ML20126E832

Insp Repts 50-456/92-23 & 50-457/92-23 on 921013-1130. Violations Noted.Major Areas Inspected:Ler Review,Outages, Radiation Protection,Operational Safety Verification,Monthly Surveillance Observation & Rept Review

ML20126E832
Person / Time
Site:	Braidwood
Issue date:	12/17/1992
From:	Farber M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20126E829	List: ML20126E826 ML20126E830 ML20126E832
References
50-456-92-23, 50-457-92-23, NUDOCS 9212300007
Download: ML20126E832 (12)

See also: IR 05000456/1992023

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III .

Reports No. 50-456/92023(DRP); 50-457/92023(DRP)

Docket Nos. 50-456; 50-457 Licenses No. N?F-72; NPF-77

Licensee: Commonwealth Edison Company

Opus West III

1400 Opus Place

Downers Grove, IL 60515

Facility Name: Braidwood Station, Units 1 and 2

Inspection At: Braidwood site, Braidwood, Illir.ais

Inspection Conducted: October'13 through November 30, 1992

Inspectors: S. G. Du Pont

J. R. Roton

G. M. Hausman

Approved By: M.Farbh,C M 7!92-

Reacto/ProjectsSectionlA Date

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Inspection Summary-

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Inspection from October 13 throuch November 30. 1992 (Reports No. 50-

456/92023(DRP): 50-457/92023(DRP))

Areas Inspected: Routine,. unannounced safety inspection by the resident and

regional inspectors of licensee action on previously identified items;

licensee event report review; outages; radiation protection; operational

safety verification; mont!.ly surveillance observation; and report review.

Results: Three violations were identif':ed in one of the six areas inspected.

In the remaining areas, no violations were identified.

The. following is a summary of the licensee's performance during this

inspection period:

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L Plant Ooerations

The licensee's performance in this area for this inspection period.was1

good. Shift briefings continued to provide sufficient'information for.

l planned evolutions to be performed during the shift. The. inspectors

have raised several questions involving operability determinations

associated with the Main Steam Line Code Safety Valves.-

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Radiolooical Controls-

Three violations were issued due to the licensee's failure to adequately

control the addition of SF -to the steam generators (two violations) and

the failure to adhere to the posting requirements' of Radiologically

Controlled' Areas'. Additionally, the report discusses activities

associated with safety evaluations for the SF, and a chloride excursion.

One was an example of good efforts producing a detailed evaluation and

-the other was an example of a failure to perform an evaluation.

Safety Assessment /0uality Verification

The one LER reviewed during this inspection period appears to have

appropriate corrective actions to preclude similar events. The

licensee's evaluation of the Unit I chloride excursion is a good example

of a detailad and comprehensive safety assessment. However, the failure

to conduct a similarly comprehensive evaluation for the sulfur

hexafluoride addition indicates that the sensitivity to and

understanding of the need for safety assessments is not uniform

throughout the licensee's organization.

Enaineerina and Technical Support

Due to the inspectors limited review in this area, the licensee's

performance was not assessed for this inspection period.

Maintenance and' Surveillance

The licensee's performance in maintenance and surveillance during this

inspection period was good.

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DETAILS-

1 -. Persons Contacted ,

Commonwealth Edison Comnany (CECO)

• K. L. Kofton,-Station Manager ,

G. R. Masters, Project Manager

G. E. Groth, Production Superintendent

D. E. O'Brien, Technical' Superintendent

D. E. Cooper, Assistant Superintendent - Operations '

R. J. Legner, Services Director

A. O Antonio, Nuclear Quality Program Superintendent .

• R. Byers, Assistant Superintendent Work __ Planning

G. Vanderheyden,. Technical Staff Supervisor

S. Roth, Security Administrator

K. G. Bartes, Nuclear Safety Supervisor '

A. Haeger, Regulatory Assurance Supervisor

• J. Lewand, Regulatory Assurance

S. Hunsader, EQ Supervisor Design Support - Nuclear Engineering

K. C. Radke, Technical Staff System Engineer

• Denotes those attending the exit interview conducted on November 30,

1992.

