ML20151T790

Insp Repts 50-327/85-47 & 50-328/85-47 on 851206-860105. Violation Noted:Work Plan 11850 Inadequately Implemented in That Disassembly of Valves 2-VLV-67-786A & 786C Not Documented

ML20151T790
Person / Time
Site:	Sequoyah
Issue date:	01/31/1986
From:	Jenison K, Linda Watson, Weise S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20151T743	List: ML20151T737 ML20151T766 ML20151T790
References
50-327-85-47, 50-328-85-47, NUDOCS 8602100363
Download: ML20151T790 (10)

See also: IR 05000327/1985047

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jfdir UNITED STATES

g 'o NUCLEAR REGULATORY COMMISSION

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p REGloN il

y j 101 MARIETTA STREET, N.W.

• '- !e ATLANTA, GEORGI A 30323

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Report Nos.: 50-327/85-47, 50-328/85-47

Licensee: Tennessee Valley Authority

6N38 A Lookout Place

1101 Market Street

Chattanooga, TN 37402-2801

Docket Nos.: 50-327 and 50-328 License Nos.: DPR-77 and DPR-79

Facility Name: Sequoyah Units I and 2

Inspection Conducted: December 6 1985 thru January 5, 1985

Inspectors: I__I M S

Date Signed

K. M. Jenison, Seng Refident inspector

MA

L. J. Watson, Res en nspector

/N/%

Dat Si ned

' Approved by: 3/ N

Date 71gned

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S. P. Weise, Section Chief

Division of Reactor Projects

SUMMARY

Scope: This routine, announced inspection involved 276 resident inspector-hours

onsite in the areas of operational safety verification including operations

performance, system lineups, radiation protection, security and housekeeping

inspections; surveillance and maintenance observations; review of previous

inspection findings; followup of events; review of licensee identified items; and

review of inspector followup items.

Results: One violation for failure to follow procedure was identified with three

examples.

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PDR ADOCK 05000327

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REPORT DETAILS

, 1. Licensee Employees Contacted

H. L. Abercrombie, Site Director

• P. R. Wallace, Plant Manager

• L. M. Nobles, Operations and Engineering Superintendent

• B. M. Patterson,- Maintenance Superintendent

• J. M. Anthony, Operations Group Supervisor

• R. W. Olson, Modifications-Branch Manager

• M. R. Sedlacik, Electrical Section Manager, Modifications Branch

• H. D. Elkins, Instrument Maintenance Grouo Manager

• M.~ A. Scarzinski, Electrical Maintenance Supervisor

• M. R. Harding, Engineering Group Manager.

D. C. Craven, Quality Assurance Staff Supervisor

• D. L. Cowart, Quality Surveillance Supervisor

• D. E. Crawley, Health Physics Supervisor

• G. B. Kirk, Compliance Supervisor

M. L. Frye, Compliance Engineer

H. R. Rogers, Compliance Engineer

• R. C. Birchell, Compliance Engineer

• D. H. Tullis, Mechanical Maintenance Group Supervisor

J. H. Sullivan, Regulatory Engineering Supervisor

• R. J. Griffin, NSRS Sequoyah Site Representative

• C. L. Wilson, Nuclear Engineer, NSS

'T *E. W. Whitaker, Nuclear Licensing Engineer, TVA-NLB

• C. E. Chmielewski, Nuclear Engineer 2, NSS

• L. D. Alexander, Mechanical Section Manager, Modifications Branch

• W. Murhead, QA Specialist, QAB

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• E. Bosley, QA Evaluator, QAB

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Other licensee employees contacted included technicians, operators, shift

engineers, security' force members, engineers and maintenance personnel.

• Attended' exit interview

2. Exit Interview (30703)

The inspection scope and findings were summarized with,the Plant Manager and

members of his staff on January 8,1986. One violation hith three examples

described in paragraphs 5 and 7 were discussed. The licensee acknowledged

the inspection findings. The inspectors acknowledged that . the safety

significance of the violation examples was low, but pointed out that they

were indicative of either lack of attention to detail or lack of an

( understanding of the administrative control . requiremants for procedure

changes versus adherence. The licensee did not identify as proprietary any

of the material reviewed by the inspectors during this inspection. During

the reporting period, frequent discussions were held with the Site Director,

Plant Manager and other managers concerning inspection findings. At no time

during the inspection was written material providad to the licensee by the

inspector. I

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3. Licensee Action on Previous Inspection Findings (92702)

(Closed) Violation 327,328/85-17-01. Failure to Follow Procedure for

Installation.of Vital Battery V. The inspector reviewed MR A-545415 and MR

A-545416 and verified.that intercell and end spacers were installed with no

more than one quarter inch clearance.

