ML20202F264

SALP 6 Board Rept 50-155/86-01 for Nov 1984 - Mar 1986. Category 1 Rating Maintained in Area of Emergency Preparedness.Regional Insp Activities for Emergency Preparedness Will Be Reduced

ML20202F264
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/10/1986
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20202F249	List: ML20202F246 ML20202F264
References
50-155-86-01, 50-155-86-1, NUDOCS 8607150126
Download: ML20202F264 (39)

See also: IR 05000155/1986001

Text

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SALP 6

SALP BOARD REPORT

U. S. NUCLEAR REGULATORY COMMISSION

d'

REGION III

1

SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE

50-155/86001

Inspection Report

Consumers Power Company

Name of Licensee

Big Rock Point Plant

Name of Facility

November 1, 1984 through March 31, 1986

Assessment Period

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8607150126 860710

PDR ADOCK 05000155

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I. INTRODUCTION

The Systematic Assessment of Licensee Performance (SALP) program is an

integrated NRC staff effort to collect available observations and data on

a periodic basis and to evaluate licensee performance based upon this

information. SALP is supplemental to normal regulatory processes used to

ensure compliance to NRC rules and regulations. SALP 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 construction and operation.

A NRC SALP Board, composed of staff members listed below, met on May 23,

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.

SALP Board, for Big Rock Point:

Name Title

J. A. Hind Director, Division of Radiological

Safety and Safeguards

E. G. Greenman Deputy Director, Division of Reactor

Projects

W. G. Guldemond Chief, Reactor Projects Branch 2

L. R. Greger Chief, Facilities Radiation Protection

Section

E. R. Schweibinz Chief, Technical Support Staff

M. Schumacher Chief, Radiological Effluents and

Chemistry Section

B. Snell Chief, Emergency Preparedness Section

D. H. Danielson Chief, Material and Process Section

R. B. Landsman ProjectManager,ReactorProjects

Section 2D

T. Rotella Big Rock Point Project Manager, NRR

S. Guthrie Senior Resident Inspector

D. A. Kers Plant Protection Analyst

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II. CRITERIA

Licensee nerformance is assessed in selected functional areas, depending

upon whetner the facility is in a construction, preoperational, or

operating phase. Functional areas normally represent areas significant to

nuclear safety and the environment. Some functional areas may not be

assessed because of little or no licensee activities, or lack of meaningful

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observations. Special areas may be added to highlight significant j

observations.

One or more of the following evaluation criteria were used to assess each

functional area.

1. Management involvement and control in assuring quality

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

standpoint

3. Responsiveness to WRC initiatives

4. Enforcement history

5. Operational and Construction events (including response to, analyses

of, and corrective actions for)

6. Staffing (including management)

However, the SALP Board is not limited to these criteria and others may

have been used where appropriate.

Based upon the SALP Board assessment each functional area evaluated is

classified into one of three performance categories. The definitions of

these performance categories are:

Category 1: Reduced NRC attention inay be appropriate. Licensee management

attention and involvement are aggressive and oriented toward nuclear safety;

licensee resources are ample and effectively used so that a high level of

performance with respect to operaticnal safety and construction quality is

being achieved.

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

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management attention and involvement are evident and are concerned with

nuclear safety; licensee resources are adequate and are reasonably

effective so that satisfactory performance with respect to operational

safety and construction quality is being achieved.

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

management attention and involvement is acceptable and considers nuclear

safety, but weaknesses are evident; licensee resources appear to be strained

or not effectively used so that minimally satisfactory performance with

respect to operational safety or construction quality is being achieved.

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III. SUMMARY OF RESULTS

The overall regulatory performance of the Big Rock Point Plant has

continued at a satisfactory level during the assessment period.

Improved performance in the area of Licensing Activities is noted.

However, performance in the areas of Plant Operations and Surveillance and

Inservice Testing declined from a Category 1 to a Category 2. Performance

in the area of Outages is rated a Category 3 this period. This rating is

a reflection of the breakdown of administrative controls over the outage

process and resulted in a Severity Level III violation during the middle

of the SALP period.

July 1, 1983- November 1, 1984-

Functional Area October 30, 1984 March 31, 1986

A. Plant Operations 1 2

B. Radiological Controls 2 2

C. Maintenance / Modifications 2 2

D. Surveillance and

Inservice Testing 1 2

E. Fire Protection 2 2

F. Emergency Preparedness 1 1

G. Security 1 1

H. Outages *

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I. Quality Programs and

Administrative Controls

Affecting Quality 2 2

J. Licensing Activities 2 1

K. Training and Qualification

Effectiveness *

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• Not Rated (new functional areas for SALP 6)

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IV. PERFORMANCE ANALYSIS

A. Plant Operations

1. Analysis

Portions of eight routine inspections by the resident inspector

reviewed plant operations. The inspections included observations

of control room operations, reviews of logs, discussions of

operability of emergency systems, and reviews of reactor building <

and turbine building equipment status. During the evaluation

period the following violations were identified:

a. Severity Level IV - Failure to perform required surveillance

on the Reactor Depressurization System (RDS) (155/84017).

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b. Severity Level V - Delay in notification to NRC of a plant

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shutdown required by Technical Specifications (155/84017).

By not performing the surveillance when required, the unit

entered a Limiting Condition for Operation statement and a unit

shutdown was necessary. The licensee also failed to recognize

the reporting requirements of 10 CFR 50.72 associated with a

forced shutdown and was late in making the required notification.

The Big Rock Operations Department is adequately staffed with

licensed and non-licensed individuals who are dedicated to

safe and efficient operation of the reactor. Observation of

operators in the control room and on tour in the plant indicates

they are generally conscientious in both routine and off-normal

activities. They make regular use of drawings and procedures to

plan and perform evolutions. Control room decorum is adequate,

and a cooperative, results-oriented attitude is apparent among

the operators and in the operator's dealings with maintenance

men, radiation protection technicians, and the engineering staff.

Operators are well trained. Shift-manning is accomplished

without excessive use of overtime and the number of individuals

in training and requalification programs appears adequate to meet

future needs.

The operations staff performs well during startups and

shutdowns of the reactor, refueling operations, and performance

of surveillances. During surveillances, operators appear to

understand the objective of the test and the impact of their

actions on plant equipment. Operators appear capable of

dealing with abnormal and emergency situations,' indicating

adequate training and a functional understanding of plant

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equipment and systems interrelationships. Examples include

operator action that prevented reactor scrams on two occasions,

including a potential scram on high pressure when the Initial

Pressure Regulator (IPR) cover was lowered onto the IPR linkage,

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and two near scrams on low vacuum during installation of a pipe

patch on bypass line piping. In another instance, operators took

conservative action by scramming the unit manually on indications

of a major steam leak. Additionally, during the period of

reactor vessel inventory loss following the incorrect disassembly

of VRD305, operators on the refueling deck were quick to identify

the cause of the decreasing level and took proper corrective

action.

Of the six Reactor Protection Systems actuations resulting in a

scram signal during the evaluation period, two were attributable

in part to operator error. On January 7, 1985, the scram was

caused by failure to reset the FRV and on December 7, 1985, the

scram was precipitated by operators who earlier had left a valve

mispositioned prior to startup. The other four scrams resulted

from spurious actuations of the RPS system because of electrical

noise affecting the circuitry at low power, a known operating

characteristic of little safety significance.

In spite of the ability of the operations staff to operate the

plant reasonably well the department experienced a series of

human errors throughout the assessment period which detracted

3' from the safe operation of the facility. Most were attributable

to inattentiveness or lack of thorough attention to detail.

Operator inattentiveness on two occasions resulted in misposi-

tioned control rods, though one instance was influenced by

inadequate management direction and cumbersome administrative

controls over several available rod withdrawal sequences.

