ML20127E517

Insp Repts 50-325/85-05 & 50-324/85-05 on 850301-31. Violation Noted:Failure to Follow Surveillance Procedure PT 1.17PC Re APRM

ML20127E517
Person / Time
Site:	Brunswick
Issue date:	04/30/1985
From:	Fredrickson P, Garner L, Hicks T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20127E462	List: ML20127E459 ML20127E467 ML20127E517
References
50-324-85-05, 50-324-85-5, 50-325-85-05, 50-325-85-5, NUDOCS 8505200105
Download: ML20127E517 (9)

See also: IR 05000324/1985005

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UNITED S FATES

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NUCLEAR REGULATORY COMMISSION

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101 MARIETTA STREET, N.W.

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ATLANTA, GEORGI A 30323

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Report Nos.: 50-325/85-05 and 50-324/85-05

Licensee:

Carolina Power and Light Company

411 Fayetteville Street

Raleigh, NC 27602

Docket Nos.: 50-325 and 50-324

License Nos.: DPR-71 and DPR-62

Facility Name:

Brunswick 1 and 2

Inspection Conducted: March 1-31, 1985

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Inspectors:

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L.W. Garner ~,ActingSeniorResidentInsptor

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Date Signed

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Approved by:

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P. E. Fredrickson, Section Chief

Da'te Si'gned

Division of Reactor Projects

SUMMARY

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Scope: This routine safety inspection entailed 280 inspector-hours on site in

the areas of. surveillance, maintenance, operational safety verification, ESF

System walkdown, in-office and on-site Licensee Event Report review, independent

inspection and modification review.

Results:

One violation was identified in one area - " Failure To Follow

Surveillance Procedure PT 1.17PC" (paragraph 6).

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850506

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0505200105ADOCK

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REPORT DETAILS

1.

Persons Contacted

Licensee Employees

C. Blackmon, Superintendent - Operations

• L. Boyer, Director - Administrative Support

• J. Chase, Manager - Operations

• G. Cheatham, Manager - Environmental & Radiation Control

R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)

• C. Dietz, General Manager - Brunswick Nuclear Project

W. Dorman, QA - Supervisor

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• K. Enzor, Director - Regulatory Compliance

W. Hatcher, Security Specialist

• R. Helme, Director - Onsite Nuclear Safety - BSEP

• B. Hinkley, Manager - Technical Support

W. Hogle, Engineering Supervisor

J. Holder, Manager - Technical Support

P. Hopkins, Director - Training

• P. Howe, Vice President - Brunswick Nuclear Project

L. Jones, Director - QA/QC

' R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)

J. Moyer, I&C/ Electrical Maintenance Supervisor (Unit 1)

D. Novotny, Senior Regulatory Specialist

G. Oliver, Manager - Site Planning & Control

• J. O'Sullivan, Manager - Maintenance

R. Poulk, Senior NRC Regulatory Specialist

L. Tripp, Radiation Control Supervisor

V. Wagoner, Director - IPBS/Long range Planning

• J. Wilcox, Principle Engineer - Operations

B. Wilson, Engineering Supervisor

Other licensee employees contacted included technicians, operators, and

engineering staff personnel.

• Attended exit interview

2.

Exit Interview-

The inspection scope and findings were summarized on April 2,1985, with

those persons indicated in paragraph one above.

Meetings were also held

with senior facility management periodically during the course of this

inspection to discuss the inspection scope and findings. The licensee did

not identify as proprietary any of the materials provided to or reviewed by

the inspectors during this inspection.

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3.

Licensee Action on Previous Enforcement Matters

(Closed) Unresolved Item 325/84-31-02.

This item involved two concerns

relative to the operation of the Unit 1 Standby Gas Treatment System (SGTS).

Further investigation was necessary in order to verify the original design

requirements (see Inspection Report 84-35).

The first concern dealt with the operation of the Unit 1 SGTS train A and B

inlet and outlet dampers (B,

C,

E and G-BFV-RB).

It was initially

understood that the Final Safety Analysis Report (FSAR), described the

dampers as having automatic open capability. After reviewing correspondence

between the licensee and the A/E (designers of the SGTS), along with

original startup data, it can be verified that the original design was to

have these dampers normally open during operation and that no requirement

existed to have automatic open capability.

The dampers serve only as

maintenance isolation valves. The licensee does intend to clarify both FSAR

and the system descriptions regarding the operation of these dampers. This

concern is resolved.

