IR 05000382/1986006

Insp Rept 50-382/86-06 on 860301-31.No Violation or Deviation Noted.One Unresolved Item Re Deficiencies in Cooling Water Valve Lineup List Identified.Major Areas Inspected:Followup on Potential Generic Problems

ML20197E439
Person / Time
Site:	Waterford
Issue date:	04/30/1986
From:	Bundy H, Constable G, Luehman J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20197E432	List: IR 05000382/1986006 ML20197E427
References
50-382-86-06, 50-382-86-6, NUDOCS 8605150258
Download: ML20197E439 (8)

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APPENDIX U. S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-382/86-06 License: NPF-38 Docket: 50-382 Licensee: Louisiana Power & Light Company (LP&L)

317 Baronne Street P. O. Box 60340 New Orleans, Louisiana 70160 Facility Name: Waterford Steam Electric Station, Unit 3 (W3 SES)

Inspection At: Taft, Louisiana Inspection Conducted: March 1-31, 1986 Inspectors-

'C (E Lueliman, Senior Resident Inspector Dat'e

=% w 9Y$o/k H. F. Bundyi Project Insp(ctor, Project Date Section C, Reac ~ ts Branch Approved: - o (

G. L 4Wnftable, Chief, Project Section C, D'at(

Reactor Projects Branch Inspection Summary Inspection Conducted March 1-31, 1986 (Report 50-382/86-06)

Areas Inspected: Routine, unannounced inspection of: (1) Followup on Potential Generic Problems, (2) Plant Status, (3) Licensee Event Report (LER) Followup, (4) Followup on Previously Identified Items, (5) Monthly Maintenance, (6) Monthly Surveillance, (7) ESF System Walkdown, (8) Routine Inspection, and (9) Inspection & Enforcement (I&E) Circular Results: Within the areas inspected, no violations or deviations were identified. One unresolved item was identified in paragraph 9 involving j deficiencies in a component cooling water valve lineup lis PDR ADOCK 05000382 G PDR ..

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DETAILS Persons Contacted ,

Principal Licensee Employees G. W. Muench, Acting Director, Nuclear Operations

• R. P. Barkhurst, Plant Manager, Nuclear T. F. Gerrets, Corporate QA Manager

• S. A. Alleman, Assistant Plant Manager, Plant Technical Services

• N. S. Carns, Assistant Plant Manager, Nuclear, Operations and Maintenance J. N. Woods, QC Manager A. S. Lockhart, Site Quality Manager R. F. Burski, Engineering and Nuclear Safety Manager K. L. Brewster, Onsite Licensing Engineer G. E. Wuller, Onsite Licensing Coordinator T. H. Smith, Maintenance Superintendent, Nuclear

• Present at exit interview In addition to the above personnel, the NRC inspectors held discussions with various operations, engineering, technical support, naintenance, and administrative members of the licensee's staf . Followup on Potential Generic Problems The NRC inspector made the licensee aware of problems involving the design adequacy of the auxiliary pressurizer spray system at Palo Verde Nuclear Generating System. At Palo Verde, the power supplies to certain valves were nonsafety grade, the auxiliary pressurizer spray system was lost during a loss of power tes No violations or deviations were identifie . Unresolved Items An unresolved item is a matter about which more infonnation is require'd to determine whether it is acceptable or may involve a violation or deviatio One unresolved item was identified during this inspection and is discussed in paragraph . Plant Status On March 7,1986', at 12:00 p.m. (CST)<the licensee commenced a functional test of the reactor power cutback system (RPCS). With the plant at full power, one of the two main feedwater pumps was secured. As expected, the

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RPCS dropped a control element assembly (CEA) group to reduce reactor l power. The combination of an automatic turbine runback and steam bypass

valve actuation stabilized the secondary plant. Following the successful completion of the test, the reactor was shut down for a maintenance outage

with cold shutdown being reached at 11

26 p.m., March 8, 198 .

On March 22, 1986, after completing the 2 week outage, the licensee placed I

the plant in Mode 4. Because of electrical problems with two control

element assembly drive systems, the plant was again placed in Mode l After the required repairs were made, the plant again entered Mode 4 on i

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March 27, 1986, the reactor was taken critical on March 30, 1986, and synchronized with the grid at 7:20 p.m. the same da No violations or deviations were identifie i Licensee Event Report (LER) Followup

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The following LERs were reviewed and close The NRC inspectors verified

! that reporting requirements had been met, that causes had been identified,

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that corrective actions appeared appropriate, that generic applicability

had been considered, and that the LER forms were complete. Additionally,

the NRC inspectors confirmed that no unreviewed safety questions were

involved and that violations of regulations or

Technical Specification (TS) conditions had been identified.

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(Closed) LER 382/85-21 - Automatic Actuation of Reactor Protective System.