Inattention to detail and an assumption that other plant

personnel or administrative systems would compensate for

failure to assume personal responsibility for plant safety were

at the root of errors associated with tagging and isolation of

components involving work on the recirculation pump, and in a

separate incident, the incorrect disassembly of a control rod

drive system check valve. Inattention to detail and a

willingness to circumvent administrative controls (see

Section IV.H) resulted in an incorrect pipe being severed during

construction of Alternate Shutdown systems. Additionally, errors

resulted in the incorrect tagging of an electrical breaker, and

in missed surveillances detailed in Section IV.D of this report.

Finally, failure to follow local tagging procedures resulted in

the repair of Valve VNS143 without tagging or isolation and is

believed to be a factor in a major steam leak and subsequent

. scram on December 7, 1985. Licensee management, in response to

increased frequency of errors, has emphasized attention to

detail, counseled individuals, retrained personnel, and

implemented revised administrative controls on control rod

manipulations.

Management demonstrates a thorough understanding of the plant's ,

operation, reflecting extensive experience with this facility. '

Management personnel are often present in the control room area

and tour the plant regularly. Management direction and control,

however, was considered to be deficient in several instances.

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While the composition of the operations department staff

continued in transition from a group of older operators with

many years of plant specific experience to a mixed group with

many operators who are relatively new to the plant, management

showed a reluctance to compensate by incorporating lessons

learned into plant operating procedures. For example, the

bypass valve has a history of erratic behavior in automatic

operation at low steam flows, resulting in two plant scrams in

1984. Management had not provided specific guidance to operators

on when to remove the valve from automatic control, leaving it up

to the individual operator's discretion, even though a disparity

in theories and practices existed among operators because of

differences in experience levels. The Reactor Scram on

January 7, 1985 was caused by failure to reset instrument air

to the Feedwater Regulating Vaive (FRV) following air system

maintenance. Older operators surveyed were aware of the valve's

characteristic of failing on loss of air, but the less experienced

operators performing the startup were not. Newer operators could

have benefited from an expanded component identification program

throughout the plant. Finally, the licensee's revised admini-

strative controls over control rod movement, "hich employed

laminated cards and was implemented as corrective action

following two instances of mispositioned control rods, went

into effect with insufficient management direction. As such,

the card system went unused until repeated requests from the

resident inspector prompted management to publish guidance

requiring consistent and regular use by operators for all rod

motion.

Some reluctance to respond to NRC initiatives was in evidence

throughout the assessment period. Examples include responses to

inspector inquiries about operability of the acoustic monitor

during plant startup and operation, the need for a second Control

Rod Drive Pump to meet the requirements of Appendix R, the need

to test the availability of one electrical power source prior to

removal of another, and the recommendation to label the contain-

ment escape lock operating handles. However, the quality and

quantity of communication and cooperation with regulators

steadily improved over the 17 months. In the closing months of

the evaluation period the licensee demonstrated a willingness

to respond positively to NRC initiatives and a concern for safety

by operating the Diesel Fire Pump continuously when its starting

reliability was in question.

The licensee also exhibited several instances where a conservative

approach to resolution of a technical issue was chosen. Examples

include conservative declarations of inoperability on the RDS

system because of a detensioned hanger and on one tube bundle of

the emergency condenser based on a barely detectable indication

of a primary to secondary leak. The circumstances surrounding

the event discussed earlier in this section point to a licensee

decision to emphasize production over safety, but is an isolated

example not representative of the licensee's approach to

technical issues throughout the remainder of the period.

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2. Conclusions

The licensee is rated category 2 in this area with a declining

trend based on increasing frequency of human errors and difficulty

in implementing administrative controls over practices and

procedures important to plant safety.

3. Board Recommendations

To avoid future declines in this functional area licensee

management should address problems with administrative controls,

particularly as they relate to nutage management, and reduce the

frequency of human error in plant operations.

B. Radiological Controls

1. Analysis

Evaluation of this functional area is based on routine

assessments by the resident inspector during implementation of

the resident inspection program and six inspections by Region III

specialists. These inspections covered radiation protection,

radwaste management, disposal of low-level radioactive waste,

chemistry and radiochemistry, and confirmatory measurements.

One violation and one deviation were identified as follows:

a. Severity Level V - Failure to conduct a quality control

program to assure compliance with waste classification and

waste characteristic requirements (155/85006).

b. Deviation - Failure to implement the Radiation Safety Plan

by the date specified in the licensee's August 19, 1982

supplemental response to the Health Physics Appraisal

(155/85003).

The violation and the deviation were the results of inadequate

procedures; the licensee's corrective actions were timely.

Responsiveness to NRC initiatives has been generally adequate.

In response to inspector concerns regarding mask-fit testing of

BioPak 60-P respirators, the licensee replaced these respirators

with open circuit Self Contained Breathing Apparatuses (SCBAs).

Also, inspector concerns identified related to laboratory

performance are often acted upon by the end of the inspection.

However, the licensee was somewhat slow in correcting an error

in a 1984 semiannual effluent report brought to their attention

by the inspector, and was also slow to complete an evaluation and

request for approval concerning retention of contaminated soil

onsite following a break in an underground line to the condensate j

storage tank. The contaminated soil issue was closed by an -

Environmental Assessment and Findings of No Significant Impact

published in the Federal Register (May 5, 1986 - 51FR16596). New i

RETS technical specifications and the ODCM were implemented 1

during this assessment period. I

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Staffing in chemistry and radiation protection appears adequate,

with no changes in key supervisory personnel. The relatively

small technician staff has recently experienced a high turnover,

with six of 12 technicians replaced during this assessment

period; however, the inspectors have not observed a significant

effect on licensee performance. All of the replacement

technicians have completed the specified basic training course,

and assigned responsibilities appeared to have been commensurate

with the level of training. Supervisory personnel appear to have

a good understanding of their areas of responsibility.

Management involvement has been adequate to assure acceptable

quality in the functional area. There is adequate ALARA program

support and involvement by all levels of management. Records

are generally complete and well maintained. Procedure adherence

has been generally adequate, and management policy encourages

worker identification of problems to help with timely corrections.

However, inspectors have noted a significant number of instances

which indicate the need for more management attention, including

persons not frisking at exit points, radioactive materials stored

outside posted areas, contaminated area postings with inadequate

or confusing instructions to workers, and area monitor calibra- i

tion sources carried through office areas without appropriate

restrictions to personnel access to the area. Quality Assurance

(QA) involvement in the health physics activities during 1984 was

marginal. This shortcoming was exacerbated by the fact that

the formal plant surveillance program required by the licensee's

Radiation Safety Plan (RSP) had not yet been implemented. In

February 1985, NRC inspectors noted that the formal reporting

system for minor radiological occurrences required by the RSP

had also not been implemented. The recent implementation of

these RSP programs should improve overall management involvement

in this functional area.

Although the licensee's approach to the resolution of radiological

technical issues has generally been technically sound, thorough,

and timely during this assessment period, instances of poor

performance have occurred. In late 1984, a policy was

implemented which established a routine decontamination program;

however, no dedicated decontamination workers were assigned.

Despite this decontamination program, the licensee has

experienced problems in contamination control, especially during

outages. The addition of two contractor decontamination workers

following the 1985 outage resulted in a major improvement in

j plant cleanliness. Formerly inaccessible areas are now

accessible. The ALARA program has shown improvement during

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, and since the 1985 outage. The licensee has committed to an

, ambitious program of person-rem reduction that will stress job

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preplanning and new fuel pool cleaning equipment. With regular

- use of decontamination personnel the licensee intends to reduce

annual exposure by approximately one-third. The licensee is

generally conservative in resolution of potential safety and

environmental concerns. Relocation of a storm drain release

path to the lake, necessitated by high lake level, was

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accomplished by routing it in a manner to ensure that releases

would be monitored by the discharge canal monitor. The licensee

has also performed extensive testing during shutdown to locate

the source of a minor primary to secondary leak that developed

in the emergency condenser during operation. When the leak

source could not be identified, an augmented sampling program

was instituted upon restart to ensure that regulatory limits

are met. Corporate management is involved in the station's

effort to develop a method of measuring minor airborne releases

via this pathway.