The second concern was relative to the disparity between the SGTS damper

operation for each unit. Unit 2 SGTS train A and B inlet and outlet dampers

do have automatic open capability. The FSAR makes no statement regarding

this difference. The A/E explained that during construction of Unit 2 (Unit

2 was built before Unit 1), a modification was made to the Unit 2 SGTS in

order to allow the system to automatically isolate itself from the drywell

during an accident.

Included in this modification was the installation of

automatic open circuits for the train's inlet and outlet dampers.

Subsequent to the modification, an Engineering Review altered the Unit 1

SGTS design from that of Unit 2 during initial construction. This review

determined that the automatic open capability for these two dampers (per

train) was not necessary. Consequently, the Unit 1 SGTS was built with the

new automatic isolation capability but, the automatic opening function of

the train inlet and outlet dampers was deleted. Again, the proposed FSAR

change will clarify the disparity between units.

This concern and the

unresolved item are considered closed.

No violations or deviations were identified.

4.

Review of Licensee Even't Reports (92700)

The below-listed Licensee Event Reports (LER) were reviewed to verify that

the information provided met NRC reporting requirements.

The verification

included adequacy of event description and corrective action taken or

planned, existence of potential generic problems and the relative safety

significance of the event.

Onsite inspections were performed and the

inspectors concluded that necessary corrective actions have been taken in

accordance with existing requirements, licensee conditions and commitments.

These reports are considered closed.

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LER .1-83-03 - An under reactor vessel inspection revealed that detector

cables were_ separated from their associated detectors.

LER 1-83-06 - Insert / withdrawal positions and drive power indication for

SRM's and IRM's ~were not working. Control power supply fuse blown.

LER 1-83-07 -- One fuel bundle was located around each of the withdrawn

control rods.

LER 1-83-08 - HVAC System exhaust inboard isolation damper open limit switch-

sticking.

LER 1-83-15 - Well water isolation valves to both Standby Gas Treatment

System Train's Deluge Systems were closed rendering both deluge systems

inoperable.

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LER 1-83-19 - Standby Liquid Control System Relief Valves lifted at 1321 psi

and 1592 psig, respectively.

LER 1-83-20 - IRM

"A" was showing- instrument upscale indication from

moisture accumulation.

No violation or deviation was identified in this area.

5.

Maintenance Observations (62703)

Maintenance activities were observed and reviewed throughout the inspection

period to verify that activities were accomplished using approved procedures

or the activity was within the skill of the trade and that the work was done

'by qualified personnel.

Where appropriate, limiting conditions for

operation were examined to ensure that, while equipment was removed from

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service, the Technical Specification requirements were satisfied.

Also,

work activities, procedures, and work requests were reviewed to ensure

adequate fire, cleanliness and radiation protection precautions were

observed, and that equipment was tested and properly returned to service.

Acceptance criteria used for this review were maintenance procedures and

Technical Specifications.

Outstanding work requests that were initiated by the operations group for

Units 1 and '2 were reviewed to verify the licensee is giving priority to

safety-related maintenance and not allowing a backlog of work items - to

permit a degradation of system performance.

1No violations or deviations were identified.

6.

Surveillance Testing (61726)

Selected surveillance tests were analyzed and/or witnessed by the inspector

to ascertain procedural and performance adequacy.

the completed test

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procedures examined were analyzed for embodiment of the necessary test

prerequisites,

preparations,

instructions,

acceptance

criteria

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sufficiency of technical content.

The selected tests witnessed . were

examined to ascertain -that current, written approved procedures were

available and in use, .that. test equipment in use was calibrated, that test

prerequisites were met, system restoration was completed and test results

were adequate. The selected procedures attested conformance with applicable

Technical' Specifications, they appeared to have received the required

administrative review and they were performed within the surveillance

frequency prescribed.

- Acceptance criteria for evaluating surveillance tests were 10 CFR, ANSI

N18.7Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..and Technical Specifications.

During the performance of Surveillance Test PT-1.1.7PC, Average Power Range

Monitor (APRM's) Channel Calibration, techniciahs' caused individual APRMs to

be inoperable without placing them in bypass as required by the procedure.

This action caused a half Reactor Protection System trip (half scram) when

one APRM sensed only 10 Local Power Range Monitor (LPRM) inputs vice the

minimum 11. Unit I was operating at approximately 60% of power.

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Technical Specification 3.3.1, requires that each ApRM channel have at least

- two LPRM inputs - per level and eleven total LPRM inputs in order .to be

considered operable.

Less than 11 inputs will cause an APRM channel to

trip.

There are six APRM channels divided into two Reactor Protection

System (RPS) channels which have three APRM's each. Technical Specification 3.3.1, also requires that at least two operable APRM's be in service for

each RPS channel.

