ML20197J144

SALP Rept 50-320/85-99 for May 1984 - Feb 1986

ML20197J144
Person / Time
Site:	Three Mile Island
Issue date:	04/03/1986
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20197J131	List: ML20197J126 ML20197J144
References
50-320-85-99, NUDOCS 8605190379
Download: ML20197J144 (47)

See also: IR 05000320/1985099

Text

o.

...

1

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE

GPU NUCLEAR CORPORATION

THREE MILE ISLAND UNIT 2

MAY 1, 1984 to FEBRUARY 28, 1986

-;

April 3, 1986 ,

-

l

.

k

.m -

8605190379 860508  :

PDR ADOCK 05000320 E.

.G_ PDR , ta

..;;.

l

.. . - . . . - . - . .-.. - , - - , - - - - - - - - . . . . - - - . . -- - - -- -- - - -

. . , .

. .

.

TABLE OF CONTENTS

.Page

I. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . . . 1.

A. Purpose and Overview . . . . . . . . . . . . . . . . . 1

B. SALP Board Members . . . . . . . . . . . . . . . . . . 1

C. Background . . . . . . . . . . . . . . . . . . . . . . 2

t

II. CRITERIA. . . . . . . . . . . . . . . . . . . . . . . . . . 5

III. SUMMARY OF RESULTS. . . . . . . . . . . . . . . . . . . . . 7

A. Facility Performance . . . .............. 7

B. Overall Summary. . .................. 7

4

IV. P ER FO RMAN C E ANA LY S I S . . . . . . . . . . . . . . . . . . . . 9

A. Shutdown Plant Operations /Defueling Preparation. . . . 9

B. Radiological Controls. . . . . . . . . . . . . . . . . 14

4 C. Effluent Monitoring and Control . . . . . . . . . . . . 18

D. Quality Assurance. . . . . . . . . . . . . . . . . . . 20

E. Maintenance. . . . . . . . . . . . . . . . . . . . . . 23

F. Design, Engineering and Modifications. . . . . . . . . 25

G. Emergency Preparedness . . . . . . . . . . . . . . . . 28

H. Security . ......................29

I. Licensing Activities . . . . . . . . . . . . . . . . . 30

j V. SUPPORTING DATA AND SUMMARIES . . . . . . . . . . . . . . . 32 I

A. Licensee Event Report Tabulation and Causai Analysis . 32

B. Investigation Activities . . . . . . . . . . . . . . . 32

C. Escalated Enforcement Actions. . . . . . . . . . . . . 33

D. Management Conferences During the Assessment Period. . 33

1

TABLES

Table 1 -

Inspection Hours Summary. . . . . . . . . . . . . .T1-1

Table 2 -

Inspection Activities . . . . . . . . . . . . . . .T2-1'

Table 3 -

Enforcement Data. . . . . . . . . . . . . . . . . .T3-1 .

-

J

Table 4 -

LER Synopsis. . . . . . . . . . . . . . . . . . . .T4-1

Table 5- --

Summary of Significant Licensing ' Actions and

Supporting Activities . . . . . . . . . . . . . . .TS-1

. . . . . - -. - -

. .- .-,- . .-

.

.

I. INTRDDUCTION

,

A. Purpose and Dverview

The Systematic Assessment of Licensee Performance (SALP) is an

integrated NRC staff effort to collect the available observations

and data on a sampling and periodic basis and to evaluate licensee

performance based upon this information. The SALP is supplemental

to normal processes used to ensure compliance to NRC rules and

regulations. It is intended to be sufficiently diagnostic to

provide a rational basis for allocating NRC resources and-to provide

meaningful guidance to the licensee's management to promote quality

and safety of plant operations and modifications.

An NRC SALP Board, composed of the staff members listed below, met

on April 3, 1986, to review the collection of performance

observations and data to assess the licensee's performance in

accordance with the guidance in NRC Manual. Chapter 0516, " Systematic

Assessment of Licensee Performance." A summary of the guidance and

evaluation criteria is provided in Section II of this report.

This report is the SALP Board's assessment of the licensee's safety

performance at the Three Mile Island (TMI) Nuclear Station, Unit 2

for the period May 1, 1984 through February 28, 1986. This report

takes into account the unique mode of operation for defueling and

the necessary preparations required for safe movement of fissile and

radioactive waste material generated as a result of the March 1979

accident. Some of the areas rated have not been previously

evaluated. However, they were incorporated to make the SALP

evaluation more reflect ive of the licensee activities at TMI-2 and

the utility's responsibilities.

B. SALP Attendees

Chairman:

T. T. Martin, Director, Division of Radiation Safety and Safeguards

(DRSS)

Board Members

,

S. D. Ebneter, Director, Division of Reactor Safety (Part-time)

W. D. Travers, Director, TMI-2 Cleanup Project Directorate

W. J. Johnston, Deputy Director, DRS

W. F. Kane, Deputy Director, DRP

R. R. Bellamy, Chief, Emergency Preparedness and Radiation

Protection Branch, DRSS

C. J. Cowgill, Chief, TMI-2 Project Section

M. T. Masnik, Project Manager, TMI-2

R. J. Cook, Senior Resident Inspector, TMI-2

).

, ,, , . . - - .-, .

_

.

2

Other NRC Attendees

J. M. Bell, Senior Radiation Specialist, TMI-2

T. A. Moslak, Resident Inspector, TMI-2

C. Background

(1) Licensee Activities

'

At the beginning of the period, the reactor plant was in cold

shutdown. The reactor vessel head was installed and shutdown

margin was being maintained by boron in the reactor coolant.

Major licensee activities centered around preparing for reactor

vessel head lift and conducting extensive decontamination

activities in both the reactor and auxiliary buildings.

By early July 1984, the reactor vessel (RV) head closure studs

were detensioned and removed without complication. The head

was lifted on July 24, 1984 and positioned on the storage stand

at the south end of the 347' elevation in the reactor building.

The lift was controlled from a shielded control station located

on top of the "A" D-ring. Dose levels were less than

anticipated (about 3R/hr) and contamination was controlled.

Some problems were experienced with polar crane malfunctions .

and limited reach capability during the head lift and work

platform and internal indexing fixture (IIF) installation.

During August 1984, the Reactor Building Polar Crane was taken

out of service for routine maintenance. During this

maintenance the licensee discovered that one of two redundant

crane brakes was inoperable due to a loose part on the hand

release mechanism. The crane was restored to service after

maintenance. Subsequent review identified the fact that the

hand release mechanism modification (installed in 1982) was not

made in conformance with licensee procedures. As a result, the

crane was again declared inoperable in September pending a

complete review of all maintenance and modification activities

performed on the crane. During the fall of 1984, GPU conducted

a thorough inspection of the polar crane and conducted a

complete review of all crane modifications.

In mid-Novemoer 1984, four hydraulic jacks equipped with

mechanical followers were installed around the reactor vessel

upper plenum. On December 6,1984, the plenum was jacked 2

inches without binding or measurable increases in reactor

building radiation levels. The licensee inspected the plenum

using television cameras. Later in December the assembly was

jacked to 7 inches.

)

I

_ .

3

On January 9,1985, after extensive review, the NRC approved

use of the polar crane to its load rating capacity of 170 tons.

On February 20, 1985, a small camera was lowered into the lower

reactor vessel to perform examinations below the core support

structure. The pictures showed debris in the form of a gravel

pile. Unlike the upper core rubble, the material in the lower

reactor vessel head was not identifiable.and appears to consist

of previously molten core material. Most rubble pieces

measured from two to four inches long and about half as wide. A

few larger pieces, however, appeared to be eight inches or more

in length. The debris in the lower vessel head is estimated to

be about 30 inches deep and contain approximately 15 to 20

tons.

On May 17, 1985, the licensee removed the reactor vessel plenum

and placed it in storage underwater in the deep end of the fuel

transfer canal. No significant problems were experienced

during the transfer.

The NRC licensed five individuals as Fuel Handling Only Senior

Reactor Operators in October 1985. This followed an extensive

training program conducted by the licensee over the previous

nine months.

The licensee experienced QA problems with the manufacture of

filter, knockout, and fuel shipping canisters. This required

the licensee to take additional measures to assure that

canisters were fabricated to design. specifications. As a

result, the initiation of defueling slipped approximately two

months. The licensee has ordered canisters from two other

manufacturers in response to these problems.

To date, canisters have been loaded using a pick and place

technique for identifiable pieces such as end fittings, spiders

and fuel pins. The bulk of the remaining debris removed from

the core has been loaded using a hydraulically operated spade

bucket. Visibility in the reactor vessel is poor (2-3 inches

using T.V. cameras) because of biological growth in the reactor

coolant system. The licensee continues to evaluate methods for

eliminating these organisms.