Due to continuing fuel cladding problems, radioactive gaseous

releases during this assessment period were about a factor of

six higher than normal but have remained well below regulatory

limits even when operating at full power. Licensee efforts to

minimize releases and to eventually eliminate the problem

included removal of identified fuel leakers and use of a new

design replacement fuel. Release rates since the November

restart have been running at about three to four times the

normal rate. Liquid radioactive releases were below average

for U.S. boiling water reactors. The activity in liquid releases

has apparently stabilized during this assessment period following

several years of gradual decline. The solid radioactive waste

volumes in 1984 and 1985 were significantly less than in recent

years due, in part, to the implementation of a segregation

program for dry active waste (DAW). No transportation problems

were identified during this assessment period.

Personal exposures were about 120 and 270 person-rem in 1984

and 1985, respectively. These exposures are below the station

average over the previous five years (approximately

300 person-rem).

The licensee performed generally well in confirmatory measurements

with 34 agreements in 36 comparisons with Region III during the

assessment period. The disagreements were both for iodine

collected on a charcoal cartridge, with the licensee's values

about 20% lower than the NRC's. Recalibration following a

similar disagreement during the previous SALP period did not

resolve the difficulty owing, apparently, to differences of

activity distribution between the licensee's standard and

plant samples. The licensee readily agreed to use a correction

factor until another recalibration could be accomplished.

2. Conclusions

The licensee is rated Category 2 in this functional area. This

is the same rating given the previous SALP period.

3. Board Recommendations

None.

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C. Maintenance / Modifications

1. Analysis

Portions of eight routine inspections by the Resident Inspector

reviewed maintenance activities. One violation discussed in

Section IV.H, Outages, reflects on the licensee's ability to

conduct maintenance work during outages. In addition, two

Regionally based inspections'were performed. The inspections

included reviews of normal maintenance and modification activities

to ensure that approvals were obtained prior to initiating work,

activities were accomplished using approved procedures, post

maintenance testing was completed prior to returning components

or systems to service, and parts and materials were properly

certified. In addition, work planning and scheduling was

reviewed as well as the effectiveness of administrative controls

to ensure proper priority is assigned. No violations or

deviations noted.

During the evaluation period the licensee interrupted plant

operations for nine unscheduled maintenance outage periods

ranging from one to 11 days. Three outages were required to

repair Reactor Depressurization System (RDS) valves due to the

degraded condition of the system preventing successful performance

of quarterly surveillances. These included one forced shutdown

required by Technical Specifications unidentified leak rate

limitations. Two outage periods of one day each were required

to successfully repair IA-60B, seal leakage to heat exchanger

for Reactor Recirculation Pump No. 2. Also, two outages of three

and four days each were required to diagnose and correct steam

leakage from the reactor vessel head o-rings. One outage period

of four days was used to replace a recirculation pump seal, and

a one day outage was required to correct steam leaks associated

with the plant scram on December 7, 1985.

Proper planning and outage control was generally evident for the  !

nine unscheduled outages. Although unplanned, the licensee in

the case of the RDS and recirculation pump outages had sufficient

warning to plan activities, prepare parts and procedures, and

perform other maintenance work that fell within the scope and

time limitations of the forced outage. Repair to RDS valve top

assemblies have become commonplace to the point that the licensee

routinely overhauls spare top assemblies. The licensee did not

, overhaul the spare recirculation pump seal in advance of the

outage and was still rebuilding the seal as the plant was being

shutdown to perform the replacement, even though the pump had been

idled for two weeks prior to shutdown. The licensee made

extensive use of vendor consultants and pump experts from the

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General Office for the seal replacement, resulting in a refined

and useful procedure for rebuilding and installation. Outages

for RDS and recirculation pump repairs were well planned and

executed. Outages to repair IA-60B represented an operational

situation that offered little warning and first attempts at

repairs were unsuccessful. The reactor vessel o-ring offered

no warning prior to failure, but successful repairs were delayed

when the problem was misdiagnosed. Once the decision was made

to perform the vessel head removal and ring replacement the

physically demanding job was successfully completed with

conservative consideration to ALARA and personnel safety.

Maintenance work (including mechanical, electrical, and

instrument / control) at Big Rock Point is performed by generally

competent repairmen who exhibit craftsmanship and a general

familiarity with the facility and the equipment. The amount

of unsuccessful repair attempts resulting in rework is generally

small. Repairmen generally are cognizant of procedural require-

ments associated with their assigned task, communicate effec-

tively with operators and health physics technicians, a'd reflect

concern for ALARA considerations. While the input repairmen

provide to machinery history is often marginal, communication

with co-workers and supervisors indicates genuine interest in

continued safe and successful operation of the reactor. The

mechanic who performs the work, for example, often participates

in post maintenance testing. While the retirement of older,

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experienced maintenance department personnel during the period

had a negative impact on performance as documented further

! in Section IV.H, Outages, the maintenance staff demonstrated

flexibility and dedication throughout the evaluation period.

l The size of the maintenance staff is generally adequate for all

periods other than major refueling outages. A gradually

increasing backlog of maintenance orders over the period is

explained in part by increased emphasis on skills training which

over the short term reduces staff size availability.

Like the Operations Department the loss of older experienced

personnel due to retirement or other duties has altered composi-

tion of the maintenance staff. While the I & C group remained

unchanged, in the mechanical maintenance group of 12 men, five

were added during the assessment period. Because hiring and

promotion is heavily influenced by Labor Relations agreements

that emphasize seniority, newly added staff members generally

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have little or no experience with nuclear powered generating

plants in general or Big Rock Point specifically. Altbough the

licensee has long recognized the need for maintenance staff

training, no training was provided until February 1986, when a

regular program of skills training offsite was initiated. The

skills training is general in nature and is not nuclear plant

specific. No nuclear plant system or concepts training is

provided.

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First line supervision in the maintenance department reflects

adequate technical skills and managerial competence. During the

1985 outage, the maintenance department overcame the loss of

staff experience, inadequate outage planning, and parts procure-

ment to accomplish a relatively large number of modifications,

repairs, and preventive maintenance tasks.

Throughout the evaluation period several recurring problems were

not successfully repaired or adequately addressed. Valve M0-7067,

Turbine Bypass Isolation Valve, was not declared operable for

much of the evaluation period, based on difficulties with the

valve operator. Reactor Depressurization System (RDS) valves

exhibit inherent design deficiencies that have resulted in three

forced shutdowns during the assessment period and a long history

of problems dating back to their installation. Management,

however, has not placed a high priority on a comprehensive

solution and as a result the RDS system was not improved over

the period. Problems with the Emergency Diesel Generator (EDG)

fuel pump were allowed to continue and a design change to the

pump mounting bracket scheduled for completion during the 1985

refueling outage was deleted in an effort to return the plant to

an operable status. Shortly thereafter the pump failed again,

placing the EDG in an action statement for the generator's

Limiting Condition for Operation. Finally, the licensee made a

commitment to verify, prior to startup from the 1985 outage,

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Limitorque Switch settings on 18 Limitorque Valves the licensee

considered important to safety. As of this date only 15 have

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been checked. The torque settings for valve M0-7067 have been

reset on three different occasions, indicating a lack of decisive

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direction on problems with Limitorques Operators that goes back

to September, 1984, as was addressed in SALP 5.

