.

No physical damage found.  ;

Suction Isolation Valves - Operated RHR at full flow back in'to RCS.

Inspection of loops and reactor vessel revealed no filter debris.

4.4 Liquid Waste Processing System a

Reactor Coolant Drain Tank - The tank suction strainer was inspected and no filter debris was found. Tank was not inspected since no other direct path exists to the tank.

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l 5.0 EVALUATION OF INSPECTION RESULTS The following section provides the engineering evaluations of the inspection results shown in Section 4 of this report.

l 5.1 Reactor Vessel - No damage was found during the inspection of the Reactor

, Vessel; therefore, the Reactor Vessel has not been affected by the filter debris.

5.2 Reactor Internals 5.2.1 Upper Internals - No physical damage was reported on the components of the upper internals; therefore it can be concluded that the margins of safety of the ui.,

internals have not been affected by the filter debris.

5.2.2 Lower Internals - No physical drmage was reported on the components of the lower internals; therefore it can be concluded that the margins of safety of the upper internals have not been affected by the filter debris.

5.3 Control Rod Drive Mechanisms - No physical damage was reported on the control rod drive mechanisms; therefore, they have not been affected by the filter debris.

5.4 Steam Generators - No physical damage was reported to the steam generator pressure boundary components on the primary side, which includes the channel head divider plate, tube to tube sheet welds, and tube sheet cladding. The superficial marks on the interior surface of the tubes are consider-ed insignificant. Therefore, the steam generators have-not been affected by the filter debris.

2433s/052987 10 5-1

5.5 Pressurizer - No physical damage was reported to the pressurizer; therefore, it has not been affected by the filter debris.

5.6 Reactor Coolant Pumps - During normal operation, the wire debris would be pumped through the pump without consequence. Only under Loss of Injection (LOI) conditions could filter debris migrate into tne pump internals.

Loss of injection did not. occur during HFT and s therefore the possibilicy of debris in the seal area is very remote. Some questions had been raised as to whether seal injection was on during filling and venting of the RCS. To ensure the cleanliness of the pumps, Pump C seals were disassembled and the shaft annulus was inspected.

No filter debris was found. Based on the inspections performed and the known operating performance, the Reactor Coolant Pumps have not been affected by the filter debris.

5.7 Primary System Piping - No physical damage due to the filter debris has been found to the primary system piping including thermowells; therefore, it has not been affected by the filter debris.

5.8 Auxiliary Systems - No physical damage has been reported on the Auxiliary systems; therefore, they have not been affected by the filter debris, 2433s/052987 10

6.0 DETAILED ASSESSMENT OF FILTER SCREEN MATERIAL 6.1 Inventory and Mapping Record of Intact and Recovered Filter Debris 6.1.1 Scope of Filter Degradation Inspection discovered 57 of the 192 filter screens degraded and/or filter screen missing. Of these, 24 had experienced degradation of various degrees as listed below:

4 filter screens completely missing i filter screen with 75% of material missing i filter screen with 50% of material miss:ng 18 filter screens with 5% to 25% of materias missing 24 total degraded filter screens

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The other 33 filters had only a piece or single strand of wire missing but (

were otherwise intact. Figure 4 maps the locations where the 24 filter screen degradation occurred. The percentages indicate the approximate percentage of material missing. A few, indicated as 0.0%, experienced large tears but had a very small amount of material missing.

Filter Screen Inventory Total screens installed 192 Initial estimate of number of 7.07 missing ec;uivalent screens l

l Number of equivalent screens 5.46 recovered to date Number of equivalent screens 1.61 unrecovered 2423s/CM987:10 g.{

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FIGURE 4 PERCENTAGE MISSING OF DEGRADED FILTERS i

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5.1.! Calculation of Unrecovered Filter Debris All of the 135 intact screens were removed and weighed. The 33 slightly affected screens were also weighed with missing wire replaced with matching pieces from collected filter debris. The total weight from the 168 screens generated an average mass per screen of 120.76 grams. Using this average, the following steps are used for the unrecovered material.

Oz. Grams a) Total mass of 24 complete screens 102.22 2898.24 b) Hass of material remaining in 24 71.98 2044.9 degraded screens c) Total mass of missing screen to 30.10 853.34 be recovered d) Mass of filter debris collected 23.26 659.4 e) Mass of unrecovered filter debris 6.84 193.94

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f) No. of missing screens (193.94/120.76): 1.61 '

g) Percentage of screen material retrieved . 77.3 l

The amount of fi' ter screen believed to be missing is approximately 194 grams.

l Characterization of Retrieved Filter Debris Figures 5 and 6 show an overall view and close-ups of damaged filter screens.

The largest sections of filter debris were found in the upper core plate holes below the support columns and guide tubes. Figure 7(lowerpictures)shows q l

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FIGURE 5 FLOW SCREEN DEGRADATION AFTER HOT FUNCTIONAL TESTING mi, an ' '

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filter screen segments from plate locations K-8 and J-4. Filter debris found in the steam generator drain holes and tube entrances, shown in the upper right photo were retrieved from the C steam generator. Most of the filter debris was discovered in the entrance of the tubes and are approximately the size of the tube inside diameter. Some filter debris resembling a ball of wires was lodged .25 inches up in a tube in the D steam generator. The upper left photograph shows filter debris retrieved from the lower internals after hydrolazing. The filter debris found in the RHR heat exchanger channel heads consisted of small segments approximately the size of the screen mesh square length (.125 inches). Filter debris found in any location downstream of the steam generators was in the form of individual wire strands less than 1.0 to 1.5 inches in length. Filter debris retrieved from the upper internals guide tubes after hydrolazing have varied from wire .25 inches long to filter debris measuring 3 x 1.75 inches in size.

6.2 Metallurgical Examination and Fabrication Investigation of Screens Metallurgical investigation was conducted to establish the mechanism and the cause of the degradation of the flow screens. The evaluations were centered on two degraded and two unaffected screens from the hot functional test and one shop screen (engineering prototype) which was not used in the hot functional testing. The evaluations included surface examination of tne screens in the as received condition by visual means, by low power light optical microscopy, and by scanning electron microscopy (SEM) techniques.

Other examinations included in metallographic examinations on the traverse sections through the wires and wire crossings by light optical and SEM techniques, fractographic examinations of the degraded screen wires by light optical and scanning electron fractography techniques and chemistry evaluation of the wire and braze materials by microprobe analysis technique. SEM examinations revealed that on all of the screens, including the new screen, most of the brazed joints at the wire crossings were cracked. The brittle Ni alloy braze was cracked by various degrees, from approximately 50 percent to 100 percent through.

