ML20154P570

Insp Repts 50-259/88-05,50-260/88-05 & 50-296/88-05 on 880301-31.Violations Noted.Major Areas Inspected:Operational Safety,Maint Observation,Surveillance Testing Observation, ROs & Plant Operations Review Committee

ML20154P570
Person / Time
Site:	Browns Ferry
Issue date:	05/11/1988
From:	Ignatonis A, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20154P562	List: ML20154P560 ML20154P564 ML20154P570
References
50-259-88-05, 50-259-88-5, 50-260-88-05, 50-260-88-5, 50-296-88-05, 50-296-88-5, NUDOCS 8806030382
Download: ML20154P570 (39)

See also: IR 05000259/1988005

Text

,, . . -. . _ . .

UNITED STATES

/gan af og% WUCLEAR REGULATORY COMMISSION

[\- , REGION 11

3 ,, , j 101 MARIETTA STREET.N.W.

• t ATLANTA,GLORGI A 30323

s,++v ...

/

Report Nos. 50-259/88-05, 50-260/88-05, and 50-296/88-05

Licensee: Tennessee Valley Authority

6N 38A Lookout Place

1101 Market Street

Chattanooga, TN 37402-2801

Docket Nos. ~50-259, 50-260, and 50-296

License Nos. OPR-33, DPR-52, and DPR-68

'

Facility Name: Browns Ferry Nuclear Plant

Inspection Conducted: March 1-31, 1988

Inspectors: (/.d M mn A T M// /ef

p G. L. Pau@ 5enicrJResident Inspector Date Signed

Accompair.edby: C. R. Brooks, Resident Inspet'.or

E. F. Christnot, Resident Inspector

W. C. Bearden, Resident Inspector

A. H. Johnson, Project Engineer  ;

Approved by: 8 d, ma o 5 /R F -

A. J. Ignat.onis, Seftlon Chief Date Signed j

Inspection Programs,

TVAProjectsDivision

SUMMARY

Scope: This routine inspection was in the areas of operational safety,

maintenance observation, surveillance testing observation, reportable

occurrences, Plant Operations Review Committee (PORC), restart testing and

Q-List.

Results: Five violations were identified: (1) two examples of failure to

follow procedures for Standby Gas Treatment System iodine removal efficiency

testing and PHI 15.4, Unique Reporting Requirements involving containment spray

header nozzles; (2) failure to control the issuance of documents, including

changes thereto; (3) failure to adhere to Technical Specification Administra-

tive controls involving the PORC activities; (4) failure to implement an

,

inspection of activities affecting quality involving the installation of a  !

1

check valve; (5) failure to provide regulatory compliance and QA/QC retraining

j to foremen and craft personnel.

.

.

'

'8806030382 880520 .

PDR ADOCK 05000259

-Q DCD

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REPORT DETAILS

1. Licensee Employees Contacted:

• J. G. Walker, Plant Manager

P. J. Speidel, Project Engineer

• J. D. Martin, Assistant to the Plant Manager

• R. M. McKeon, Operations Superintendent

• T. F. Ziegler, Superintendent - Maintenance

D. C. Mims, Technical Support Superintendent

J. G. Turner, Manager - Site Quality Assurance

M. J. Nay, Manager - Site Licensing

• J. A. Savage, Compliance Supervisor

A. W. Sorrell, Site Radiological Control Superintendent

R. M. Tuttle, Site Security Manager

L. E. Retzer, Fire Protection Supervisor

• H. J. Kuhnert, Office of Nuclear Power, Site Representative

T. C. Valenzano, Director - Restart Operations Center

• Attended exit interview.

Other licensee employees or contractors contacted included licensed '

reactor operators, auxiliary operators, craftsmen, technicians, public

safety officers, quality assurance, design and engineering personnel.  :

2. Exit Interview (30703)

The inspection scope and findings were summarized on April 1 and April 8,

1988, with the Plant Manager and/or Superintendents and other members of

his staff. A special management meeting was held on April 8,1988, with

OSP and licensee management to discuss technical concerns with the Q-list '

and PORC implementation prograins. The following new items were identified

during this inspection:

a. Violation (259,260,296/88-05-01), Failure to control the issuance of

documents and changes thereto. Paragraph 6.a.

b. Violation (260/88-05-02). Failure to follow a program for inspection  !

of activities affecting quality. Paragraph 7.

c. Violation (259,250,296/88-05-03), Two examples of a failure to follow i

procedures and a lack of attention to detail. Paragraphs 8 and 12. '

d. Violation (259,260,296/88-05-04), Technical Specifications (TS),

Section 6.0, Administrative controls. Failure to properly designate

PORC alternate chairman and alternate members. Paragraph 10.

e. Unresolved Item * (259,260,296/88-05-09), Designation of proper ,

altenate PORC chairmen per Plant Managers Instruction (PMI) 7.1. l

Paragraph 10.

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2

f. Violation (259,260,296/88-05-08), Failure to provide regulatory

compliance and QA/QC retraining for foremen and craft personnel.

Paragraph 14.

g. Inspector Followup Item (259,260,296/88-05-05), Closecut of MR's

generated by the Restart Test program.

h. Inspector Followup Item (259,260,296/88-05-06), Two trains of ESFAS

actuating two separate sets of two dampers through one relay.

Identified as part of restart test program,

i. Unresolved Item * (260/88-05-07), Unit 2, Phase I, Q-List programmatic

deficiencies. Paragraph 13.

j. Inspector Followup Item (259,260,296/84-38-02), This item was

initially opened on Unit 3, involving shutdown board room air

handling system and has been expanded to include all three units.

Paragraph 4.

The licensee acknowledged the findings and took exception with one of

three examples in Violation A, regarding PORC composition. The licensee

maintained that TS 6.5.1.2.b allowed the PORC Chairman to designate anyone

as an alternate PORC Chairman. The inspector stated to the licensee this

specification was not intended to authorize any additional alternate

chairman other than those clearly stated in TS 6.5.1.2.a, but instead was

intended to require that those individuals be clearly designated, by name,

in a written document. This contention is subject to further evaluation

by the NRC management. The NRC will determine whether or not the licensee

is meeting the intent of TS 6.5.1.2.b by designating alternate PORC

chairmen in the PMI 7.1, which are other than those specified in TS 6.5.1.2.a. Thus, this example of the violation is being reclassified as

an Unresolved Item * (259,260,296/88-05-09).

The licensee did not identify as proprietary any of the materials provided

to or reviewed by the inspectors during this inspection. I

3. Licensee Action on Previous Enforcement Matters (92702)

(CLOSED) Deviation (259,260,296/87-02-09), Neutron 00simetry Around New i

Fuel Vault. During a review of criticality margin for storage of new '

fuel, the inspector identified that the licensee had not provided neutron I

dosimeters for the new fuel storage vault contrary to FSAR paragraph l

10.2.4. The licensee had not utilized the new fuel storage vault for fuel  ;

stcrage during the period around the time this item was identified. i

As a result of this item the licensee has revised the FSAR, Technical

Instruction (TI)-14, Special Nuclear Material Control, and General ,

Operating Instruction (G01)-100-2, New Fuel Operations, to reflect the

l

i

• An Unresolved Item is a matter about which more information is required to  !

determine whether it is acceptable or may involve a violation or deviation, j

.

.l

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3 l

decision to eliminate the use of the new fuel storage vault as a fuel i

storage facility. The FSAR was revised in Amendment 5 to require that '

each vault would be provided with neutron dosimeters whenever new fuel

is stored. TI-14, Rev 4, Section 3.3, contains the current list of

authorized Item Control Areas where SNM is stored. The list does not

include the new fuel storage vault. G01-100-2, Rev. 2, states that'Radeon

personnel must be contacted to install Neutron Detection Dosimetry in the

new fuel storage vault, prior to placing any fuel bundles into the vault.

(CLOSED) Violation (259,260,296/85-57-05), Failure to Perform

10 CFR 50.59 on Damper Timing. During review, regarding stroke times on

- Secondary Containment Isolation Dampers, the inspectors identified that

the Secondary Containment isolation damper function had been altered

through acceptance of longer damper closure times. This condition

resulted in a change by default to a system as described in the FSAR.

Although the Commission eventually agreed that the change in damper

closure time did not constitute an unreviewed safety question, timely

action by the licensee was not initiated to make the evaluation required

by 10 CFR 50.59.

The inspector reviewed TVA's supplemental response dated April 7,1987,

which stated the reasons for the violation and outlined the proposed

corrective actions. The licensee identified two separate causes for the

violation. The first cause was specific to this case while the second

cause was related to more general principles associated with applications

of 10 CFR 50.59.

The inspector determined that the licensee has revised Program Manual

Procedure, PHP 060.04, Evaluation of Changes, Tests, and Experiments, to

add the requirement that a discrepancy between the as-built facility or-

procedures and the description in the FSAR reported in a Condition Adverse

to Quality Report (CAQR) shall be reviewed or a safety evaluation aer- l

formed to evaluate the existing condition in accordance with th's  !

procedure and to support the proposed corrective actions. Additionally

The Nuclear Quality Assurance Manual (NQAM), Part 1, Section 2.16, Quality

Notice-Corrective Action was revised to add the requirement to require

10 CFR 50.59 screening and/or USQ determination for a CAQR that documents

discrepancy between the plant and the FSAR.

The completed corrective actions should be adequate to preclude recurrence

of the original violation. This item is closed.

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(CLOSED) Violation (259,260,296/87-33-07), This violation resulted from

inspection review of Special Test (ST) 8723 which was performed to obtain  !

data on diesel generator excitation system. The violation stated that the l

Plant Operations Review Committee's (PORC's) approved results of the test

could not be located.

Plant Managers Instruction (PMI) 17.1 Conduct of Testing was revised to

place special tests under the administrative control of the plant pro-

cedures staff. Condition Adverse to Quality Report (CAQR) number 8FP 87

i

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0844 was written and dispositioned to indicate that the tracking and

filing of Special Tests will be turned over to document control following

a review by Technical Services. A review of Special Test Records

indicated that ST-8723 results have been approved by PORC turned over to

document control and microfilmed.

The licensee has addressed the inspectors concern as stated in the

original inspection report and corrective action should be adequate to

preclude recurrence. This violation is closed.