SALP 5 expressed concern that the Prevention Maintenance (PM)

program may be inadequate to address aging equipment. At the end

of this assessment period the PM program continues to be reactive

in nature, relying heavily on visual inspections that do not

involve disassembly or physical measurements, and on the obser-

vations of operators monitoring noticeable changes in component

l operating characteristics. There continues to be no program to

analyze for trends in failures or any other measurable parameter

other than pump capacity on certain pumps. The licensee has not

responded to NRC initiatives to upgrade the PM program to incor-

porate vendor recommendations and industry experience. The plant

continues to rely on surveillance tests to identify problems that

may be in some advanced stage of development due to aging

equipment. At the close of the assessment period the licensee

assigned an engineer to develop a program of predictive analysis

focusing on vibration and lubricating oil analysis. Evidence of

problems associated with aging of plant equipment during the

assessment period included:

a. Several examples of end of service life for solenoid valves

on the turbine stop valve, diesel fire pump (DFP), and the

exhaust ventilation downstream isolation valve.

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b. Deterioration of fuel delivery system on the DFP.

c. Failure of several motor operated valves to operate on

demand, including the turbine bypass isolation valve, the

recirculation pump suction valve, turbine stop valve, and

the shutdown system reactor isolation valve.

A regionally based inspection performed in response to a

, declining performance trend identified in SALP 5, pointed out

! weaknesses in the PM program including failure to update the

progri.a based on plant experience, inadequate root cause analysis,

and inadequate consideration of the generic implications of

maintenance action. The report recommended a more comprehensive

method of evaluating potential end-of-service-life failures.

Another regionally based inspection assessed the adequacy of the

licensee's response to Generic Letter 83-28 and determined the

licensee was generally meeting the requirements in the areas of

vendor interface and post maintenance testing. The report noted

the lengthy delays in implementation of the vendor interface

program and inadequacy of post maintenance testing instructions

and documentation.

For the last half of the assessment period the site engineering

group has functioned under the Maintenance Department, an

organizational move intended to improve coordination between the

engineering and maintenance functions. The engineering group

seems slightly overburdened, a situation compounded by lack of

consistent prioritization of project assignments. Engineers were

regularly redirected from one project to another based on manage-

ment's sense of urgency over a given engineering project. The

licensee, at the end of the assessment period, performed an

inventory of all engineering projects and has devised a system

of consistent prioritization which should alleviate this problem.

The quality of modification packages prepared by the site

, engineering group is consistently high, reflecting the group's

extensive familiarity with the facility and a genuine interest in

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the safe and successful operation of the plant. Sound engineering

judgement that stresses safety and reliability is evident. Some

members of the staff do not consistently identify and incorporate

into their proposals and designs the quality requirements derived

from the various codes and regulations, relying instead on review

, by the Quality Assurance group to identify all the requirements.

! The deficiencies in the Nuclear Operations Department Standards

(N0DS) discussed in Section IV.I contribute to this problem.

A communication barrier exists between members of the engineering

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and mechanical maintenance staffs, and the knowledge of mechanics

is not routinely conveyed to engineers or factored into design

decisions. A notable example is information gathered by mechanics

during disassembly and cleaning of RDS valve top assemblies which

never made its way to the engineer in charge of the project.

This resulted in the repeated failure of the RDS valves.

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_

.

.

While licensee management is generally informative and cooperative

with NRC inspectors, there is only a marginal level of respon-

siveness to NRC initiatives displayed. Compliance with regulatory

requirements is generally adequate, but mediocrity or deficiencies

in performance or programs is tolerated and often justified by

citing budgetary and manpower constraints. Management action in

the areas of preventive maintenance, mechanical training

upgrading, and resolution of long standing engineering projects

is marginal. Management's lack of effective control of the

outage process was a major factor in the events during the 1985

outage discussed in Section IV.H. The reorganization of both the

maintenance and engineering functions under one Superintendent

appears to be too much activity for any one individual to

effectively manage, contributing in part to the licensee's

commencement of the 1985 outage with incomplete engineering

projects, inadequate scheduling of maintenance activities, and

deficient material procurement to support planned work.

2. Conclusions

The board rates the licensee Category 2 with a declining trend

based primarily on insufficient management control over the

maintenance process.

3. Board Recommendations

The board notes that this is the second consecutive assessment

period of declining performance and special management attention

is needed to offset the effects of aging equipment.

.

D. Surveillance and Inservice Testing

1. Analysis

During this evaluation period the resident inspector regularly

observed licensee performance in this area. These inspections

included observations of technical specifications required

surveillance testing to verify adequate procedures were used,

that instruments were calibrated, and that test results conformed

with technical specifications and procedure requirements. In

addition, all or part of four regional inspections were conducted

in this area. These inspections reviewed startup core perfor-

mance, Containment Integrated Leak Rate Tests, intergranular

stress corrosion cracking, and inservice testing.

Big Rock Point uses a manual tracking system to schedule

performance of operational surveillance of mechanical, electrical,

and Instrumentation and Control (I & C) components and systems.

Each surveillance procedure is sponsored by a knowledgeable

individual, and the mechanism exists for revision to the

procedure based on performance experience. Surveillances are

generally taken seriously by those performing the test and not

run to simply satisfy a requirement. Two surveillance tests

were overlooked during the evaluation period, including daily

15

_ _ _ _ - - _ _ _ .

._ __ _ . - __ .

.

.

control rod drive exercises and test of fire detectors in the

recirculating pump room. Cumbersome administrative controls

over fire detector tests contributed to the pump room detector

,

-

omission.

1'

During the evaluation period, one unresolved item resulted from

a concern over the frequent lack of detail in instructions and

documentation of post maintenance testing when work orders and

equipment outage requests are used to meet the post maintenance

testing requirements of Generic Letter 83-28, Sections 3.1 and

3.2.

An inspection reviewed the licensee's Inservice Inspection (ISI)

program after the corporate ISI group was disbanded in favor of

inder endent program administration at each plant. The inspection )

'

reviewed the Big Rock 1985 ISI Examination Program Plan and the

licensee's outage plan and found them to be acceptable.

Observations of ISI activities shows plant personnel have an

adequate understanding of work practices and adhere to procedures

that are generally well defined. Records are generally complete,

and indicate that equipment and material certifications are kept

current.

This inspection also reviewed the licensee's inspection program

to detect intergranular stress corrosion cracking (IGSCC) in

i

large diameter recirculation system piping to verify that the

actions set forth in Generic Letter 84-11 were performed. The

inspection determined the acceptability of inspection procedures

and techniques, documentation, and examiner qualifications.

4

In reviewing the licensee's containment integrated leak rate test

(CILRT), the inspector noted that the activity was adequately

staffed with knowledgeable individuals experienced in the Big

Rock unit. No specific training of the participants had preceded

the event and the licensee's familiarity with Type A testing

requirements was weak. Licensee management involvement in

supplemental verification testing was considered marginally

acceptable as evidenced by efforts to complete the Type A test ,

,

before acceptable supplemental verification test data was '

1

obtained.

In the area of startup and surveillance testing programs

subsequent to the refueling outage, the inspector concluded that

licensee personnel appeared to understand technical issues and

had a genuine interest in plant operations, providing timely and

thorough responses to inspector identified concerns. Procedures

appeared to be well written and employed a technically sound

methodology.

l

2. Conclusions

The licensee is rated Category 2 in this area. This is a decline

in performance from the last assessment period based primarily on

the missed surveillances.

16

,

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

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

.

i .

3. Board Recommendations

None.

E. Fire Protection

1. Analysis

I

i During this assessment period the resident inspector routinely

observed licensee activities in the fire protection functional

area, including routine housekeeping. One special inspection was

conducted by Region III personnel to assess the licensee's

compliance with 10 CFR 50, Appendix R, close out previously

identified open items, and verify compliance with routine fire

2

protection program requirements. The report has been delayed

while the staff completes the review of Big Rock Point's '

compliance with Appendix R, and will be addressed in SALP 7.