(2) NRC Activities

NRC oversight and inspection is performed by a staff of two

resident inspectors, three radiation specialists assigned from

the Region I office and a.Section Leader,and three project

engineers from the office of Nuclear Reactor Regulation. There

is a Senior NRC manager onsite who is_ responsible for overall

4

.

c.

4

coordination of NRC activities. In addition, there is a

Region I Section Chief assigned onsite. Periodic specialist

inspections were conducted in security and modification activi-

. ties and to evaluate specific problems that have surfaced.

'

NRC continues to approve certain operational and maintenance

procedures as well as Safety Evaluation Reports -for major

.defueling activities. Senior site NRC management changed

.

'

during the period. The NRR Project Director changed in August

1984 and the Region I Section Chief changed in December 1984. .

1

N

.

1

4

'

.)

l

.

1

l

' 'l

i

,

,

I

i

i

'

i:

'

!

__

G

'

5

II. CRITERIA

The following evaluation criteria were applied to each functional area:

1. Management involvement in assuring quality.

2. Approach to resolution of technical issues from a safety standpoint.

3. Responsiveness to NRC initiatives.

4. Enforcement history.

5. Reporting and analysis of reportable events.

6. Staffing (including management).

7. Training effectiveness and qualification.

To provide consistent evaluation of licensee performance, attributes

associated with each criterion and describing the characteristics

applicable to Categories 1, 2, and 3 performance were applied as

discussed in NRC Manual Chapter 0516, Part II and Table 1.

The SALP Board conclusions were categorized as follows:

Category 1: Reduced NRC attention may be appropriate. Licensee

management attention and involvement are aggressive and oriented toward

nuclear safety; licensee resources are ample and effectively used such

that a high level of performance with respect to operational safety or

construction is being achieved.

Category 2: NRC attention should be maintained at normal levels.

Licensee management attention and involvement are evident and are

concerned with nuclear safety; licensee resources are adequate and are

reasonably effective such that satisfactory performance with respect to

operational safety or construction is being achieved.

Category 3: Both NRC and licensee attention should be increased. ,

Licensee management attention or involvement is acceptable and considers i

nuclear safety, but weaknesses are evident; licensee resources appeared i

strained or not effectively used such that minimally satisfactory I

performance with respect to operational safety and construction is being j

achieved.

)

l

The SALP Board also assessed each functional area to compare the  !

licensee's performance during the last quarter of the assessment period

'

to that during the entire period in order to determine the recent trend

for each functional area. The trend categories used by the SALP Board

are as follows:

Improving: Licensee performance has generally improved over the last

quarter of the current SALP assessment period.

Consistent: Licensee performance has remained essentially constant over

the last quarter of the current SALP assessment period.

--

- . .

_

- _ . ,_ . .

?.x

.

6

Declining: Licensee performance has generally declined over the last

quarter of the current SALP assessment period.

4

I

')

.

O l

'

7

III. SUMMARY OF RESULTS

A. Facility Performance

Functional Area Category Category Recent

Last Period This Period Trend

(October 1, 1981 - (May 1, 1984 -

September 30,1982) February 28,1986)

1. Shutdown Plant Operations /

Defueling Preparations 2 2 Improving'

2. Radiological Controls 1 1 Consistent

3. Effluent Monitoring and Control 1 1 Consistent

4. Quality Assurance 1 1 Consistent

5. Maintenance 2 1 Consistent

6. Design, Engineering and

Modifications 2 2 Consistent

7. Emergency Preparedness 1 No Basis- No Basis

8. Security 1 No Basis No Basis

9. Licensing Activities 2 1 Consistent

B. Overall Summary

This assessment is based on licensee performance over a period when

complex and, in some cases, unprecedented activities have been underway

to recover from plant conditions created by the March 1979 accident.

Activities during this period have been primarily directed towards

initiating reactor vessel defueling and on50ing decontamination of

building and equipment surfaces. In order to evaluate licensee

performance relating to defueling and decontamination, NRC assessment has

focused on activities including; personnel training, management controls,

plant modifications, radiological controls, maintenance, quality-

assurance and_ operations.

Overall, the licensee has carried out its cleanup and shutdown activities -

in a safe and technically competent manner. The licensee's emphasis on

safety has been demonstrated by a conservative approach, and a. generally

high degree of management involvement in TMI-2 issues. Licensed Fuel

Handling and Control Room operators have carried out their responsibili-

ties effectively and professionally. Some difficulty, early in the

assessment period, was experienced in conducting major evolutions. 'These

.

'

8

problems can be attributed to some inattention to detail which, as

demonstrated by more recent activities, has been improved upon by the

licensee.

As a result of the accident, radiation protection has presented a complex

and difficult challenge at TMI-2. The licensee has an appropriately

large number of resources dedicated to protecting workers and the

environment from radiation. The radiological controls organization has

performed effectively as evidenced by; 1) substantial progress in the

decontamination of building and equipment surfaces, and 2) the low

radiation doses incurred by cleanup workers.

A problem area which warrants improved licensee attention involves the

preparation of detailed procedures for carrying out cleanup activities.

Due to the unique nature of the cleanup, the licensee is required to

submit some detailed procedures to the NRC for review and approval. The

number of flawed procedures initially disapproved by the NRC indicates

that the licensee's review process is not effective in assuring all

procedural details are correct.

The site quality assurance (QA) department remains strong in their

involvement in all functional areas. The QA department has been

particularly effective in providing oversight for maintenance and design

engineering activities. The department appears to be a significant factor

in the success of major maintenance activities and in the design of

specialized modifications and tooling required for defueling. The multiple

levels of review incorporated into the QA organization are expected to

provide adequate support for remaining cleanup activities.

The licensee's engineering organization is an integrated staff composed

of engineers from the licensee and Bechtel in Gaithersburg, Maryland.

Early in the assessment this staff experienced problems conducting

adequate reviews of planned plant modifications. This was attributed to

d general lack of understanding of the licensee's administrative review

requirements and certain applicable regulatory requirements. The licensee

implemented an aggressive training program to correct this deficiency. NRC

observations later in the period showed subst&ntial improvement in this

area.

The licensee has been successful in carrying out cleanup and shutdown

operations safely and has generally operated in conformance with NRC

regulatory requirements.

l

l

l

l

i

i

1

)

.

.

IV. PERFORMANCE ANALYSIS

A. Shutdown Plant Operations /Defueling Preparation

Analysis-

a. Shutdown Plant Operations

Shutdown operations have been under continuous scrutiny by NRC. The

-plant is in a shutdown recovery mode with highly borated primary-

coolant at ambient temperature. Incore temperature is monitored by

the remaining operable incore. instruments. The Reactor Coolant

System is vented to the reactor building atmosphere with the. reactor

vessel head.and plenum assembly removed.from the reactor vessel.

Reactor coolant system cooling is by natural heat loss to the

reactor building atmosphere. A defueling platform is installed over

the reactor vessel an.d defueling operations are in progress. The

licensee has~ established two command centers located outside the

control room. One controls the reactor building. evolutions and the

other controls fuel handling activities.

The licensee's major shutdown operation emphasis has been

maintaining reactor coolant system water chemistry and level,

containment-integrity, controlled and assayed liquid discharges,

prevention of uncontrolled airborne releases, minimizing and

controlling personnel exposure, processing of. contaminated water,-

performance of routine plant surveillar,ces, maintaining security,-

maintaining industrial safety, and plant housekeeping. Two

violations were identified in this area during the assessment

period. The violations were assessed by the NRC to be isolated

~

cases of operator error and were not indicative of a programmatic

breakdown in procedural control.

The plant operators are well tra'ined and' qualified and perform their:

duties in a professional manner. However, one problem with operator

training was identified early in the assessment period. During

operator licensing examinations, weaknesses were noted in

candidates' overall knowledge and understanding of administrative -

and operational procedures.- An examination conducted 'later in the

-

period noted a marked improvement in candidate knowledge in these

areas. This improvement is-attributed to the--increased licensee '

management attention to operator training p.rograms.

The control room is maintained quiet' and orderly to provide an

appropriate atmosphere for monitoring-plant parameters and

evolutions. The control room operators monitor and interface with

the command centers which have direct control of' operations in the.

reactor building. In general, these interfaces have been smooth and-

.

no problems have been identified. The Command. Center and the Fuel-

)-

._ . - . _ - _ _ . . _ . . _ . _ .. -_ ._ - . . _ . _ _ __-

.

l -

10

i ,

'

Handling Control Center have been maintained in an appropriate

manner. Personnel access is restricted and operator communication

j_ i s formal.

