IR 05000498/1987055

Insp Repts 50-498/87-55 & 50-499/87-55 on 870901-30.No Violations or Deviations Noted.Major Areas Inspected: Licensee Action on Previous Insp Findings,Nrc Bulletins, Incore Instrumentation Sys,Unit 2 Test Program & Site Tours

ML20237A226
Person / Time
Site:	South Texas
Issue date:	11/04/1987
From:	Carpenter D, Constable G, Hildebrand E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20237A222	List: IR 05000498/1987055 ML20237A221
References
50-498-87-55, 50-499-87-55, IEB-85-001, IEB-85-1, NUDOCS 8712140345
Download: ML20237A226 (11)

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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-498/87-55 License: NPF-71 50-499/87-55 Construction Permit: CPPR-129 Expiration Date: December 1989 Dockets: 50-498 50-499 Licensee: Houston Lighting & Power Company (HL&P)

P. O. Box 1700 Houston, Texas 77001 Facility Name: South Texas Project, Units 1 and 2 (STP)

Inspection At: STP, Matagorda County, Texas Inspection Conducted: September 1-30, 1987 Inspectors: .

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[ D. W Chrfenter, Senior Resident Inspector

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Project Section D, Division of Reactor Projects

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. Mildebrand, Resident Inspector, Project Date '

Section D, Division of Reactor Projects Accompanying Personnel: J. P. Clausner, French Commissariat A L'Energie Atomique, Institut De Protection Et De Surete Nucleaire

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Approved: _- _ /////[((7 G N ~ Con' stable, Chief, Project Section D Date Division of Reactor Projects 8712140345 871209 PDR G ADOCK 05000498 PDR

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Inspection Summary {

Inspection Conducted September 1-30, 1987 (Report 50-498/87-55; 50-499/87-55)

Areas Inspected: Routine, unannounced inspection including license action on previous inspection findings, NRC Bulletins, incore instrumentation system, status of incomplete preoperational test, Unit.2 test program, component cooling water (CCW) heat exchangers, power operated relief block valves, station accountability drill, allegation followup, initial control rod drop testing and rapid refueling demonstration, pressure code safety valves, and site tour Results: Within the areas inspected, no violations or deviations were l identifie i

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3 DETAILS Persons Contacted

• L. Kern, Manager Nuclear Security

• D. F. Bednarczyk, Nuclear Assurance

• M. A. Ludwig, Manager NPOD Maintenance

• G. L. Jarvela, Manager NP00 Technical Support

• G. L. Parkey, Manager Plant Engineering

• L. G. George, Nuclear Security

• W. H. Kinsey, Plant Manager

• A. McBurnett, Manager Site Licensing In addition to the above, the NRC inspectors also held discussions with various licensee, architect engineer (AE), constructor, and other contractor personnel during this inspectio * Denotes those individuals attending the exit interview conducted on October 2, 198 . Licensee Action on Previous Inspection Findings (Closed) Open Item 498/0708-60 This open item concerned several inadequacies with the Emergency Operating Procedures (E0Ps). The subject E0Ps have been revised and all procedures noted as requiring action by this open item have been corrected and reviewed by_the NRC inspector. This open item is considered close (Closed) Open Item 498/8708-62 This open item concerned several discrepancies from the requirements of the E0Ps writers guide. Four E0Ps had been reviewed and several problems were noted resulting in this open ite The licensee has taken action to revise the E0Ps to conform with the requirements of the writers guide and the subject E0Ps have been reviewed by the NRC inspector. The requirements and intent of the writers guide has been discussed with the licensee's representative and this open item is considered close (Closed) Open Item 498/8723-03 This open item addresses a concern that a procedure did not exist for checking pump and valve indication at the remote shutdown panel. The licensee has prepared and approved Station Procedure 1 PSP 03-ZG-0005, l

" Remote Shutdown System Operability Test-Cold Shutdown." This procedure ;

has been reviewed and adequately checks pump and valve indication at the remote shutdown panel. This open item is considered close . .