4. Unresolved Items

Unresolved items are matters about which more information is required to

determine whether they are acceptable or may involve violations or

deviations. One unresolved item identified during this inspection is

discussed in paragraph 9.

5. Operational Safety Verification (71707)

a. Plant Tours

The inspectors observed control room operations, reviewed applicable

logs, conducted discussions with control room operators, observed shift

turnovers, and confirmed operability of instrumentation. The

inspectors verified the operability of selected emergency systems,

reviewed tagout records, verified compliance with Technical Specifi-

cation (TS) Limiting Conditions for Operation (LCO) and verified return

to service of affected components. The inspectors verified that

maintenance work orders had been submitted as required and that

followup activities and prioritization of work was accomplished by the

licensee.

Tours - of the diesel generator, auxiliary, control, and turbine

buildings and containment were conducted to observe plant equipment

conditions, including potential fire hazards, fluid leaks, and

excessive vibrations and plant housekeeping / cleanliness conditions.

During walkdowns of the Auxiliary Building and Turbine Buildings, two

cases were identified where pipe hangers were inadequate. One case

involved a pipe hanger on Safety Injection System piping that had

apparently been removed for valve maintenance during the current outage

and had not been replaced. The second case involved two missing nuts

on the pipe supports for feedwater bypass line on the Main Feedwater

System. Both of these items have been referred to NRC regional

specialist inspectors for followup.

The inspectors walked down accessible portions and/or reviewed control

room indications of the following safety-related systems on Unit I and

Unit 2 to verify operability and proper valve alignment:

Residual Heat Removal System (Units I and 2)

Charging Pump Flowpath (Units 1 and 2)

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Control Room Ventilation Chlorine Detection System (Common)

Auxiliary Building Gas Treatment System (Unit 1 and 2)

Auxiliary Control Air (Units 1 and 2)

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The inspectors conducted a walkdown to review Abnormal Operating

Instruction, AOI-27, Control Room Inaccessibility, Units 1 and 2. The

inspectors noted the following items:

(1) The reactor operator. providing information on the procedure was

well informed.in the use of the procedure and had participated in

annual drills of the procedure.

(2) The procedure stated that certain material (controlled copies)

were to be maintained in the Auxiliary Control Room ( ACR). These

materials were maintained in a locked cabinet outside the ACR.

The licensee committed to correct this administrative error in the

procedure. The list of material included telephone directories.

The inspectors noted that these directories were not up-to-date.

In addition, an administrative error consisting of the entry of a

procedure revision into the wrong procedure was noted. The

licensee stated that these items would be corrected. Followup on

corrective action for these administrative errors is identified as

Inspector Followup-Item (327,328/85-47-01).

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(3) A caution in the procedure, which would require an operator to

swap suction for the turbine driven Auxiliary Feedwater pumps from

the Condensate Storage Tank to the Essential Raw Cooling Water

System at 11.5 psig, was incorrect. The reactor operator stated

that the value had been changed from 11.5 to 13 psig. The inspec-

tors asked the licensee to provide additional information on

the methods used to update procedures when critical parameters

change. The inspector reviewed Administrative Instruction AI-19,

Plant Modifications, which requires the review of-plant procedures

when modifications .are made to the plant or critical parameters

are changed. For Operations procedures, the licensee stated ths.t

an SRO or RO reviewed procedures that he/she determined to be

affected. This omission appears to be an isolated case,

therefore, the item will be identified as Inspector Followup Item

(327,328/85-47-02) for additional NRC review of control of pro-

cedure revisions due to modifications and update of appropriate

procedures to reflect the correct setpoint discussed above.

No' violations or deviations were identified.

b. Security

During the course of the inspection, observations relative to protected

and vital area security were made, including access controls, boundary

integrity, search, escort, and badgin3 No violations or deviations

were identified.