NRC reviewed the licensee's surveillance testing program routinely

during the evaluation period. Licensee performance continues to be

i

excellent with a well established program. The program is implemen-

ted using a computerized scheduling system for routine surveillances

j and supplemented with a manual system for non-routine surveillances.'

j Weekly schedules are reviewed and used by each department in produc- '

ing their work lists to assure that all required surveillances are

performed in a timely manner. Daily planning meetings serve to  !

minimize conflicts in the scheduling and conduct of surveillances.  ;

i Surveillance procedures are detailed and consistent with the ,

licensee's administrative controls and current technical specifica-  ;

4

tion requirements.  !

'

l

Management has aggressively pursued identification and changes of

j those Technical Specifications (TS) which are no longer appropriate for

, present plant conditions and has received approval-from the NRC to

delete unnecessary requirements. Several systems which are no

longer required to maintain plant safety have been removed from the

4 TS. However, the licensee has maintained integrity and operability i

for some of these systems to maintain a backup capability for

j systems currently in use.

The licensee is required by Technical Specifications to submit

certain procedures with safety significance to the NRC for review

prior to implementation. A large number of these procedures have

been initially disapproved by the NRC. The reasons for NRC

j disapproval range from relatively minor editorial adjustments, to

major equipment and/or system line-ups deficiencies which have-a

potential safety impact, or result in violations of the Technical

j Specifications. The lack of procedural adequacy was addressed in

<

the previous SALP evaluation. Examples of the above problems

include one procedure that was submitted with mathematical errors i

and one procedure that would have violated Techr.ical Specification '

l requirements in that the procedure used a check valve for

containment isolation instead of a manual valve as required by the

Technical Specifications. Both of-these procedures had received the

'

required licensee reviews including those performed by the Safety

Review Group before being submitted to the NRC for review. These

i problems are indicative of'probTemsFiffthe licensee's ~techriical

~

2

ire _v.iew of procedures. 7 . Z .[ Z T _

.

E ' _11~ _ -

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

" * * * " " ' ' " ^ * ' ~~

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

'

- .-- .

e, m . . / =

-- .. - _. -__ -~ ~

~ = =e.ne- e s -e . - + . se

a.... - -

- .

i _. , _ , _ -

,

l'

,g - . w*r . - , - ,- = mimr-vre-w-ta----.-e-.- ----=".e,y,-.,,.- y .-se,-.c," *----av-=w v.r-*-.-+-M -

ww . mv-e-v e.-*.*-e-wr ms -v - --rm*=---*-wag-++ - - * = e-a- ~w +we *d- *wr.a-w*wr

. . _ _ _ _ _ _. _ __ . . _ _ _ - .

.

-

11

b. Defueling Preparation

The licensee's primary efforts during the evaluation period have

centered around preparations for and conduct of initial defueling

activities. Management involvement in all activities remains

strong. Senior management is involved in daily planning, project

status meetings, and makes frequent plant tours to maintain a close

working relationship with the technical staff. The licensee remains

committed to a safe expeditious cleanup of the facility. A lengthy

defueling checkoff was completed, reviewed and approved by senior

licensee management prior to the start of defueling. Technical and

Safety Advisory Groups-are still retained by the licensee to provide

independent review of activities. The licensee has a continuing

effort to review and upgrade existing procedures to reflect changing

needs. This review has identified a number of areas where errors or

conflicting requirements existed. The licensee has consolidated,

simplified, and streamlined the procedures and this effort has

expedited the defueling process.

The first major evolution leading to defueling was reactor vessel

head lift. During the preparations the licensee installed lead

blankets around the periphery of the head for shf aldf r.g. The

original hangers for the lead blankets experienced weld failure

before the lead blankets and hangers were completely installed.

Even with this experience, the licensee installed new hangers

without first verifying weld quality. This resulted in the need to

make an entry into the Reactor Building to verify weld quality after

blanket installation. During the head lift a number of relatively

minor problems were encountered which extended the time required to

complete this evolution. This rest.? ted, at least in part, from the

licensee's failure to utilize information gained from previous

experience moving the head.

The licensee conducted a detailed inspection of the polar crane

prior to each major lift and made repairs as necessary. NRC also

inspected the crane and identified several items that had been

missed in licensee inspections. These items ranged from loose

equipment on the crane to missing nut locking devices and loose

cables. While individual discrepancies were not significant, the

number of discrepancies identified were relatively large. The

i licensee subsequently assigned a systems engineer and revised their

preventive maintenance program including management review. These

improvements appear effective. (Additional discussion in

Maintenance, Section 5.) The above problems are indicative of a

lack of attention to detail during preparation for and conduct of-

major evolutions.

'

In conducting later major evolutions including plenum jacking, lower ,

reactor vessel head exploration and the subsequent plenum lift and

storage in the Fuel Transfer Canal, licensee preparations were more ]

thorough and showed appropriate attention to detail and resulted in

smooth evolutions.

i

,

l

]

.

.

12

1

i

The licensee conducted a vigorous training program to qualify Fuel

Handling Senior Reactor Operators (FHSR0s) to assure that they would

be licensed by the NRC to handle fuel. The training program was

closely monitored by the NRC, including witnessing reactor operation

at the Penn State University pool reacter. The licensee was

selective in choosing the FHSRO candidates with experience being a

key prerequisite. Five of the six candidates were granted-FHSR0

licenses by the NRC. The sixth individual was subsequently granted

a license after reexamination. Similarly, existing . control room

SR0s were trained in defueling operations during their

requalification program. NRC review indicated that the overall

program was conducted in a thorough, professional manner.

On October 30, 1985, the licensee initiated preliminary defueling

activities. These activities involved rearrangement of core debris

within the reactor vessel to allow complete installation and

uninterrupted rotation of the Canister Positioning System. The

licensee had in place an organization that included a Defueling

Director and engineering and radiological controls support to assist

the defueling crews in solving problems while defueling was in

progress. By the end of the evaluation period, 16,000 pounds of

core debris had been removed. NRC observation of initial defueling

efforts indicated that the licensee has taken a deliberate, cautious

!

approach to each new evolution. Operator performance and

_

! coordination of activities has been good.

Late in the period, the licensee reorganized the Unit 2 project

organizational structure to incorporate a newly formed Defueling

Group reporting directly to the Director of TMI-2. The new group is

managed by a Defueling Manager and consists of a Defueling

Operations Section, a Defueling Engineering Section and a Defueling

Support Section. The new group was formed to place management

emphasis on increasing productivity.

In summary, the licensee has safely maintained the plant in a safe

shutdown condition while preparing for and conducting initial

defueling. The licensee continues to have problems during technical

review of new procedures. The licensee has pursued conservative

solutions to technical problems which have arisen during initial

4

defueling and has conducted operations in a cautious manner. The

licensee has modified its organization as necessary to meet changing

needs.

Conclusion

i

Rating: 2

Trend: Improving

)

.

' -

13

Recommendations

Licensee: Review system for development and review of procedures to

ensure technical adequacy.

!{RC: Perform inspection of the licensee's procedure development and

review process,

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

.

14

B. Radiological Controls

Analysis

Radiation Protection and Waste Management are under continuous review

by siw radiation specialists. Additional specialist inspections were

performed by both site and region-based inspectors.

Radiation Protection

The licensee's radiation protection program has been effective

throughout the appraisal period. Management commitment to the

maintenance of a high quality program is evidenced by its active,

timely and decisive involvement. Worker concerns and suggestions

are accommodated by means of the Radiological Awareness Report (RAR)

and the Incident / Event Report (IER) systems. RAR and IER reviews

are generally timely and corrective actions appropriate. In

addition, radiological concerns are discussed at " Radiological

Awareness" meetings open to all workers. These meetings are

conducted about once per month.

The Radiological Controls Department (Rad Con) has been generally

effective in impleinenting the ALARA principle at the task level by

means of pre-work reviews, radiation work permits and personnel dose

followup. Examples of the application of this process include cork-

seal replacement work in the Control and Service. Building, valve

alley flushing (decontamination), and auxiliary building reactor pit

decontamination. Pre work reviews resulted in the specification of

l

' such things as appropriate protective clothing, training (including

mockups), monitoring, high radiation areas to be avoided based on

appropriate survey results, the use of local shielding and eye

,

,

protection for high energy beta. In the case of work in the makeup ,

i

valve rooms, analysis of pre-work survey data resulted in the use of

l

locally fabricated plastic hoods that resulted in the avoidance of i

substantial dose from high energy beta radiation. Post-job reviews

compared cumulative dose estimates with the cumulative dose

received, analyzed the causes of higher than expected doses and made

recommendations for improvements for future work. As an example, an

evaluation of conditions experienced during decontamination work in

the auxiliary building elevator pit resulted in recommendations for

improvements in tooling, exhaust air filtration, personnel shielding

for high energy beta, and respiratory protection / personnel stress i

reduction.  !

NRC reviews of the personnel selection, qualification and training

indicated that formalized training programs are being implemented

for Rad Con technicians, Rad Con Supervisors, and Rad Con Support-

Technicians. Additionally, the licensee is developing ALARA

training for engineers and managers. Positions are adequately

)

.