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(Closed) Open Item 498/8737-01 Station Procedures 1 PEP 04-ZG-0008, " Thermal Expansion Monitoring Test,"

and 1 PEP 04-ZG-0010, " Vibration Monitoring Test," were not available for review during the inspection period April 13 through June 6, 198 The licensee has prepared and approved both procedures. Both procedures have been reviewed by the NRC inspector and are satisfactory. This open item is considered close (Closed) Open Item 498/8750-02 This item addressed internally cracked injector nozzle tips discovered in the Unit 1 Standby. Diesel Generators (SDGs) and a cracked cylinder head on SDG 13. The licensee has conducted a complete inspection of all fuel }

injector nozzle tips for all Unit 1 and 2 SDGs. The cracked nozzle tips were traced to Bendix Lot No. 001124. A total of 176 injector nozzles were manufactured from this lot number in 197 The licensee's inspection revealed that 41 nozzles from this lot number were installed in Unit 1 engines. Of the 41 nozzles, 20 exhibited internal cracks. All 41 nozzles have been replaced. No nozzles from this lot number were installed in Unit 2 engines and no cracked nozzles were discovered in Unit 2. No engine failures have occurred at STP due to this type of cracked injector nozzle tip. These events occurred prior to receipt of an-operating license for Unit 1; therefore, the licensee has issued a 10 CFR 50.55(e) report (dated October 23,1987) for both Units 1 and 2. In addition, a 10 CFR Part 21 report was issue The cracked cylinder head from SDG 13 was sent to Cooper-Bessemer for analysis. Cooper-Bessemer reports that cracks such as this occasionally occur due to stresses incurred during the casting process at the foundr Cooper-Bessemer considers the crack a nuisance proble The cracked head did not allow fluid leakage and did not cause an engine failur The licensee's actions upon discovery of these problems was thorough and professional. The cracked cylinder head was evaluated and determined to be not reportabl The NRC inspector has monitored the licensee's actions to resolve this problem and finds them acceptable. This open item is considered close . NRC Bulletins (Closed) Bulletin 85-01, Steam Binding of Auxiliary Feedwater Pumps NRC Inspection Report 50-498;499/87-23, paragraph 9, documented the closure status of NRC Bulletin 85-01 with the exception of procedure approval and alarm point verification. The NRC inspector has reviewed the approved Plant Procedure IP0P04-AF-0001, " Auxiliary Feedwater Discharge Header High Temperature," Revision 0, and found it acceptabl Additionally, the NRC inspector verified that Alarm Points AFTA-7523,

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AFTA-7524, AFTA-7525, and AFTA-7526 do have set points of 200 F on the Emergency Response Facility Data Acquisition and Display System. This item is considered close . Incore Instrumentation System Ouring the performance of Plant Procedure 1 PEP 04-ZL-0021, "Incore Movable Detector System Checkout," anomalies were noted during the measurement of the top of core setpoint by running the dummy cable to the end of the thimble guide tube. Only 20 percent of the paths were in accordance with the criterion (three consecutive measurements with a tolerance of 0.2 inches). The licensee initiated a complete cleaning of all of the thimble guide tubes. After cleaning, about 60 percent of the measurements were in accordance with the criterio After reperformance of the procedure, the status was as follows:

Detector No. of Bad Thimbles Thimble No A 5 2, 3, 4, 5, 10 B 5 6 through 10 C 5 2, 3, 4, 5, 10

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E 4 2,3,6,7 l F 2 6, 7

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l The licensee suspects the design of the thimble guide tubes from the drive l unit to the vessel is the cause of this problem. Several bends in the tubing (horizontally and vertically) between the drive units to the vessel increase the friction and may disrupt the progression of the cable near the top end of the thimble guide tub After initial performance of Plant Procedure 1 PEPO 4-ZL-0021, the grease

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was removed from the drive unit slip clutches under the direction of a I

vendor representativ The slip clutch adjustment set point was then reverified in accordance with Plant Procedure 1 PEP 04-ZL-0021, Addendum The licensee obtained four readings with the digital force gauge for each drive unit to be sure the slip clutch adjustment was within toleranc Work on the system is still in progress (approximately 65 percent complete) and is being monitored by the resident inspector No violations or deviations were identifie f 8

5. Status of Incomplete Preoperational Tests ,

The following is an update of Unit 1 preoperational test completion status during this inspection period. These test procedures were identified as being incomplete in NRC Inspection Report 50-498;499/87-47, paragraph 3 and updated in NRC Inspection Report 50-498;499/87-50, paragraph .