The inspectors also interviewed several other licensee employees.

2. Licensee Action on Previously identified Items (92701. 927021

a. Ol en item

LClosed 50-456/92017-02(DRP): 50-48i7 /92017-02(DRPl: Failure to

Mlow Posting Requirements of a Radiologically Controlled Area.

Inspection Report 92017 details the f ailure of a Radiological

4 Protection Technician (RPT) to adhere to the posting requirement

to conduct 'a whole body frisk prior to exiting _ a radiologically

controlled area (RCA). In their followup review, the inspectors

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discovered that-two weeks prior to this incident, a RPT had failed

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to verify the decontamination of the 1A letdown heat exchanger

room before removing the posting. As a result, the RPT and one

- Electrical Maintenance Department person were contaminated when

they entered the room to replace light bulbs. At:ditionally, there -

has been one other incident since the open item was identified. In

this incident, _two Mechanical Maintenance Department personn_el

failed to adhere to the posted requirements-for entry into a RCA

and were subsequently contaminated. These failures to adhere to

the posted requirements for conducting work within a RCA are

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violations of Braidwood Technical Specification 6.11, " Radiation

Protection Program," as detailed in Braidwood Radiation Protection

Procedure 1110-3, " Radiological Postings, Labels, and Controls,"

(50-456/92023-01(DRP); 50-457/92023-01(DRP)).

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(Closed) ODen Item (456/88010-Ol(DRS):457/880ll-01(DRS)):

Adequacy of fire protection for several unprotected structural

steel columns and auxiliary steel attachments. The columns were

located in the fuel building between V and W at coordinates 17,

18, and 19, and the attachments were in the auxiliary building on

column P-21 at elevation 401'-0". The~ licensee was to provide the l

methodology used for selection and_ identification of the fire 1

protected structural steel components and the technical 1

justification that column P-21 met the specified fire rating. l

For the culumns, the licensee stated that because of the low fire

loading and the large open volume of the area, a credible fire

would pose no hazard to the structural steel columns, therefore,

fire proofing of the columns was not required. Justification was

provided in Sargent & Lundy Engineers letter dated May 20, 1988.

The columns support the slab at elevation 451' 0", a portion of

which carries a fire rating. The calculated fire loading for the

area, which includes an allowance for transient combustibles, is

5000 Stu/ft' (Fire Protection Report, Subsection 2.3.12.1). This

equates to a fire severity of under four minutes duration (NFPA-

Fire Protection Handbook, Chapter 9, Section 7). Therefore, a

credible fire would pose no hazard to the structural steel

columns.

For the auxiliary steel attachments, the licensee stated that- the

additional heat transfer into the fire protected column from the

unprotected auxiliary steel attachments did not degrade the fire

rating for column P-21 below the specified three hour rating.

Justification was supplied in Sargent & Lundy Engineers letter

dated May 20, 1988. The column was protected by a fire-proof

material, Pyrocrete 102 (7/8" thick), in accordance with

applicable installation drawings, which designated a three hour

fire rating according to Underwriters Laboratory (UL) Detail

X-719. The UL rating was based on tests conducted on a W10x49

column. P-21 was a W14x342 column, which had a cross section

seven times as massive as the VL tested column. The American Iron

and Steel Institute (AISI) had performed extensive research and

tests on a wide range of column sizes including sections which

were more massive than the UL tested W10x49 column. These tests

were summarized in AISI publication " Design Fire Protection for

i Steel Columns," Third Edition, March 1980, which indicated that

L the effective fire rating of the W14x342 column was more than

t twice that for the W10x49 column. Therefore, ample margin was

L provided to compensate for the additional heat input from a

i potential fire due to the unprotected auxiliary steel attachments.