15

staffed and vacancies were filled with qualified experienced

individuals. The licensee has experienced a high turnover in

contractor technician positions primarily due to competition within

the industry for personnel'. When vacancies have existed, the

licensee has acted promptly to obtain qualified replacements and to

ensure that personnel are properly qualified and trained. Inspector

observations of formal training, interaction with plant workers, and

participation in special task briefings indicate that personnel at

all levels are aware of their responsibilities with respect to

radiation protection.

The quality of the licensee's radiation protection program is

evidenced by major. reductions in airborne, surface, and general area

radioactivity and radiation levels, accreditation of its personnel

I

dosimetry program by the National Voluntary Laboratory Accreditation

Program of the National Bureau of Standards, improved personnel

contamination detection and assessment (including testing and

installation of state-of-the-art personnel frisking devices),

upgrading of its respiratory protection program, and the development '

of special purpose radiation detection and measurement instruments.

The licensee has been particularly effective in minimizing personnel

exposures associated with defueling operations. This has been

accomplished by coordinated and complimentary efforts including; 1)

shielding design of the work platform, 2) general area decontamina-

tion, 3) training, and 4) continuous monitoring of radiological

conditions by health physics technicians. Radiation fields on the

defueling work platform, which has the highest worker occupancy in

the reactor building, are the lowest in the reactor building.

One significant licensee-identified radiation exposure event

occurred where three persons exceeded the licensee's administrative

skin exposure limit during an entry into- the Seal Injection Valve

room in the Auxiliary Building. NRC limits were not exceeded during

this event. The causes of this problem were inadequate radiological ,

engineering review, inadequate prejob briefing and mockup training '

and inadequate performance by a Rad Con technician. Licensee

investigation into the incident and subsequent corrective actions

were aggressive and thorough. Although there were major l

4

deficiencies associated with this event, NRC review of subsequent  !

activities in high radiation areas conclude that there was not an

overall breakdown in the conduct of the licensee's radiological- l

controls' program, i

l

Some ;*oblems have occurred which indicate that communications

between workers and Radiological Controls technicians could be

improved and that lessons learned from these problems could be more

effectively included in technician training. Examples include (1)

in August 1984, a worker received internal contamination because of I

poor communication between supervisors; (2) in May-1985, unclear

direction was provided with respect to decontamination of a steel .

I

l

l

..

, - - , ,

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

C

16

plate; and (3) in June 1985, a worker received internal

contamination in the Reactor Building annulus because of confusion

, over where work was to be performed.

Radioactive Waste Management

4

The licensee is implementing an effective radioactive waste

management program. The responsibilities and functions of the rad-

'

waste organization are adequately described in administrative proce-

dures. However, there was some confusion regarding communication

among the personnel involved in radioactive waste system operations;

1

shipping and packaging; equipment decontamination; and laundry and

respiratory protection maintenance because they reported to different

supervisors within the Plant Operations Department. An example of

this problem involves misclassification of a waste shipment. Three

separate groups did calculations to determine the classification of

the affected packages but the groups did not properly coordinate

their efforts and as a result the misclassification went unnoticed.

In responding to these problems, the licensee consolidated most of

the radwaste functions under one experienced manager reporting to the

Director of Site Operations in September 1985. This change has

resulted in better communications and smoother radwaste operations.

Subsequent review by the NRC has identified no similar problems.

I Personnel at all levels in the radwaste organization.have been

i

adequately trained and are qualified with regard to each individual's

4

functions and responsibilities within the organization.

A significant problem was identified by the licensee late in the

period in the assay of solid samples for Strontium-90 content. The

licensee was using unverified computer programs and did not have an

interlaboratory QA/QC program to detect analysis errors. This

resulted in underestimating the value of Sr-90-reported on some

waste shipments by a factor of two and caused one shipment to be

improperly classified (not related to the above problem). Licensee

action on this error once identified, was prompt, thorough and

should preclude similar errors in the future. The laboratory

program was examined by a contractor, and some additional steps

recommended. The licensee has begun a QA/QC interlaboratory

program.

The NRC routinely inspects radwaste shipments leaving the site. In

general, the shipment documentation is adequate and the shipments

are properly prepared. Licensee Operations Quality Assurance

reviews all shipments of greater than Type A quantities. No

i

significant problems have been identified.

Some radwaste operational problems occurred during this assessment

period: including; a resin liner becoming jammed in a concrete

shield; slight contamination of two workers while clearing a hose

line; and the incomplete solidification of one liner, requiring

extra handling of the liner and some additional exposure of workers.

4

- .

t

'

17

These problems were the result of personnel errors and in the case

of incomplete solidification, a_ deficiency in the contractor's

process. NRC review has concluded that the~se were isolated problems-

and not indicative of a programmatic breakdown.

Decontamination efforts by the licensee have resulted in improved

radiation exposure conditions and personnel access to large areas

within the plant, especially access to the chemical addition area,

the hallways of the 281' level auxiliary building, and the auxiliary

building elevator pit; the latter resulting.in restoration of the

elevator to general use. The waste management group has also helped

in the dose reduction effort by scabbling the floors of the reactor

building, and by shielding and decontamination-of the 281 fuel

handling building valve alley.

Radwaste Management has implemented a plan to reduce the amount of

radwaste stored outside and exposed to weathering. Stored radwaste

has been largely removed from the " paint shed" at_ South East Acres.

Radwaste packages are now being olaced in an enclosed Interim Solid

Waste Storage Facility. A radwsste handling and packaging facility

is under construction. This building will provide spaces for

decontamination, survey, compaction and disassembly of contaminated

items so as to reduce the amount of radwaste for burial.

'

Conc?usion

Ratirg: 1

Trend: Consistent

,

Recommendations

i

Licensee: The licensee should continue efforts to maintain

effective communications between supervisors and workers.

NRC: None

f

T

_ ._i_'__._._._.____ - . . . _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ . - _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ . _ _

'

,

,

n

a '

18

C. Effluent Monitoring and Control

Analysis

The NRC onsite inspectors routinely reviewed the licensee's Effluent

Monitoring and Control Programs. In addition, one inspection of the

Radiological Environmental Monitoring Program (REMP) was performed

by a regionally based specialist inspector. <

The licensee maintains a strong, well-managed REMP. ' Personnel are

knowledgeable, well trained and thorough. The quality of analytical

reasurements is maintained by a good quality assurance program. For

example, licensee contractor laboratories are required to participate

in the Environmental Protaction Agency (EPA) Cross-Check program and

do periodic duplicate analyfes'. In addition, the licensee splits

samples from certain Radiological Environmental Monitoring (REM)

sampling stations between two laboratories. The licensee partici-

pates in the International Environmental Intercomparison Project,

sponsored by the Department of Energy (DOE) and EPA for dosimeters

(TLD). These comparisons indicate that the licensee's measurements

are accurate. \

The licensee maintains more than twice the number of TLD radiation

monitoring stations required by the Technical Specifications. In

addition, the licensee maintains pressurized ion ~ chamber gamma

detectors and cryogenic krypton samplers at selected REM sampling

stations which are not required by Technical Specifications, and

collects more than the required number of environmental samples

'(milk, soil, leafy vegetables). In support of the defueling, the

licensee is doing analyses of transuranics in soils from~ eleven

_

locations and high volume air samples from four offsite locations.

The licensee has an extensive on and off site well monitoring

program. Groundwater aquifers are monitored by wells on two sides-

of the site. In addition, several wells are monitored onsite to

detect leaks before aquifers are affected. As a result of t.his

onsite monitoring program, several problems have been identified and'

corrected including several leaks on the borated water storage tank

valves.

There were no unplanned liquid or gaseous releases to the

environment during the assessment period. The NRC reviewed GPU

Quarterly Dose Assessments and Semi-Annual Radiological

Environmental reports for the period. The reported effluent

releases and corresponding calculated doses in each report were

small fractions of the TMI-2 Technical Specification limits for

annual averaged releases and were consistent.with the results of

environmental and effluent monitoring systems independently

monitored by EPA and NRC. When problems are identified, the

licensee evaluated the problems for potential generic implications.

For example, as a result of a licensee identi_fied potential

i ,

i)

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

..- . -.. _ - - .

,

I

• l

l

19

unmonitored release pathway to the environment, an extensive-

investigation was initiated to determine if other similar release

paths exit.

The licensee has improved its sump sampling program during the

assessment period. Certain sump samples had not previously been

representative due to lack of recirculation' procedures. In response

to NRC concerns, permanent. recirculation piping was installed during

late 1984 and early 1985 in most sumps to' improve sump sample

quality.