1 On the degraded screen, approximately 30 percent to 50 percent of the w1re  !

I crossections were also worn away at many of the joints where the braze j material had completely degraded. It is assumed that the primary purpose of un. - no 6-7 I

l the braze was to keep the fretting wear from occurring as well as providing-additional strength to the screens. Fractographic examinations using scanning electron microscopy showed the fracture mechanism to be attributable to high -

cycle fatigue. Examination of the broken wire ends revealed crack initiations I began at the worn surfaces of the reduced wire crossectional areas at the crossing joints and propagated through due to high cycle fatigue. Numerous fatigue striations could be clearly seen across the fracture faces. Wire o fractures examined from several locations on the degraded screen consistently supported these observations. Metallographic examinations revealed that the surface of the wires were in a sensitized state where the Ni braze was applied. This was considered to be a possible fracture initiation contributor through stress corrosion cracking, but close examinations found that the sensitization depth was too shallow to affect the locations where the fracture initiations began. Therefore the wire sensitization was eliminated as a contributing cause of the degradation. It was concluded that the fracture initiation was primarily caused by the stress concentrations produced by the narrowed wire crossections.

SEM examination of the non-degraded screens discovered a wide variation of the braze application. One screen had little braze material while the second one had a much heavier braze application. This has also been not hed on one other .

new screen (not the one examined here). One half of the new screen was almost bare c# the braze material allowing the screen to easily flex, while the other half was much stiffer with a heavier braze coating. Also, examination of the y braze on the flat flange of the.two non-degraded screens revealed nc braze cracking, contrary to that .seen everywhere on the domed regions of the screens. Surface fretting wear examinations on the screen flanges revealed.no vibratory wear marks. This indicated that the non-degraded screens were l

clamped well into the filter blocks which helped to prevent any additional ,

flow induced vibration of the screens. This appears to~be an important consideration in light of the fact that one of the non-degraded screens had a poor quality braze application. Detailed surface wear examination of the' i degraded screen at the shiny spots along the flange near the dome intersection suggested material loss due to impact loading against the clamping rings presumably induced during vibration of the screens.

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The following material evaluation .'were performed omthe screen wit'e:aud braze l l

coating. Chemistry evaluation of the screen wire material..by ~micrcprobe- l 1_ analysis confirmed that the material meets Type 304 stainless!. steelx,. ..

\

L requirements. The hardne'ss of the 364l stainless steel wire end Ni aljoy braze a were measured with the following results.

3 Material, Knoop Hardness ~

l 304 SST Wire - 160 ,,  !

Ni Alloy Braze 500 From the examinations performed, the following screen. degradation sc'enario can' be postulated. Flow induced vibration of the screens caused continued ', ,

degradation of the cracked braze joints .ind. subsequent fretting <near of'tho-loose wire joints. Fracture initiated by Mress concentrations produced at the worn wire joints then propagated through by high cycle fit 1gue.- No 1

finding of rupture due to impact of debris was found from any of the observations, and did not appear to be a factor ~in the initiation of the

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screen degradations.

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3 2432s/000187.10 3 6-10

7.0 PLANT OPERATION WITH UNRECOVERED FILTER SCREEN DEBRIS For operation with unrecovered filter debris, the following engineering evaluations were performed assuming conservatively, that the entire 194 grams of unrecovered filter debris are in the primary coolant system. From the geometry of the recovered filter debris found in the systems, two bounding forms of the filter debris were used in the evaluations.

1. Filter debris in the form of wire of 3/8 inch length.

2. Filter debris in the form of a ball of wire .66 inches in diameter  !

(as retrieved from the steam generator tubes).

l The filter debris is composed of approximately .038 inch diameter stainless steel wire with the remnants of nickel allow braze. Hardness values are as described in Section 6.2 of this report.  !

7.1 Effect of Unrecovered Filter Screen Debris on Nuclear Fuel l

The unrecovered filter screen debris in individual wire form is within the ]

range of the types of debris which are known to be capable of causing fuel cladding breach. The potent'ial safety consequences of fuel cladding breach have been reviewed and are evaluated in FSAR safety analyses. Primary coolant radioactivity levels are monitored regularly according to Technical 1 Specification guidelines. It is expected that the monitoring requirements of I the Technical Specifications will preclude the possibility of exceeding design basis coolant radioactivity limits during operation. Though small amounts of debris in the primary system can lead to increased coolant radioactivity levels, it is believed that there is no unreviewed safety question which could result from fuel cladding breach due to the unrecovered filter debris.

From a thermal and hydraulics viewpoint the unrecovered filter debris is not expected to result in any operational issues from a flow blockage / departure from nucleate boiling (DNB) standpoint. Particles of filter debris reaching the core inlet which are larger than approximately 0.25 inches in diameter, such as the 0.66 inch diameter balls found in the steam generators, would be sm."'" " 7-1

trapped either by the fuel assembly bottom nozzles or bottom grids. Computer models indicate that when flow is completely blocked to one fuel assembly at the inlet, full recovery of the flow occurs about 30 inches downstream of the blockage and the effect of this blockage is not limiting with respect to DNB.

It is considered extremely unlikely that more than one fuel assembly would be completely blocked in a localized region of the core inlet. For example, based on the estimate of the amount of unrecovered filter debris, assuming 4 l inches of wire per ball, there could be no more than 430 balls which could l block no more than two fuel assembly bottom nozzles in all. Therefore, fuel assembly inlet blockage by balls of filter debris is not a safety concern.

Filter debris smaller than approximately 0.25 inches in width could pass through the fuel rod array and one or more fuel assembly grids. It is considered unlikely that enough filter debris would become lodged in one location to significantly affect the core DNB. Tests performed on open lattice fuel assemblies show tnat a blockage of up to 41% at a grid location is acceptable, with disappearance of the stagnant zone behind the flow blockage after 1.65 L/De (length / equivalent hydraulic diameter). This blockage is equivalent to a particle of 0.25 inch diameter placed in each ,

space between four fuel rods. Local flow blockages of this type have little effect on subchannel enthalpy rise and cause only minor perturbations in local mass flow velocity. In reality, these blockages promote turbulence and, thus, l

would likely not adversely offect DNB at all. j i

In conclusion, it is highly unlikely that potential core flow blockage due to filter decis will adversely affect the DNB evaluations for the South Texas core. The 2nrecovered filter debris is capable of causing fuel cladding breach. is could result in an increase in coolant radioactivity levels; {

however it is expected that normally performed monitoring will preclude the l possibility of exceeding design basis coolant radioactivity limits during I operation and it is believed that there is no unreviewed safety question which could result from fuel cladding breach due to unrecovered filter debris.

1 2433s/060167 10 7-2

7.2 Effect of Unrecovered Filter Screen Debris on Reactor Vessel and Reactor Internals )

1 7.2.1 Reactor Vessel If all of the debris is made up of individual wires no foreseeable adverse effects would occur on the reactor vessel cladding. The energy absorption baseplate has a large opening in the center where the stagnate flow area would be located in the lower plenum region. The wire debris would most likely collect there and not lodge between the baseplate and vessel. The worst case scenario is if a ball ~of wire .66 in, in diameter lodges against a lower head BMI penetration and wore from flow induced vibration. The wire is coated with j up to 5 to 6 miles of Ni braze. Certain wear data, though, has indicated that j a Ni base alloy (Nucalloy) very similar in composition and hardness to the braze has a wear coefficient no greater than 304 SS for this wear pair.