(CLOSED) Violation (259,260,296/87-27-01), ' Accountability of byproduct

material. An intensive records search of power store and radioactive

material areas was conducted. A complete physical search of all byproduct

material storage locations and work areas was completed. A complete

inventory was developed. A formal critique involving designated manage-

ment and key personnel was held on October 26, 1987, to discuss program-

matic problems followed by a discussion of program improvements instituted

siace the NRC inspection. The critique was held to assure all problems

had been identified and addressed in the program improvement plan. BFN

procedures SDSP 23.2 and Surveillance Instruction 4.8.E was revised to

clarify responsibilities, define sealed source, and revise the byproduct

inventory control form. Administrative control of non-exempt byproduct

and source material was transferred to the RADCON section (RCI-21

applies), TVA central program guidance for byproduct and source material

control was issued under ONP Standard 5.7.11 (Controlling Byproduct and

Source Material). A Technical Specification 6.6.B change is listed on the

Corporate Commitment Tracking System to clarify radioactive material

inventory requirements. The T.S. change is scheduled for submission

in December 1988.

(CLOSED) Violation (259/84-34-02) Core Spray Check Valve FCV-75-26

inoperable during core spray overpr,essurization event due to inadequate

maintenance. The testable check valve was held open by its test actuator

during power operation and may not have closed if a line break outside 1

containment had occurred. The root cause of the event was attributed to I

inadequate maintenance activities during the air solenoid rebuild activi-

ties and incorrect methodology used to verify valve operability on return

to service. Mechanical Maintenance Instruction 51 (Maintenance of CSSC/

NON-CSSC Valves and Flanges) was upgraded to address the concerns of this i

violation. l

4. Followup of Open Inspection Items (92701)

(CLOSED) Inspector Followup Item (250,260,296/8.7-02-08), New fuel

criticality precautions (SIL No.152). During a review of criticality

margin for stcrage of new fuel, the inspector identified inconsistencies

with caution signs associated with the new fuel storage vault. The

licensee had not been utilizing the new fuel storage vault for fuel l

storage during the period this was identified.

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5

As the result of this item the licensee has revised Technical Instruction

(TI)-14, Special Nuclear Material Contrcl to reflect the decision to

eliminate the use of the new fuel storage vault as a fuel storage

facility. TI-14, Rev. 4, Section 3.3, contains the current list of

authorized Item Control Areas where SNM is stored. The list does not

include the new fuel storage vault. The licensee has adequately addressed

the inspector's concerns on this item. This item is closed.

(CLOSED) Inspector Followup Item (259,260,296/86-22-03), This item

concerned three inspector identified cable pelling instruction deficien-

cies concerning General Specification 38. The licensee addressed and

incorporated these concerns into Modification / Addition Instruction 44,

Cable Pulling For Insulated Cables Rated Up to 15,000 Volts. This item is

closed.

(CLOSED) Inspector Followup Item (259,260,296/86-40-06), This item

concerned the standby gas treatment system (SBGT) surveillance instruction

(SI) acceptance

setting criteria

requirements. Thenot meeting's review of the FSAR showed that thethe tec

licensee

SI acceptance criteria met the design bases. The licensee's TS change was

approved by the NRC on January 19, 1988. This item is closed.

(OPEN) Unresolved Item (259,260,296/87-02-07), This item concerned

inspector identified issues during the licensee's performance of

Surveillance Instruction 4.7.C, Secondary Containment Integrity; all

issues were addressed except for the following:

a. The licensee did not address the standby gas treatment system

relative humidity heaters where train B did not energize during the

performance of the instruction.

b. The licensee did not address the lowering of flow calculation

acceptance criteria by an immediate temporary change which was

subsequently approved by the Plant Operations Review Committee.

l

This item remains open.

(CLOSED) Inspector Followup Item (259,260,296/87-46-01), This item was j

opened to track review by the inspector of a materials substitution '

evaluation performed by the licensee in response to the inspector l

concerns. The Reactor Building Equipment Access Lock inflatable seals '

were replaced with Presray Corporation EPDM compound E-603 as opposed to

the originally installed neoprene elastomer. An acceptable analysis was

performed (TVA document B44 '880106 001) which evaluated the hardness,

tensile strength, elongation, tear strength, radiation resistance and

aging properties of the new material. This item is closed.

(CLOSED) Inspector Followup Item (..',260,296/87-12-01), This item

concerned the incorporating into plant procedures of the Restart Test

Program (RTP) plan requirements for system status punchlist and outage

system turnover packages.

_ _ _ _ _ - _ - _ _ _ - _ _

6

a. Provide the documentation and bases that will be used for declaring

systems operable which will not be included in the Restart Test

Program Plan, will not have a system test specification developed and

receive minimal or no testing.

A review of Site Directors Standard Practice (SDSP)-12.1, Restart

Test Program, indicated that the plant systems are divided in groups

as follows:

Group 1: Systems critical to safe operation or shutdown of plant

will be included in this group. Testing requirements are

determined primarily by Design Baseline Evaluation Program.

Group 2: Systems which provide support to plant operation are

categorized as Group 2. Few test requirements specified by

Design Baseline Evaluation Program; the majority of test

requirements are determined by the RTP system review.

Group 3: Systems not directly supporting plant operation and not

important to safety. Generally, no testing will be

required.

Attachment 1 of the SDSP-12.1 and Attachments 1 and 2 of RTP

submittal, dated July 13, 1987, lists each system by group.

b.Section II A lists plant systems for which tests' specificatinns will

be written but do not include ECCS initiation signals or automatic

depressurization system.

The RTP Tests include various logic functional tests and initiation

using jumpers, etc. as well as a common accident test for all diesel

generators and various surveillance instructions. The RTP includes

not only the loss of power / loss of coolant accident (LOP /LOCA) test,

but also the backup control test and an integrated cold functional

test. It is a very comprehensive testing program which provides

initiation signals with emphasis on surveillance instructions,

c. The establishment of those administrative controls which will be used

to conduct the test program and document test results are not defined

in the RTP plan. Examples of these are:

(1) Test Organization

The RTP submittal of July 13, 1987, contains Figure 1 which

j indicates the Restart Test Program Organization chart.

(2) Program for reviewing and approving tests documents and tests

results.

A review of SDSP 12.1, Section 6.2, RTP Test Instruction Review

and Approval, and Section 6.7, RTP Test Results Package

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Preparation and Review, as well as Figure 2 of the submittal

dated July 13, 1987, indicated that adequate controls were

established, which involved the Joint Test Group (JTG) and Plant

Managers reviewt and approvals.

(3) Interface between Licensee Organizations

A review of SDSP 12.1, Section 6.1, RTP Test Instruction Format

and Section 6.3, Test Conduct, as well as Figure 2 of the

submittal dated July 13, 1987, indicated that various ifcensee

organizations such as operations, maintenance, modifications and

nuclear engineering must coordinate their activities with the

RTP group in order to support testing.

(4) Controls for handling test interruptions, test discrepancies,

corrective actions and test requirements are addressed in SDSP

12.1, Sections 6.1, 6.3 and 6.6, Test Exceptions and SDSP 12.2

Daelopment of System Test Specifications. The inspector's

review of these documents and direct observation of tests being

written and conducted indicated that adequate procedural

controls are in place and are being utilized. This includes:

the periodic activities of the JTG; the use of Condition Adverse

to Quality Re

malfunctions, ports (CAQRs);

i.e. pump motcradequately documenting

fails to start equipment

or pump motor

grounds out; critiques conducted after malfunctions; and the use

of test exceptions.

(5) Existing plant administrative controls that will be applicable

to the restart test program

A review of SOSP 12.1 indicated that various reference, source

and implementing documents such as the Nuclear Quality Assurance

Manual, FSAR, Technical Specifications, Nuclear Performance

Plan, Volume III and numerous SDSPs, Browns Ferry Standard

Practices, etc. as well as in field observation of testing

activities, and the day to day monitoring of the program also

indicated that the existing plant administrative controls

applicable to the RTP are adequately listed and implemented.

(6) Controls for scheduling test activities

The daily RTP status meeting, use of the punchlist, the Restart

Operations Committee (War Room) meetings and use of the P2

scheduling program demonstrate adequate controls for scheduling

test activities,

d. Existing plant implementing procedures for the RTP do not provide

methods for handling system status punch list and system turnover

packages.

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4

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A review of SDSP 12.1, Section 6.7, RTP Test Results Package

Preparation and Review and RTP-SIL-005, System Punch List Program,

and periodic attendance at JTG and RTP Punch List meeting indicated

that implementing procedures exist and are being utilized for system

status punch list and system turnover packages (results packages).

The licensee has adequately addressed the inspectors concerns on this

item. This item is closed.

(CLOSED) Inspector Followup Item (259,260,296/87-12-02), This item

concerned the deficiencies in Site Directors Standard Practice (SDSP)

12.2, Restart Test Program. The deficiencies and remedial actions are

listed below:

a. Provide restart test organizational chart.

TVA letter from Director, Nuclear Safety and Licensing to US NRC,

dated July 13, 1987, cor. Mins figure I which indicated an organiza-

tional chart titled Restart Test Program Organization,

b. Document qualifications of test engineers and test directors.

SDSP 12.1, Section 6.9 and RTP Section Instruction Letter (SIL)-002,

Training and Qualification of Restart Test Personnel outline the

cualifications of test personnel and which in turn reference ANSI

h45.2.6 Qualifications of Inspection, Examination and Testing

Personnel for Nuclear Plants and Regulatory Guide 1.58 Qualification

of Nuclear Power Plant Inspection, Examination, and Testing

Personnel,

c. Establish on going training for the restart test group.

RTP-SIL-002 assigns the ongoing training of restart test personnel

responsibility to the RTP Manager. A review of group records

indicated that ongoing training is in progress, including the RTP

manager,

d. Establish methods to ensure that the latest test procedure revision

and changes are being used.

A review of SDSP 12.1 and periodic attendance of the Joint Test Group

(JTG) meetings indicated that no intent changes or test revirions are

allowed without JTG approval.

e. Establish controls which require RTP group approval prior to

conducting work on a system under test and remain in effect

throughout test until system is turned over to operations.

l A review of SDSP 12.1, SIL-005 and SIL-006, System Checklist

'

preparations as well as continuous attendance at the RTP status

meeting indicated that sufficient controls are in place to ensure

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that the RTP Engineers are adequately apprised of the status of their

assigned systems throughout the pre-testing, testing and post testing

phases.

The licensee has adequately addressed the inspectors concerns on this

item. This item is closed.