Conclusion

Rating: 1

Trend: Consistent

Recommendations ,

,

,

Licensee: None

1

NRC: None

1

1

I

s

a

l

,

,

,- - . , . _ - . . , - . , , , , , ,,.v... , - ,..,e , , ..-s,-,. .- - - ~ s w. . - -en - -

-g- ,, . - 4 -- ,

w

.

lb -

1/ 20

-

~,

D. Quality" Assurance

.

Analysis '

During this report period, a special inspection of the Recovery

Quality Assurance (QA) Plan was performed by a team consisting of

Resident,- I&E Headquarters, and contracted personnel; and the

.results of the' Performance Appraisal Branch inspection of the QA

Program, conducted between February 27 and March 28, 1984, were

received. Implementation of the program was routinely reviewed by

resident inspectors. There were no violations identified in this

area during the assessment period.

NRC review of this area determined that there is an extensive,

-detailed, and largely effective QA program. Many strengths were

identified. One.of the'mgre .significant strengths was the extensive

support for QA throughout the organization. Management support at

the highest levels was articulated in policy >Latements, corporate

procedures and memorandum addressing specific QA issues. This

backing was reflected at all- three levels (quality control,

operational monitoring, and' audits) of the QA Department in a

comprehensive goals ano: objectives program that addressed areas for

improvement. Specific action was taken by the Vice President,

Nuclear Assurance Divisiin to form an'Important to Safety (ITS)

working group to consider the classification of activities as they

relate to safety. Senior management has also strengthened and

clarified the management escalation program by better defining the

thresholds that trigger upper manag~ement action on longstanding

deficiencies; i.e. audit findings, Quality Deficiency Reports (QDR),

Material Nonconformance Reports (MNCR) and Quality Assurance

Monitoring Reports (QAMR).

The QA organization hcs used a graded approach to apply QA oversight

to facility operations. This graded approach allows QA to apply an

appropriate level of evaluation to activities, components, ,

procedures or systems b2$5d on their r 'clative contribution to plant

!

sa.fe ty. The licensee's i;11cy directsjthe QA organization to j

evaluate all site at i r? ~es including those not considered safety

significant. Fraq e t1 ,cheduled" audits and monitoring activities

complement the Q.s.ity .,' trol fuactions. Audits are thorough and  !

timely. The monitoring program .is aggressive and ensures proper  ;

implementation of administrative controls. Reports are timely and I

are distributed to appropriate senior _ management and review groups

within and outside the QA department. ' Corporate QA management is

frequently' involved in site activities. Quarte.rly presentations are

made by the Director, Quality Assurance to the GPUNC Board of

Directors on the status of site related QA issues.

-

s

%

w

C '-

- - - . -

.

.

21

Corporate QA has maintained an oversight role in evaluating the

performance of the Bechtel QA organization and its primary

suppliers. This oversight role involved routine visits to contractor

facilities and evaluating contractor compliance with a QA program

that had been approved by the licensee. During the initial phases

of fuel canister fabrication, this oversight was ineffective. NRC

inspections conducted in the Summer of 1985 of one contractor found

a number of problems with the implementation of the contractor's QA

program. Specifically, regulatory requirements were not satisfied

in several areas including use of unapproved vendors, failure to

identify and tag material, failure to perform receipt inspection,

failure to properly store material, and failure to maintain

up-to-date manuals. These findings precipitated a more aggressive

monitoring of the contractor QA programs by GPU Nuclear. Immediate

measures included increasing the frequency of visits to contractor

sites by QA personnel and using more detailed check lists in

evaluating the manufacture and testing of canister components.

Based upon the results of follow-up NRC inspections at contractor

facilities, these changes appear effective.

NRC evaluations determined that the onsite QA staff is experienced,

technically competent and sufficient to perform the functions

documented in the QA plan. The monitoring staff has issued over

1,000 QAMRs per year, covering the various aspects of licensee

activities; i.e. operations, surveillance, chemistry, maintenance,

radiological controls, etc. An example of the licensee's strong

commitment to quality was evident in the extent of action taken, in

response to NRC concerns, to ensure the quality of welds in the

Canister Storage Module. During the receipt inspection, a sampling

plan was developed to assure a representative sampling of the welds.

This plan involved spending approximately three man weeks performing

dye penetrant testing and visual examination of the welds, and then

adding as corrective action, an additional 1,900 stitch welds to the

four storage modules to ensure structural integrity.

QDRs, MNCRs, QAMRs, and Audit Findings were trended and evaluated to

provide recommendations for plant improvements. The program

consists of a four phase process that collects information in a l

computerized data base, evaluates the data to identify potential i

trends, investigates trends to determine underlying causes and I

implements corrective action to improve plant operations. A review

of QA records indicates steadily decreasing instances of procedural

problems such as failure to implement a procedure or failure to have

a procedure properly reviewed and approved. Through the trending of

QAMRs, QA identified that many procedures contained redundant

requirements. To improve these procedures, QA has tasked the

applicable departments to identify and remove needless duplication.

- _ _ _ _ _ _ - _ __

.

'

22

The licensee's commitment to the QA/QC program is also apparent from

the cooperation and extensive interface between the QA department

and plant personnel. NRC interactions indicate that licensee

personnel felt that QA unearthed substantial findings, made valid

suggestions for improvement, and enhanced the performance of their

line programs.

Conclusion

Rating: 1

Trend: Consistent

Recommendations

Licensee: None

NRC: None

I

e

__ . . . . .

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

.

23

'

E. Maintenance

Analysis

Resident Inspector review in this area centered on maintenance of

equipment required by Technical specifications, selected corrective

and preventive maintenance activities and major maintenance support-

,

ing defueling preparation.

The licensee has a well developed maintenance control program which

prioritizes, assigns and schedules work activities. There appears

to be effective communication between the various groups involved.in

planning, coordinating and scheduling various maintenance

4

activities. The NRC has noted that the licensee considers

interdepartmental communications important and are supported by all

levels of management. The licensee conducts a daily planning

,

meeting where maintenance activities and coordinated support are

discussed and scheduled.

j-

NRC review of the annual diesel generator maintenance identified

that the persons performing activities were knowledgeable of

existing administrative controls requirements and technical aspects

of the maintenance to be performed. Procedures were detailed and in

1 all but one case accurate. During maintenance it was noted that the

licensee maintained good records, took accurate data and performed

appropriate quality reviews.

Early in the period substantial problems were encountered with the

quality of the polar crane maintenance and inspection. In response,

the preventive maintenance program was expanded and upgraded. NRC

reviewed the licensee's revised preventive maintenance program for

, the crane. Procedures are detailed. the data sheets are reviewed,

compared to established acceptance criteria and trended by the

-electrical maintenance engineer with responsibilities for the polar

crane. The licensee has also developed a~ lubrication schedule which

is tailored to the TMI-2 polar crane.

The NRC routinely observed major maintenance items. An example was

the replacement of a Neutralizer Tank (Radwaste) Transfer System

Filter due to high differential pressure and increased radiation

level. The NRC witnessed activities. including review of the Unit

Work' Instruction, valve cycling, opening of breakers and subsequent

tagging. Supporting departments, including Radiological Controls,

Waste Management, Plant Operations, and Plant Engineering had

effectively coordinated and consolidated their efforts during this

evolution. As a result, the replacement was effected without

4

incident even though' working conditions were difficult due to

working in a high radiation area.

)

. - , . . - . . -~ -

-_ -

.

.

24

As a prerequisite for initiating flooding of the deep end of the

refueling canal, the licensee inspected and subsequently refurbished

components of the fuel transfer system. During the NRC reviews it

was determined that the Startup and Test Group performed the testing

in accordance with instructions and procedures, Plant Engineering

performed required safety analyses for "like kind" replacement of

components, and that there was adequate QA and QC coverage and

review for the fabricated parts. NRC review of the Unit Work

Instruction, Preliminary Safety Evaluation, Engineering Safety

Analysis, QC Plant Inspection Report, Radiation Work Permits and

Test _ Procedures and other associated documentation revealed that the

documentation was complete and accurate to support the work. The

work was performed well and problems identified during conduct were

corrected.

There were two violations in this functional area during the

evaluation period. These were isolated cases and no programmatic

breakdowns were identified in either case.

Conclusion.

Rating: 1

Trend: Consistent

Recommendations

Licensee: None

NRC: None

,

4

.m.____._..______.____._.___.___..._________.____._i..___._____-__.__'__..___.._________.___

.

.

25

F. Design, Engineering and Modifications

Analysis

The analysis in this area is based primarily on the results of

technical and safety evaluations and regulatory oversight reviews

performed by the NRR technical staff of major plant design change

and modification packages. There has been no programmatic

inspection performed in this area, though indepth reviews of the

licensee's implementation of the Quality Assurance (QA) Plan did

evaluate some aspects of the design change / modification program.