Therefore, wear rates can be assumed to be essentially equal for the wire material on the 304 SS penetration. If the ball is shaped out of individual wires, the geometry would reasonably generate, at most, a density of one half that of solid metal. From these considerations, if a .66 in, ball completely were away against a cylindrical BMI penetration (1.5 0.D. x .75 I.D.), a wear depth approximately nalf the ball diameter could be generated which would not result in wear through the wall of the penetration. The likelihood of this j ever occurring would be very remote. Wear through of the .125 to .22 in. I vessel clad is feasible if there is a normal loading mechanism. The.only possible place would be under the baseplate. Here the ball would deform and i wear slightly until loading would be relieved. No concern, therefore, is seen from this standpoint.

7.2.2 Reactor Upper Internals J

The upper internals were evaluated for the effect of unrecovered. filter debris in the form of a wire ball of .66 inch diameter, such as that found in the steam generators, and for a single wire with a length of 3/8 inch. l j

4 2433s/060187.1C

f. 3

l For the case of the wire ball, it is considered extremely unlikely that a j piece of filter debris of this size would migrate into the upper internals l since it would probably be filtered out by the fuel. However, if a piece of filter debris of this size managed to get through the fuel there would be no j impact on the structural margins of safety of the upper internals since the l ball .is deformeble and the resulting loads would be inconsequential. If the I ball of filter debris gained access into the Upper Internals Guide Tube the potential for jamming a Rod Cluster Control Assembly (RCCA) exists. For a a RCCA to become med, it would be necessary for the ball of filter debris to either become lodged in one of the holes of the guide tube sheath, tubes or guide cards'and a RCCA rodlet or that a ball rested on top of a guide card slot so as to interfere with the RCCA spider. The probability of this occurring is considered remote due to the filtering action of the fuel and the complicated pathway and associated flow conditions necessary to orient the ball of filter debris as described above. Additionally, it is expected that the technical s?ecification requirements of periodic exercising of the RCCAs will detect any significent increase in drag force should the ball of debris i

become jammed between the guide tube and RCCA.

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For the case of a single wire of filter debris, the only component that could potentially be affected during operation it, the upper internals guide tube. A i single wire of 0.038 inch diameter, and 3/8 inch long, could be postulated to rest on a guide tube card slot and thereby potentially jam the RCCA spider during its downward movement. This is considered unlikely during operation since a single wire 3/8 inch long weighs approximately .05 grams and, most likely, would either migrate between the guide tube cards as a result of the flow velocities in the guide tube, or exit the guide tube through the flow windows in the enclosure, or become lodged between the o rd weld nd enclosure slot. Additionally, it could be postulated that a sir - wire could become jammed between a RCCA rodlet and r holes in the guit ' ube sheaths, tubes or guido cards. However, as prevbusi, stated, it is expected that normal RCCA exercising shculd datact high drag forces should a piece of filter debris become jammed between the RCCA and guide tube.

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7-4

l In conclusion, the unrecovered filter debris will not effect the structural integrity of the reactor upper internals. The potential exists for the i unrecovered filter debris to become entrapped in the upper internals guide tube. However, due to the complicated pathway and associti s flow conditions {

in these regions it is considered highly improbable that the unrecovered filter debris will affect RCCA movement. Furthermore, technical specification l requirements to exercise the RCCAs should detect any significant increase in RCCA drg force as a result of filter debris restricting the RCCA.

7.2.3 Reactor Lower Internals l The lower internals were also evaluated for the effect of unrecovered filter debris in the form of a wire ball and single wire as described above. The evaluation determined that a ball of wire will have no affect on the margins of safety or functional operation of the lower internals due to the size and construction of the lower internals compared to the deformable wire ball and its inconsequential impact force. Additionally, a singl.e wire has no impact on the structural margins of safety or functional operation of the lower internals.

7.3 Effect of Unrecovered Filter Screen Debris on Control Rod Drive Mechanisms The operation of the Control Rod Drive Mechanism (CRDM) has been evaluated for the effect of the unrecovered filter debris. Due.to tFe restriction formed by the drive rod and thermal sleeves, whole sections of filter cannot enter the CRDM internals during plant operation. Therefore, the CRDM has been evaluated for the effects of single wire strands of filter debris.

The design of the CRDM latch assembly can be characterized as a series of close fitting cylindrical plunger magnets and support tubes, and two bar linkages. The linkages translate the axial plunger motion into radial motion to grip the drive rod. The linkages are ruggedly constructed to withstand the large dynamic loads placed on the components during normal CRDM operation. If the wire strands from the filter entered this area, the only damage which would occur to the mechanism would be a minor amount of cosmetic scratching.

mwomem 7-5

The stepping operation ef the CRDM may be affected if wire strands " hangup" two interfacing components. Three potential areas have been evaluated; 1) the radial clearances between cylindrical components, which are too small for the wire strands to become lodged, 2) the side clearances between the latches, linkage and :,urrounding parts, which are also too small for wire strands to become lodged, and 3) wire strands entering the latch assembly through the slots in the concentric cylinders and coming to rest under the actuating plunger pieces, this case is discussed below with respect to normal CRDM operation and Rod Holdout Device (RHD) operation.

Normal operation of the CRDM with a few pieces of wire resting under the j plungers will not be affected. If a significant number of wire pieces become {

located at this location, the latch arms would not be able to retract. This j could cause mistepping and could also interfere with the scram performance of l the RCCA. Additionally, no mal exercising of the RCCA, as required by Technical Specification, shculd identify any mistepping which may result from filter debris.

The RHD is operated only during refuelvigs performed in the RAPID mode. It j provides the capability of locking the drive rod and RCCA in the withdrawn position when no electrical power is supplied to the CRDM. If a strand of filter debris was carried into the slots of the concentric cylinders of the stationary gripper part of the latch assembly, it could become lodged between the RHD lock ring and the back of the latch arms. This could prevent positive i lock-out of the drive rod /RCCA assembly.

The probability of problems with normal CRDM operation or RHD operation are considered extremely remote due to the significant amount of material necessary to be present to affect operation of the CRDM and the relative isolation of the CRDM internals from the high flow regions of the reactor vessel. This isolation was demonstrated by the results of the hydrolazing of the CRDMs which revealed that no filter debris was found in the CRDMs following hot functional testing when the greatest quantity of filter debris was present and circulating in the system. Also, the CRDM internals will be further isolated during normal plant operation since the drive rod will be in mwoswc 7-6

place, while during hot functional testing it is net. Additionally, normal exercising of the RCCAs, as required by Technical Specification, should l identify any mistepping which may result from unrecovered filter debris.

1 In conclusion, the probability of the unrecovered filter debris causing CRDM operational problems is considered extremely remote.

7.4 Effect of Filter Debris on Steam Generators The effect of a wire ball of filter debris, such as that found following hot functional testing, has been evaluated and determined that it.does not impact the structural margins of safety because the. deformable wire results in small l loadings on components impacted by the ball of wire. However, over a period >

of- time, a wire ball may be a source site for corrosion if it is stuck in a f steam generator tube during operation. Also, scratches greater than 0.002 inch could also serve as a potential site for eventual stress corrosion crack Mg. The potentia l for these scratches exists since tests have shown that the hardness of the filter screen braze material is higher than that of the Alloy 600 steam generator tubes. The significance of scra'tching is that it may serve as a potential site for eventual stress corrosion cracking. It  ;

is believed that eddy current surveillance based on the technical l

specification campling plan should be sufficient to permit monitoring of steam j generator tube integrity.  !