'

The licensee's ability to respond to fire alarms both in

preplanned drills and actual alarms was observed on several

occasions with satisfactory results that indicated the effective-

ness of fire brigade member training and fire response procedures.

Licensee personnel received hands-on fire training and training

in use of the self contained breathing apparatus. Licensee

personnel are generally knowledgeable about fire prevention.

'

Housekeeping has improved from the previous assessment period.

Housekeeping during plant operation is generally of high quality

and accessible plant areas of the facility are routinely policed.

Maintenance workers generally clean up after their job is

i complete. Cleaning lockers and assigned areas are part of a

'

housekeeping system that is incorporated into the daily routine.

Supervisors are regularly in the plant and monitor cleanliness

levels, taking action as appropriate. Housekeeping during

!

extended outage periods, as it relates to both fire protection

and contamination control, declines noticeably from periods of

normal operation. While plant appearance deteriorates during

extended outages, post-outage cleanup is generally prompt and

thorough and reflects management involvement.

During NRR visits to the plant, the staff was impressed with

the clean, well ordered appearance of the plant. Even during

construction of the Alternate Shutdown Building, cc struction

materials and supplies were well controlled. The Control Room

appeared very well run and well organized in terms of reference

materials and drawings. '

Throughout the evaluation period the licensee experienced

difficulties with the dependability of the diesel fire pump

(DFP), including end of service life for a solenoid coil on the

fuel supply shutoff valve, sluggish behavior and slow start

times that required cleaning in the fuel delivery system, and

a leaking fuel filter. The greatest cause for concern about the

3

DFP reliability arose during February, 1986, when excessively

,

17

.

.

long start times and erratic starting behavior were corrected by

further cleanings, tightening, and adjusting. Efforts to

diagnose and correct problems were hampered by the age of the

engine and the unavailability of parts and diagnostic instruments.

These facts combined with a shortage of vendor representatives

experienced on older engines make future repairs unlikely. These

factors, combined with the DFP's Core Spray function, make

replacement of the DFP a high priority for the 1986 outage.

The licensee has committed to replace the engine at that time.

2. Conclusions

The licensee is rated Category 2 in this area. There is

improvement noted in housekeeping.

3. Board Recommendations

None.

F. Emergency Preparedness

1. Analysis i

Three inspections were conducted during the assessment period to

evaluate the licensee's performance with regard to emergency

preparedness. These included two routine inspections of the

emergency preparedness program and observation of the licensee's

annual emergency preparedness exercise. Two violations were

identified as follows:

a. Severity Level V - Failure to evaluate the adequacy of

interfaces with State and local governments as part of the

annual audit as required by 10 CFR 50.54(t) (155/84014).

b. Severity Level V - Failure to conduct Health Physics drills

in 1983 as required by the Site Emergency Plan (155/84014).

The above violations were the result of isolated administrative

breakdowns in the emergency preparedness program and not indica-

tive of any major programmatic problem. In both cases the  ;

licensee took prompt corrective actions to resolve the violations

and ensure that they would not reoccur.

Management involvement in assuring quality is evidenced by the

fact that corrective actions are effective as indicated by the

lack of repetition of identified weaknesses. Management support

is also shown through the significant corporate assistance in the

training program and in the planning and conducting of exercises.

During the emergency preparedness exercise, licensee management

demonstrated an above average command and control capability

and were effective in carrying out their assigned emergency

responsibilities.

18

l

)

.

.

The licensee continues to be responsive to NRC concerns.

Violations and weaknesses that are identified are almost always

resolved in a timely manner and demonstrate technically sound and

thorough approaches. This is evidenced by the fact that few

issues of concern are identified by the NRC, and those that are

have generally been resolved by the next inspection.

Staffing of key emergency response positions has been adequate

with the authorities and responsibilities of personnel identified.

The licensee has a Senior Nuclear Emergency Planning Coordinator

position at the site, which has been generally adequate to main-

tain the daily emergency program activities at an acceptable

level of performance. Knowledge and capability of personnel to

carry out their assigned emergency response duties and responsi-

bilities was demonstrated during both the annual emergency

preparedness exercise and through walkthroughs of personnel

during the routine inspections. The licensee's performance in

these areas is indicative of an effective training program that

has adequately prepared personnel to carry out their emergency

response assignments. Examination of the training program and

observation of several training sessions during the last routine

inspection determined that the program was sufficiently thorough

and well conducted.

However, several events during the assessment period indicated

awkwardness with interpretation of reporting requirements and

emergency event classification. An example of this was the

notification to NRC Headquarters on May 25, 1985 of Unit

shutdown, which did not advise of the declaration of the

Unusual Event. During these events the licensee's capability

to interpret reporting requirements and classify the events was

less than the level of performance demonstrated during drills,

exercises, and inspection walkthroughs.

2. Conclusions

The licensee is rated Category 1 in this area. The licensee was

rated a Category 1 in this area in the last two SALP periods

which reflects the continued effectiveness of the emergency

preparedness program.

3. Board Recommendations

None.

G. Security

1. Analysis

Two inspections were conducted by region based inspectors during

this assessment period. The resident inspector also conducted

periodic observations of security activities. No violations were

noted during the inspection efforts. l

19

. ___ _ _ _ - _ _

.

.

Several allegations pertaining to alleged deficiencies with the

licensee's security program were received from a member of the

public during this evaluation period. The investigation and

resolution of the allegations have extended beyond the close of

this evaluation period and will be addressed in a future

inspection report.

The licensee has been generally responsive to resolving NRC

concerns. An inspection conducted early in the assessment period

(November 1984) identified the need for revision of the security

plan and some supporting implementing procedures. The most

significant concern pertained to training methods for newly hired

security force officers. These concerns did not constitute

violations or enforcement issues and were generally administrative

in nature. However, they were indicative of security management's

need to more closely monitor the administrative aspects of the

security program. All of the concerns were reviewed during a

February 1986 inspection, and the licensee's actions were

considered adequate to resolve the concerns. The site and

corporate security staff have provided timely and sound technical

solutions to inspection findings.

The February 1986 inspection noted that the morale of the security

force was low but had not deteriorated to the point where job

performance was affected. The primary cause for the morale

,

concern was attributed to long-term labor relation concerns

beyond the immediate control of the licensee. Licensee management

was aware of the concern and was addressing the issue, within

existing labor relation constraints. Deterioration of certain

security equipment was also noted and the licensee committed to

resolve the issue in a timely manner. The licensee needs to

, continue to be sensitive to required maintenance for aging

security equipment.

Only one security event was reported during the assessment period.

The event pertained to degradation of a vital area barrier and

did not constitute an enforcement issue.

Training and performance of the security force continued to be

maintained at a high and consistent level during this assessment

period as evident by the excellent enforcement history and lack

of reportable events caused by personnel error. Supervision of

day-to-day operations appears strong.

Corporate security support appears adequate. Licensing issues

are responded to in a timely manner and analysis of such issues

are generally thorough and technically sound. Inspection results

are closely monitored by the corporate security office and the

corporate office responds in a timely manner to help resolve

1 inspection findings and concerns. Audit functions by the

corporate security office appear adequate.

'

,

20

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

-

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

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

.

.

2. Conclusions

The licensee is rated Category 1 in this area based on

j demonstrated good performance by the uniformed security force

'

members and no violations being cited during this assessment

period. In spite of that the trend is declining based on the

,

aging security equipmen and continued low morale of the guard

~

force.

3. Board Recommendations

None.