Test IR 50-498/87-50 Current (September 30,1987)

1-BR-P-01 75% 99% field complete 1-CN-P-01 100% 100%

1-EW-P-05 -

100% retest required 100% retest required 1-FW-P-01 66% 99% field complete 1-HB-P-01 99% 99%

1-NK-P-01 0% 60%

1-PS-P-01 99% 99%

1-RA-P-04 100% 100% '

1-RA-P-18 100% 100%

1-HC-A-02 100% 100%

1-RC-P-06 99% 99%

1-RC-P-07 99% 99%

1-RC-P-08 99% 99%

1-RC-P-11 98% 98%

1-SP-P-03 95% 95%

1-WL- P-02 0% 0%

1-WL-P-03 99% 99% field complete 1-WS-P-01 0% 0%

1-WS-P-02 75% 99% field complete 1-CV-A-01 100% 100% ,

1-CV-A-01 75% 100%

1-FW-A-01 90% 99% field complete 1-LA-A-02 30% 50%

1-LA-A-04 100% 100%

1-LA-A-05 90% 100%

1-LA-A-06 100% 100%

1-LA-A-08 75% 100%

Some of these tests are restrained by plant conditions. The required plant operating mode must be established before testing and data collection / reduction can be completed. The NRC inspectors are monitoring the licensee's specific activities on selected tests onl No violations or deviations were identifie . Unit 2 Test Program STP Unit 2 construction is about 81 percent complet Nuclear Steam -

Supply System (NSSS) systems have almost completely been turned over from construction to startup (about 95 percent turned over). Many systems are in the flush phase and preparations are being made to flush the reactor coolant system into the reactor vesse _ _ _ _ _

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Many preoperational tests have been written and are available for NRC review. Prerequisite testing is in progress on many systems. Most NSSS pump runs have been completely satisfactory. The first preoperational testing is scheduled to commence in October 1987 on the reactor makeup water _ syste Work is in progress on the CCW heat exchangers to repair tube failures as described in Section 7 belo No violations or deviations were identifie . CCW Heat Exchangers STP uses' CCW heat exchangers manufactured by Strnthers Wells Cc. in Warren, Penn.; size No.66-588, type TENA-CEN. They are single pass, straight titanium tube design with essential cooling water (ECW) on the tube side and CCW on the shell side. The CCW is the higher pressure flui Two of the six heat exchangers have exhibited tube failures. The Unit 1

"A" Train heat exchanger has two ruptured tubes and further inspection by eddy current testing and boroscoping revealed several other tubes with'

thinning wall The failed and deteriorating tubes are all centralized in the areas above and below the CCW inlet shell side impingement plate. A total of 74 tubes have been plugged. The heat exchanger was leak checked but still extIbited leakage. Inspection revealed that 20 of the 74 plugged tubes still lea Repairing of the three Unit I heat exchangers is scheduled to be completed before going to Mode The Unit 2 "C" Train heat exchanger has had two tube failures similar in scope and location as the Unit 1 "A" Train heat exchanger. The Unit 2 heat exchangers have had only limited use for flushing and testing. The licensee has experienced difficulties in providing an adequate solution to this problem. Maintenance is still in progress. The NRC inspectors have been monitoring the licensee's actions during the identification, evaluation and repair activities. Initially, the licensee had difficulty in scoping the extent of the problem and in initiating an acceptable repair plan. This was evidenced by improper removal of failed tubes in the Unit 1 Train A het exchanger. This action required the welding of plugs in the tube sheet). This welding was in a difficult location and with nonstandard material (aluminum-bronze). The licensee undertook this welding without qualifying the welder on a mockup prior to the repair (the normal industry practice). This led to a degradation of the tube sheet area. Further, the licensee's corrective actions were slow in developing, partially due to an apparent lack of management direction during the early stages of the investigation and repair. At the end of ,

this inspection period, repair and retesting activities were progressing acceptably. The NRC inspectors will continue to monitor the licensee's actions on this issu __