In conclusion, balls of wire or single strands of filter debris will not have a structural impact on the margins of safety of the steam generators since  ;

loads due to impacting of the filter debris are considered insignificant due to the small mass of the pieces. The potential for scratching of the inside surface of the steam generator tubes exists but this is considered to be more of a, functionality concern later in life for the operation of the steam generator rather than a structural concern. Normal in-service inspections with eddy current is expected to detect any major flawi or crevices that might eventually occur as a result of scratches or small gauges from filter debris.

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l' 7.5. Effect of Filter Debris on the Pressurizer-A wire ball or single wire of filter debris will have no impact on the structural margins of safety or functional operation of the pressurizer due to the small mass of the filter debris.

The pressurizer safety and relief valves are open only for certain system transient conditions. In the normal operating mode, the valves are in the closed position and there is no flow through the valves. The filter debris is not expected to have any impact on the pressurizer safety.and relief valves.

7.6. Effect of Filter Debris on Reactor Coolant Pumps The unracovered filter debris in the form of a wire ball, .such'as-that found in the steam generator,. or a single wire 3/8 inch long will have no . impact on  !

the reactor coolant pump (RCP) during normal operation. The filter debris would pass through the normal flow path without consequence. The possibility exists that slight scoring may occur on the surfaces of hydraulic components i but this possibility is remote and any scoring would' be insignificant.

In the event of a loss of injection (LOI) condition, however, the possible presence of filter debris in the RCS poses a serious condition with the possibility of bearing and/or seal failure. During LOI,-loop water rises into the shaft a.nnulus, carrying with it debris present in the system. An isolated penetration of a single piece of bare wire (no braze material on the surface) is unlikely to cause any problems as the shaft labyrinths would-tend to " chew up" the foreign debris. It is the possibility of multiple foreign debris penetrations compounded by the presence of .the hard Ni alloy brazing that rive-rise to concern for the bearing and seals.

Possible migration paths during LOI for the wire debris and'the impact of.the migrations on the RCP depend on the configuration and size of the debris material. Two possibilities are evaluated here: a screen ball 0.66 inch in diameter and multiple pieces of wire approximately 3/8 inch in length.

am,mmm 7-8

1 i

The postulated screen ball could migrate to the labyrinths'at the upper seal [

wear ring and become lodged. Grooving of the wear ring might then occur due to the hardness of the brazing material compared to the stainless steel impeller. This grooving would have no affect on RCP operation. The screen l ball would eventually work itself free and flow through the normal' flow passages. 'The size of the ball would prevent it from migrating into areas ,

where it could cause serious damage.

The postulated wire pieces present the worst problems to the'RCP internals during an LOI condition. Because of the hard Ni alloy braze on the wire, the debris will tend to " chew up" the labyrinths rather than be ground up by them. While this damage to the labyrinths and the mating surfaces on the impeller and shaf t will not significantly affect the structural integrity or '

operation of the RCP, the clearances could be opened up and allow more debris into the shaft annulus. }

Once inside the shaft annulus, the foreign debris can follow several pathways. The particles could settle in the open space below the bearing where they would be of no consequence, or migration of particles into the heat exchanger might occur. Here, the particle would become entrapped in the cooling coil stack. The subsequent loss in heat transfer, as one might expect, would be minute.

Another possible pathway, and much more serious, for the wire debris is up through the bearing, between the journal and the graphitar rings.

Westinghouse experience has shown that an isolated penetration of a.small foreign particle into the bearing has no impact on pump operation.- Upon disassembly and inspection, superficial scoring in a spiral pattern on the.

graphitar bore might be visible. In this event, the bearirig is typically replaced as a precautionary measure only.

If foreign debris penetration is more significant, i.e., more or larger particles, the likelihood of bearing failure is greatly increased. The bearing graphitar would completely disintegrate and dissipate graphite throughout the pump internals. The event is serious and damage is extensive.

unmuao 79  ;

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l If particles make it through the bearing graphitar or flow holes, they could settle in the annulus above the bearing where they would be of little consequence. Or, particles could work up to the No. 1 faceplates. As in the case of foreign debris penetration through the bearing, an isolated penetration through the No.1 faceplates might not caure a problem. The likelihood of seal failure increases with the number and size of wire strands introduced in the area.

In summary, the filter debris will have no effect on normal RCP operation.

DuringanLOIcondition,however,thedebrisinthesystemcouldmigrateinto the pump internals and cause serious damage to the bearing and seals. It is Westinghouse recommended procedure to have seal injection on at all times while the pump is running and also during filling and venting of the RCS.

This recommendation is made even stronger and more emphatically with the known possibility of wire debris in the RCS.

7.7 Effect of Filter Debris on Primary System Piping Inspections after hot functional testing of the accessible areas of the reactor coolant piping and penetrations (thermowells) have revealed no damage

.nd only limited superficial scratches. Any additional impact from the remaining filter debris would bs negligible primarily because a greater quantity of filter debris was present in the RCS during hot functional testing. Thus it can be reasonably postulated that the limited quantity of debris now present will have little if any impact. Also.the remaining fi.lter debris is characterized by both relatively small size and mass, and consequently possesses little ability to produce any mechanical adverse affect.

7.8 Effects of Filter Debris on Bottom Mounted Instrumentation The unrecovered filter debris has been evaluated to determine its effect on the bottom mounted instrumentation (BMI) system. The type and size of the expected debris will not interfere with the operation of the BMI flow '

limiter. The only other component of the BMI system that may potentially come in contact with the unrecovered filter debris is the flux thimble. The flux thimble is a completely sealed tube which passes through the lower internals 1

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• 8 " o 7-10

instrumentation columns and the fuel assemblies. Since the thimbles are sealed and have no external moving parts, the filter debris will have no impact on the component. To prevent vibration / wear of the flux thimbles, flow limiters and lower internals instrumentation column sleeves were added.

Evaluation shows that filter debris will not impact the intended function of these devices. Therefore, thimble wear due to unrecovered filter debris is not a concern.

7.9 Effects of Filter Debris on Auxiliary Systems The following i.s the evaluation of the effects of the unrecovered filter debris on the CiC and RHR systems. This evaluation was limited to these systems since these were the Westinghouse systems directly connected to the l RCS which were in service during hot functional testing.

CVCS Note: Except for isolation capability, the function of the CVCS l'etdown line is not safety-related.