The NRC inspector verified that Station Modification SMP-464 has been completed. This modification installed a time delay in the main feedwater pump low suction trip logic which should prevent pump trips due to momentary perturbations in the condensate and condensate polisher system Because of similar events reported in LERs 85-10 and 85-20, the licensee is also pursuing the corrective actions which are detailed in those i reports. The actions include procedure changes as well as a modification

of the vibration trip for the feedwater pump ! (Closed) LER 382/85-48 - Control Room Isolation. This event is one of

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several similar events the licensee has reported. Sometimes during normal planned plant stack releases the wind direction is such that sogie of the j released gases are directed into the control room ventilation intake.

Although such events are not desirable, the amount of gas is small and the

events infrequent. The licensee has been examining what changes, if any,

! can be made to correct the problem. However, location of permanent plant j structures seems cc play a major role and these cannot easily be change ,

No violations or deviations were identifie . Followup on Previously Identified Items I (0 pen) Violation 382/8520-04 - The NRC inspector has reviewed the licensee's response as well as the Region IV reply to that lette Because of problems with both the control room ventilation ammonia and

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t-4-chlorine monitors, the licensee continues to operate t,he control room ventilation in the recirculation mode for extended periods of time. This configuration is permitted for an indefinite amount of time by both Technical Specifications (TS) 3.3.3.7.1 and 3.3.3.7.2. When operating in this configuration, the control room emergency outside air intakes are kept shut for toxic gas considerations by the problem ammonia and chlorine monitors, thereby preventing the automatic pressurization of the control room with the ventilation in the high radiation mode of operation. This being the case, the licensee has analyzed t%e consequence of dispatching an operator to manually open the emergency' air intakes, if require Though this task has been analyzed and can safely be accomplished, the licensee has not proceduralized the evolutio If the licensee is forced to continue to ' operate in control room recirculation for long periods of time, procedures should be updated to ensure the emergency air intakes are cpened manually if a toxic gas sipal prevents immediate automatic switchover. The Office of Nuclear Reactor 3 Regulation (NRR) is currently evaluating control room habitability including these normal, recirculation and pressurized modes of operatio In reply to the licensee's response to this violation, W Region IV suggested the licensee review the process used to perform 10 CFR t

Part 50.59 reviews. The licensee responded by performing various QA audits of different aspects of the proces Observation (c) of Audit SA-W3-QA-85-58A indicated that no formal 10 CFR Part 50.59 evaluation training has been given to those individuals tasked with makinq

'; or approving the reviews. The NRC inspector discussed this observation with the assistant plant manager for technical services. The need to provide such training is being evaluate ~'

(Closed) Open Item 50-382/8602-02 - This item involved correction of deficiencies identified in the startup report by the NRC inspector The licensee submitted the required changes to the NRC Regional Administrator s

on March 14, 1986, via Cover Letter W3P86-004 (Closed) Deviation EA 85-10 - This deviation, involving failure to complete certain hydrostatic tests in accordance with FSAR commitments, was included in an enforcement package. The licensee responded individually to it by letter on June 21, 1985, and Region IV accepted this response on October 8, 198 The NRC inspector verified that all actions had been satisfactorily completed with the exception of issuance of the FSAR change It is closed on the basis that the FSAR changes have been made and placed in the suspense fil No violations or deviations were identifie . Monthly Maintenance Station maintenance activities affecting safety-related systems and, ,

components were observed / reviewed to ascertain that the activities were

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-5-conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with T The NRC inspector observed a portion of the work done on Valves SI-4128 and SI-125B under Procedure ME-7-008, " Motor Operated Valves." In addition to verifying calibration of measuring and test equipment (M&TE)

used, the NRC inspector reviewed the procedure with the electrician to verify his familiarity with the work in progres No violations or deviations were identifie . Monthly Surveillance The NRC inspectors observed / reviewed TS required testing and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation (LCO) were met, and that any deficiencies identified were properly reviewed and resolve On the morning of March 7,1986, the A Emergency Diesel Generator (EDG)

tripped while loaded for Surveillance Test OP-903-068, " Emergency Diesel Generator Operability Verification." The NRC inspector verified that the requirements of Technical Specification (TS) 3.8.1.1, Action a, were perfonned by observing the operability test of the EDG and by reviewing the verification done of the breaker alignment between the offsite transmission network and on,ite Closs IE distribution (0P-903-066).

The NRC inspector reviewed a number of the plant surveillance procedures including OP-903-069, Revision 3, " Integrated EDG/ Engineered Safety Features." Step 7.5.1.1 of that procedure requires that the EDG be run loaded for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the load for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 4840 KW. This portion of the surveillance procedure is conducted to meet the requirements of Technical Specification (TS) Requirement 4.8.1.1.2.d.7. W3 SES Final Safety Analysis Report (FSAR) states, in paragraph 8.3.1.1.2.13.k, that 4840 KW is the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> out of 24 supplementary power rating with 4400 KW being the continuous power rating. The NRC inspector brought to the licensee's attention that the surveillance procedure has no upper bound on EDG power output when conducting the surveillance. Potentially, the EDG could be run in an overload condition while performing the test. The EDG power and supplementary power ratings illustrate that there is significant drop in allowable time with only a 10 percent increased load. The NRC inspector suggested that LP&L consider how far above 4840 KW the EDG can be safely run and for how long and whether a change of the EDG surveillance procedure to incorporate an upper limit on EDG output is warrante No violations or deviations were identifie l l