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c. Radiation Protection

The inspectors observed Health Physics (HP) practices and verified

implementation of radiation protection control. On a regular basis,

radiation work permits (RWPs) were reviewed and specific work

activities were monitored to assure the activities were being conducted

in accordance with applicable RWPs. Selected radiation protection

instruments were verified operable and calibration frequencies were

reviewed.

On January 2, 1986, the inspectors observed three TVA employees clean-

ing borated water from the spill basin surrounding the boric acid ~

transfer pumps. RWP 02-0-86668 required the workers to wear a top and

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bottom plastic suit. Two of the three individuals failed to meet the

' RWP requirement .in that they had unfastened the top front portion of

the plastic suit to increase air circulation to their upper bodies.

Because of the nature of the work that was involved (bailing contami-

nated water with buckets) the above actions increased the potential of

skin contamination. In addition, the foreman who was in charge of the

cleanup activities, was standing just outside of the contaminated area

and was observing the workers. When questioned as to what the

requirements of the RWP were, he did not appear to know. Technical

Specification 6.11 states that procedures for personnel radiation

protection shall be prepared consistent with the requirements of

10 CFR 20 and shall be approved, maintained and adhered to for all

operations involving personnel radiation exposure. Sequoyah Nuclear

Plant Radiological Control Instruction RCI-14, Radiation Work Permit

(RWP) Program, requires'each individual to wear all required protective

clothing and equipment as required when entering an RWP area. Failure

to properly wear protective clothing required by an RWP is a Violation

(327/328-85-47-03).

On December 12, 1985, a sheetmetal worker was internally contaminated

while working on auxiliary building ventilation ductwork. He was

decontaminated by the SNP Health Physics staff and his body burden is

being evaluated. This item will be reviewed by an NRC Region II Health

Physics specialist during the next scheduled inspection (Inspection

Report 50-327, 328/85-04) and will be tracked as Inspector Followup

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Item (327, 328/85-47-04).

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6. i Monthly' Surveillance Observations (61726)

The inspectors observed Technical Specification (TS) required surveillance

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testing and verified that testing was performed in accordance with adequate

procedures; that test instrumentation was calibrated; that Limiting Condi-

tions for Operation were met; that test results met acceptance criteria

requirements and were reviewed by personnel other that the individual

directing the test; that deficiencies were identified, as appropriate, and

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that any deficiencies identified during the testing were properly reviewed

and resolved by management personnel; and that . system restoration was

adequate. - For . complete tests, the inspector verified that testing

frequencies were met and tests were performed by qualified individuals.

The inspector witnessed / reviewed portions of the following surveillance test

activities:

SI-2, Operations Shift Log .

Functional Testing of 1-FCV-63-22 in accordance with WP 11866

.SI-7, Electrical Power System: Diesel Generator

IMI-99, Reactor Protection System

No violations or deviations were identified.

7. Monthly Maintenance Observations (62703)

Station maintenance- activities of safety-related systems and components

were observed / reviewed to ascertain that they were conducted in accordance

with approved procedures, regulatory guides, industry codes and standards,

and in conformance with TS.

.The following items were considered during this review: LCOs met while

components or systems were removed from service, redundant components

operable, approvals obtained prior to initiating the work, activities

accomplished using approved procedures and inspected as applicable, pro-

cedures adequate to control the activity, troubleshooting activities

controlled and repair records accurately reflected what actually took place,

functional testing and/or calibrations performed prior to returning com-

ponents or systems to service, quality control records maintained, activ-

~1 ties accomplished by qualified personnel, parts and materials used properly

certified, radiological controls implemented, QC hold points established

where required and observed, fire prevention controls implemented, outside

contractor force activities controlled in accordance with the approved

Quality Assurance (QA) program, and housekeeping actively pursued.

The following maintenance activities were observed / reviewed by the

inspectors:

WP 11473 Installation of Fire Protection Piping to Correct Areas of

Adequate Spray

WP 11850 Torquing of ERCW Check Valves to Upper Containment Compart-

ment Coolers (2-VLV-67-786A, 786C)

.WP 11850 required that four check valves be installed in the ERCW lines

.to the upper compartment coolers. During the inspection of WP 11850, the

inspector noted that the' work plan required, "at completion of the check

valve installation, QC inspector shall verify that valve internals meet

Class D cleanliness requirements per TI-70."