(CLOSED) Inspector Followup Item (259/86-32-12), Deficiencies noted

during annual emergency drill of September 24, 1986. Several meetings

were held with licensee representatives to address the concerns identified '

by this item. Licensee action is summarized below in regards to each of

the 28 concerns,

a. Phone communications: Call override on the Technical Support Center

(TSC) dimension phones was added to assure riority calling was

available. The concern on the ENS phone ri ng in the TSC when

being used by another station would require c rcuit re-engineering

when recuested by the NRC. When the TSC is staffed there is no need

for a r< ng in the TSC, since the plant emergency procedures require

continuous manning of ENS. The inspector conducted operability

checks on selected communications equipment in the TSC

satisfactorily,

b. Communication of Reactor Water Level: This item was addressed and

discussed in inspection report 87-39.

c. TSC assessment of key plant aarameters: Observance of the 1987 drill

indicated improvement in th

Building:

on this item was clarified.TheThisemergency

item wasplant

also procedure (EPIPthe

observed during 14) providing g'

1987 drill and closed out in inspection report 87-39.

u. Operators slow to recognize unisolable leak from Scram Discharge

Instrument Volume drain: This item was addressed and closed out in '

inspection' report 87-39.

v. Conflict between Security and Health Physica personnel at entry to

reactor building: Security force dressouts should be required for

3 drills to enhance emergency response training,

w. Local and Perimeter Environmental monitoring data not used during

drill. Since the drill TVA has disconnected the monitors and will

I not use the data henceforth.  !

x. TSC status board not up to date. This item was addressed and closed

out in inspection report 87-39.

,

y. Control room logs illegible: Observance of the logs during the 1987

drills indicated no deficiencies as noted in inspection report 87-39,

although unofficial logs were maintained for drill purposes.

z. Improper use of E0I's: System-based E0Is should be more closely

followed during drill and real scenarios. This item was closed in

inspection report 87-39.

aa. Shift -Engineer wireless phone usage inadequate: This item was

observed and closed out in inspection report 87-39. ,

bb. Use of plant procedures difficult due to numerous cross-references.

l

Procedures were upgraded and reviewed during 1987 drill. This item

J was closed in inspection report 87-39.

(CLOSED) 2nspector Followup Item (259,260,296/87-33-08), This item

resulted < rom an inspector review involving the Chernobyl accident lessons

learned ud the specific item being control of special tests. The

.

procedure. Plant Managers Instruction (PMI) 17.1 Conduct of Testing,

l controls special tests was classified as NON-SAFETY RELATED although it

not only adninistratively controlled special tests it also controlled

surveillance instructions. A review of temporary change to become

permanent numeer 17.1-08 to procedure PHI 17.1 indicated that the

classification of the procedure was changed to SAFETY-RELATED. This item

is closed.

(OPEN) Open Item (296/84-38-02), Shutdown Board Room A/C Test. The

inspectors had monitored testing of the clectrical shutdown board room

s ventilation system which occurred on September 8, 1984. Numerous 3roblems

.

were observed associated with cooling capacity and obtaining requ' red air

i flow for the Unit 1 and 2 units. Unit 3 testing was not performed and

1

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12

planned for a later time. This item was opened to track the resolution of

problems after completion of testing.

The original corrective action was to correct the various problems with

the system as designed by work under numerous maintenance requests

including such work as increasing fan speeds to increase air flow and

improve cooling capacity. The systems were intended to be left as

designed.

The inspector reviewed documentation provided by the licensee and

determined that the licensee has abandoned the original intention to

repair the ventilation systems as designed. DCR 2344 was developed to

remove the existing equipment and install for each unit a new single

seismic /anvironmentally qualified air conditioning and ventilation system.

Each system will consist of two redundant full capacity trains each

capable of providing cooling requirements for both of the units shutdown

board rooms. ECN P0956 was written to implement the design change for

Unit 2. The licensee stated that the design work associated for ECN P0956

is partially field complete and scheduled to be complete and tested prior

to unit startup. Units 1 and 3 systems will be modified prior to restart

of the respective units. Additionally, the licensee stated that post

modification testing would occur after field work is complete and that

routine testing of the system would occur under revised versions of

Electrical Maintenance Instruction (EMI)-97, Quarterly Maintenance on

Electrical Board Room Emergency Air Conditioning, and Technical

Instruction (TI)-81, Shutdown Board Room Emergency Cooling System

Performance Check. This item will remain open pending completion of field

work, post-modification testing, restart test, and the necessary

procedural changes. This item will be placed on the Unit 1 and Unit 2

open list to allow tracking and closecut for each unit individually

Inspector Followup Item (259,260,296/84-38-02).

(CLOSED) Inspector Followup Item (296/86-28-01), This item resulted from

a routine tour of the plant by the inspector and a diesel generator fuel

oil transfer valve 3-18-611 was discovered unlocked and closed. Plant

drawing 47W840-1 indicated the valve's normal position as open. The

original intent was to change the drawing; however, on further examination i

the licensee decided to change procedure 0-01-18, Fuel Oil System l

Operating Instructions, Attachment 1, Fuc1 Oil System Valve Lineup

Checklist, to indicate the required position of valve 18-611 as open.

,

l

This item is closed.  !

(CLOSED) Inspector Followup Item (259,260,296/86-40-11), This item

concerned location of the responsible organization and program which would

evaluate significant changes in the site surroundings such that original

licensing assumptions remain valid. The specific example cited changes in l

hazardous material transported by barge past the plant site. As described {

in TVA's letter to the NRC dated March 1, 1988, in response to a similar i

deficiency, a corporate rSAR update program is being developed. This l

program will assign responsibility for and require an annual review of 4

hazardous material passing or located near the facility. The Office of l

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13

Natural Resources and Economic Development (ONRED) has been tasked with

performing this function. This item is closed.

(CLOSED) Unresolved Item (259,260,296/86-40-10), Revaluation of the

control room habitability following a postulated chlorine release. The

inspector questioned the original analysis of this NUREG-0737 Item

III.D.3.4 since it was based upon an assumption that no chlorine was ,

shipped by barge past the site. The inspector learned through contacts '

with the Army Corps of Engineers that indeed several chlorine barges

passed near the site each month. In response to this information, the

licensee undertook a fairly comprehensive search for sources of chlorine

river traffic and concluded that about 26 shipments per year could be

expected. Since this is below the 50 per year cut-off in Regulatory Guide

1.78, no remedial protective measures were required. The licensee placed

this information on the docket with it's letter to the NRC dated May 26,

1987. The letter requested concurrence with the conclusion that Item

III.D.3.4 of NUREG-0737 remains closed in light of the revaluation. This

unresolved item is closed and any further activity will be coordinated

through the licensees letter of May 26, 1987.

5. Unresolved Items *

A new unresolved item related to Q-list programmatic deficiencies is

identified in paragraph 13.

6. Operational Safety (71707,71710)

t

The inspectors were kept informed of the overall plant status and any

significant safety matters related to plant operations. Discussions were

held as required with plant management and various members of the plant

operating staff.

The inspectors made routine visits to the control rooms. Observations

included instrument readings, setpoints and recordings; status of

operating systems; status and alignments of emergency standby systems;

onsite and offsite emergency power sources available for automatic

operation; purpose of temporary tags on equipment controls and switches;

annunciator alarm status; adherence to procedures; adherence to limiting

conditions for operations; nuclear instruments operable; temporary

alterations in effect; daily ournals and logs; stack monitor recorder

,

j

traces; and control room mann ng. This inspection activity also included

numerous informal discussions with operators and their supervisors.

General plant tours were conducted on at least a weekly basis. Portions of

the turbine building, each reactor building and outside areas were

visited. Observations included valve positions and system alignment;

snubber and hanger conditions; containment isolation alignments;

instrument readings; housekeeping; proper power supply and breaker;

alignments; radiation area controls; tag controls; on equipment; work

activities in progress; and radiation protection controls. Informal

j

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_ _

. _ . _

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14

discussions were held with selected plant personnel in their functional

areas during these tours,

a. Temporary Alteration Control

On March 10, 1988, the licensee initiated a Temporary Alteration

Control Form (TACF) to allow substitute bolts to be installed in

place of two failed bolts on the Reactor Building overhead crane.

,

During the licensee's review and 'ap3roval of the TACF (Number

3-88-001-111), another deficient condition developed regarding damage

to the threads of a different bolt on one of the crane rail supports.

The damage was so extensive that repair was not possible and the nut

could not be replaced. A revision to the TACF was initiated on

March 15,1988, in order to accept the deficient bolt and allow

limited operation of the crane. PHI 8.1, Temporary Alterations,

contains the administrative controls for processing TACF's. Only the

following guidance on revising TACF's is contained in the

instruction:

Revision to an existing TACF is permitted provided:

1. The original is replaced with a copy marked "original".

.

2. The revised and reapproved original is returned to the

control room within 7 days.

As a result, some approval and review signatures were not re-obtained

following addition of revised information on the TACF form.

1. The Operations Supervisor's concurrence signature was dated

March 13, 1988.

"

2. The Shift Engineer's approval of the TACF was dated Harch 13,

1988.

!

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3. The File Clerk made and distributed copies of the TACF on

March 15,1988, although four signatures on the TACF were dated

March 16, 1988.

-

The above information was only available on the original TACF form

maintained in the Shift Engineer office and was the condition of the

TACF on March 17, 1988.

j

l A review of the TACF files for similar problems yielded the

following:

1. TACF 2-85-50-24 - Revision 1 added more tags; however, the Shift

Engineer signature certifying that the tags were hung was not

,

updated nor was the signature for changing the as-built drawings

updated within the required 30 days. The Unreviewed Safety

1

1

15

Question Determination (USQD) for this alteration was also not

updated to ensure the original evaluation remained valid.

2. TACF 2-84-097-57 - Revision 1 changed the number of tags. The

sign-offs for installing these new tags were not obtained.

3. TACF 2-84-101-64 - Revision 1 added a restriction on the load

carrying ability of some temporary framework and required the

installation of an absolute filter. The USQD was not updated

for this change.

4. TACF 2-85-039-064 - Revision 1 chang"ed the system status after

installation of the alteration from operable" to "inoperable."

The appropriate approval signature from the Operations Superin-

tendent was not obtained.

No provisions are made for modification of the TACF number to signify

the revision level nor is the TACF index appropriately annotated.