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-6- ESF System Walkdown On March 18-21, 1936, the NRC inspectors reviewed the component cooling water system standby valve lineup (OP-2-003, Revision 3, Attachment 8.1), ,

reviewed the applicable plant drawings (LOU-1564 G-160, Sheets 1-3) and performed a walkdown of selected essential and accessible portions of the system to verify operabilit At the completion of the system inspection, the inspectors had the following observations: OP-2-003, Revision 3, Attachment 8.1, has many incorrect drawing j sheet and grid number reference The procedure contains numerous typographical and other nomenclature

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error A number of valves are listed to be checked more than once on the l valve lineu '

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j Tne following valves are not on the valve lineup, but are on the

drawings

CC-309A CC-309B CC-4148 CC-613 CC-937B CC-1791B CC-1891A CC-1891B CC-1981B CC-2001 CC-5351A CC-5251B CC-6012 CC-6013 CC-8021 CC-8061 CC-8062 CC-8063 CC-9011B CC-80310B CC-80313B It appears many of these valves have been added to the drawing by station modifications (SMs). CC-309A and B do appear on the monthly lineup check, OP-903-049, Attachment 1 The following valves are on the valve lineup but appear not to be on the drawings:

CC-199B CC-4178 CC-560 CC-7201 CC-8036A CC-8038A CC-8039A CC-8039B All main flow path valves were correctly identified in the valve lineup and found correctly positioned during the walkdow The above deficiencies are ihntified as an unresolved item (URI)

50-382/8605-0 No violations or deviations were identifie . Routine Inspection By observation during the inspection period, the NRC inspectors verified that the control room manning requirements were being met. In addition,

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j -7-the NRC inspectors observed shift turnover to verify that continuity of system status was maintained. The NRC inspectors periodically questioned shift personnel relative to their awareness of the plant condition Through log review and plant tours, the NRC inspectors verified compliance with selected TS and limiting conditions for operation During the course of the inspection observations relative to protected and vital area security were made including access controls, boundary integrity, search, escort, and badgin On a regular basis, radiation work permits (RWP) were reviewed and the specific work activity was monitored to assure the activities were being conducted per the RWP Selected radiation protection instruments were periodically checked and equipment operability and calibration frequency were verifie The NRC inspectors kept informed on a daily basis of overall status of plant and of any significant safety matter related to plant operations.

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Discussions were held with plant management and various members of the operations staff on a regular basi Selected portions of operating logs and data sheets were reviewed dail The NRC inspectors conducted various plant tours and made frequent visits of the control room. Observations included: witnessing work activities

in progress; verifying the status of operating and standby safety systems

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and equipment; confirming valve positions, instrument and recorder readings, annunciator alarms; and housekeepin Additionally, during the maintenance outage, the NRC inspector made a tour of the reactor containment. During this tour, area cleanliness and work in progress were observed. A brief inspection was made of the pressure boundary closure studs on one of the reactor coolant pumps (RCPs) to assess the extent of boric acid buildup due to possible primary fluid leakage from gasket joints on the RCP. Some boric acid buildup was observed but it was judged not to be excessive. The 2B RCP has apparently had an appreciable amount of leakage at the seal cooling heat exchanger and consequently Combustion Engineering, Inc. (CE) recommended inspection of that gasket joint during the maintenance outage followed by inspection of all bolted closures on the seal cooling heat exchangers and seal cartridge joints auring refueling outages. Further, CE recommended that, as long term corrective actions, the present stainless steel gaskets be replaced with Inconel/Grafoil gaskets and that the present carbon steel closure studs be replaced with studs made of 17-4 precipitation hardened (PH) stainless steel. The new gaskets are felt to be better because they have a greater springback capability and are more likely to remain leak tight in an area that has some differential in expansion / contraction during heatup/cooldown. The 17-4 PH stainless steel

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studs and those made of Inconel 718 are superior to the present studs because increased resistance to boric acid wattage. The 17-4 PH stainless steel studs are recommended over the ones made of Inconel 718 chiefly because of the smaller probability of galling problems.

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The licensee has inspected all the RCPs and found the smaller diameter studs (which are stainless steel), such as those on the control bleedoff line, to be in good condition. The amount of boric acid buildup around

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the nuts on the larger carbon steel studs was found to be acceptable. The

! licensee presently has plans to replace the gaskets and carbon steel studs beginning the first refueling outage.

No violations or deviations were identifie . Inspection & Enforcement (IE) Circulars

The licensee's actions on the following IE Circular has been reviewed and the IE Circular is considered closed because it is not applicable to

J W3 SES.

j (Closed) IEC 78-14 - HPCI Turbine Reversing Chamber Hold Down Bolting.

i No violations or deviations were identifie . Exit Interview i

The inspection scope and findings were summarized on March 31, 1986, with

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those persons indicated in paragraph 1 above. The licensee acknowledged the NRC inspectors findings. The licensee did not identify as proprietary any of the material provided to or reviewed by the NRC inspectors during this inspection.

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