LETDOWN STOP VALVES (LCV-465, -468)

RCS pressure boundary (safety related), but redundancy exists imprc.bable trap due to valve design (gate)

REGENERATIVE HX performance unaffected long-term wear potential of trapped debris small (small mass)

LETDOWN ORIFICES straight pieces will pass through (high velocity) j performance unaffected I i

I mwoeomo 7-11

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LETDOWNORIFICEISOLATIONVALVES(8121,8122,8123) short travel (less than 1/2 in.), curled wire segment could be caught letdownLisolation ability available elsewhere supplied by Bechtel j

i LETDOWN ORIFICE TRIM VALVES (8124, 8125, 8126) locked in position, no need to close performance unaffected supplied by Bechtel LETDOWN LINE CONTAINMENT ISOLATION VALVES (8133, 8134) safety-related, but redundant improbable trap due to valve design (gate)

BTRS REHEAT HX THROTTLE VALVES (TCV-381A, B) not safety related BTRS not normally used (TCV-381B full open)

LETDOWN HX performance unaffected '

long-term wear potential of. trapped debris small LETDOWN LINE PRESSURE CONTROL VALVE (PCV-135) 4 not safety related not required to isolate flow LETDOWN LINE TEMPERATURE CONTROL VALVE (TCV-143) -

not safety related small debris pieces not likely to affect divert function

. un,mcwo 7-12

LETDOWN LINE FLOW ELEMENT (FE-132) not safety related small wire segments will pass through (bore - 2.5 in.)

LETDOWN FILTER / REACTOR COOLANT FILTER -

designed to trap small particles debris penetration into CVCS not considered past this point RHRS RHR H.L. SUCTION ISOLATION VALVES (9000A-C, 9001A-C)

RCS pressure boundary (safety related), but redundancy exists improbable trap due to valve size an~d type (12" gate) -

RHR PUMP used for plant cooldown (safety related) minimum wear clearances 9-28 mils; therefore wire not expected to become lodged in wear areas vane to casing clearance approximately 1/2 inch; debris will pass through with pumpage RHR PUMP DISCHARGE FLOW ELEMENT (FE-867, -8, -9) small wire segments will pass through (bore = 6.0 in.)

RHR HX perfort ance unaffectad long-term wear potential of trapped debris small (small mass) unvoeciano 7 13

l RHR BUTTERFLY VALVES (HCV-864, -5, -6; FCV-851, -2, -3) not safety .alated j improbable trap due to valve size (8") and type RHR/LHSI C.L. FLOW ELEMENT (FE-851, -2, -3) -

small wire segments will pass through (bore = 5'.5 in.)

RHR RETURN LINE/LHSI C.L. INJECTION LINE VALVES (8901A, B, C, 8902A, B, C, 8948A, B, C)

I -

M0V's normally open for SI i

improbable trap due to valve sizes and types (8" gates and checks,12"' checks) -

The operability of the TGX CVCS and RHRS is expected to be unaffected by the form and amount of the unrecovered core filter debris. l In addition to the above, the effect of filter debris on the operation of the safety injection and containment spray systems in the recirculation :aode of operation, assuming that the filter debris is swept from the RCS into the emergency sumps following a design basis accident was evaluated. This evaluation is based on the postulation that a " screen ball of 0.66 inch in i diameter and/or wire in the length of 3/8 inch in length" enters the auxiliary systems. The wire diameter is assumed to be in the range of .035 to .046 inch.

i Per section 6.2.2.2.3 of the TGX FSAR, the second stage of screen in the emergency sump is designed to limit the small particles which enter the ECCS pump suction line to less than 1/4 inch in diameter. Particles of this size cannot clog the containment spray nozzles (3/8 inch orifice diameter), which are the smallest restrictions found in any system served by the sump. On the basis of this sump screen size, the entrance of the postulatsd 0.66 inch 2433s/063187.10 71

diameter screen ball into the SIS / CSS is precluded. Small lengths of wire which are postulated to pass through the sump screen end first are assumed to also pass through the remainder of the ECCS, including the containment spray nozzles. ECCS operability is therefore expected to be unaffected by the form and amount of unrecovered core filter debris.

Also evaluated was the effect of the unrecovered filter debris on the sensing lines for RCS flow, pressure and level instrumentation in which the fluid is static. Based on the assumption that the sensing line is 1/4 inch tubing, the following assessment of the effect of wire debris en the function of the instrument is made.

Due to the static nature of the sensing line and typical line layouts in which the instrument line ties in to the process line in a vertical run or above the midplane in a horizontal run, fouling of the sensing line by wire debris is unlikely. However, even assuming that wire does enter an instrument sensing line, the form of the postulated wire debris cannot completely block the line so as to prevent the transmission of the fluid pressure to the instrument. A partial reduction in the cross-section area of the sensing line by a single wire segment should have no effect on the function of the instrument. RCS instrumentation (most of which are redundant) is therefore expected to be unaffected by the form and amount of the unrecovered filter debris.

7.10 EFFECTS OF FILTER DEBRIS ON REACTOR COOLANT SYSTEM MATERIALS There is no anticipated short or long term metallurgical effect expected from the unrecovered filter debris in contact with the materials which make up the reactor coolant system with the possible exception of the steam , generator tubes which is discussed in paragraph 7.4 above. Any minor scratches which may occur to the surfaces of the components, constructed of or clad with stainless steel or inconel, would not constitute a safety issue.

7.11 CONCLUSIONS As a result of the engineering evaluations performed for operation with unrecovered filter debris, essentially no potential exists for the filter debris to reduce any structural margins of safety or cause major operational 2433s/06018T10 7 }g

concerns for any of the equipment in the primary coolant and auxiliary systems. Operational concerns are minimal. Of the areas evaluated, it was noted that the standard technical guidelines relative to RCCA centrol rods and reactor coolant pumps sould be followed during operation. Successful operation of the safety related isolation valves should not be impaired due to redundancy and design of the valves.

The effect of filter debris on the fuel integrity has been addressed from cladding breach and thermal and hydraulic standpoints. In conclusion, the filter debris is capable of causing fuel clad breach which could result in increased coolant radioactivity levels. Normally performed monitoring is expected to preclude the poscibility of exceeding design basis coolant radioactivity limits. It is also highly unlikely that potential core flow blockage due to filter debris will adversely affect the DNB evaluations for the fuel core.

It should be noted tnat it is believed that the actual amount of unrecovered filter debris is less than the amount assumed for these evaluations due to the thorough inspections and filter debris retrieval performed on the system. A majority of the remaining amount of filter debris may have been filtered out during the hot functional testing or removed during system drainage.

Therefore, these evaluations are considered conservative and represent the effects of the bounding filter debris amount condition.

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5/22/87 APPENDIX A SAFETY EVALUATION (WCAP-11506, FULL-FLOW FILTER REC 0VERY AND EQUIPMENT ASSESSMENT REPORT, SOUTH TEXAS PROJECT UNIT 1) 2432s/060 t 47.10 A-1

0 TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 SAFETY EVALUATION 2.1 REACTOR COOLANT SYSTEM 2.2 AUXILIARY SYSTEMS 2.3 INSTRUMENTATION

4.0 REFERENCES

u n uosoierin A-2

1.0 INTRODUCTION

The completion of the second phase of the corrective action, as noted in the Interim Report (Ref.1), resulted in the following basic conclusions:

1. The results of the detailed inspection of the reactor coolant and interfacing systems have indicated that there was no damage to any of the components.

2. As a result of the thorough inspection and flushing of the reactor coolant and interfacing systems it is believed that essentially all of the filter debris has been removed from these areas.