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PT-1.1.7PC requires that LPRM's, be calibrated prior to calibration of the

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associated APRM.- Each LPRM is calibrated individually by placing the LPRM

card selector switch to "By" (Bypass), which then permits the technician to

perform the necessary adjustments.

However, this action also removes that

LPRM from its' associated APRM. To account for this, the procedure includes

a step to bypass each APRM'while its associated LPRMs are being calibrated.

- Failure to have _ an APRM . bypassed when the eleventh LPRM was bypassed -

. resulted in the half scram.

The root cause of the problem is the way in which the testing crews handle

the turnover of surveillance tests which carry over from one shift to the

next.

PT-1.1.7PC was begun on March 12, 1985, during day shift.

It had

been continued through the swing shift but stopped before midshift.

The

test was then continued on the following day shift. Howaver, one of the

prerequisites for continuing the test (which had previously been met) was to

ensure Lthat ~ operators bypassed the applicable APRM at the Reactor Turbine

Generator Board-in the Control Room prior to the LPRM calibrations. This

step was not re performed when the test was restarted. Although the shift

operators had given permission to continue the test, the _ technicians

informed them that no APRM's would be made inoperable. This information was

incorrect.

The consequences of this action was to place APRMs out of service without

shift operating personnel permission or knowledge. Each APRM already had at

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1 east one LPRM level -with only two inputs remaining (due to various

unrelated problems). Consequently, when the LPRM calibration was performed,

each APRM was sequentially made inoperable and then, subsequently, returned

to service. At no time was more than one APRM inoperable in an RPS channel.

Technical Specification 6.8.1(c) requires that written procedures be

implemented for surveillance and test activities of safety related

equipment. Contrary to TS 6.8.1(c), requirements described in PT 1.1.7PC

step VII D.3.(a) were not met upon resumption of the surveillance test after

an eight-hour postponement in that the APRMs were not bypassed during the

individual LPRM calibration (violation 325/85-05-01).

The immediate

corrective actions were - to: (1) Revise PT 1.1.7PC to add precautions

concerning minimum. numbers of LPRM channels required and to better define

steps which require APRMs to be bypassed; (2) Conduct training of testing

crews as to the relationship between APRMs and LPRMs as well as general APRM

operation; (3) Counsel the technicians involved and testing crews regarding

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the necessity to review previously performed sections of any surveillance

test they start which had been temporarily interrupted. These actions have

been completed.

One violation was identified in this area.

7.

Operational Safety Verification (71707, 71710)

The inspector verified conformance with regulatory requirements throughout

the reporting period by direct observations of activities, tours of

facilities, discussions with personnel, reviewing of records and independent

verification of safety systems status.

The following verifications were

.made:

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Control Room Observations - The inspectors verified that control room

manning requirements of 10 CFR 50.54, and the Technical Specifications

were being met.

Control room, shift supervisor, clearance and

jumper / bypass logs were _ reviewed to obtain information concerning

operating trends and out of service safety systems to insure that there

were no conflicts with Technical Specifications Limiting Conditions for

Operations. Direct observations were conducted of control room panels,

instrumentation and recorder traces important to safety to verify

operability and that parameters were within Technical Specification

limits. In addition, the inspectors observed shift turnovers to verify

that continuity of system status was maintained and questioned shift

personnel relative to their awareness of plant conditions.

The

inspectors verified the status of selected control room annunciators

and were assured that the control room operators understood the reasons

why impo-tant annunciators were lit.

In addition, periodic verifi-

cations were conducted to insure that corrective actions, if appro-

priate, were initiated and completed in a timely manner.

ESF Train Operability - Operability of selected ESF trains was verified

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by insuring that; each accessible valve in the flow path was in its

correct position; each power supply and breaker, including control room

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fuses, are aligned for components. that must activate upon initiation

signal; removal of. power from those ESF motor-operated valves so

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-identified by Technical Specifications was completed; there was no

leakage.of major components; there was proper lubrication and cooling

water available; a condition did not exist which might prevent

fulfillment of the train's functional requirements.

In addition,

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instrumentation essential to system actuation or performance was

verified operable by. observing on scale indication and proper

instrument valve lineup, if accessible.

The High Pressure Coolant

Injection System-(Unit 1) was verified operable.

Radiation- Protection Controls - The inspectors' verified that the

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licensee's health physics policies / procedures are being followed,

including . area surveys, RWP's, posting a'nd calibration of selected

radiation protection instruments in use.