There are additionally no controls on retrieval of superseded copies

or notification of revision to organizations such as the Nuclear

Safety Review Board (NSRB) who are on distribution for copies. These

deficiencies have been identified as a violation of the document

control requirements defined in Criterion VI of 10 CFR 50,

Appendix B. (259,260,296/88-05-01).

7. Maintenance Observation (62703)

Plant maintenance activities of selected safety-related systems and

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

accordance with recuirements. The following items were considered during

this review: the 1"miting conditions for operations were met; activities

were accomplished using ap3 roved procedures functional testing and/or

calibrations were performec prior to returning components or system to

service- quality control records were maintained; activities were

accomp1}shed by qualified personnel; parts and materials used were

properly certified; proper tagout clearance procedures were adhered to;  ;

Technical Specification were adhered to; and radiological controls were j

implemented as required.

l

1

Maintenance requests were reviewed to determine status of outstanding jobs

and to assure that priority was assigned to safety-related equipment  !

maintenance which might affect plant safety. The inspectors observed the I

below listed maintenance activities during this report period:

l

a. RHRSW Dresser Coupling Replacement,

b. Unit 2 Condenser Tube Bundle Replacement.

c. Reactor Protection System (RPS) Motor generator set motor recondi-

tioning and cleaning.

!

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16 i

.

d. Refueling Bridge preventive maintenance.

e. Refuel floor blowout panel inspection,

f. Standby Gas Treatment System (S8GT) heater conduit trouble shooting

and blower motor mount repairs.

,

No deviations or violations were identified for the areas inspected above.

On March 15, 1988, the licensee discovered that check valve number

2-67-659 was installed backwards. This condition prevented the flow of

Emergency Equipment Cooling Water (EECW) from the north EECW header to the

Unit 2 Residual Heat Remc, val (RHR) pump seal cooler and the RHR pump room

) cooler. This condition went undetected since an adequate supply of

cooling water was available from the south header. The licensee

researched the valve's maintenance history and found that the last

activity pertained to Maintenance Request (MR) 792717 on November 20,

1987. This activity removed the check valve and installed a blank flange ,

for hydrostatic testing. Following the hydro, the valve was reinstalled. '

The only sost maintenance test requirements documented on the MR was a QC

verificat< on of cleanliness and a leak check at the flanged connection.

The leak check was marked "not applicable" because another Surveillance

Instruction (SI 3.3.14. A.2) was to be performed. There was no QC

inspection required to ensure proper valve installation. The Browns Ferry

Maintenance Program contains adequate controls to prevent this situation

had they been used. Plant Managers Instruction (PMI) 6.2 Conduct of

Maintenance, Section 4.14.3 contains the program guidelines,on independent

verification. These verifications are to be performed in order to provide

a high degree of assurance that a maintenance activity was performej '

correctly and to eliminate personnel errors affecting work where there is

l

a chan:e for an error to degrade a safety function. This instruction

j references Standard Practice 3.11, Second Person Verification, which in

turn references Standard Practice 8F-3.2, Quality Control In.pection

Program. BF-3.2 section 5.2.1 contains examples of activities that should -

!

be verified by using QC inspection holdpoints. Satisfactory operation of

a valve following maintenance is one of the examples. Failure to properly ,

verify by inspection that the maintenance activity conformed with the

instruction on MR 792717 which required that the valve be reinstalled in

1 the proper orientation is a violation of 10 CFR 50 Appendix 8, Criterion X

j (260/88-05-02). 7

8. Surveillance Testing Observation (61726)

The inspectors observed and/or reviewed the below listed surveillance l

'

activit'es. The inspection consisted of a review of the procedures for i

technical adequacy, conformance to technical specifications, verification

! of test instrument calibration, observation on the conduct of the test, ,

i removal from service and return to service of the system, a review of test

i

data, limiting condition for operation met, testing accomplished by

,

!

! qualified personnel, and that the surveillance was completed at the 1

] required frequency.  !

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.

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17

a. EECW Flow Verification

The inspector observed Surveillance Instruction (SI) 4.5.C.1(4),

Emergency Cooling Water (EECW) Flow Verification, performed on

March 10, 1988. Previous problems with this particular SI are

documented in Inspection Reports 86-28 and 86-25. This SI was

performed on 83 EECW pump as a post-maintenance test following repair

of its associated discharge check valve. The major problem with this

attempt, as observed previously, was the failure of the three valves

to open. These valves cross-connect EECW to the Reactor Building

Closed Cooling Water (RBCCW) heat exchangers in the event of a Raw

Cooling Water (RCW) system failure. The Unit 1 and Unit 3 valves

(1-FCV-67-51 and 3-FCV-67-51) would only open to an intermediate

position. The Unit 2 valve failed to open at all. As a result of

this, the required flow rate of 4500 gpm was not attained. The

previous Unresolved Item on this SI will remain open pending a

s6tisfactory, trouble-free performance (259,260,296/86-25-13). Other

problems noted during the SI are listed below:

(1) Operators do not understand Precaution statement 3.9 which

states that RCW pressure should be monitored and maintained

below 50 psig. They were unsure what action would need to be

initiated if RCW pressure approached the limit.

(2) The reason for precaution statement 3.10 which states that all

diesel generator coolers should be operable was not apparent to

the operators and in fact one cooler was out of service.

(3) Procedure step 7.1.2 requires a sign-off that all prerequisites

are met even though each one of the prerequisite steps requires

an individual sign-off. Duplicate sign-offs make the procedure I

cumbersome and inefficient.

(4) Completion of this procedure was unnecessarily burdensome to the ,

Reactor Operator thereby distracting him from his primary l

duties. This procedure was intended to be a quick post-

maintenance test to support declaring the B3 pump operable

following check valve maintenance (a simple flow verification).

The SI consumed the better part of a shift. The operators can't i

afford to have this type of distraction when more important l

licensed duties are required during reactor startup and

operation.

(5) When the required 4500 gpm was not reached, the operators felt a

need to go outside of the procedure steps and adjust the RBCCW

temperature control valve (TCV) in order to increase the flow.

Although this was not contrary to any procedure it is indicative

of the responsibility that operators feel are imposed on them to

4 make a procedure work. Several examples of this type of

activity have been recently observed.

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18

(6) The number of phone calls received by the Reactor Operators is

excessive. The operator was observed to be interrupted by phone

calls during every face to face discussion with maintenance

technicians, other operators and engineers during the period of

observation by the inspector.

The above items were discussed with the EECW system engineer and

operations superintendent as appropriate.

b. SBGTS Charcoal Testing

As required by Technical Specification 4.7.8.2.a, the licensee

sampled and analyzed charcoal from the Standby Gas Treatment System

(SBGTS) following a fire in the Unit 2 Drywell in November 2,1987.

The charcoal was s!.!pped offsite to an approved contractor for methyl

iodide removal testing in accordance with ASTM 03803. Results were

received on January 12, 1988, for trains B and C and on February 16,

1988, for train A. The inspector's review of the data detected

examples of noncompliance with the required testing procedure. ASTM

0 3803-1979, Standard Test Method for Radioiodine Testing of

Nuclear-Grade Gas-Phase Adsorbents, contains five optional test

methods. The first methoc .ieasures the ability of the carbon to

capture methyl iodide under conditions approximating inside

containment under normal conditions. The second method approximates

the conditions faced by a SBGT system (outside containment) following

a design basis accident (DBA). The third method approximates a

system located inside containment following a DBA. The last two

tests use elemental iodine as a test gas and are not relevant to this

discussion. The licensee did not specify which option was required

in the procurement documents; however, the test Sarameters more

closely approximate method C. This would be simu'ating a DBA with

the system located inside containment. Method B (the SBGTS post-0BA

simulation) would have been more appropriate in this case: however,

the Technical Specifications require a temperature of 130*C which

essentially requires that method C be performed. Table 1 of ASTM

D3803 specifies the test parameters for method C which include 130 C,

95% RH, 60 minute feed time, and 240 minute elution time. The

contractor's test report indicated compliance with the temperature

and humidity requirements; however, the feed time was 90 minutes and

the elution time was 90 minutes. The feed period consists of flowing

humid air at 1.75 mg/m3 of methyl iodide past the charcoal bed.

During the elution period, air flow without the tracer gas is

maintained in order to evaluate the ability of the carbon to hold the

adsorbate once it is captured. The licensee contacted the vendor for

a justification of this noncompliance after the inspector raised the

issue. The contractor contends that the longer feed time was a

conservative test in that the ecuilibrium concentration is more

closely approached as the feed t me lengthens. He further stated

that the results are not affected either way by a shorter elution

time. The contractor arovided test data to cupport this. The

contractor didn't prov< de his motivation for deviating from the

19

provisions of the AJ1. Other problems with the conduct of the

surveillance are desci d below:

(1) Section 1.2 of ST 4.7.8.6, Standby Gas Treatment System - Iodine

Removal Efficiency, contains an erroneous statement that the

methyl iodide concentration should be 0.5 to 1.5 mg/m3. The

actual tested concentration and the concentration required by

ASTM D3803 was 1.75 mg/m3.

(2) Section 2 of SI 4.7.B.6 failed to list ASTM D3803 as a

reference.

(3) Section 2.8, NRC Commitments, failed to reflect a commitment

made to the NRC in response to a violation issued in Inspection

Report 86-11. The commitment was to obtain a certificate of

compliance to ASTM 0;803 from the testing organization.

(4) The procurement documents sent to the testing organization

failed to specify which version of the ASTM was required. The

licensee committed to the 1979 version and confusion with the

latest version (1986) is a real possibility.

(5) The certificate of compliance sent by the contractor certified

that the service conforms to the "client spec.". Since the TVA

procurement documents failed to be specific with regard to the

method and version, the validity of the certificate was

questionable. A clear statement of compliance with method C of

ASTM D3803-1979 should have been required from the contractor.

The SI was silent on what individual was responsible for

verifying acceptability of the certificate. It was therefore

not poss ble to judge whether that individual wes sufficiently

trained in QA requirements to make that judgement.

The record of noncompliances with this technical specification shows a l

lack of attention to the detailed test parameters. A violation for

failure to comply with procedures has been issued (259,269,296/88-05-03).