3. The unaccounted for filter debris, which has been calculated to be

~ 194 grams, is believed to have been released from these systems with no identificable return path.

Based on the above, no situation more severe than previously analyzed in the FSAR was identified.

Since, as stated in 3 above, there is ~ 194 grams of unaccounted for filter debris, for purposes of evaluation, it ias been postulated that if some of the filter debris remained in these systems, it would be expected to be in the form of small pieces of 0.038 inch diameter wire, approximately 3/8 inches long, or a ball of the wire material approximately 0.66 inches in diameter.

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l 2.0 SAFETY EVALUATION The Safety Evaluation, assuming filter debris remaining, is divided into three areas of discussion. These are the Reactor Coolant System, Auxiliary Systems f and Instrumentation.

2.1 REACTOR COOLANT SYSTEM 2.1.1 Nuclear Fuel The filter debris has two potential effects on the nuclear fuel:

1. partial flow blockage of fuel assemblies due to the filter debris loo 3i ng in the fuel assembly flow path,

2. clad wear due to filter debris becoming lodged within the assembly or between two assemblies.

From a fuel mechanical design viewpoint, the filter debris should not pose a safety issue provided the fuel assemblies can be seated properly on the core plate. A portion of the filter debris would be trapped by the bottom nozzle or the bottom core plate due to dimensional considerations.

i From a thermal / hydraulics viewpoint, results of the evaluation show that the presence of the unrecovered filter debris is not expected to result in any operational issues from a flow blockage / Departure from Nucleate Boiling (DNB) standpoint. '

Based on the size of unrecovered filter debris, the evaluation involved three basic conditions, including:

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1. the effects of filter debris entrapped by the bottom nozzle plate,

2. the effects of filter debris carried upward into the fuel assemblies, un,nectoio A-4 L l

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3. the effects of filter debris on core thermal and hydraulic performance Filter Debris Entrapped by the Bottom Nozzle Plate Particles of filter debris reaching the core inlet which are larger than about 0.25 inches in diameter such as the 0.66 inch diameter balls could be trapped either by the fuel assembly bottom nozzles or bottom grids. Computer models indicate that when flow is completely blocked to one assembly at the inlet, full recovery of the flow occurs about 30 inches downstream of the blockage and the effect of this blockage is not limiting with respect to DNB. It is unlikely that more than one fuel assembly would be completely blocked in a localized region of the core inlet.

Filter Debris Within the Fuel Assemblies The filter debris, in the form of individual wire, is within the range of the types of filter debris which are known to be capabla of causing fuel cladding breach. Previous investigations have shown this to be the result of fuel rod fretting wear from debris trapped at fuel assembly grid elevations. Primary coolant radioactivity levels are monitored regularly according to Technical Specification guidelines.

It is expected that the monitoring requirements of the Technical Specifications will preclude the possibility of exceeding design basis coolant radioactivity limits during operation.

Thermal and Hydraulic Performance Filter debris smaller than about 0.25 inch in width is expected to pass through the fuel rod array and one or more grids. Some of this filter debris might pass entirely through the core, particularly, short straight segments of filter debris, but some filter debris could be expected to be entrapped by any of the fuel assembly grids. I However, it is not likely that enough filter debris would become lodged in one location to significantly affect the core DNB. Tests

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performed on open lattice fuel assemblies ~show that a blockage.of_up to 41%.at a grid location is acceptable, with disappearance of the.

stagnant zone behind the flow blockage after 1.65 L/De (length / equivalent hydraulic diameter).. Local flow blockages of this-type have little effect on subchannel enthalpy rise-and cause only minor perturbations in local mass flow velocity. In reality, these blockages promote turbulence and thus would likely not adversely affect DNB at all.

2.1.2 _ Reactor Vessel and Internals An extensive inspection of the Reactor Vessel (RV) and core support components, and.RV internals .(upper and lower packages) has been conducted to evaluate the condition-of these components af ter the hot functional testing during which time a portion of the full flow filter failed.-_ It was during the hot functional testing that the greatest quantity of filter debris was present in the RCS. These inspections have revealed no significant physical or structural change, and the margins of safety'for these components remains l unchanged, i

2.1.2.1 Reactor Vessel The RV lower head, core barrel ledge, and upper head were inspected and cleaned following the recent hot functional testing. These inspections '

revealed no adverse affect, and the vessel margin of safety'is judged to be uncompromised. The remaining filter debris is also considered to not challenge the integrity of the RV. It'is thought that a portion of the' debris will collect in the relatively stagnant flow area of the lower plenum region  !

where the possibility for wedging between compenents would be reduced.

The lower head Bottom Mounted Instrumentation (BMI) cylindrical penetrations were considered as a possible area where flow induced v'bration could result in wear if the debris were to orient itself b such a fashion. However preliminary calculations indicate that complete wear through of a wall of a BMI penetration would be highly unlikely based on the expected geometry of the >

debris.

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2.1.2.2 @oer Internals Inspection results and cleaning of the upper internals package have revealed no physical effect on the guide tubes, support columns, keyways, upper support plate, core plate, locking caps on threaded fasteners, welds, and thermocouple columns, conduits and brackets. Any scratches that might appear on the stainless steel surfaces of the reactor internals are not believed to be of concern for cracking, because of the superior resistance of this material to Stress Corrosion Cracking (SCC) as well as high toughness and ductility. It is evaluated from the inspection results that the margins of safety of the upper internals components and structures have not been reduced and that the functions of these components and structures will not be affected.

The remaining quantity of filter debris in the RCS is not expected to adversely impact the RV upper internals. The only component of the package that could potentially be affected by the filter debris during operation is the guide tube. The sheath tubes and guide tube cards represent areas where the filter debris could potentially become trapped. However, due to the complicated pathway and associated flow co~nditions in these regions it is highly improbable that the filter debris will affect control rod movement.

Furthermore, Technical Specification requirements to exercise the control rods will detect any significant increase in rod drag force as a result of any filter debris restricting movement. Appropriate action will be taken if high drag forces are detected.

2.1.2.3 Lower Internals Inspection results and cleaning of the lower internals package have revealed no physical effect upon the core barrel edge, spray nozzles, specimen basket, baffle-former area, BMI columns, keyways, fuel pins, lower support plate, neutron pads, fastener locking cups and, welds. As with the upper internals, scratches that might occur en the stainless steel surfaces of the lower internals are not believed to be of any concern. It is considered from the inspection results that the margins of safety of lower internals has not been reduced and the fun:tions of these components will not be affected.

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The present quantity of filter debris in the RCS is not expected to adversely impact the RV lower internals. The size and mass of the filter debris preclude any consequential impact force.

2.1.3 Control Rod Drive Mechanisms The relative isolation of the Control Rod Drive Mechanism (CRDM) aternals from the high flow regions of the RV suggest that extensive migration of the filter debris up into the CRDM internals during hot function testing is unlikely.. Results of the'CRDM hydrolazing conducted on site support this scenario. However, a portion of the remaining filter debris could possibly.

migrate to the CRDM latch region and in significant quantity affect the latch arms ability to retract. This unlikely event could cause misstepping of the CRDM and could also interfere with scram performance 'of the control rod. It is felt that the above scenario represents an improbable occurrence. This conclusion is based partly on the results of the hydrolazing which revealed that no filter debris was present in the CRDM after hot functional testing.