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A QC inspector signature was found for valves 2-VLV-67-786B and -786D.

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The blanks had not been signed by a QC inspector for valves 2-VLV-67-786A

and -786C, for which torquing had just been completed. The WP had not been

followed in that a QC inspection had not been performed, however, Mainte-

nance Instruction MI 11.4, Maintenance of CSSC Valves, had been used by the

pipefitters to verify cleanliness.

The inspector discussed the problem with the cognizant engineer. The

cognizant engineer had been notified that the Work Plan had not been imple-

mented and determined that since the piping and valves were Class C that a

QC signoff was not required and therefore there was no reason to reverify

cleanliness. The engineer was processing an intent change to delete the QC

holdpoint.

The inspector reviewed Administrative Instructions AI-19, Plant Modifica-

tions, and AI-4, Plant Instructions - Document Control. The instructions

state that upon discovery of a discrepancy, the cognizant engineer will be

notified and will resolve the discrepancy. The engineer's actions in this

case appear to be appropriate. The inspector noted that the procedures do

not specifically address the corrective action to be taken or how it will

be documented. The inspector brought this to the attention of the licensee.

Failure to follow a procedure for modification to a safety-related system

-is a violation; however, since the failure to follow procedure was identi-

fied by the licensee and corrective action was being taken, this is con-

sidered licensee identified and will not be cited.

Also during the review of WP 11850, the inspector determined that data

sheets documenting the disassembly .of the valve had not been completed

as required by MI 11.4. Failure to follow the procedure for installation

of the ERCW check valves is identified as an example of Violation (327,

328/85-47-03).

MR A 563333 RHR Pump 2A-A Minimum Flow Valve 1-VLV-74-12 MOVATS.

WR 102907 Qualification Maintenance Data System maintenance on

Rosemount Pressure Transmitter 1-FT-3-147, Steam Generator

1. 3 Auxiliary Feedwater Inlet Flow Transmitter

MR A 593499 Replacement of NAMCO limit switch on Steam Generator

Blowdown Isolation Valve

Maintenance Instruction MI 10.37, NAMCO Limit Switches, was used during the

performance of Maintenance Request A-593499. Measuring and Test Equipment

(M&TE) prescribed in step 5.2 of MI 10.37 included a 0-30 inch pound torque

screwdriver. The torque screwdriver used in step 6.1.1.2 of MI 10.37 to

replace the bottom cover gasket was not a 0-30 inch pound torque screw-

driver, and a "non-intent" change was not made per AI-4 to reflect as such.

However, the value to which the cover was torqued appeared to be within the

accuracy range of the tool actually used. This is identified as a failure

to follow procedure and is a further example of Violation (327, 328/85-

47-03).

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8. Licensee Event-Report (LER) Followup (92700)

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The lfollowing LERs were reviewed and closed. The inspector verified that:

reporting requirements had been met; .causes had been identified; corrective

actions appeared appropriate; generic applicability had been considered; the

LER forms were complete; the licensee had reviewed the event; no unreviewed

safety questions were' involved; and violations of regulations or Technical

Specification conditions had been identified.

LERs Unit 1

327/83007 Auxiliary Feedwater Level Control Valve Inoperable

327/83009 . Control Rod Position ~ Indication Inoperable

327/83154 Waste Gas Decay Tanks With High Oxygen Concentration

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327/83167 Waste Gas Decay Tanks With High Oxygen Concentration

327/84019 Ice Condenser Ice Weights Below Design Requirements

327/84044 Diesel Generator Start During Surveillance Testing

327/84062 Control Room Isolation

327/84065 Auxiliary Building Isolation

LERs Unit 2

328/82115 Containment Personnel Air Lock Inoperable (Rev. 1)

328/83049 Inoperable Intermediate Ice Condenser Door (Rev. '1)

328/83093 Condenser Vacuum Flow Rate Monitor Inoperable (Rev. 1)

328/84017 Reactor Trip on Low-Low Steam Generator Level

328/84018 Three Reactor Trips on Low-Low Steam Generator Level

9. Event Followup (93702, 62703, 61726)

a. On December 9, 1985, with Unit 2 in Mode 5 at 127F and 150 psig,

instrument technicians were calibrating the Unit 2 reactor coolant loop

1. 3 wide range pressure transmitter, 2-PT-68-66. . Surveillance

Instruction, SI-484, Periodic Calibration of the Reactor Vessel Level

Instrumentation (RVLIS) and RCS Wide Range Pressure Channels (P-403,

P-406) (Refueling Outage), was being utilized.