1

H. Outages

,

1. Analysis

The Resident Inspector performed routine inspections during

outage periods and one inspection by a Regional Inspector

reviewed refueling activities. These inspections included

observation of maintenance activities including administrative

requirements, review of planning activities, refueling activities,

plant modifications, and post outage testing. One violation was

issued as follows:

Severity Level III - this violation combined in the aggregate

seven identified violations stemming from three separate examples

during the 1985 outage of supervisory personnel, repairmen, and

operators circumventing or ignoring administrative requirements

and not exercising sufficient care and attention to detail to

ensure plant safety. Contributing to the situation was the lack

) of component identification throughout the facility, the absence

1 of a single point supervisory contact to direct the activities of

i travel repair crews, inadequate management involvement in

i directing maintenance activities during the outage, and evidence

of a lackadaisical attitude on the part of certain operators

,

toward adherence to procedural requirements.

During the assessment period the licensee conducted one refueling

'

outage. Originally scheduled for 53 days, the outage was extended

10 days due to delays associated with repairs to feedwater and

poison system valves, turbine alignment troubles, and the dis-

assembly of incorrect valves which was the subject of the

violation noted above. Despite the delays, a significant

number of major outage activities were successfully completed,

including ISI/IGSCC inspections, electrical equipment environ-

mental qualifications modifications, and installation of the

alternate shutdown panel. The licensee completed 1100 main-

tenance orders, eight facility changes and 18 specification

field changes.

!

21

4

-

- . - . _ __ . _ _ ___

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

- . ._ _ . _. -._ _ ___

.

.

Operations Department personnel performed fuel handling

operations for the 1985 refueling outage. Fuel handling was

safely conducted by adequately trained individuals in accordance

with approved procedural requirements. Staffing on both the

reactor deck and in the control room was adequate, and communi-

cation between the two areas was effective. Management involve-

ment in refueling activities was evident. Tool control and

status board maintenance was adequate. Licensee responsiveness

to NRC initiative was evident by their prompt action to correct

procedural deficiencies in data recording and in relocation of

bagged equipment that had obstructed access to the. refueling

deck status board.

During the 1985 refueling outage several incidents occurred

which demonstrated inadequate management control over the outage

process. The incidents involved:

• Repeated examples of contractors and licensee travel crew

j personnel, not normally assigned to Big Rock Point,

performing work on the wrong component or system, pointing

i to inadequate control over the activities of travel crews

and contractors.

• Repeated examples of supervisors, maintenance, operations,

and engineering personnel, and travel crew personnel,

circumventing or failing to adhere to administrative

requirements, particularly those related to component

tagging and isolation.

• Repeated examples by individuals, throughout the

organization, of inattention to detail and failure to

exercise sufficient care in performance of outage related

work to ensure plant safety.

Several factors contributed to the breakdown in the outage

'

management process:

• Throughout the facility, components, valves, and syste;ns

identification was generally inadequate, with many compo-

nents unlabeled. The licensee had not acted upon earlier

requests from the Resident Inspector to improve component

identification and discounted warnings on the potential for

mishaps.

• Forced retirement of several older key members of the

licensee staff, including the Operations Superintendent,

the coordinator of the ISI program, an experienced Shift

Supervisor, and a Maintenance Supervisor who in the past

had acted as a coordinator and single contact point for

control of travel crew personnel. The impact of the loss

3

of these individuals two months prior to commencement of

22

-. ._ __ _ _ - - - .

.-

-

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

-. - - = . .. . .- - - - - . . - _ _ . .

.

.

the outage was exacerbated by a major reorganization of

the plant staff with reassignments of functional depart-

ments, creation of new departments, and redistribution of

duties within the Maintenance and Operations Departments

immediately prior to the outage.

,

• In the absence of a single point contact to direct and

coordinate the activities of travel crew personnel, manage-

l ment involvement in directing maintenance work was

j inadequate.

'

• Training provided for travel crew and local licer.see

'

personnel on tagging and isolation was inadequate.

l * Outage planning, including parts procurement and job

sequencing of specific work activities was inadequate.

Design work on many facility changes was incomplete at

outage commencement.

• Licensee travel crews were inadequately supervised and did

not display the same level of concern for reactor safety

4

normally in evidence among Big Rock personnel.

• Work crews assigned a particular task often were comprised

entirely of travel crew members without the guidance and

experience of Big Rock employees.

! * Travel crew supervisors invested too little time and effort

in inspecting and planning a specific job activity and in

instructing their workmen on the job's performance.

'

.

l * A lackadaisical attitude on the part of certain personnel

i toward attention to detail was a major contributing factor

l in the events.

1

I' The licensee has enjoyed decades of safe and successful reactor

operation resulting primarily from the professional attitude

i

displayed by talented and experienced individuals. The plant's ,

,

limited staff and small physical size makes the outage process a

J manageable activity. The events of the 1985 outage appear to

4

have impressed upon the licensee the need to aggressively manage

outages. The corrective actions in response to the 1985 events

have been comprehensive and include:

l * An expanded component identification program.

!

• A photograph book of the plant to aid in job planning.

j

• Counseling and disciplinary action for personnel involved in

the problems.

l * Expanded training for Big Rock and travel crew personnel.

!

23

r

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

-- _ _ _ _

.

.

• Procedure upgrading for valve preventive maintenance.

• Controls on future outage activities of travel crews,

including single point contact and prejob planning.

In addition, the licensee has created new groups to handle

material procurement and control and outage planning and

scheduling.

The licensee's responses to NRC concerns in response to the

incidents showed a desire to communicate and cooperate. The

Plant Superintendent made separate visits to regional and NRR

management to present licensee programs to correct programmatic

and management deficiencies. Prior to the close of the

evaluation period no outage activities were conducted that

would permit evaluation of the licensee's corrective action.

2. Conclusions

The licensee is rated Category 3 in this area.

3. Board Recommendations

The licensee should aggressively implement the corrective

actions noted previously.

,

I. Quvlity Programs and Administrative Control Affecting Quality

1. Analysis

Throughout the rating period the Resident Inspector routinely

reviewed the activities of the Quality Assurance (QA) and Quality

Control (QC) groups. This included administrative controls for

maintenance and operations as well as deviation reports and

quality control department involvement in accordance with the

QA Plan. In addition, this functional area was examined by the

region addressing the adequacy of site QA staffing levels and

qualifications in light of increased work load and the impact of

Nuclear Operation Department of Standards (NODS) deletion at the

facility. The QC activities associated with disposal of low

level radioactive waste under 10 CFR 20 and 10 CFR 61 were

reviewed in Inspection Report 155/85006(DRSS), resulting in one

violation discussed in section IV-B. The violation resulted from

the inadequacy of the licensee's procedure governing the shipment

& of radioactive materials to provide for determination of the

correct waste classification, although worksheets used to

classify the three waste shipments made since the regulation

became effective resulted in the correct waste classification.

The site QA and QC staffs are comprised of a generally adequate

number of qualified individuals with site experience who demon-

strate a high degree of professional conduct and integrity.

24

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

.

.

During the evaluation period there was eviJence that the site QA

staff was in danger of becoming overburdened by assignment of

'

several functions formerly performed by the corporate QA group.

Those added duties were subsequently completed or reassigned

elsewhere and the site staff appears adequate for the remaining

workload. The site QA staff communicates effectively with plant

management and is persistent in pressing for management action

to resolve audit findings. The Plant Review Committee (PRC)

considers the quality aspects of technical and safety issues.

In turn, plant management generally demonstrates their regard

for the significance of findings and comments from the QA staff.

Site QC inspectors are generally thorough and conscientious and

draw heavily on their plant experience. Both the QA and QC site

i staff are responsive to NRC initiatives and inquiries.