It is auring hot functional testing that the greatest quantity of filter debris was present and circulating in the system. Secondly, the drive rods will be in place during normal operation. This will further isolate CRDM internal cavities from the upper internals to a greater extent than during hot functional tests, which are dono without drive rods. Finally, the amount and orientation of screen debris required tc, lodge in the CRDM internals and produce misstepping preclude the possibility of this occurrence. Based on the above, the chances for a stuck rod as a resul.t of screen debris in the CRDM internals are judged to be remote.

2.1.4 Steam Generators Because of hot functional testing with the filter debris present in the RCS, an assessment was undertaken for the steam generators. Visual inspection has revealed no evidence of any physical effect on the steam generator channel head divider plate, the tubesheet cladding, and the tube-to-tubesheet welds.

Also, a FOSAR was conducted in each steam generator in an attempt to recover any filter debris from the steam generator channel head and tubing. When F0SAR efforts located filter debris in a particular tube, the inside surface un.maano A-8

l of the tube was visually inspected using a boroscope for tube scratches or-tube wall loss. No significant tube scratches or tube wall reductions were -

-observed.

Concerning subsequent plant operation, if the filter debris were to enter a steam generator channel head, the effects would be expected to be at most a peening type impact on the tubesheet cladding and channel head. The Model E steam generators have a flush tube-to-tubesheet weld neither the tube end nor ,b the tube-to-tubesheet weld would be expected to be repeatedly impa'cted by the f filter debris in the channel head. However, should ippatting occur, due to the small mass of the pieces, the effect on the primary' system pressure boundary and the steam generator structural components is insignificant.

f It is possible that filter debris could enter the steen generator tubes. Due to the small mass of the filter debris, impacting of the loose filter debris on a tube inside surface as it travels through a tube is judged not to significantly affect the integrity of the tubing. As a result of the recent assessment to the steam generator tuoes, the relative hardness of the filter debris and braze material (which may be attached to the filter debris) was compared to the Alloy 600 tubing hardness. The Inconel tubing hardness is slightly higher than the wire, but less than the hardness of the braze material. Therefore, the potential exists that scratches may occur on the inner surface of a tubing as the filter debris tqavels through a tube. The significance of any scratching is that such scratching constitutes a potential site for SCC occurrence. To date, Westinghouse has no in service data linking the extent of scratches on the inner surface of a tube to the future occurrence of primary water SCC. Also, if the filter debris wer.e to become lodged in a steam generator tube, localized tube wall degradation due to a tube wear mechanism would not be expected to occur due to the lack of relative motion between the flow filter screen and the inside surface of the tube. <

t Most importantly, should unrecovered filter debris enter into the steam generator tubing during future plant operation, eddy current surveillance based on the Technical Specification sampling plan should be sufficient to '

permit maintenance of steam generator tube integrity.

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2.1.5 Pressurizer Inspection results of the pressurizer pressure boundary, spray nozzles, and outlet screen report no physical effect on such components. Consequently, it can be confidently assumed that the margins of safety for the pressure bounda v and the functionality of the pressurizer were unaffected by the filter debris circulating through the RCS during hot functional testing. This conclusion will also hold true for operation with the remaining quantity of filter debris present in the RCS, It is considered unlikely that any filter debris will interfere with operation of the Power Operated Relief Valves (PORV) and Safety Valves due to the physical separation of these valves from the RCS. Furthermore, redundancy is provided in the system through the utilization of two PORV's and three safety j valves at the pressurizer. I I

2.1.6 Reactor Coolan+ ' .g

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It is expected that 4 - t Coolant Pump (RCP) will not be affected by _the filter debris during n..r .lant operation. The filter debris should pass through the pump with no change in pump vibrational characteristics or increase in locked rotor accident probability. Any significant mechanical impact to the pump impeller is considered to be highly unlikely due to the size and quantity of the filter debris.

The RCP seal Loss Of Injection (LOI) event poses a condition in which the possioility exists for RCP bearing and/or seal failure. This possible-conditier. exists because during a LOI event reactor coolant in the loop rises into the pump shaft annulus, carrying with it debris that may be in the RCS.

It is conceivable that any piece; af filter debris, coupled with the nickel alloy brazing found on some pieces of the recovered filter debris, could challenge the integrity of the bearing and seals. Should this condition occur the increases in RCP shaft vibration and in temperature of the bearings and no.1 seal would alert the operator and allow for adequate recognition of the event. Once identified, the operator could isolate the pump and proceed with .

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I standard event response procedures. In this case seal failure beyond the no.

1 RCP seal should not occur and any leakage would be routed to the RCS drain 1 tank. In the event that high RCP shaft vibration and/or the seal / bearing temperature condition is allowed to persist some potential for a spill to

! containment exists. However, isolated single pump spills are well within the capacity of the plant safety systems. l It should be noted that during the recent hot functional testing the seal injection was on, as reported by the site, and as a consequence the above situation could not occur. To provide added assurance, the seals in RCP C were disassembled and the shaft annulus was inspected. The inspection revealed no damage or filter debris. Since the time the RCP was running during hot functional testing, the greater quantity of filter debris has been recovered from the various plant systems. The reduction in quantity of filter debris, as a result of the recovery effort, is expected to lessen the probability that the required amount and type of filter debris would enter the RCP internals.

2.1.7 Primary System Piping inspections after hot functional testing of the accessible areas of the reactor coolant piping and penetrations (thermowells) have revealed no damage and only limited superficial scratches. Any additional impact from the remaining filter debris would be negligible primarily because a greater quantity.of filter debris was present in the RCS during hot functional a sting. Thus it can be re.asonably postulated that the limited quantity of debris now present will have little if any impact. Also the remainingLfilter debris .is characterized by both relatively small size and mass, and consequently possesses little ability to produce any mechanical adverse affect.

2.1.8 Materials i

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There is no anticipated short or long term metallurgical effect expected from the filter debris in contact with the reactor coolant system materials. Any I miner surface scratches to the cladding would not constitute a safety issue. l l

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2.2 AUXILIARY SYSTEMS-

=2.2.1 Chemical'and Volume Control Sys, tem (CVCS)

Note: Except for isolation capability, the function'of the CVCS letdown line is not safety-related.