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No violations or deviations were identifie . Power Operated Relief Block Valves During a joint tour of the Reactor Containment Building (RCB) by the STP l plant manager and NRC resident inspector, it was noted that the power operated relief block valves exhibited body-to-bonet leakage. The reactor coolantsystem(RCS)wasbeingmaintainedatlowpressure(about390psig)

to support required plant conditions for control rod drop test <,n Further investigation by the licensee revealed that the body-to-connet fasteners were torqued to below their specified value The licensee'immediately initiated an inspection of both valves to determine if any additional problems existed. No additional problems were note The "A" Train valve has been cleaned, reassembled with a new gasket, and torqued to the proper value. The "B" Train valve has been left disassembled to provide o vent path for the RCS required by current plant conditions. The reason for inadequate fastener torque is not known at this time and is still being investigated by the licensee. Review of the licensee's investigation findings on this inadequate fastener torque application is considered an open item (498/8755-01).

No violations or deviations were identifie i 9. Station Accountability Drill On September 18, 1987, an accountability drill was held at STP to demonstrate that station personnel have the capability to account for or identify as unaccounted, all station personnel by name and badge number within a 30-minute period. The Emergency Response Organization (ERO) was activated followed by a simulated Site Area Emergency with all nonessential personnel ordered to exit the protected area. All ERO accountability sheets were received at the east gate house and a manual verification completed in 27 minutes., There were 11 individuals unaccounted for. Their names and badge numbers were available for search and rescue, if required. Of the 11, 7 were accounted for on a recheck of the accountability sheets. The drill included an alert and assembly of the Unit 2 personne The overall exercise would have been successful except that after the drill was completed, it was noted by the licensee that a group of fire watchers, which remained in the plant for industrial safety considerations, were not counted and, therefore, should have been listed as unaccounted for. The license considered the drill unsuccessful based on this information and will conduct another drill the first week in Octobe The NRC inspector made an observation that the public address system could not be heard in the corridor of the main plant entrance between the .

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security doors at the 35-foot level and the first fire door at the 45-foot level of the Mechanical Auxiliary Building. This is the main entrance and exit for the power block personnel and, as such, should not have a large dead spot in the public address system. Resolution of this public address system dead spot is considered an open item (498/8755-02).

No violations or deviations were identifie .10. Allegation Case 4-87-A-058 (Technically 'losed)

A letter stated that in 1985 an alleger supplied, to the then resident '

inspector, information about a meeting between a top executive of bl&P and South Texas Nuclear Plant (STNP) Quality Control (QC) inspectors wherein the top executive made remarks that were interpreted by the inspectors as intimidating and designed to discourage vigorous inspections. The alleger states that there was never any documentation of the outcome of the investigatio During a telephone conversation, the alleger indicated that further information on the statement in the letter could be gained by referring to the Operating License (0L) Hearing transcript. This appears in the transcript of the hearing (beginning at Page 15132) on Tuesday, August 13, 198 Review of the transcript indicates that a above information was apparently(given inspector to the, now deceased, former NRC senior residentSRI) at STP d December 1983. NRC Inspection Report 50-498/83-25 indicates that this meeting took place on December 1,1983, rather than 1985 as the alleger states in his letter. The transcript also indicates that the former SRI did look into the issue and subsequently informed the alleger of the results but that he did not document these findings in an inspection repor The allegation is substantiated in that the NRC inspectors review of the allegation was not documented. Current NRC policy requires that all allegations be properly documented. The NRC inspector informed the Chief, Project Section D, while at the hearing in 1985, that he did not think the issues were an allegation, but rather he thought the alleger was providing helpful information. NRC guidance to inspectors in 1983 was not clear on when to classify information supplied to an inspector as an allegatio The former SRI went on to say that the HL&P executive had, in fact, had the meeting referenced by the alleger. The SRI recalled that the HL&P executive had apparently informed the QC inspectors that "anyone that slows the project down will hit the gate" which could be taken to be intimidating. The NRC inspector related that he did not believe the comment was intended to cause the QC inspectors to do poor work but rather to encourage them to do their job properly. He based this view on discussions that he had with others present at the meeting.