Step 1.1 of Appendix A.1 of the procedure requires the removal of input

wires to PS-68-66B/E to~ prevent closing of valve 2-FCV-74-2, the

Residual Heat Removal (RHR) suction valve. This interlock protects

against overpressurization of the RHR System. During performance of

the calibration, -operators noted that the RHR suction valve,

2-FCV-74-2, started closing. The operator promptly removed the Train A

RHR pump from service and notified the instrument technicians of the

valve closure. The technicians re-terminated the input wires and

stopped the work. The system was restored to operation in 8 minutes.

There was no increase in RCS temperature. Maintenance Request (MR)

AS46561 was written to determine the source of the problem. The

reactor operators took RHR out of service for approximately 50 minutes

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during the troubleshooting work. The licensee simulated the event by

relifting the input wires and determined that when the wires were

lifted together that a circuit was created that caused a valve ~ closure-

signal. .The licensee is revising the procedure to prevent a similar-

. occurrence in the future. Notification of all-Westinghouse plants was

Lperformed through the Westinghouse news letter.

No1 violations or deviations were identified.

b. :0n December 13,-1985, it was . discovered that vendor axial flux W(z)

curve data for Unit 2 incore calculations was entered incorrectly into

the prime' computer code. This W(z) curve data had-been entered for use

in Unit 2 - Cycle 3 axial flux and local power peaking factor

calculations. After the error' was identified, preliminary data. was

analyzed and .it was determined by the licensee that no safety limits

were violated. The error was identified by the licensee during the

performance of Surveillance Instruction SI-126, Incore Calculation.

NRC review of whether or not a safety limit was violated is Unresolved'

Item.(327,328/85-47-05).

c. On December 18, 1985, the non-safety-related station air supply to a

door gasket, located in the barrier door separating the spent fuel pool

and the~ transfer canal, was lost. The deflated door began to leak

spent fuel pool. water into the transfer canal resulting in a two foot

' drop in spent fuel pool level in approximately forty minutes. Thr. leak

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was essentially stopped by repressurizing the door seal with safety

related air. Spent fuel pool level was subsequently returned to a

normal level of greater than twenty three feet above the irradiated

fuel. The practice of inflating the door with non-safety-related air

and operating with the transfer canal dry will be reviewed as Inspector

Followup Item (327, 328/85-47-06).

No violations or deviations were identified.

10. Inspector Followup Items (92701)

(Closed) IFI 327,328/85-16-05. Surveillance Testing for Diesel Generator

Batteries. The inspector reviewed the performance testing of the four

diesel generator batteries. The performance tests were conducted using

Surveillance Instruction SI-238.2, Diesel Generator Battery Capacity Test,

Units 1 and 2, for batteries IA-A, IB-B, 2A-A and 2B-B on October 18,

September 11, October 19, and September 13, 1985, respectively. The results

of these tests.were reviewed. One error was identified in the calculation

of the capacity of battery 1A-A. The capacity had been entered as 97% when i

the actual capacity indicated (with the chosen test parameters) was 91%.

This value would not have resulted in additional testing per IEEE 450-1980.

'It should be noted that the test was term 1nated at a conservative value of

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105 volts when IEEE 450-1980 would have allowed a discharge to 99 volts.

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In addition, the inspector reviewed SI 238.2 -against IEEE 450-1980. The

licensee has an internal commitment to conduct diesel generator _ battery

service and performance discharge testing in accordance with this standard.

.Two typographical errors involving a reference to an incorrect step number

, and an incorrect title for a key' were noted which did not affect the results

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of the tests. Also, one reference to a table in IEEE 450-1975 in lieu of

l ' the table in IEEE 450-1980 was noted. The cognizant eagineer stated that

the 1975 table was utilized to eliminate requirements for calculations. of

temperature corrected test voltage levels prior to the test. Instead, the

test results are corrected for temperature after the test using the method

in IEEE 450-1975.

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