Physical Security Plan - The inspectors verified that the security

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organization is properly manned and that_ security personnel are capable

.of performing their assigned functions, that persons and packages are

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checked prior to . entry into the Protected Area (PA), vehicles are-

properly authorized, searched and escorted within the PA, persons

within the PA display. photo identification badges, . personnel in vital

areas are authorized, that effective compensatory measures are employed

when required, and that security's response to threats or alarms

appears adequate.

Plant . Housekeeping - Observations relative to plant housekeeping

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identified no unsatisfactory conditions.

Containment Isolation - Selected containment isolation valves were

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verified to be in their correct positions.

Radioactive Releases - The inspectors verified that selected liquid and

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gaseous releases were made in conformance with 10 CFR 20 Appendix B and

Technical Specification requirements.

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No violations or deviations were identified.

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' ADS Valve'Not Connected to Accumulator per Design (37700)

On March .12, 1985, with Unit 2 shutdown, the licensee discovered that

~ Automatic Depressurization System (ADS), valve B21-F013C solenoid, was being

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supplied 'with instrument air. from a line which did not contain an

accumulator. The- condition was found when maintenance personnel attempted

to- isolate the solenoid on the

"D"

ADS valve to repair an air leak.

Isolation of valves specified in the procedure, failed to remove air f rom

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"D" _ ADS valve.

Walkdown of the instrument air tubing between the

-accumulators and the ADS valve's revealed the following:

a.

"C" ADS valve was receiving air from a line which had no accumulator.

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b.

"D" ADS valve was receiving air from the accumulator tagged for "C" ADS

valve.

c.

"E" manual _ Safety Relief Valve (SRV), was re.ceiving air from the

accumulator tagged for "D" ADS valve.

d.

ADS valve "H"

and "J" had their air tubing connected to accumulators

for "J" and "H"

respectively.

e.

- Manual SRV "G" and "F" had their air tubing interchanged.

Of>these, only item a. has any safety significance in that an ADS valve did

-not have an accumulator to supply capability'to open the valve and-hold the

valve open upon' failure of the normal air supply,'as described in the Final

Safety Analysis Report, paragraph 5.2.2.4.

As documented in the safety

evaluation entitled ' Verify Qualification of Accumulator on ADS Valves'

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dated June 15, 1984, the NRC found acceptable that (1) the accumulators are

used only as snubbers in the system and are not relied upon to maintain

pressure to the ADS valve actuators and (2) the standby compressor system

will supply the required air pressure under postulated accident conditions.

Hence, item a. has only. minor safety significance.

Apparently these problems were caused by installation of modification 80-086

in the summer of 1982. Mod 80-086 changed the SRV's from three stage.to two

stage valves. The modification procedure 80-086 did not address the removal

or _ reinstallation of the air tubing to the solenoids which allow remote

actuation of the SRV's. Specifically, the tubing had been disconnected at a

point at which the tubing had been bunched into groups to allow penetration

through the floor grating. Vhen they were reconnected, several of the tubes'

were mismatched.

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The root cause of the event was attributed to personnel failing to follow

ENP-03, Plant Modification Procedure, in that work'was performed as part of

the modification but was outside the scope as defined in Plant Modification

80-086, a condition not authorized by ENP-03.

This is a violation of

Technical Specification 6.8.1.(a) which requires that procedures be

implemented.

However this event occurred prior to the -enhancements

-associated with the Brunswick Improvement Program which resulted in

clarification and upgrading of management expections in the areas of-

adherence to administrative controls and attention to details while

-performing tasks. These were communicated to all staff members and have

.been incorporated into the daily conduct of business at the site. As part-

of.the general employee training, each employee sees a film' emphasizing

corporate commitment to quality and procedural compliance. As a result of

- the discovery- of this event, a review of current training . practices was

. conducted of the affected organizations. No changes to.the current programs

were deemed necessary.

Therefore because corrective action has been-

accomplished as part of the Brunswick Improvement Program 'and because the

event had minor safety significance in that only a redundant design feature

was unavailable which did not render an ADS valve inoperable, the event is

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classified as a licensee identified violation per 10 CFR 2 Appendix C

paragraph IV.A.

At the time of discovery Unit I was operating; ha aver, the event was

considered as not applicable to Unit 1 in that all SRV's, both manual and

ADS, have accumulators.

No violation or deviation was issued in this area.

9.

'Onsite Review Committee (40700).

The inspectors attended several special Plant Nuclear Safety Committee

meetings conducted during the report period.

The inspectors verified the following items:

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Meetings were conducted in accordance with Technical Specification

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requirements regarding quorum, membership, review process and personnel

qualifications;

Corrective actions, recommendations and decisions were completed as

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assigned.

No violations or deviations were identified.