Based upon the above, the inspectors concluded that the

methodology and technical justification provided were acceptable '

and the inspectors had no further concerns. This item is closed.

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b. Unresolved Items

(Closed) Unresolved Item (456/90Q19-02(DRSl: Physical separation

between fuel oil overflow, supply, and vent lines associated with

the opposite train emergency diesels. The licensee conducted a

detailed review'and analysis ~that determined the installed piping

arrangement, although, not consistent with the configuration

described in the fire hazard analysis-(FHA), posed no immediate

operational concerns for the opposite train diesel.. The FHA

stated that all equipment, cables, and piping in the diesel

generator room, the diesel oil day tank room, and the diesel oil-

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storage tank room would be associated with only one ESF division.

However, physical inspections revealed that some of the diesel oil

piping on each train was routed through the three rooms of the

opposite train.

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A detailed review of the Byron /Braidwood Fire Protection Report

(FPR) was made to identify all other inconsistences related to

physical separation of safety related equipment, cables, and

piping. This consisted of a zone-by-zone review of the FHA

(Section 2.3 of the FPR), a review of the safe shutdown analysis

(Section 2.4 of the FPR), and a review of. FPR Section 3.0, which

addressed conformance to the Standard Review Plan. No other

inconsistences from the FPR were identified. Minor changes were

completed to address the inconsistences as described in Sargent &

Lundy Engineers Letter dated November 29, 1990, and'Transmittcl

DIT-88-EXT-0124, " Assessment of Diesel Oil Piping Routed in

Opposite Train Diesel Generator Rooms" dated February 25, 1992.

The inspectors reviewed the licensee's detailed analysis and

discussed the results with NRR. The discussion concluded that the

licensee's analysis was acceptable and that Appendix R concerns

were adequately addressed, since offsite power would be available

and the diesel oil piping would remain intact during a postulated

fire. . Corrective actions taken by the licensee indicated prompt.

actions were performed including an expanded scope of review and

analysis which included the Byron Station, as-well-as, evaluating

the probability of missile affects on the opposite train diesel

piping. The inspectors noted, however, that during the

performance of two minor changes numerous field problem. reports

(FPRs) were generated relating to interference and clearance

. problems indicating that planning was not effective. The

inspectors had no further concerns and considered this item

closed.

c. Violation

(Closed) 50-456/92017-03(DRP): 50-457/92017-03(DRP): Technical

Specification 6.8.1 was violated when the licensee failed to-

convert Nuclear Work Requests to Temporary Alterations in

accordance with Braidwood Administrative Procedure 2321-18. 'The

licensee's response to-this violation was prompt and thorough.

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Actions taken appear adequate to preclude recurrence of this or

similar events. This item is closed.

One violation was identified.

3. Licensee Event Report (LER) Review (92700)

LER: were reviewed and closed based on the following criteria:

• Reportability requirements were met.

• Immediate corrective actions were accomplished.

• Corrective actions to prevent recurrence has been or will be

initiated per technical specifications.

No violations or deviations were identified.

(Closed) 457/01006: Reactor Trip Due to Valve Hispositioning. This

event is discussed, in detail, in Inspection Report 50-456/92020; 50-

457/92020, Paragraph 4. In addition to those corrective actions

previously identified, the inspector also notes the involvement of GN

Venture contractor personnel in the assessment of root cause

determination and corrective action. This item is closed.

4. Outanes (86700)

No violations or deviations were identified.

. On November 3,1992, Unit 1 Main Generator was synchronized to the grid,

ending the licensee's planned 66-day refueling outage seven days ahead

of schedule. In addition to completing ahead of schedule, the 185.0 Rem

of exposure and 88 personnel contaminations were well under the ALARA

goals of 208.5 Rem and 126 personnel contaminations established for the

outage. From a budgetary standpoint, initial estimates show the outage

to be $1.8 million under the approved budget. The outage was -

accomplished without major incidents and difficulties. The lessons-

learned must be carried forward into the refueling outage for Unit 2,

which begins in March 1993.