Inspection Reports 85-57 and 86-11 document the recent history on this

test,

i

1

9. Reportable Occurrences (90712,92700)

The below listed licensee events reports (LERs) were reviewed to determine  ;

if the information provided met NRC requirements. The determination I

included: adequacy of event description, verification of compliance with

technical specifications and regulatory requirements, corrective action

taken, existence of potential generic problems, reporting requirements

satisfied, and the relative safety significance of each event. The

,

'

following licensee event reports are closed:

4

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LER No. Date Event

259/86-32 11/10/86 Standby Gas Treatment

Not Meeting Technical .

'

Specification Requirements

259/87-17 and Rev. 1 7/17/87 Improper Byproduct Material

Inventory And Records

Retention Results In Failure

To Meet Technical

Specification Requirements

260/87-06 8/21/87 Fire Watch Function Was Not

Fulfilled While Grinding

Activities Were In Progress

Because Of Personnel Error

260/87-10 9/24/87 Fire Watch function Was Not

Fulfilled While Grinding .

'

Activities Were In Progress

Because Laborer Was Sleeping

On Duty

296/85-06, Rev. 2 2/13/85 Mismatch Of Reactor Water

Level Indicators

The licensee determined that the Standby Gas Treatment (SBGT) Surveillance

Instruction (SI) acceptance criteria did not meet the Technical Specifi-

cation (TS) trip setting requirements (LER 259/86-32). A review of the

FSAR showed that the SI acceptance criteria met the design basis. The

SBGT TS trip setting change was approved by the NRC on January 19, 1988. l

1

The NRC resident inspectors discovered improper byproduct material and

record retention (LER 259/87-17) during their inspection. The licensee in

turn found additional problems in this area. The accountability of

byproduct material and source material was shif ted to the Radiological

Control section. Procedures for the control of radioactive byproduct and

source material were revised.

The first event found a fire watch asleep while grinding was in progress.

The second event found grinding in progress without a fire watch present

(LER 260/87-06). Disciplinary action was taken against the sleep'ng fire l

watch. The craft personnel and their supervisors were counseled on the '

requirements and importance of the fire watch function. A critique of the

events were provided to personnel who may be involved in work requiring

fire watches.

A fire watch was discovered asleep (LER 260/87-10) during a housekeeping

team inspection, while grinding work was in progress. Disciplinary action l

was initiated against the fire watch. A critique of this event was

'

provided to personnel who are presently involved in work requiring fire <

watches.

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.

A mismatch between GEMAC reactor water level indicators developed while at f

low system pressure and the situation was misdiagnosed by the operators,

and the appropriate technical specifications were not implemented (LER 296/85-06). Reactor water level instrument response training was given 1

and placed into the requalification program. I

'

10. Plant Operations Review Committee (40700)

Amendment Number 134 to the Unit 2 TS significantly changed the function

of the PORC. The changes were implemented on December 10,.1987. The

'

major impact was to remove some of the administrative burden of procedure

reviews imposed on PORC in order to focus the PORC activities on it's

operational oversight role. The safety evaluation associated with this

amendment heavily references Regulatory Guide 1.33, Quality Assurance '

Program Requirements (Operation) which in turn endorses ANSI-

N18.7-1976/ANS-3.2, Administrative Controls and Quality Assurance for the 1

Operational Phase of Nuclear Power Plants. In order to address current

standards, the inspector compared certain aspects of the Browns Ferry

program with the current,1982 version of ANSI /ANS 3.2. The Browns Ferry

Program is implemented throu Plant Operations Review Committee.

Duringthislimitedreview,ghPMI7.1}ngproblemareaswerenoted:

the follow

1

a. PMI-7.1 designates three alternates for PORC chairman who are not

i

authorized by TS 6.5.1.2.a. These are the Maintenance Superintendent

! and two Unit Su3erintendents. When brought to the licensee's

j attention, the licensee maintained their position that TS 6.5.1.2.b

'

allows the PORC Chairman to designate alternate PORC Chairmen other

than stated in TS 6.5.1.2.a The applicable regulatory requirement .

(i.e., TS on PORC composition and alternates) requires i

i interpretation. This item will remain open as an Unresolved Item '

j (259,260,296/88-09) pending further NRC management evaluation,

b. One individual (the acting Maintenance Superintendent) has been

'

designated PORC alternate chairman on the backshift duty list for

many months and in fact chaired a PORC meeting on March 10, 1988,

1

although this individual has not been appointed in writing by the

, PORC chairman per TS 6.5.1.2.b.

c. Contrary to ANS-3.2-1982 the use of alternates were not restricted

i to legitimate absences of principals. In fact, the rotating

j assignment of PORC members on the backshift duty list ensured that .

random assignments for expeditea PORC meetings were made without f

j regard to availability of the principals.  ;

j d. Contrary to ANS-3.2-1982 alternative means for conducting meetings  !

such as telephone conference calls were being made in the absence of  ;

.

extenuating circumstances for which it is inipractical to convene

[ formal meetings. Four recent examples of telephone conference PORC .

meetings (2/20/88, 3/10/88, 3/13/88 and 3/15/88) were convened only  ;

l for the purpose of avoiding delays in the performance of maintenance [

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22

or modification activities. These delays would have been a day or

two of the current three year outage had the issues been held for the

next routinely scheduled PORC meeting. The inspector noted that no

guidance is contained in PMI-7.1 on who may cai1 for an expedited

telephone PORC meeting or what criteria is to be used in evaluating

the appropriateness of a telephone PORC.

e. An expedited telephone conference PORC meeting was held on March 10,

1988, with unqualified alternates acting as the Chairman and Health

Physics Supervisor. The Shift Supervisor who was designated the

Acting Health Physics Supervisor on shift was surprised that he was

being called in connection with a PORC meeting and questioned his

selection as a PORC member. This individual had received no PORC

training and had never attended a PORC meeting. He initiated a

discussion with his supervisor the following day which resulted in a

decision to void the PORC meeting. The hand written PORC meeting

minutes were discarded. After this event, the inspector obtained a

copy of the Safety Evaluation Review and Approval Form (tracking

number 111-C-88010) which was discussed during the telephone PORC.

This was a PORC reviewed and Plant Manager approved safety evalua-

tion. In addition, MR No. 721014 had it's PORC approval sign-offs

filled in based on the invalid PORC meeting. Thus, although the PORC

meeting was voided and no meeting minutes existed, the documents

which were aparoved by the invalid PORC were not retrieved and -

voided. The inspector noted that the chairman for this PORC meeting

was the acting Maintenance Superintendent who is not designated in

PHI 7.1 as an alternate PORC chairman. This deficiency was not

identified by the licensee,

f. Contrary to ANS-3.2-1982, the action taken by expedited telephone ,

PORC meetings are not being reviewed at the next regularly scheduled 1

meeting. In fact, PORC did not review the events surrounding the  !

voided PORC meeting until after the issue was raised by the inspector

during a meeting with the plant manager on March 18, 1988. During

this review of the invalid PORC, the licensee failed to detect that

the PORC chairman was not an approved alternate chairman nor was the ,

fact that MR number 821014 was approved by the invalid PORC l

addressed.

)

Although the ANSI /ANS 3.2 issues discussed above are not regulatory I

requirements, the inspector recommended that these practices be addad to

the Browns Ferry PORC program in order to align it with the current

industry standards. Failure to properly designate alternate chairmen and

control alternate members on the March 10, 1988, meeting and failure to

maintain written minutes of the March 10, 1988, expedited PORC meeting is

a violation of TS 6.5.1 (259,260,296/88-05-04). The licensee took action

to eliminate having unqualified PORC members listed on the duty call-out

list and has initiated changes to PMI 7.1 and the TS in order to prevent

similar situations. The licensee has also placed more attention on the

necessity of telephone PORC meetings and has recently oisapproved several

requests for these meetings.

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,

11. Restart Test Program

The inspector attended RTP status meetings, reviewed RTP test procedures,

! observed RTP Tests and associated tests performances, reviewed RTP Test

results and attended selected Restart Operations Center (War Room) and '

, Joint Test Group meetings. The following are the RTP activities and

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associated activities' monitored and status of testing during this

reporting period:

1 a. Restart Test Status t

(1) RTP-002, Condensate (COND). Special Test-99, Condensate

Demineralizer, was completed his test involved the nine (2A

through 2J) condensate demiviiters for the clean and precoat

sequence. Additional testir y . '

depend on the availability of

Unit 2 condenser.

(2) RTP-023, Residual Heat Removal Service Water (RHRSW) '

'

The system continued to be impacted by the RHRSW/EECW systems

north header outage. Valve 2-FSV-23-56 cycled on lost power;

however, the alarm failed to actuate. This item i s 9e rg

addressed by MR 885963 and Parts Request (PR)88-143. ' ch

, requires the issuance of a Design Change Notice (DCN) a'a n a work

plan.

1

(3) RTP-024, Raw Cooling Water (RCW)  ;

The system was also impacted by the header outage. However,

-

areas not affected by the outage were tested, such as: Section '

l 5.2 - Units 1 and 2 Raw Cooling Water flow path and temperature

"

control valve operability test for control bay chillers A and 8;  !

Section 5.5 - RCW pumps automatic pressure control test - Units *

1 and 2; and Section 5.7 - RCW system minimum flow verification +

4

test. Additional testing for Sections 5.1, RCW check valve (

operability test and 5.3 RCW pressure permissive - auto start of  ;

RHRSW pumps test is being impacted by material needs.

!

(4) RTP-030, Diesel Generator and Reactor Building Ventilation (DG & i

' RX BLOG VENT) i

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Testing was performed in Section 5.10 through 5.17 which

involved the diesel generator rooms battery room hood exhaust

fans. Testing is ongoing for Sections 5.2 through 5.9 which  !

involves the diesel generator rooms exhaust fans. Several l

! problems in this area were encountered involving MR's not being  :

1 adequately addressed, which in turn resulted in several test

exceptions. The use and dispositioning of RTP MR's is *

identified as inspector followup item IFI 259,260,296/88-05-05,  !

q closecut of RTP Maintenance Work Requests. ,

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(5) RTP-031, Control Building Heating Ventilation and Air Condi-

tioning (CONT BLDG HVAC)

The system RTP Test procedure has not been released by the Joint

Test Group (JTG) for testing. However, preliminary activities

are ongoing, involving the Control Room Emergency Ventilation

(CREV) power supply modifications, shutdown board room exhaust

fan 1A being out of service, and a tear in Unit 3 duct work.

The original schedule for this system was completely rewritten

with no definite forecast date for the actual start of the RTP.