The licensee's current engineering organization has evolved through

the integration of GPU Nuclear and Bechtel staffs under one

management. This organization involves three separate, but

coordinated, staffs consisting of Plant Engineering and Site

Engineering located at the TMI site, and Design Engineering located

at the Bechtel project offices in Gaithersburg, Maryland. There was

no GPUN Technical Functions involvement in the design and

modification process. The engineering activities during this

assessment period have been primarily related to support of the

cleanup and defueling effort. Projects have involved engineering

support of plant operations and maintenance activities, plant

decontamination support, onsite management of plant modification and

construction, and detailed engineering design and procurement

related to major plant system modifications.

The organization is currently staffed with sufficiently qualified

personnel to provide an adequate level of multidisciplinary review

and management oversight. The organization is adequately defined in

procedures and policy documents which describe the functions,

responsibilities, and interorganizational-relationships of the three

engineering staffs. The organization generally assure an adequate

flow of communications, acceptable level of inter-departmental

reviews, and adequate implementation of both the GPU Nuclear

Recovery Quality Assurance Plan and the Bechtel Project Quality

Assurance Plan. Considerable GPUN and Bechtel management attention

are evident in the development of the organization and there have

been frequent management initiatives to modify the organization as

needed to best support the project as it'has evolved.

Although the present organization. appears to be functioning well,

there were several deficiencies noted early in the assessment period

that were indicative of significant problems in the licensee's

technical review process. To implement the review process, the

engineer preparing a document, performs a two step determination to

identify the quality requirements for the item. This determination

requires the use of applicable Regulatory Guides and TMI-2

Administrative Procedures. Secondly, the preparer then determines

the document's " Review Significance." This determination will

)

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

s

.

26

indicate whether the document will be independently reviewed by the

Safety Review Group subsequent to reviews by the Responsible

Technical Reviewer (RTR) and the Cognizant Engineer (CE).

Throughout this review and approval process, the Modification

Control Group is responsible for tracking, expediting, and ensuring

that the document-has the required concurrences and that additional

requirements identified during the reviews are incorporated in the

final document.

Early in the period there was a general lack of understanding of

this review process and applicable regulatory requirements,

particularly with the Bechtel organization in Gaithersburg,

Maryland. There were instances where the licensee's review

concluded that a modification did not involve an unreviewed safety

question (USQ) but the documented evaluation lacked adequate

engineering basis to support that conclusion. There appeared to be

a tendency for the engineering staff to conclude that a design change

did not involve a USQ because the change was temporary or because it

involved a portion of a system that was classified as "not important

to safety." In one case, the licensee approved an Engineering

Change Authorization (ECA) to install a temporary waste

solidification system in the Auxiliary Building that was determined

by the licensee to not involve a USQ. The basis for that

determination was that the installation was temporary. In addition,

there was lack of assessment of a modifications impact on overall

pl. ant safety. The licensee approved a Unit Work Instruction (UWI)

to remove a section of decay heat system piping inside the reactor

building. The licensee's evaluation failed to recognize the fact

that the modification would have resulted in a containment

penetration being in a configuration that did not conform to the .

'

General Design Criteria.

As a' result of NRC concerns regarding adequacy of licensee reviews,

the licensee implemented a training program which included indepth

training of all RTRs and CEs. As a result of the training program,

there has been significant improvement in this area. The licensee

has shown a better awareness of the regulatory requirements,

attention to detail and overall safety impact of' design changes.

Later in the assessment period, major engineering activities were

accomplished with extensive prior planning and a generally sound

technical approach. Examples include fuel transfer canal

modifications, stationary work platform and fuel pool modifications.

Ther.e was significant management involvement in these efforts, and

an adequate level of interdisciplinary review was evident. There

also appeared to be a much better understanding by the licensee's

staff of the internal policies and programs relating-to.the

.

engineering and design review process. The licensee demonstrated a

!

much improved awareness of the regulatory and safety aspects of

these programs and a generally conservative approach toward insuring

health and safety of both the public and the onsite work force. The

_ - _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - - . __. . _ _ _ _ _ - - - _ _ _ _ _ _ _ _ - - _ _

"s.

fi

27

licensee's technical review process showed considerable improvement

-with multiple levels of review, good flow of communication, good

feedback mechanisms within the organizations, and positive response

to resolve safety and. technical issues identified during the review

process.

Overall, the licensee has established an acceptable design,-

engineering and modifications program that provides for a sound

technical approach to solving engineering problems, adequate levels

of management control, acceptable degree of. technical review, and

resolution of technical problems in a manner that assures public

health and safety and conformance to regulatory requirements.

Conclusion

Rating: 2

Trend: Consistent

Recommendations

Licensee: Continued licensee attention to understanding the

application of regulatory requirements.

NRC: None

)-

- ..

,

'

T

l'

28

,

.G. Emergency Preparedness

Analysis

NRC observations in this area were oriented towards Unit 2 support

of Unit 1 activities and included: (1) participation and witnessing

>

of the-Unit 1 Annual Exercises on October 3, 1984 and. November 20,

1985; (2) Unit 2 support activities associated with TMI-I drills;

i (3) routine monitoring of Unit 2 quarterly departmental and

!

integrated practice drills and (4) review of emergency plan

implementation during actual events.

Management attention to viable emergency planning was evident. In

'

addition to quarterly integrated site drills and unannounced shift

- drills, the licensee has initiated quarterly ' shift walk throughs for

specific departments (e.g. operations, radiological controls,

security, maintenance, and Chemistry), to enhance emergency response

training.

1

During declared unusual events, implementation of ~ the applicable

portions of the emergency plan was appropriate. The response of

control room personnel, problem identification, corrective steps,

and required notifications were accurate and timely.

'

Emergency Preparedness is an overall site function and its

effectiveness will be evaluated in the next full Unit 1 assessment.

Conclusion

Rating: No Basis

Trend: No Basis

1 Recommendations

Licensee

None

NRC: None

l

l

l

l

-w e w y <-. ~ -

-*y-

,

,

g

29

H. Security

Analysis

The licensee has created a corporate position of Director of

Security who is dedicated to the nuclear security programs at'Three

Mile Island Units 1 and 2. A small amount of ' inspection effort has

been conducted in this area at Unit 2. More security inspection

effort has been carried out'at Unit 1. The effectiveness of the

security organization will be evaluated with the next full Unit I

assessment.

Conclusion

Rating: No Basis

Trend: No Basis

Recommendations

Licensee: None

NRC: None

)

e

t'

30

I. Licensing Activities

Analysis

Evaluation and monitoring of licensing activities during the assess-

ment period occurred primarily by routine contact between the

licensee's licensing staff and the NRC headquarters and onsite

staffs. A total of 81 licensing actions were initiated by the

-licensee during the assessment period. A breakdown of the' licensing

actions by categories is given in Table 5 of this report.

As a result of the unique and changing status of the TMI-2 facility

the number of licensing actions has been relatively large. This has

required coordination and frequent contacts between NRC -and licensee

staffs. The level of management _ involvement during the assessment-

period was significant. Of particular note has been the extensive

involvement of management in licensing activities such as head lift,

plenum removal and commencement of defueling. Licensee management

has also acted to encourage interaction between licensee and NRC

staff level personnel. This interaction has expedited NRC reviews,

by promoting early resolution of technical issues and has,

accordingly, acted to ensure that cleanup activities are carried out

promptly. Licensee management has established frequent and routine

communication with NRC management to coordinate licensing priorities.

This was particularly useful in the identification of those licensing

actions, including changes-to the Technical Specifications that were

required prior to beginning defueling.

Responses, formal and informal, to NRC requests for detailed

technical information during licensing reviews were generally

timely. While the licensee has disagreed'with some NRC technical

positions these issues were quickly elevated through the licensee's

management chain to an appropriate level for resolution.

The licensing staff has acted as the principal point of contact with

the NRC staff. This organization has been effective in coordinating,

within the licensee's organization, resolutions of NRC technical

concerns. The licensee has demonstrated a thorough understanding of

the technical issues-and NRC requirements. Resolutions to technical

problems have been conservative and sound. As an example, the

licensee carried out a detailed study to determine the best means

for preventing the possibility of recriticality in the reactor

coolant system. This work utilized, as contractors, recognized

experts in the criticality field and relied upon bounding

conservatisms in developing a strategy.

Descriptions of most proposed actions'have been detailed and

thorough. However, the licensee has on occasion failed to provide

the NRC staff with adequate initial information on their safety

evaluation for a proposed action. For example, during the review of

the technical sp'ecification request to lower the containment airlock

e-

1

31

door seal pressure a number of phone calls and a visit by an NRC

staff member was required to determine that their position was

acceptable.

In response to NRC staff requests the licensing staff has updated,

on a weekly basis, all licensing action items and has made a strong

effort to keep the NRC staff informed of the status of these items.