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I 10 The HL&P executive was an Executive Vice President at that time according to an inspection report where he was identified as an attendee at one of the SRI's-exit interview No further action is planned on this matter since.the NRC. inspector l involved is now deceased and the HL&P executive retired in Fe)ruary 198 Some additional assurance that this issue has been resolved 'can be gained from the fact that SAFETEAM came into existence late in 1984 and their records indicate that in similar cases, where supervisors have harassed, or have appeared to have harassed QC inspectors, the utility has taken strong action to articulate its anti-harassment policy and appears to have

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No violations or deviations were identifie . I_nj61 Control Rod Drop Testing and Rapid Refueling Demonstration

The license has determined that performance of the scheduled rapid refueling demonstration would not provide stfficient added assurance of

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operability to justify the time that would be required to perform the l demonstration. All elements of the rapid refueling concept have been demonstrated separately except movement of the reactor vessel head with l all control rods latched in the refueling position. The STP Final Safety Analysis Report (FSAR) does not require this be demonstrate During this inspection period, control rod drop testing was started under Plant Procedure 1 PEP 04-ZL-0026, " Initial Rod Test." The testing was terminated when during the test two control rods failed to stay latched in the refueling position. Control Rod L-13 dropped once and H-14 dropped twice. The vendor recommended increasing the latch voltage from the minimum of 255 volts to 305 volts. Before testing could be restarted and the fix verified, plant conditions changed such that core reactivity alterations could not be made. When plant conditions allow, retesting uCer Plant Procedure IPEPO4-ZL-0026, " Initial Rod Test," and IPEP04-ZL-0024, " Rod Drop Time Measurement (Cold-Full Flow)," will proceed.

! The NRC inspectors will continue to monitor the control rod testing when testing resume No violations or deviations were identifie . Pressurizer Code Safety Valves On September 5 and 6,1987, the licensee identified several anomalies pertaining to the installation of the pressurizer code safety valves (IRC-PSV-3450, IRC-PSV-3451, and IRC-PSV-3452).

During RCS fill and vent with pressure at approximately 35 psig, it was noted that Valve IRC 'SV-3451 displayed an inlet flange lea _ - - - - _ _

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f Valve IRC-PSV-3452 exhibited a slight offset (cocking) of.its discharge flange. Additionally, all three safety valves had open threaded drain holes in the discharge side of the valve bodies which were not plugge The licensee initiated an investigation and prepared nonconformance reports (NCRs). Valve and piping configuration repairs have been ;

completed. Plugs have been installed in the open body drain port l l

'Although the licensee took prompt action on the flange leak issue, they

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initially failed to provide proper plugs or even temporary pipe caps for the threaded drain holes in the valve bodies. Several days after the NCRs were issued, the licensee provided temporary plugs after being encouraged by.the NRC inspector to provide some closure for these openings. A complete technical report of these deficiencies and the

repairs which have been made are contained in NRC Inspection Report 50-498/87-58.

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l No violations or deviations were identified.

i 13. Site Tours The NRC inspector conducted site tours of the following areas:

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Unit 1 - mechanical and electrical auxiliary building (MEAB), RCB, fuel handling building (FHB), emergency diesel generator building (EDGB), isolation valve cubicle (IVC), and the ECW intake structur l

. Unit 2 - MEAB, EDGB, turbine building, and RC Results of these tours have been discussed with licensee managemen No violations or deviations were identifie . Exit Interview I

The NRC inspectors met with licensee representatives (denoted in paragraph 1) on October 2,1987, and summarized the scope and findings of the inspection. Other meetings between NRC inspectors and licensee management were held periodically during the inspection to discuss identified concern _ ___ _