Licensee corporate management detracted from the effectiveness of

Programs and Administrative controls affecting quality. Examples

include:

a. Licensee corporate management, by transferring to the site

staff several significant Quality Assurance functions with-

out a corresponding increase in available site resources,

placed a burden on the staff which resulted in QA reviews

that were less comprehensive, withdrawal of commitments to

support audit activities off site, and a virtual elimir.ation

of time available to auditors to review and observe activi-

ties in the plant. Some QA functions were performed by QC

inspectors. The reluctance of corporate management to

respond to the concerns of the site QA Superintendent in

i this regard and their poor response to NRC initiatives to

address the issue was noted.

f b. Licensee Corporate management deleted entirely fifteen N0DS,

i the document in which the licensee staff can theoretically

j be assured of finding all applicable code and regulatory

requirements compiled in one location. The N0DS are the

j means by which the licensee's Quality Assurance Program

! Description for Operational Nuclear Power Plants (Topical

! Report CPC-2A) is implemented, and results from a commitment

j made in the licensee's Regulatory Performance Improvement

Program submitted in response to a March 9, 1981 Confirmatory i

4 Order. Wholesale deletion of the N0DS without a review to

! insure all of the quality requirements contained therein

]

were already addressed in existing administrative procedures

resulted in a period when the quality requirements were not

'

available to the N0DS user. Inspectors identified at least

two examples of cancelled N0DS being referenced in other

procedures.

I

1

!,

4

25

i

.

.-, , - . - - - - , - - - ,.,r-__ - .

e- - - . - , - _ , , , - , - - - ..,,,,.,v-,-m,w,,-,--,.r--+,, ,v- -_m, v- y-- , 4,-v-,-

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

.

.

c. The findings of the licensee's team that N0DS development

was incomplete and that an inadequate review and approval

process allowed the issuance of N0DS with a significant

magnitude of deficiencies relative to CPC-2A basis documents

went unacted upon by management.

'

During the evaluation period the licensee designed and

implemented a program to reduce QA involvement with reviews of

procedures in departments where there was long term evidence of

high levels of quality performance. The program was implemented

late in the period with the licensee's stated goal of redirecting

auditor resources into areas of poorer performance.

2. Conclusions

The licensee is rated a Category 2 in this area. The exemplary

level of performance by the site QA and QC staff is offset by

our concerns with the actions of corporate management.

3. Board Recommendations

None.

J. Licensing Activities

1. Analysis

a. Methodology

,

The basis of this appraisal was the licensee's performance

in support of licensing actions that were either completed

i

or active during the current rating period. These actions,

consisting of license amendment requests, exemption requests,

relief requests, responses to generic letters, TMI items,

LER's, and other actions, are summarized below:

(1) Amendment Requests

Technical Specifications (TS) Defining Operability

for Safety Systems

Containment Pressure and Water Level Monitor TS

Reporting Requirements TS

TS Change Section 6 - Plant Staff Reorganization

TS Change for Surveillance Frequencies

Control Rod Testing Frequency

Incorporation of Byproduct License

^

Cycle 21 Reload Il Fuel TS Change Package

Administrative TS

Gamma Monitor Calibration Frequency

Control Rod Withdrawal Rate Limit TS

26

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

.

.

Post Maintenance Testing TS Change - Item 3.2.3

PRC Approval Method TS

CRD Performance Testing Frequency TS

Auto-Isolation (CV-4049) TS

Stack Gas Monitoring System TS

Organizational TS

Appendix I TS Implementation Review

Administrative TS Changes Related to RETS

Integrated Program Plan (ILS)

Appendix "R" Alternate Shutdown System TS

(2) Exemption Requests

ATWS Recirculation Pump Trip

Containment Airlocks

Reporting Requirements - Spurious RPS Actuations

Fire Protection

Equipment Environmental Qualification

High Point Coolant System Vents

(3) Relief Requests

In-Service Testing

In-Service Inspection

(4) TMI Items

I.C.1, Emergency Operating Procedures

I.D.1, Detailed Control Room Design Review

I.D.2, Safety Parameter Display System

II.B.1, Reactor Coolant System Vents

II.D.1, RV and SV Testing

II.F.1, Accident Monitoring

II.F.2.3, Inadequate Core Cooling Instrumentation

III.A.1.2, Emergency Response Facilities

III.A.2.2, Meteorological Data Upgrade

(5) Other Licensing Actions

Control of Heavy Loads

BWR Pipe Cracking

Salem ATWS Follow-up

Electrical Equipment Qualifiction

Systematic Evaluation Topics

Fire Protection Modifications

Diesel Generator Reliability

Retention of Contaminated Soil Onsite

During the SALP period, 67 licensing actions were completed

which consisted of 45 plant-specific actions, and 22 multi-

plant actions including nine TMI (NUREG-0737) actions.

27

__ -___ _______ . _ _ _

.

.

A very important licensing activity completed during the

review period was the formalization of the Big Rock Point

Integrated Assessment. License Amendment No. 82, " Plan for

the Integrated Assessment," issued February 12, 1986,

incorporates the requirement to adhere to the " Plan," as

documented in License Condition (7) of Big Rock Point

Facility Operating License DPR-6. This achievement is

noteworthy as Big Rock Point is one of the industry

leaders in terms of long-term program implementation.

In addition to these licensing activities the project

manager and other members of NRR participated in an

in progress audit of the licensee's Detailed Control Room

Design Review process as well as 10 CFR Part 50, Appendix R

related modifications,

b. Management Involvement and Control in Assuring Quality

Licensing activities for Big Rock Point show consistent

evidence of prior planning and assignment of priorities and

decision making is almost always done at a level that ensures

adequate management review. The cornerstone of the

licensee's efforts in this area is the Big Rock Point

Integrated Assessment (termed the Plan). The licensee

adopted this integrated approach to licensing issues in

early 1983. Much of the initial assessment was completed

during the last evaluation period; however, the incorpora-

tion of the Plan was completed during this evaluation period.

As part of an on going process, the licensee makes safety

judgements based on the use of the Big Rock Point Proba-

bilistic Risk Assessment as well as standard safety assess-

ment methods to ensure that plant safety is optimized in a

cost-effective manner. The Plan governs the implementation

of significant facility changes.

As presented above, there have been a significant number

of licensing actions processed, and for the most part, the

majority were completed requiring little or no additional

information or meetings. Adequate management control was

not exercised, however, in the handling of the Reactor

Depressurization System (RDS) Valve Testing Technical

Specification Change Request to reduce surveillance testing

frequency. The request showed a lack of prior planning and

the technical evaluation was not thorough. This RDS issue

has been ranked by the licensee as the most important

current facility project as described in Integrated Plan

Update No. 4. NRR agrees with the licensee's ranking and

believes a continued strong management involvement for

assuring quality on this project is needed.

An area in which Big Rock needs to focus more attention is

in their safety evaluations generated to support submittals

to NRR involving proposed license amendments. Examples of

safety evaluations which we found to be less than adequate

28

_ __

b

.

.

were the application for the incorporation of the byproduct

license and the application related to the corporate

reorganization. Both applications required extensive NRC

efforts to evaluate the impact of the proposed changes.

Also, the depth of explanation of the no significant hazards

consideration (NSHC) determinations could be improved. It

should be noted that the applications presented above were

evaluated during the first half of the evaluation period;

and we have noted improvement over the past year.

During the last half of the evaluation period, the licensee's

evaluations have been well stated, understandable, and

,

showed consistent evidence of prior planning. Most of the

,

applications received have been timely, thorough, and showed

decision making consistently at a level that ensures

'

adequate management review.

We recognize the strong improving trend; however, Big Rock

i must be keenly aware of their unique plant design and as

such should strive to fully present complete information

to the staff. The key point being that the audience to

which Big Rock is presenting their SEs, in some cases, is

not as familiar with plant-specific design features unique .

i

to Big Rock, and therefore, a conscious effort should be j

made to present more information to better understand a

', given issue.