M 5. Radiation Protection (937011

Two violations were identified pertaining to radiological work

practices, performing safety analysis, and preparing written procedures

for chemistry related evolutions.

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Addition of sulfur hexafluoride (SF.).

• Evaluation of planning and implementation (SF.).

• Inspectors conclusion (SF ).

• Apparent violations (SF.) .

• November 6,1992, chloride excursion.

• Chloride excursion safety evaluation.

Addition of sulfur hexafluoride (SF.) causes unexpected variations in

steam generator chemistry. On November 6, 1992 the Braidwood Chemistry

Department inadvertently caused a chemical variation on the Unit 1 steam

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generators. Technicians injected a large amount, Labout 15 standard

cubic; feet, of sulfur hexafluoride (SF.) gas into the condensate system.

The injection of SF, was part of troubleshooting efforts on Unit 2--steam '!

generators. _ Probable leakage in the Unit' 2. steam generators resulted in .i

unexpected chemistry leve'is. Since the condensate system contains

several connections between the units, the licensee ;uspected that these i

connections were.the leakage source.

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Shortly after injecting SF., the technicians noticed unexpect' ed

responses in the Unit 1 steam generators' chemistry. A chemistry sample

confirmed that a large variation occurred. This required entry into the

action level of the station chemistry. procedures. The sample displayed

elevated cation conductivity, phosphates, fluorine, and sulfonates. -The

technicians determined that the SF, gas unexpectedly broke down into

sulfonates and fluorine.

The inspectors evaluated the planning and implementation of _the

troubleshooting efforts involving injection of SF ._ Previously, SF,

injection was used to find leaks in the condenser water boxes. The

success of this usage influenced the chemistry department to use SF, in

the condensate.

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However, they did not evaluate the possibility of SF, going in the steam

generators. They also did not evaluate the effects of the -steam

generator water chemistry on SF, . S F, gas is only slightly soluble in

water and soluble in alkaline solutions. Since water chemistry is-

alkaline, SF, broke down into fluorine and sulfonates. This condition

is not desirable since fluorine is corrosive to the heat transfer

surfaces and sulfonates is basic. The technicians injected SF, into

the condensate system without considering th'ese effects. They also

injected the SF, without a procedure.

The inspectors concluded that the chemistry department did not follow

several procedures and requirements before and during the injection of

the SF. gas. 10 CFR 50.59, " Changes, Tests,-and Experiments," requires .'

performance of an evaluation of tests or experiments not described in

the safety analysis report. The regulation also requires the . evaluation-

for changes in the facility. _This evaluation is to determine the

possibility of an u_nreviewed safety question or a change in the

1 technical specification.

The Braidwood safety a'nalysis report describes the methods for

naintaining water chemistry in the steam generators. The addition of

ammonia in the form of ammonium hydroxide, or an equivalent amine, and

hydrozine to the condensate is in Section 10.3.5.1. The addition of SF,

gas to_the condensate is not in-the safety analysis report.

TD regulation 10 CFR 50, Appendix B, Criterion V, " Instructions,

< Procedures, and Drawings,' requires that activities affecting quality

shall be done by documented procedures. These. procedures shall be of a

type appropriate to the circumstances of the activity and followed.

Additions *,f chemicals to the condensate affects the steam generator and

can affect the quality _of the heat transfer surface.

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The Braidwood Technical Specification, Section 6, " Administrative

Controls'," requires-the_ establishing _of procedures for activities

prescribed in Regulatory Guide 1.33,. Quality Assdrance Program

Requirements. The regulatory _ guide requires the establishing of

procedures for controlling water quality. These procedures _will contain

limits for concentrations of agents that are corrosive to heat transfer

surfaces. The addition of SF. into the ccndensate was performed without--

establishing a procedure.