The licensee has an effective system to prioritize licensing

actions. Highest priority is given to issues with a potential for

directly impacting the cleanup schedule, while a lower priority is

assigned to those items that are not on the critical path. This

prioritization and close schedular coordination with the NRC staff

has generally been effective in helping to assure that licensing

actions are completed on a schedule that supports expeditious

cleanup.

Conclusion

Rating: 1

Trend: Consistent

Recommendations

Licensee: None

NRC: None

)

_

. .

.

32

V. SUPPORTING DATA AND SUMMARIES

A. Licensee Event Report (LER) Tabulation and Causal Analysis

Tabular Listing

.

Type of Events

a. Personnel Error 10

b. Design / Manufacturing / Construction /

Installation Error 1

c. External Cause 0

d. Defective Procedures 3

, e. Component Failure 10

,

f. Other 3

Total 27

t

Licensee Event Reports reviewed:

Report Nos.84-007 through 85-008

Causal Analysis

j

4

No trends were identified in the analysis performed on the LERs

z although three areas revealed some commonality.

1. Report Nos. 84-10, 84-13, 85-06, 85-07, and 86-02 involved

personnel error in implementing administrative requirements.

.

2. Report Nos. 84-09, 84-16, 84-21, 85-02, and 85-03 involved the

j inoperability irf incore thermocouples (accident damage).

$c 3. Report Nos. 84-12, 84-15, 85-08, and 85-10 involved

implementation of fire protection requirements.

B. Investigation Activities

During the SALP evaluation period one investigation was completed

pertaining to reactor coolant system leak rate data falsification

prior to the March 28, 1979 accident. The investigation culr.inated

in the Nuclear Regulatory Commission issuing an Order and Notice of

Hearing dated and served on December 18, 1985.

A report dated September 5, 1985 and titled, " Potential Willful

,

Material False Statement Concerning Polar Crane Modification" was

issued by NRC's Office of Investigations. This report ~is being

,

utilized by the agency in its consideration of appropriate enforcement

action related to polar crane violations.

,

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

._.

.

3-

.

.. .

33

' C. Escalated Enforcement Actions '

-

1. Civil Penalties

On August 12, 1985, the NRC Office of Inspection and Enforcement

issued a Notice of Violation and Proposed Imposition of Civil

Penalty of $64,000. The Notice of Violation was for the

licensee demonstrating acts of discrimination against a contract

employee for raising safety concerns and communications with the

NRC. This was a violation against 10 CFR 50.7 which prohibits

such discrimination and was considered a Severity Level II

'

.

violation. On March 4,1986, the Director of the'0ffice of

i Inspection and Enforcement ordered the licensee to pay the civil ,

'

penalty. (Not applicable to activities during the assessment.)

! 2. Orders

An Order and Notice of Hearing was issued by the Nuclear

Regulatory Commission on December 18, 1985 concerning reactor

coolant system leak rate data falsification. (Not applicable

to activities during the assessment.)

i

3. Confirmatory Action Letters (CALs)

There was one CAL issued during the assessment period which was

dated October 2,1985. The escalated enforcement issue

'

pertained to the licensee's failure to properly determine the

Sr-90 content of radioactive waste material.

D. Enforcement Conferences Held During the Assessment Period

'

During the SALP assessment period, four enforcement conferences were

conducted to discuss the licensee's corrective actions for the

i following enforcement issues.

'

One enforcement conference was held at the Region I Office on August

17, 1984 to discuss the licensee's failure to secure and deactivate

the containment isolation valve on the "A" train purge line with the

outboard isolation valve inoperable while purging through the "A"

l train (Inspection Report No. 50-320/84-13, Enforcement Conference

Report No. 50-320-84-13).

Another enforcement conference was held at'the Region I offices on

February 8, 1985 to discuss the licensee's failure-to adequately

protect Safeguards Information from unauthorized disclosure on three

recent occasions (Inspection Report No. 50-320/85-02, Enforcement

Conference Report No. 50-320/85-05).

i

i

)

T

. _ _. ._.

s

N'

34

I

4

Also,1on February 8,1985, another enforcement conference was held

in the Region I offices to discuss the NRC concerns that work

activity involving entry into the seal injection valve room was not

adequately planned in that prior to entry, proper consideration was

not given to the high energy beta sources that were known to exist

in the room (Inspection Report No. 50-320/85-03,- Enforcement

. Conference Report No. 50-320/85-07).

The fourth enforcement conference was held in the Region I offices

on October 9, 1985 to discuss an error in the determination for

, Sr-90 content in radioactive material which had existed for four

years without detection. This error resulted in the Sr-90 content

being systematically understated by'a factor of two in most radwaste

shipments (Inspection Report No. 50-320/85-20).

1

T

i

I

l

l

!

!

I

o

f

a

TABLE 1

INSPECTION HOURS SUMMARY (5/1/84 - 2/28/86)

THREE MILE ISLAND UNIT 2

HOURS  % OF TIME

1. Shutdown Plant Operations / 1935 32

Defueling Preparation

2. Radiological Controls 2937 49

'

3. Effluent Monitoring and Control 400 7

4. Quality Assurance 328 5

5. Maintenance 233 4

6. Design, Engineering and Modifications N/A N/A

7. Emergency Preparedness 131 2

8. Security 45 1

9. Licensing Activities N/A N/A

Total 6009 100%

.

I

.

4

TABLE 2

INSPECTION ACTIVITIES

THREE MILE ISLAND UNIT 2

INSPECTION REPORT NUMBER AREAS INSPECTED

84-09 Resident inspection

-10 Resident HP inspection, radwaste

-11 Resident inspection, preparation for

head lift

-12 Quality Assurance Program

-13 Review licensee's failure to close and

deactivate isolation valve on "A"

purge train with second isolation

valve inoperable

-14 Plant operations, radioactive material

shipments, Resident HP inspection,

radiological control practices

-15 Resident HP inspection, reactor

building operation

-16 Reactor operator examinations

-17 Resident inspection, plenum lift

preparations

-19 Resident inspection, polar crane

modifications

-20 Security plan and implementation of

security programs

-21 Resident HP inspection, dosimetry

-22 Operational radiological environmental

monitoring program

-23 Resident inspection

-24 Resident inspection

i

s

\

T2-2

-25 Resident HP inspection

85-01 Resident inspection

-02 Protection of Safeguards Information

-03 Unplanned radiation exposure to skin

of three individuals

-05 Enforcement Conference

-06 Security plan and implementation of

security programs

-07 Enforcement Conference

-08 Resident inspection

-09 Reactor operator examinations

-10 Resident inspection, plenum lift

preparation

-11 Defueling Water Cleanup System design

and fabrication

-12 Resident inspection

-13 Examination of the requalification

program for certification of presently

licensed SR0s to directly supervise

defueling operations

-14 Resident inspection

-15 Resident HP inspection, misclassified

radioactive waste material shipped

from site

-16 Resident inspection

-17 Reactor operator examinations

-18 Resident inspection, defueling

preparation

-19 Resident inspection, defueling

operations

>

e

\

T2-3

-20 Special inspection conducted to

evaluate method of determining Sr-90

by beta spectroscopy

-21 Resident inspection, defueling

operation

86-01 Resident inspection, defueling

operations

-02 Resident inspection

-03 Reactor operator examinations

,

b

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

'

-

g.

'\

TABLE 3

ENFORCEMENT DATA.

THREE MILE ISLAND UNIT 2

, A. Number and Severity Level of Violations and Deviations-

4

1. Severity Level

,

[ Deviations- 0

Violations S.L. I O

i Violations S.L. II 1

i Violations.S.L. III 4

Violations S.L. IV 12

Violations S.L. V 5

'

Total 22

'~ .

B. Violations and Deviations vs. Functional Area <

] .

l FUNCi10NAL AREAS I II III IV V DEV

i

1. Shutdown Plant Operations / 1* 2

Defueling Preparation

i 2. Radiological Controls 4 5 3

<

)

3. Effluent Monitoring and Control 1

.

i 4. Quality Assurance

j 5. Maintenance 1 'l

! 6. Design, Engineering and Modifications

1

7. Emergency Preparedness

8. Security 4

j

9. Licensing Activities

,

,

1 4 12 5

• Violation not applicable to activities during assessment.

4

l-

I

l

-i

i 1

. . . .