Letdown Stop Valves (LCV-465, -468)'

RCSpressureboundary-(safetyrelated),butredandancyexists l improbable trap due to valve design (gato) j (large flow area in full open position)' O letdown isolation capability available elsewhere (8121,8122,8123)

Regenerative Heat Exchanger i performance unaffected long-term wear potential of trapped debris small (small mass)

Letdown Orifices straight pieces will pass through (high velocity)  !

performance unaffected Letdown Orifice Isolation Valves (8121, 8122, 8123) 1 short travel (less than 1/2 in.), curled wire segment could be caught letdown isolation ability available elsewhere (LCV-465,-468) supplied by Bechtel l 1

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l Letdown Orifice Trini Valves (8124',' 8125, 8126)'-

- ' locked in position, no need to close performance unaffected j supplied by Bechtel l

.I Letdown Line Containment Isolation Valves-(8133, 8134) safety-related, but redundant- -

improbabletrapduetovalvedesign(gate)

(large. flow area in full open position)

BTRS Reheat Heat E:: changer Throttle Valves (TCV-381A', B) ,q not safety related BTRS not normally used (TCV-3818 full open)

Letdown Heat Exchanger function not safety related performance unaffect'ed long-term wear potential of trapped debris small l Letdown Line Pressure Control Valve (PCV-135) not safety related not required to i.solate flow Letdown Line Temperature Control Valve (TCV-143) j I

not safety related j small debris pieces not likely to affect divert function j i

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Letdown Line Flow Element.'(FE-132) not safety related small wire segments will pass through (bore = 2.5 in.)

Letdown Filter / Reactor Coolant Filter  ;

designed to trap small particles debris penetration into CVCS not considered past.this point see BRS/BTRS evaluation (p. 6) i 2.2.2 Residual Heat Removal System (RHRS)

RHR H.L. Suction Isolation Valves (9000A-C, 9001A-C) l RCS pressure boundary (safety related), but redundancy exists j improbable trap due to valve size and type (12" gate)

(large flow area in full open position)

RHR Pump used for plant cooldown (safety related) minimen wear clearances 9-28 mils; therefore wire not expected to

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become lodged in wear areas 1 vane to casing clearance approximately 1/2 inch; debris will pass ]

through with pumpage. 1 RHR Pump Discharge Flow Element (FE-867, -8, -9),

small wire segments will pass through (bore = 6.0 in) j i

RHR Heat Exchanger performance unaffected long-term wear potential of trapped debris small (small mass) mwoniano A-14 J

RHR Butterfly Valves (HCV-864, -5, -6; FCV-851, -2, -3)  ;

I not safety related improbable trap due to valve size (8") and type i

RHR/LHSI C.L. Flow Element (FE-851, -2, -3) small wire segments will pass through (bore = 5.5 in.)

l RHR Return Line/LHSI C.L. Injection Line Valves j (8901A, B, C, 8902A, B, C, 8948A, B, C)

M0V's normally open for SI improbable trap due to valve sizes and types (8" gates and checks, 12" checks)

(large flow area in full open position) 2.2.3 Boron Recycle System (BRS) & Boron Thermal Regeneration System (BTRS) l The BRS and BTRS are not safety re. lated. They take flow after the CVCS Letdown filter, which is dei,igned to trap small particles. Debris penetration into the CVCS past this filter in not considered probable. Therefore, penetration of full flow filter screen debris into the BRS and BTRS is'not considered probable or of significant consequence.

l 2.2.4 Emergency Core Cooling System (ECCS)

(Recirculation Mode) i This evaluation assumes that the full flow filter screen debris is swept from ')

the RCS into the emergency sumps following a Design Basis Accident. Per 1 section 6.2.2.2.3 of the TGX Final Safety Analysis Report, the second stage of .

screen in the emergency sump is design to limit the small particles which l enter the ECCS pump suction line to less than 1/4 inch in diameter. Particles of this size cannot clog the Containment Spray nozzles (3/8 inch orifice l diameter), which are the smallest restrictions found in any system served by I the sump.

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On the basis of this sump screen size , the cntrance of the postulated 0.66 inch diameter screen ball into the Safety Injection System / Containment Spray System is precluded. The small lengths of wire which are postulated to pass through the sump screen are assumed to also pass through'the tha remainder of ECCS, including the Containment Spray nozzles. The small lengths of wire will pass through.the high head safety injection, low head safety injection and containment spray pumps. Also,.these wires will have no impact on the safe function of the ECCS valves.

2.2.5 Reactor Vessel Head Vent System (RVHVS)

The Reactor Vessel Head Veat System (RVHVS) consists of two portions: (1) a local, manual vent line to containment atmosphere for normal filling of the RCS following refueling or maintenance and (2) a remotely operable, safety grade vent path to the Pressurizer Relief Tank for use in venting non-condensible gases from the upper head post-accident or for providing a' J safety grade letdown flowpath The safety evaluation of the Reactor Coolant System, SECL-87-154, notes that the reactor vessel upper. internals and upper head area were inspected for traces of the wire debris, with none found. Under conditions.in which the RVHVS would be expected to be utilized, core vessel hydraulics would not be I favorable to debris intrainment in the upper head (low flow or stagnant.

conditions). Therefore, wire debris is not expected to enter the RVHVS inlet i connection to the vessel head.

t However, assuming that wire debris does migrate into the RVHVS, if.the debris did interfere with the closure of the local manual valves used in the normal venting oc the RCS, repairs can b' e affected before the plant. start-up operations proceed. Therefore, this portion of the RVHVS is unaffected by any postulated wire debris.

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Should any unrecoverable wire' debris' enter the safety grade portion of the RVHVS, the series / parallel valve arrangement _ ensures that, given a single failure of any one isolation valve or throttle valve in the RVHVS vent / letdown path, flow can be established and can also be isolated.' Therefore, the safety function of- the RVHVS is expected to be' unaffected by the- form and amount of the unrecovered core filter debris, i

2.3 INSTRUMENTATION (REF.4) l The impact of unrecovered filter debris in the reactor coolant system on the  !

capability of the plant instrumentation to perform its functions has been evaluated.

Due to the static nature of the sensing lines and the typical line layouts in'  !

which the instrument lines connects to the process lines in a vertical run or above the midplane in a horizontal run, fouling of the sensing line by filter debris is unlikely. In the unlikely event that sensing-lines become fouled due to the filter debris, the plant has redundant channels installed for all required safety related instrumentation. Utilizing the redundant channels of  !

information, the operators would be expected and detect those channels that are postulated to be fouled. Furthermore, the Qualified Display Processing System (QDPS) performs quality coding on all Regulatory. Guide 1.97 Revision 2 Category 1 instrumentation channels and other variables utilized as inputs to the reactor protection system. These incluse narrow range RCS temperature and RCS flow.

As indicated in the structural evaluation, after~ a thorough inspection, no physical damage was observed with respect to the core exit thermocouples and the RCS RTD thermowells.

Hence, the performance of the plant instrumentation will continue to perform

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in a satisfactory manner even with the possibility of unrecovered filter debris in the RCS.

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3.0 CONCLUSION

The operability of the STP is expected to be unaffected even assuming I unrecovered filter debris remaining in the systems. Operation would not be.

expected to increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR or create the possibility for an accident or malfunction of a different type than that evaluated nor decrease the margin of safety as I defined in the bases for any technical specification, i Thus the unrecovered filter debris does not constitute a threat to the safe and reliable operation of the plant.

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4.0 REFERENCES

1. Ref.: ST-HL-AE-2120, 5/8/87, Letter from J. H. Goldberg to NRC relative i to 10CFR50.55(e) notification, Interim Report. 'j i

2. SECL-87-152. '

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3. SECL-87-233.

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4. SECL-87-235. r t

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