The inspectors concluded that the following activities were violations

of NRC requirements. The injection of SF, without a procedure was an

apparent violation of Technical Specifications (50-456/92023-02(DRP);

50-457/92023-02(DRP)). The failure to perform an evaluation for

unreviewed safety condition is an apparent violation of 10 CFR 50.59

(50-456/92023-03(DRP); 50-457/92023-03(DRP)).

The inspectors _ reviewed the activities associated to the September 6,

1992, Unit 1 Steam Generator chemistry variation (chloride excursion).

The 10 CFR 50.59 Safety Evaluation of Unit 1 chloride excursion

concludes there is no unreviewed safety question. Inspection Report 50-

456/92020; 50-457/92020, details the chloride excursion experienced by

Unit.1 during its shutdown to commence a refueiing outage. As required

by Technical- Specification 3.4.7, a Fafety Evaluation was completed

which addressed the potential effect of this excursion on the Reactor

Coolant System (RCS) austenitic stainless steel, Alloy 600 materials,

and Zircaloy fuel cladding.

The 10 CFR 50.59 safety evaluation was completed by Westinghouse and

concluded there was no unreviewed safety question resulting from the

excursion. The technical basis for this conclusion was:

< a. Austenitic stainless steels are potentially susceptible to

chloride stress corrosion cracking (SCC) in aqueous solutions

under certain conditions. Oxygen is necessary for chloride

cracking in austenitic steels at temperatures below boiling. No

detectable oxygen was reported in the Unit 1 RCS during the

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excursion. Hydrogen peroxide was not.added to the RCS during the

excursion. Moreover, of the 47-hour duration of the-excursion,

the RCS average temperature was above 150 F-(the acceleration

. temperature for chloride SCC) for only. 35-hours Existing

Westinghouse test data for sensitized 304 stainless steel-

indicates that-the crack initiation time for chloride

concentrations of 390 ppb (the peak concentration seen during the-

excursion)'is on the order of 12-13 months, well beyond the-35-

hour exposure for-Unit 1. The test data was conservatively based

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on exposure in a fully aerated-chloride solution of 150 F. As

previously stated, no detectable oxygen was reported-in the RCS

during the excursion. Therefore, under the conditions for Unit-1,

it was judged that the elevated chloride concentration would not

have a negative impact on-the performance of the austenitic

stainless steel present in the RCS.

b. Alloy 600 materials have generally exhibited good resistance to

chloride induced cracking. Existing test data has demonstrated .

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that. chloride induced cracking 'is not an issue for Alloy 600

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material exposed to the type of environment found in the Unit 1 -

RCS during the excursion. Alloy 600 c-ring test specimens

stressed to 2/3 of yield strength, havebeen exposed to fully-

aerated' chloride solutions with different ciloride t __ concentrations -

(100-1000 ppm) at 150*F for periods of up to 12 months _ with no

-incidents.of cracking. The test data far exceeds the chloride

level and exposure-period that the Alloy 600 materials were

exposed to during the chloride excursion. Therefore, future

performance of Alloy 600 material will not be adversely affected

because of the excursion.

c. Westinghouse Zircaloy-4 material specifications limit the chloride

content to 20 ppm maximum, but certifications for the material

typically report that values are less than 10 ppm. During reactor

operation, the protective oxide that is formed on the Zircaloy-4

fuel components is considered to contair., due to diffusion-

effects, similar impurity levels to -the base material. Assuming

therefore, that the oxide film on the Unit 1 Zircaloy-4 fuel

components contained approximately 10 ppm chloride prior to the

coolant chloride excursion, and that all of the 390 ppb chloride

from the coolant was absorbed by the oxide 'Im, the total

, resultant chloride content of the oxide fih: would show an

increase of less than 1 ppm. The resulting c.loride content would

still be considerably below the material specification limit of

10 ppm maximum. Additionally, corrosion studies performed on

. Zirconium, where Zirconium at 650 F was placed in water .containing

200 ppm chlorine gas, have shown Zirconium's corrosion resistance

to be much less sensitive to impurities in the water than.to those

in the metal. Thus, no adverse effects on the Zircaloy fuel

cladding, fuel integrity and fuel handling operations are

expected.