(6) RTP-57-4, 480 Volt Distribution System (480 V DIST)

Due to the special test being performed on the diesel generators

this system will not be tested in conjunction with system

082-Standby Diesel Generators. The test will be performed on a

total of one hundred eighteen (118) separate circuits. As of

March 31,1988, a total of three (3) have been completed with a

forecast completion date of April 27, 1988. This RTP was

impacted by the change out of Transformer TS28 Power to 480 Volt

Shutdown Board 28, which was completed and returned to service

on March 30, 1988.

(7) RTP-57-5, 4160 Volt Distribution System (4.16 KV DIST)

The system is also impacted by the special tests being performed

on the diesel generetors. Testing of the individual boards,

such as low voltage start relays, etc. is continuing. However,

the major portions of the tests depend on the individual load

acceptance test of the diesel generators.

(8) RTP-57-7, 250 Volt DC Shutdown Batteries (250 VOC BATT)

A total of six battery chargers se involved with the system,

i.e. , A, 8, C, D, Spare and 3 EB. B charger was placed back in

service minus the filter capacitors and resistors in order to

test the spare charger. Both the spare charger and the 3 E8

failed the ripple voltage specification, which was .5% (arrived

at by calculation). The Division of Nuclear Engineering raised

the ripple voltage specification to 1%, which made the spare and

3 E8 chargers acceptable. The 8 charger will be tested when the

filter network is restored.

(9) RTP-065, Standby Gas Treatment (SGTS)

Train "B" humidity heaters cable was meggered and found to be

grounded. The licensee removed the defective cable, which

contained a total of six conductors, and verbally reported that

the conduit was filled with mud. This was due to the condition

of the conductors and the large amount of force required to

remove the individual conductors. The conduit run is from the

__ ---

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.

25

diesel generator building to the SGTS building and it is under

twenty-three (23) feet of soil. It is suspected that this run

is crushed and full of mud. Theengineeringandmodifications

groups devised an interim fix using Train "C spare conduits, in

order to support the fuel reconstitution effort. Due to this

problem establishment of secondary containment and fuel

reconstitution efforts have been rescheduled for mid April 1988.

(10) RTP-067, Emergency Equipment Cooling Water (EECW)

The system continued to be impacted by the north header outage

and repairs to chillers for shutdown boards as well as thermo

walls. These outstanding items virtually stopped most testing

during practically the entire month of March 1988. However,

some testing was performed in Section 5.5.

(11) RTP-070, Reactor Building Closed Cooling Water (BRCCW)

The aretesting activities were completed involving leak rate

test' ng and hydrostatic testing. The system was partially

released for testing in order to support LOP /LOCA. However,

several outstanding MR's involving material such as caskets and

brass flanges need to be c1) sed out before partial testing can

be completed. Dry well air coolers A3. A4 and A5 have leaks and

this material is needed to repair the coolers.

(12) RTP-075, Core Spray (CS)

The system is required to support LOP /LOCA. Testing was

performed in the following areas; Section 5.1, Manual initiation

for core spray pumps 2A, 28, 2C and 20; Section 5.9, Core spray

ready logic and 2A pump power interlock test 4160 volt shutdown

board (4160 V SD Bd) A; 28 pump power interlock test 4160 V SD

Bd C; 20 pump power interlock test 4160 V SD Bd 8, and 2 0 pump

power interlock test 4150 V SD Bd 0; and Section 5.2 core spray

pump start signal to Automatic Depressurization System (ADS).

(13) RTP-092, Standby Diesel Generators (STD8Y DG)

DNE by correspondence has conditionally accepted the load

acceptance tests performed to date. However, the licensee must

perform speed governor and voltage regulator adjustment tests

and DG Icading as a result of the Sequoyah DG experience.  !

Special Test (ST) 8806 was written for 18 DG and approved by the l

PORC. The ST commenced on March 29, 1988; however, a device l

known as the EGA box did not perform as the vendor required and

a new box was installed. A ST will be written for each of the

eight DG's in Units 1, 2 and 3. After the ST for each DG is ,

performed then the load acceptance test is performed immediately l

after the ST and this also includes the RTP diesel paralleling  !

test.

_ _. _ _ _ _

_ _ _

26

b. Quality Assurance / Control Overview of RTP (40704)

The inspector reviewed Quality Surveillance Reports directly

associated with the RTP. The inspector noted QA/QC involvement in

actual- testing activities, status meetings and JTG activities over

the past few months. The review was conducted ta assure adequate

documentation of QA/QC coverage of RTP activities.

The following specific quality surveillances were reviewed in depth:

(1) QAF-S-87-0498

This surveillance was performed to observe / verify that portions

of the Restart Test Program (RTP) were being performed in

accordance with SDSP 12.1 by observing sections of 2-BFN-RTP-082

Standby Diesel Generators. The surveillance and accompanying

checklist indicated that the test director exhibited an

excellent understanding of the requirements of SDSP 12.1 and

noted that CAQR BFP-87-0945 was written as a result of Test

Exception (TE)-29.

(2) QAF-S-87-0526

This surveillance was performed inpart to verify and observe

acceptable implementation of Special Maintenance Instructions

(SMI s) performed by the Division of Power Systems operation

(DPS0). RTP is required to use existing plant procedures in

whole or in part to satisfy test requirements. The performance

of DPS0 SMI 1-3EC.4 was required to satisfy portions of 2-8FN-

RTP-57-5, 4160 volt Distribution System. The surveillance

indicated that DPS0 personnel involved exhibited an under-

standing of the SMI and communication was conducted clearly and

objectively.

(3) QAF-S-87-0534

This surveillance was performed to verify and observe that l

portions of RTP Test 2-BFN-RTP-024, Raw Cooling Water (RCW) '

system, Section 5.1.4, Heat Exchanger Check Cooling Water  :

System, Valve Closure Test, were performed in accordance with l

SDSP 12.1. The surveillance and accompanying checklist

indicated that test exceptions TE-6 was doct.mented adequately

and that the qualifications of the RTP engineer and system ,

engineer were verified. I

(4) QAF-S-88-0008

This surveillance was performed to verify and observe that  !

portions of RTP Test-024, RCW, Section 5.6, RCW Bocster Pumps

1A,18 and 2A Automatic Pressure Control were performed in

accordance with SDSP 12.1. The surveillance and accompanying

27

checklist iricluded that a total of three (3) TE's were

identified and four (4) MR's were written to correct hardware

diff. 'alties.

(5) QAF-S-88-0074

This surveillance was performed to observe and verify that

portions of 'RTP Test-057-4, 480 Volt Distribution System,

Section 5.5, Verification of Automatic Transfer of 480 V Control

Bay Ventilation Board A from Normal to Alternate Source, were

performed in accordance with SDSP-12.1. The surveillance and

accompanying checklist indicated that personnel were familiar'

with the test instructions and that three (3) test exceptions

(TE) were identified with one TE generating a maintenance work

request.

(6) QAF-S-88-0033

This surveillance was performed to observe and verify that

portions of RTP Test 024, Raw Cooling Water, Section 5.5, RCW

Pumps Automatic Pressure Control, were performed in accordance

with SDSP-12.1. The surveillance and accompanying check list

indicated that the test was performed on pumps 2A and 28. Two

test exceptions were written, TE-21 and TE-23, which documented

the aumps 2A and 28 auto started a second time when tests

requ'rements stated that neither pump would auto start a second

time and the restoration was performed out of sequence.

(7) QAF-S-88-0073

This surveillance was performed to observe and verify that

portions of RTP were being performed in accordance with SDSP

12.1, which requires that the RTP use existing plant procedures

in whole or in part to satisfy test requirements. The surveil-

lance and accompanying check list indicated the Surveillance

Instruction (SI) 4.9. A.1.b-2 was being 3erformed to support

RTP-082, Standby Diesel Generators, Sect' on 5.7, Load Run, Load

Acceptance and Miscellaneous Tests. However, during the

performance of the SI a ground fault indication on Residual Heat

Removal Pump 20 was detected and the SI was aborted.

(8) QAF-S-88-0038

This surveillance was performed to observe and verify that

portion of RTP Test-057-7, 250 V DC Shutdown Batteries, Section

5.2, Measure the 250 volt Battery Charger Voltage and determine

the ripple factor were performed in accordance with SDSP 12.1.

The surveillance and accompanying checklist indicated that the

test was performed on the 250 V Shutdown Board Battery Charger

and that the surveillance involved review of the completed test

instruction.

- -. - . . . _ _.

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(9) QAF-S-88-0127

This surveillance was performed to. observe, verify and review

acceptable implementation of RTP-065, Standby Gas Treatment

System, Section 5.7, Off-Gas Cubicle Fan Test and Section 5.8.

Off-Gas Dilution Fan test. The surveillance and accompanying

check list indicated that the several minor deficiencies were

noted such as not all signatures in the test instruction was on

the signature log. Panel 9-8-1 was identified as panel 9-8 etc.

The surveillance also indicated that all personnel involved

exhibited an understanding of the test instruction.

From the in depth review of the surveillances and the day-to-day

observations, the QA/QC oversight appears to be adequately

documented.

c. Design Deficiency Identified by RTP

RTP-65-SGTS. This design deficiency pertains to the Standby Gas

Treatment System damper closure logic. The four dampers located in

the equipment bay, between the inner and outer equipment doors, close

on a signal from either of two trains; however, the two signal trains

are wired up to one relay which closes all Sui dampers. The

licensee's representatives verbally informed the NRC that this may

not meet the single failure criteria. This item is identified as

InspectorFollowupItem(259,260,296/88-0536).

d. Joint Test Group (JTG) Meetings

The JTG held periodic meetings with the inspector attending on a

continuous basis. Meetings numbered 88-006 and 88-007 were held on

March 24 and March 31, 1988, respectively. The meetings involved

discussions and recommendations of seven changes affect ng five RTP

test procedures, and one system test specification, two RTP test

procedures, seven system test specifications and two system check

lists. The inspector noted that during a discussion involving

2-BFN-RTP-099, Reactor Protective System RTP Test, a cuestion arose

about a test specification for the Mode Switch. A mem)er of the JTG

inquired as to wnether a test of the mode switch was necessary. The

member of the JTG was informed that it was a test requirement

established by the Base Line Test Requirements Document (BLTRD).

However, the discussion still continued as to the necessity of the

Mode Switch test. The inspector was later informed that the JTG as a

group understands that Base Line, by program, establishes test

requirements and that JTG and Restart Test Group are tasked with

meeting these requirements.