..

e

T3-2

C. Summary

Inspection Severity Summary of

Report No./Date Level / Functional Area Violation

84-09 V Failure to post a

4/24/84-5/30/84 Rad Con Notice of Violation

involving radiological

working conditions as

required by 10 CFR Part

19

84-10 V Failure to calibrate a

5/14/84-5/29/84 Effluent Monitoring radioactive process

effluent monitor in -

accordance with written

procedures

84-13 I" Violation of Technical ,

5/25/84-6/29/84 Operations Specification limiting

condition of operation

84-21 IV Failure to control

10/16/84-11/26/84 Security Safeguards Information

84-21 IV Failure to adhere to

10/16/84-11/26/84 Rad con Rad Con Procedure

85-01 V Failure to attach Do

1/12/85-2/28/85 Maintenance Not Operate tag'to

inoperable breaker

85-01 -V Failure to cneck

1/12/85-2/28/85 Rad Con available dose for

individual entering RWP

area

85-03 III Failure to perform

1/14/85-1/22/85 Rad. Con adequate rad survey

85-03 III Failure to adhere to

1/14/85-1/22/85 Rad Con RWP

85-03 III Doses above

1/14/85-1/22/85 Rad Con administrative limits

85-06 IV Failure to properly

3/4/85-3/8/85 Security control photo ID badges

L___--__--_-_ _-- - - - ------------ --- -- --- ---- - - - - - - - - - - - - - - - - -- --

- . _

o

S

T3-3

,

85-06 IV Failure to complete

'3/4/85-3/8/85 Security annual physical fitness

tests by three

protection personnel

85-08 IV Failure to comply with

3/1/85-4/19/85 Maintenance procedures regarding

ventilation system

maintenance practices

85-10 IV Failure to comply with

4/11/85-5/17/85 s

Security procedures regarding

security practices

85-15 III Failure to properly-

7/31/85-8/5/85 Rad Con package and classify

waste in accordance

with Reg Guide

85-16 V Failure to inspect

8/6/85-9/6/85 Rad Con respiratory equipment

85-20 IV Failure to properly

9/30/85-10/2/85 Rad Con indicate Sr-90 content

on rad material

shipping papers

85-20 IV Failure to properly

9/30/85-10/2/85 Rad Con classify radwaste

shipment

85-20 IV Failure to properly

9/30/85-10/2/85 Rad Con indicate Sr-90 activity

on rad material

shipping papers

85-21 IV Failure to comply with

10/7/85-11/8/85 Operations a double isolation

valve requirement

85-21 IV Failure to perform

10/7/85-11/8/85 Rad Con adequate surveys prior

to entry

I&E Headquarters & II Personnel harassment

01 Report Operations

l

1

!

u

a

i

!

TABLE 4

LER SYNOPSIS

THREE MILE ISLAND UNIT 2

i

May 1, 1984 - February 28, 1986

LER Number Summary Description

84-007 Operation of RB purge system outside bounds of

Technical Specifications84-008 Examination and evaluation of pressure and level

transmitters for core flood tanks

'84-009 Incore thermocouple N-4 was declared inoperable

84-010 Operation of containmer.t isolation valves without

proper administrative approval

~

84-011 Exceeded timeclock without entering action statement

after defeating the Control Room air inlet radiation

monitor, HP-R-220, interlock

84-012 Failure to test within the required interval a Fire

i

System detector located in Control Building

84-013 Reactor Building airlock doors were operated using a

Temporary Change Notice that had not been submitted

to the NRC for approval

84-014 Representative reactor coolant samples were not

obtained due to failure to verify opening 'a manually

operated valve

]

84-015 Hourly fire watch was not performed when a fire door

was breached l

84-016 Incore thermocouples M-3, 0-10 and F-3 were declared

inoperable

84-017 Sections of submerged toe of thetdike exhibited

evidence of degradation.84-018 "A" diesel generator was out of service to . replace

cylinder assembly and exceeded timeclock

)

.. . .. ..

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

.

. __

..

A .

V .

c

i T4-2

l

84-019 Action statement was not entered when the Fuel

Handling Building ventilation exhaust flowrate

dropped below the minimum limit of the Technical

Specifications84-020 The RCS sample performed through the TNS system was

non-representative because sample line not adequately

purged

t 84-021 Incore thermocouples F-12, 0-5, and G-11 were

declared inoperable

85-001 Minor degradation of the Flood Protection dike

85-002 'Incore thermocouple E-4 declared inoperable

85-003 Incore thermocouple G-13 declared inoperable

85-004 Reactor Building Internal Pressure Indicator

i Registered Value in excess of Technical Specification

! limit

85-005 Reactor Building Internal Pressure Indicator

Registered Value in excess of Technical Specification

limit

85-006 Administrative error in operating procedure stating

wrong valve position

85-007 Containment isolation valve operated without an NRC

approved procedure

85-008 . Fire Suppression System inoperable for Emergency

Diesel Generator

85-009 .0pened Containment Isolation Valve

85-010 Failure to test the TMI-2 Fire Suppression Water

System Valves ,86-001 Inoperability of Emergency Diesel Generator exceeds

l

7-day timeclock

86-002 Containment Isolation Valve opened without

identifiable cause

l

l

l

!

i

. ._--- _ _ _ ______________________- _ ___ _ _____ _____ - -___-_-__ _ __________-_-___ _ ____- --_ ___.__.-_ - _ _

um:w

a

.'o

TABLE 5

SUMMARY OF SIGNIFICANT LICENSING ACTIONS

AND SUPPORTING ACTIVITIES

THREE MILE ISLAND UNIT 2

This table provides a summary of significant licensing actions and related

activities during the SALP evaluation period from the 1st of May 1984 through-  !

the 28th of February 1986.

1. Updates to the Programmatic Environmental Impact 3tatement (1 completed) l

. Issuance of a Final Supplement Dealing with Occupational Radiation

Dose

2. Major Cleanup Evolutions (21 completed, 6 ongoing)

. Reactor Pressure Vessel Head Removal

--

Head Removal SER

. Reactor Pressure Vessel Plenum Removal

--

Plenum Removal Prep Activities SER I

--

Plenum Jacking SER

--

Reactor Building Heavy Loads Analysis for Plenum Lift

--

Polar Crane Auxiliary Hoist

--

Plenum Lift SER

--

Plenum Load Drop Analysis

. Defueling Actions

--

Preliminary Defueling Activities

--

Early Defueling SER

--

Defueling Canister TER *

--

Boron Dilution Hazards Analysis

--

Fuel Storage Rack TER

--

Defueling Water Cleanup System TER, Rev. 4

--

Heavy Load Handling SER for Defueling

--

Filter Canister QA Approval *

--

Fuel Pool "A" Refurbishment SER

--

Reactor Building Decontamination and Dose Reduction Program SER

--

Internals Indexing Fixture Processing SER

--

Modifications of Fuel Canisters

--

Modifications of Fines / Debris Vacuum System

--

NES Canister Closeout

--

Use of Debris Canisters >

--

Canister Handling - Preparations for Shipping *

--

Reactor Building Sump Criticality *

--

Temporary RCS Filtration System

--

SDS TER & SD Update *

i

--

Defueling Water Cleanup System TER, Rev. 7-9 *

)

4

i

i

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

4

-

e

T5-2

3. Ancillary SDs, TERs and SERs (5 completed, 1 ongoing)

. SER for Core Stratification Sample Acquisition *

. Relocation of Missile Shields SER

. Purification Demineralizer SER - Cs Elution-

• Ongoing as of February 28, 1986 '

4

. Submerged Demineralizer System TER Update

! . Once Through Steam. Generator Layup TER

. Containment Air Control Envelope TER

4. Other Documents in Support of Cleanup (6 completed, 2 ongoing)

. Applicability of Seismic Design Critaria to GPUNC Recovery Efforts

. Clarification of the General Project Design Criteria

. Containment Access Control Envelope Design Criteria

<

. Solid Waste Staging Facility TER *

,

. Waste Handling Packaging Facility *

. GPU/0RNL Criticality Report

i .

Fuel Characterization in the Lower Vessel Head

. PWST and Recycle System / System Description

i 5. Exemptions (10 completed, 2 ongoing)

j .

Waste Classification of EPICOR II Resin Liners

!

. In-Service Inspection *

,

, Core Accountability

. Fire Protection *

<

. Pressurized Thermal Shock

<

. Residual Head Removal System and Testing Requirements for the ECCS

i . Code Safety Valves

i . Seismic Monitoring Requirements

i . FSAR Updating Requirements Relative to the QA Revisions

1 . Containment Penetration Design

, . Seismic Requirements for Containment Penetrations

! . Containment Isolation' Valves

6. Changes to the Technical Specification

! . Issued: 8

. Ongoing: 0

,

,

7. Recovery Operations Plan Change Requests '

. Issued: 10

. Ongoing: 1

,

i 8. License Amendments

. Issued: 2

. Ongoing

2

i 9. Organization Plan Changes

. Issued

4

. Ongoing: 0

4

I

r.

'

n

.

g> -

o

T5-3

e

/

10. Licensee /Commissior. Briefings - 2 i

.

Cleanup Schedule and Funding

• Ongoing as of February 28, 1986

t

I