6. Onerational Safety Verification (71707)

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The inspectors verified that the facility was being operated in

conformance with the licenses and regulatory requirements and that the

licensee's management control system was effectively cerying out its

responsibilities for safe operation. No violatiens or deviations were

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identified.

• Untested code safety valves.

• Inspectors' concerns.

o inspectors' review of Technical specifications.

• Determination of having proper lift setpoints.

• Unit 1 outage to investigate generator cooling system.

• Westinghouse _ recommendations for generator cooling syst.

• Unit I return to 100% reactor power.

Untested Main Steam Line Code Safety Valves raise questions regarding

the ability of Unit I to proceed with modo change to Mode 3. On

October 29,1992, Un.c 1 entered Mode 3 following completion of its

' refueling outage. During the outage, five Main Steam Line (MSL) Code

Safety valves .were modified or repaired. The Restart Onsite Review

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identified the need to test the lift setpoint pressures of these valves

at nominal operating pressure (N0P) and temperature (NOT), while in

Mode 3, to close the Unit 1 outstanding Nuclear Work Requests (NWRs).

In completing the checklist required before entry into Mode 3, the

Unit 1 Supervisor noted that four of the MSL Code Safety valves were

located on the same MSL. Based on the Unit Supervisor's interpretation

of Technical Specification (TS) 3.7.1.1, " Safety Valves," it was

questioned whether the mode change could be made with four untested

(inoperable) code safeties on one main steam line. Table 3.7-1 of the

IS only provides direction for continued power operation witn a maximum

of three inoperable code safety valves.

Through a series of conference calls, the licensee determined "a review

of NWRs associated with these valves provides a reasonable assurance of

croper lift setpoint." Therefore, the mode change could occur since the

valves, although not tested, were not ir. operable.

On October 30, 1992, with Unit 1 at NOP and NOT in Mode 3, the five MSL

Code Safety valves were tested per Braidwood TS Pr edure BwVS 7.1.1-1,

' Main Steam Safety Valves Operability Test." All five valves f ailed the

surveillance, were adjusted and retested satisfactorily.

In reviewing this event, the inspectors raised the following

questions / concerns:

a. Was the interpretation of the TS correct?

Based on the licensee's interpretation, the inspectors then questioned:

b. How was the conclusion that the untested MSL Code Safety valves

had a " reasonable assurance" of having proper lif t setpoints

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c. Why were the MSL Code Safety valves not considered inoperable and

an operability determination conducted per Braidwood

Administrative Procedure BwAP 330-10, " Operability Determination

of Safety-Related Equipment?"

In reviewing TS 3.7.1.1, che inspectors determined that the provisi'>ns

of TS 3.0.4 applied and the Mode 3 change with four untested /inopr:rable

code safety valves on the same MSL was not prohibited. AltE~igh .he TS

applies to Modes 1, 2, and 3, it is based on the operability of the MSL

Code Safety valves to ensure that secondary coolant system will be

limited to within 110% (1320 psia) of its design pressure of 1200 psia

during the most sever anticipated system operational trans:ent. The

maximum relieving capacity is associated with a turbine trip from 102%

rater thermal power coincident with an assumed loss of condenser heat

sink (i.e., no steam dumps to the condenser). Therefore, since the

ability to provide relieving capacity is not an issue in Mode 3, the

mode change could be made with all five MSL Code Safety valves, on any

single steam line, inoperable. To allow entry into Mode 3, where the

code safety valves can be tested and set, tha provisions of TS 3.0.4

(the mode change provision) are applicable to TS 3.7.1.1,

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The determination that the MSL code Safety valves had a reasonable

Assurance of having proper lift setpoints was, in the inspectors

opinion, weak. While a review of the NWRs showed the valves were

assembled correctly, they had not been bench tested. Additionally,

historical data and the engineering judgment of the Technical Staff and

Engineering and Nuclear Construction staff did not support the

conclusion reached. The inspectors questioned how the available

information and engineering judgment was weigFed in reaching a

determination of " reasonable assurance."