12. Containment Spray Header Examination For Rust Blockage of Spray Nozzle

The NRC requested the licensee to inspect the Unit 2 containment spray

headers for rust deposits that might affect system operability. There is

L

29

!

no history of leaks into the containment spray header or inadvertent l

actuations of this system on the ifcensee files. The containment spray

header cozzles are tested in accordance with Surveillance Instruction

4.5.2a (using air). This SI is performed every 5 years. There have been

no indications of blockage of the nozzle heads found during the past

performance of this SI.

On June 25, 1987, the Unit 2 containment spray header nozzles at elevation

601', azimuth 45 and 315* were removed and inspected for rust per MR

(maintenance request) A-710489. There was a slight amount of rust found l

which was uniformly distributed and which tightly adhered to the inside of

the carbon steel piping (i.e.10" Spray Header and 1-1 1/2" Standard

Nipples). The brass nozzle heads had no signs of loose rust and no signs

of oxides. Therefore, based on these preliminary findings, the licensee

reported to the NRC that no concern existed. l

On January 22, 1988, with the RHR System aligned for a hydrostatic

pressure test, water was observed leaking from several spray nozzles on i

the Unit 2 lower containment spray header. The water which was leaking

past the containment spray outboard isolation valve (2-FCV-74-60) was  !

drained from the header and the header was inspected. Rust particles were

found in the lower spray header and several of the spray nozzles inspected ,

were noted to be partially clogged with the rust. I

Stagnant water found in the normally dry header was the apparent cause of l

the rust.  !

The design of a 170 circumferential section of the Unit 2 lower contain-

ment spray header allowed water which had leaked into this normally vented ;

area to remain until it evaporated. Subsequent visual inspections I

revealed most of the nozzles on the 170 circumferential section of the

lower header to be 30% clogged with rust. Additionally, loose rust

particles were noted in the spray header which could affect system

operability during initiation of the sprays. A CAQR (Condition Adverse to

Quality Report) BFP880052 was written by the licensee to address the

concerns of this problem. During the initial discovery of this problem by

the licensee no Licensee Reportable Event Determination (LRED) action was l

initiated as required by plant procedures, j

The shif t supervisor was involved in the reportability determination on

the rust in the containment spray header. When the condition was

initially discovered the shift supervisor was involved in securing the

leak and removing the water frw the header. The shift supervisor or STA,

due to these distractions, did not initiate a LRE0 on the discovery date.

When the CAQR was reviewed by the Plant Operating Review Staff (PORS), a

copy was taken to the shift supervisor on February 3,1988, at 10:15 a.m.

The shift supervisor still did not initiate a LRED because the CAQR did

not provide any additional information to him. A LRED was initiated on

March 4, 1988, after discussing the condition with the resident inspectors

on March 3,1988. The shift supervisor determined the condition as

described to him was not reportable.

-.

30

A review of the CAQR by the resident staff indicated that an LRED should

have been initiated no later than February 3,1988. The CAQR states

explicitly that unit operability was affected. Although, there were

apparent disagreements within the TVA staff as to system operability or

affects, a LRED is still required to be initiated by Plant Managers

Instruction 15.4 (Unique Reporting Requirements) Section 4.3. Failure to

initiate an LRED to determine reportability or generic applicability is a

second example of the violation against 10 CFR 50, Appendix B, Criterion V

(260/88-05-03).

13. Q-List

a. The inspectors reviewed the status of the Q-List imalementation

program which was effective February 26, 1988, for Unit 2. The CSSC

list is no longer effective for Unit 2. The following concerns were

identified to the plant management in regards to this initial

inspection.

(1) Preliminary review of TVA design documents (e.g. 8FEP PT 87-52,

Development and Control of Browns Ferry Nuclear Plant Unit 2

Phase I Q-List, and 50SP-3.10) and discussion with TVA 3rogram

management shows that the Phase I program does not ful'y meet

10 CFR 50, Appendix B requirements for all required equipment.

(2) Generic Letter 83-28 requires the licensee have available for

review an equipment classification program which includes the

broader class of structures, systems, and components "important-

to-safety" required by GDC-1 (defined in 10 CFR Part 50,

Appendix A, "General Design Criteria, Introduction"). A request

by the inspector for this data indicated this Generic Letter

item had not been completed.

(3) No data was available to indicate how the General Electric

contractor who developed the Q-list justified the inclusion or

exclusion of equipment from the list. Concerns exist that the

source data used to compile the list may not be all inclusive.

For example as an outcome of GE's work, leading to the develop-

ment of Unit 2 Phase I Q-List, the Standby Liquid Control System

was designated as having primary functions that are nonsafety-

related in which only a few components of that system are

safety-related.

(4) Items listed as non-safety related on the Q-list do not require

any 10 CFR 50, Appendix 8 controls during any facet of its

maintenance or operation. This includes items required to be

operable by the plant TSs. An example is the Standby liquid

Control System which has been dropped from all QA requirements

as indicated by TVA program management. The inspector requested

justification for this analyses. Also, a listing of all

equipment "Important to Safety" should be developed and

. _ . - _ .. . - . , . - _- .. .

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31

available for review by the NRC of all equipment not currently

under any QA controls.

(5) The licensee's memorandum which 3romulgated the Q-List (R01

870805 935) stated that the orig:nal objective of the Q-List

Project was to eliminate the Critical Structures Systems and

Components list (CSSC list) with the initial issue of the

Q-List. It further stated that due to scope restrictions on the

program's input sources, the CSSC list would be retained until

completion of the Phase II effort. In spite of this, the Unit 2

CSSC list was cancelled with promulgation of the Phase I Q-List.

(6) The same memorandum referenced in (5) above documented that

ancillary supporting devices for components which were

determined to be safety-related for the Control Rod Drop

Accident and the Refueling Accident were not included in the

Q-List source documents. The phase II activity will classify

and include these devices.

(7) The FSAR Chapter 14 analysis of "special events" was not

included as part of the Q-List development process. Licensee

representatives indicated that this deficiency was largely

responsible for the lack of inclusion of any reactivity control

functions of the SLC system in the Q-List.

(8) The following documents contained contradictory lists of the

source documents used to develop the Q-List.

(a) The BFNP Nuclear Performance Plan, Section 14.1.2

(b) Q-List promulgation memorandum (R01870805 935)

(c) Site Directors Standard Practice 3.10, Use of the Q-Lists

(d) DNE Project Instruction 87-52, Development and Control of

the BFNP Phase I Q-List

(e) "Unit 2 Phase I Q-List" Notes associated with Drawing 47A

302-1, Rev 000

In summary, the Q-List cannot be relied upon to ensure that all

components or systems which are safety-related or important to safety

are included. The reasons for this conclusion are:

(1) Source data used to compile the list were restricted to the

Design 8tseline Program and the Environmental Qualification

Program. These were not sufficient to ensure integration of the

complete TVA QA program for BFNP into the Q-List.

(2) The FSAR Chapter 14 Analysis of "Special Events" was not

evaluated.

(3) Two Design Basis Accidents (DBA's) were not properly evaluated

to determine all ancillary support devices as was done for the

other DBA's

32

Deficiencies

be tracked as anidentified withItem

Unresolved the(259/260/296

Q-list prog /88-05-07). ram implementation w

b. Program Description

The Browns Ferry Phase I Q-List was developed to satisfy the

regulatory requirements contained in NRC Generic Letter 83-28 and

10 CFR 50, Appendix 8, to implement the commitments made in the BFN

Nuclear Performance Plan and partially implement NQAM Part I, Section

2.2, ID-QAP-2.7, "Q-List". The references, assumptions, definitions

and procedures used by DNE to develop the Phase I Q-List and the

procedures DNE will use to control and revise the Q-List are

contained in BFEP PI-87-52. One of the objectives of the Q-List

program is to replace the CSSC listing contained in Standard Practice

BF-1.11 with the Q-Lists and cancel BF1.11. The Q-Lists are

unitized. Therefore, as each unit Q-List is issued it will supercede

and replace the CSSC listing for that unit. It should be noted that

the Q-List program used design output documents to produce the

Q-List. The Q-List is a listing of safety-related systems, equipment

and components with their identifications and classifications derived

from design output documents. The Q-List was primarily produced to

satisfy NRC regulatory requirements to have such a listing for

centralized identification and control only. The benefit of the

Q-List is the ability of a number of BFN organizations to use the

Q-List to assist users in system, equipment and component classifi-

cations required by, as an example, design, procurement, modifi-

cation, inspection and special program efforts. Q-Listdevelopment

and implementation will be in two phases. The Phase I Q-List does

not list any limited QA or special management controls components or

non-safety related components in safety-related systems. The Phase I

Q-List provides partial fulfillment of the requirements of 10 CFR 50,

Appendix B, Criterion II and NQAM Part I, Section 2.2, ID-QAP-2.7.

The Phase II Q-List will expand and improve the Phase I Q-List and

will provide complete fulfillment of the requirements of 10 CFR 50,

Appendix B, Criterion II and NAQM Part I, Section 2.2, ID-QAP-2.7 and

will be final and complete BFN Unit 2 Q-List.

The Unit 2 Phase I Q-List included compilation and analyses of data

from the following sources:

- Safe Shutdown Analyses

- System Retirement Calculations

- Design Basis Commitments / Requirements Tracking System

- BFN-2 Final Safety Analysis Report (FSAR)

- Component Master List (CML)

- Environmental Qualification Walkdown Team (EQWT) 81/82

Analyses Block Diagrams

- EQWT 81 Analysis Block Diagrams - ECNs Pending

- Torus Integrity Long-Term Program, Plant Unique Analysis

Report

-- . - .- -_. - - - - - _

33

.

- Master Component Electrical List (MCEL)

- Mechanical Control Diagrams

c. Definitions of-Terms

(1) Q-List - This document is a design output unitized list of

aermanent plant systems, structures and components that have

3een identified as being SR, non-SR or Q-L. For BFNP, the

Q-List stipulates for each item the safety classification, main

and safety functions, description and source document. The

Q-List is a verified and controlled equipment listing derived

from design output documents developed to satisfy the NRC

regulatory requirements contained in NRC Generic Letter 83-28

and 10 CFR 50, Appendix B and the TVA commitments contained in

NQAM Part I, Section 2.2,10-QAP-2.7. BFNP Q-List development

and implementation will be in two phases.