Regarding the issue of operability, the inspectors feel there can be no

question that the valves were clearly inoperable and required an

operability determination made per BwAP 330-10.

The inspectors have discussed these questions and concerns with the

licensee and will follow their resolution closely.

Unit I enters planned forced outage to investigate elevated Delta T on  !

the Generator Cooling system. On November 20, 1992, Unit 1 entered Mode

3 to commence a planned forced outage. The outage was required tc

investigate and repair the cause for the elevated temperature

differential (Delta T) of approximately 10 C between the stator-rator

cooling water inlet and outlet temperatures.

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A review of the refueling outage activities completed on the tator-

rotor cooling svstem did not identify a possible cause of the high '

Delta T. The licensee performed a fiber-optic inspection of the inlet

manifold, tock flow measurements of individual coils while conducting a

reverse flow flush on the stator, and inspected the coil hoses for

blockage. These efforts failed to identify a root cause for the

condition.

Based on Westinghouse recommendations, the alarm setpoints were raised

to 14 C for high Delta-T and 16oC for high-high Delta-T. Additionally,

Westinghouse performed a balance of the #5 turbine generator bearing.

The balancing successfully reduced the vibration form approximately 5.9

mils to approximately 1.1 mils. All other bearing vibrations are less 1

than 2.1 mils. The licensee also repaired various steam, water, and oil M

leaks which had developed since the unit was returned to service 3

following the refueling outage.

On November 24, 1992, the unit entered Mode 1 and is currently at 100%

uactor power. The licensee and the inspectors will continue to monitor

tne cooling system performance, the inspectors will evaluate the

licensee's corrective actions during a subsequent inspection based on

the system's continued performance.

7. Monthly Surveillance Observation G1726)

The inspectors observed several of the surveillance testing required by

technical specifications during the inspection period and verified that

testing was performed in accordance with adequate procedures, that test

instrumentation was calibrated, that results conformed with technical

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specifications and procedure requirements _and were reviewed, and that

any deficiencies identified during the testing were properly resolved.

No violations or deviations were identified.

The following surveillance activities were observed and reviewed: ,

• Eddy current inspection.

• 174 tubes were plugged.

• 21 tube plugs were replaced.

Braidwood Technical Surveillance 4.5.0-1, Steam Generator Eddy Current

Inspection. During Unit l's ret'ueling outage, a steam generator eddy

current inspection _ was performed on 100% of the tubes in all four steam

generators from the hot leg tube end through the U-bend. In addition, -

50% of the tubes were examined full length from tube end to tube end. A

total of 174 tubes were plugged due to indications at the hot leg

support plates and antivibration bar (AVB) wear.

In addition to the tubes plugged, a total of 21 Inconel 600 mechanical

tube plugs were replaced with Inconel 690 mechanical tube plugs. This

was accomplished in accordance with NRC Bulletin 89-01, " Failure of

Westinghouse SG Tube Mechanical Plugs."

8. Report Review

During the inspection period, the inspector reviewed the licensee's

Monthly Performance Report for October 1992. The inspector confirmed

that the information provided met the requirements of Technical

Specification 6.9.1.8 and Regulatory Guide 1.16.

The inspector also reviewed the licensee's Monthly Plant Status Report

for September 1992.

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No violations or deviations were identified.

9. Exit Interview (307031

The inspectors met with the licensee representatives-denoted in

Paragraph I during the inspection period and at the conclusion of the

inspection on November 30, 1992. The inspectors summarized the scope

and results of the inspection and discussed the likely content of this

inspection report. -The licensee acknowledged the information and did

not indicate that any of the information disclosed during the inspection

could be considered proprietary in nature.

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