(2) Phase I Q-List - For BFN, the term which identifies the initial

Q-List development and implementation. The Phase I Q-List

includes Unit 2 safety-related components and a limited number

of non-safety related components with the UNID, system

description, functions and classification identified. The

Unit 2 Phase I Q-List does not contain any limited QA components

(Q-L) and the majority of NSR components located in SR systems.

The Phase II Q-List Program will incorporate these components.

The Phase I Q-List will provide aartial fulfillment of the

requirements of 10 CFR 50, Append'x B, Criterion II and NQAM

Part I, Section 2.2, ID-QAP-2.7.

(3) Phase II Q-List - For BFN, the term which identifies the final

Q-List development and implementation that will expand and

improve the Phase I Q-List to incorporate the limited QA

components and NSR components in SR systems. The Phase II

Q-List will provide the complete BFN Unit 2 Q-List and complete

fulfillment of 10 CFR 50, Appendix B, Criterion II and NQAM Part

I, Section 2.2, ID-QAP-2.7. The Phase II Q-List is scheduled

for issue in FY 89.

l

(4) Safety-Related (SR) - Each item listed on the Unit 2 Phase I

Q-List is identified as Safety-Related or Non-safety related

depending on the item's function and/or requirements. Safety-

Related items are those structures, systems, and components

necessary to ensure: (1) integrity of the reactor coolent

pressure boundary, (2) capability to shutdown the reactor and

maintain it in a safe shutdown condition, or (3) capability to i

prevent or mitigate the consequences of accidents which could '

result in potential offsite radiation exposures com3 arable to

guideline exposures of 10 CFR Part 100 which are included in the i

j

Q-List and classified as safety related due to requirements '

resulting from TVA commitments. )

Q

1

34

(5) Non-Safety-Related (non-SR) - Those items which are not safety-

related and have no limited QA requirements.

(6) Limited QA Program (Q-L) - Refers to a system of special

management controls which are applied to ONP special programs

and special features in order to ensure that they are

appropriately controlled.

14. Regulatory Performance Improvement Program Items

In a letter from R. L. Gridley of TVA to J. N. Grace, dated March 18,

1986, the licensee proposed action to close Confirmatory Order EA 84-54

issued on July 13, 1984, which pertains to the implementation of

commitments made by(RPIP).TVA

Improvement Program The TVAinstatus

the Browns Ferryshort-term

of remaining Regulatory andPerformance

long-term open RPIP items is provided in Appendix A of the Browns Ferry

Nuclear Performance Plan (NPP), Volume 3, Revision 1. The last inspection

status of these items is described in Inspection Report 86-32. A

following inspection on selected open RPIP items was performed with

emphasis placed on short-term items. Current status is shown below:

Short-Term

Item I-3.4 Training for Craftsmen

The required licensee action is to continue with procedure

and systems training for craft personnel. This item

remained open because the specialized training procedure

BF-PMI-4.3 did not specify the frequency of retraining.

The inspector reviewed the licensee's lesson plans and l

schedule for craft training on several systems operation l

and procedures. The latter course has been changed to a l

Conduct of Maintenance course. Course content was found to l

be comprehensive and thorough. Training reported.that most i

of their craft people (greater than 90 percent) in the

disciplines of I & C, electrical, and mechanical have

completed the initial courses. Further, training initiated

a 3-day pilot systems refresher course in January 1988.

This course was well received by the employees and the

licensee plans to proceed with this program. Per SDSP-4.7,

the licensee also implements continuing training for annual

craft personnel. This includes a reading list for all

personnel to review which documents procedure changes. The

inspector concluded that the retraining frequency for craft

is not a concern and this RPIP item can be closed.

(CLOSED)

1

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i

Item - 3.5 Systems Engineering

The specific action and intent of this item was to assign

system responsibility to section engineers and provide

specific training to each. This item was -addressed -in

Inspection Report 86-32 and remained open because BF-PMI-

6.4 does not specify frequency of retraining. Current PMI

4.3, retraining frequency for this item is specified "as

required by supervisor.' The inspector reviewed BF-PMI

4.3, Specialized Training, Attachment 1, Item 1.3 and

required systems training for all engineers. The following

courses are required: Systems 101, BWR - Systems for

Engineers; QAT 003, Regulatory Requirements; EGT 121;

Technical Specifications, Plant Modifications, Work

Control; and EGT 119, Print Reading and Reference Material.

Standard Practice BF 4.14 requires all technical staff and

managers newly assigned to the site to complete the above

orientation phase training within the first 18 months

assignment and before being allowed full unreviewed respon-

sibility in areas concerned with the training to topic.

8F 4.14 also requires all incumbent technical staff and

managers to have completed the orientation 3hase training

by August 1,1987, or have an approved wa ver for the

training topics. The inspector reviewed the training

status for the affected technical staff and managers. A

large number of the technical staff were noted to have

completed the training and the licensee informed the

inspector that only two incumbent individuals did not

satisfactorily complete the training as required by

August 1,1987. These two were in repeat training at the

time of inspection. The inspector also reviewed selected

training records for three engineers to confirm that

engineers completed the training or had an approved waiver.

The records were in order and no discrepancies were found.

In addition, each section supervisor or manager receives

Regulatory Compliance Supplemental Training Bulletin to

review on a monthly basis. The type of information that is

provided includes issued CAQRs, NMRG findings, Radiological

Incident Reports, and NRC Inspection Report Summaries.

Finally, the cognizant engineer concept described in the

RPIP was replaced by a system engineering organization.

Based on the above, the inspector concluded that the

licensee is providing appropriate training for the tech-

nical staff and that a specific retraining frequency need

not be provided in the PMI 4.3. The inspector considers

this RPIP item closed. (CLOSED)

Item - 3.7 Unreviewed Safety Question Determination (USQD)

Preparation Training

-.

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36 ,

l

The required licensee action for this item was to provide I

training in preparation of safety evaluation techniques to

those individuals assigned such duties. This item was

addressed in Inspection Report 86-32 and remained open i

because BF PMI 4.3 did not require this training for  !

licensing staff or other engineers processing USQDs. l

During this inspection and a previous look in October 1987, a

the inspector was informed that the Site Licensing person-

nel are inc.luded in the USQD training and GF PMI 4.3,

Revision 2, Item 1.4 requires all engineers preparing USQDs

to receive USQD training. Towards ~ the end of 1987,

approximately 100 TVA personnel have taken and passed USQD

training. The Training Department plans to administer more

training sessions. Further, Section 5.1 of procedure PHP

0604.04, Revision 1, Evaluation of Changes, Tests, and

Experiments, states that the organizations assigning

individuals to approve screening reviews or perform safety

evaluations shall ensure that the individuals are techni-

cally qualified and are trained in accordance with Section

6.6 of the same procedure. The inspector considers this

RPIP to be closed. (CLOSED)

Item - 3.8 Quality and Compliance Awareness

The required licensee action for this item was to ensure

that the craft personnel understand their responsibility

regarding quality awareness and regulatory compliance. In

response to this item the licensee is providing G.E.T.

training (i.e. , G.E.T.4, QA/QC and G.E.T.6, Plant Pro-

cedures) and R.P.I.1.383 Regulatory Compliance training

for all personnel. R.P.I. 1.384 is designated as a

retraining course for regulatory compliance. Per BF

PMI-4.3, Item 1. 7, foremen are required -to receive

retraining on regulatory compliance annually and all other

employees on a biannual basis.

The inspector reviewed the licensee's implementation of

administering quality and compliance training to craft

personnel. Tracking records were reviewed for 12 arbi-

trarily selected craft personnel, including machinists,

steam fitters, electricians, instrument mechanics,. foremen,

and general foremen. Of the sampled training records, the

inspector found three foremen, one general foreman and one

craft person delinquent in receiving regulatory compliance

retraining (R.P.I.1.384), and three craft people delin-

quent in receiving GET 4 QA/QC retraining. For some people

the ' Straining delinquency was greater than one year. The

finding was essentially the same to a similar inspection

performed back in October 1987. This constitutes a

violation in that the licensee did not adhere to the

frequency retraining requirements specified in PMI-4.3,

._

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37

Item 1.7 and PHP . 0.202.04, Nuclear Plant General Employee

Training (GET) Program (259,260,296/88-05-08).

In response to the above finding the licensee has stated

that they recognized this problem and plan to have correc-

tive actions implemented by June 1,1988. To improve -

timeliness -of retraining, the training de3artment is

developing's

individual retraining / qualification dates to fall duea training cyc

around the same time each year. Furthermore since there~

is overlap of course content in G.E.T. 4, QA/QC; G.E.T. 6,

Plant Procedures and Instructions; and R. P. I. 1.383,

Regulatory Compliance, the training department plans to

combine all three. of these into one course. This RPIP

remains open. (OPEN)

Item 4.5 Organization

Licensee action for this item is to review and revise

appropriate documents to define and describe the new-

organization.

The inspector reviewed the following documentation to

determine if they accurately reflect the new corporate and

BFN site organization: TVA Topical Report, TVA-TR75-1A,

Revision 9; BFN TS; Nuclear Performance Plan (NPP) Volume 3

Revision 1, Figures II-1 and II-2; Site Director's Standard

Practice (SDSP) No.1,- Site Organization; SDSP-22.2,

Emergency Response Organization; Standard Practice BF 3.12,

Organization; the Office of Nuclear Power, Nuclear Quality

Assurance Manual (NQAM) Part I, Section 2.1; and the Final

Safety Analysis Report. The ins 3ector found the TVA

Topical Report, BFN TS, and NPP Vol. 3 organization figures

to be out-of-date. Also, the inspector did not find any

information in SDSP Section No. 1, entitled Site

Organizaton. The BFN site reorganized at the plant

manager's level in February 1988. This item will remain

open pending the revision of at least the TVA Topical

Report and BFN TS documentation. (OPEN)

Item I-4.11 Independent Safety Engineering Group

Engineering Group (ISEG). TheLicensee action

current site is to

ISEG) establish a Site I

ccmposi-

tion consists of the supervisor and a full-time engineer. l

The licensee's 31anned three staffing level calls for three  ;

full time engineers. Per diseussion with the ISEG l

supervisor the 4 1spector was informed that his department ,

is attempting to obtain two more full-time engineers. The

BFN ISEG staffing level and experience was previously l

addressed in Inspection Report 87-14. The inspector was j

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informed that TVA plans to respond to this concern. Thus j

RPIP item remains open pending satisfactory resolution. J

(OPEN) I

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