!

c. Approach to Resolution of Technical Issues from a Safety

3

Standpoint

l

The licensee generally demonstrates understanding of the

technical issues involved in licensing actions and proposes

technically sound, thorough, and timely resolutions.

, However, there have been issues where the licensee's

approach was good, but the licensee did not thoroughly

understand NRR staff guidance. Once the staff guidance  :

was fully explained, the licensee proposed timely resolutions l

l which were technically sound and exhibited proper conserva-

tism. For a few issues, full explanation of the staff

guidance required an above average amount of staff effort. ,

Examples of such issues are Incorporation of Byproduct

'

. License, RDS Valve Testing, and Environmental Equipment

Qualification.

It should be noted, however, that the issues presented above

were evaluated rarly in the evaluation period. During the

'

last half of the evaluation period, the licensee has

demonstrated a clear understanding of the issues, appropriate

conservatism when the potential for safety significance

existed, and generally sound and thorough approaches. This

reflects positively on Big Rock Point's willingness to work

closely with the staff. l

I

i

29

.

.

d. Responsiveness to NRC Initiatives

The licensee's initial responses to NRC initiatives almost

always contain acceptable resolutions, provide for timely

resolution of issues, always met deadlines and were generally

sound and thorough. Although the assessment for this

attribute was determined to be near average for the first

half of the evaluation period (due to the Incorporation of

the Byproduct License, RDS Valve Testing, and Environmental

Qualification of Electrical Equipment), the performance of

the licensee for this attribute during the second half of

the evaluation period was excellent. We attribute this,

in part, to the willingness of the plant manager to take

control and ensure mutual goals are attained.

e. Enforcement History l

1

This area is addressed in other functional areas of this

report.

f. Reporting and Analysis of Reportable Events

The Big Rock Point plant operated at power during most of

the report period, except for about two months of refueling

outage from September 6, 1985 to November 7, 1985, and short

periods of shutdown for other causes. In a period of about

eight months (from January 1, 1985 to September 6, 1985),

the plant operated with a Reactor Service Factor * of 82%.

In the 17 months covered by this SALP evaluation, the

licensee reported eight** events to the NRC Operations

Center as required by 10 CFR 50.72. Three unusual events I

concerning mechanical and electrical failures were also I

reported. One of the unusual events reported dealt with

the shutting down of the unit from 91% power on December 31,

1984 due to failure of the reactor depressurization system

(RDS) valves to pass as 'urveillance test. Failure of the

RDS valves was noted in the previous SALP report on this

plant. The repetition of the RDS valve failure suggests

that the licensee needs to give more attention to follow-up

analyses and actions. Two of the three unusual events,

including RDS valve failure, resulted in entry into limiting

condition for operation (LCO) action statements. During

this report period, 12 Licensee Event Reports (LERs) per

10 CFR 50.73 were received.

• Reactor Service Factor = (Hours of Critical Reactor

Operatior./Possible Hours) x 100%.

• • The number of events reported to the operations center may

not be the same as the number of Licensee Event Reports

because of different reporting criteria and in some cases

an event initially reported to the operations center may be

reassessed as not reportable.

30

.

.

Of the eight 50.72 reports, two reports involved reactor

scrams which occurred in 1985. These scrams were manually

performed at 10% and 15% power. This reactor trip frequency

of two per year compares favorably with the current national

average frequency of 5.9 trips per year.

Of the remaining six 50.72 reports, four reports involved ,

reactor protection system actuations due to a spurious '

signal resulting from electrical noise affecting power level

instrumentation at low power levels. Two of the spurious

RPS actuations involved no rod movement, while a third

occurred during control rod drive testing and resulted in

the insertion of the single withdrawn control rod. The

fourth actuation occurred at 0.1% power while shutting down

for routine maintenance. One report dealt with the loss of

emergency notification sirens. The last of the 50.72

reports pertained to a discovery that a support hanger for

the reactor depressurization system had not been preten-

sioned after a system hydro several years ago (3-6 years)

due to what the licensee called a procedure inadequacy.

Although this incident represented a fourth unusual event,

the licensee failed to inform the NRC that an Unusual Event

had been declared until securing from that classification.

None of the reportable events was considered individually l

significant enough to warrant detailed NRR staff follow-up.

None of the events reported during the period was discussed

,

at the Operating Reactor Events Briefings.

g. Staffing

The licensee has a licensing staff which appears to be

sufficient to provide adequate and timely responses.

Positions are identified and authorities and responsi-

bilities are well defined. The CPC licensing contacts

for the NRR licensing Project Manager at the facility and

in the Corporate Office have or once held an SR0 license.

Because of the Operations experience of these contacts many

technical issues can be' resolved on initial contact with

the licensee.

Management attention and involvement was generally aggressive l

and disciplined. This was evident in both the safe and efficient i

operation of the facility. Staffing levels and quality were '

adequate. Commurication levels between the operating staff and

proper management were established and generally effective. The i

licensee has been, in most cases, effective in dealing with  !

significant problems and NRC' initiatives. The licensee's  ;

attention to housekeeping appears to have been excellent. The

licensee's efforts in the functional area of Licensing Activities

has significantly improved during this evaluation period. This

is reflected in the quality of work, attention to NRR concerns

and involvement of senior management. Big Rock was an active l

participant at the counterparts meeting of January 30, 1986, and

31 l

1

_ ._. _, ,

.

.

their plant superintendent has visited Headquarters to give an

independent perspective of this concerns, and views regarding

major issues confronting Big Rock and the utility industry.

Thus, we see several trends which have brought this utility

upward in our evaluation scale. We note room for improvement

and all indications reflect a very positive attitude toward

continued improvement.

2. Conclusions

The overall rating for the functional area of licensing activities

is a Category 1. During this period, the licensee's performance

was found to be above average to excellent overall.

3. Board Recommendations

None.

K. Training and Qualification Effectiveness

1. Analysis

The resident and regional based inspectors regularly reviewed

training and qualifications during inspection of other areas and

review of events. No violations were identified in this area.

During the assessment period, NRC examinations were administered

to five Reactor Operator candidates. All candidates passed the

examinations. This passing rate is significantly above the

national passing rate. Based on these results, the operator

licensing training program at Big Rock Point is considered

satisfactory.

", During the evaluation period several instances were identified

where specialized training was conducted prior to non-routine

operations or maintenance activities. Examples include: I

a. Prior to installation of a Control Rod Drive with a unique l

modification the maintenance crew received instructions from

an experienced Superintendent using mock-ups.

b. The licensee's maintenance staff received training in the use

of new Control Rod Drive overhaul equipment by the Vendor,

General Electric.

c. Some training was conducted for Operation Personnel prior to

installation of spent fuel pool racks.

d. Walkthrough by Operation Personnel on Emergency Operating

Procedures under preparation served to familiarize the

operators and identify weaknesses in the procedures.

32

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

.

.

e. Extensive training was conducted for Operators, Superinten-

dents, On-call Technical Advisors, and Instrument and

Control Technicians prior to startup of the Alternate

Shutdown System required by Appendix R.

f. Reviews were conducted of procedures prior to installation

of recirculating pump seals.

g. Training was conducted for all personnel in use of Scott

Air Paks.

h. Hands-on Fire Training for all personnel was conducted.

One Example where training was inadequate was in the preparation

l of travel crew workers in the use of local control (equipment

tagging) procedures.

The training department routinely incorporates into system

training and requalification cycle training the information i

contained in all LER's, IE Notices, GE-SILS, Deficiency Reports,

and industry reports. i

Effective April, 1986, the Operations department instituted an On

the Job Training (0JT) program aimed at consolidation of five

former training programs leading to the SR0 license. The program

will use qualification cards. The program's effectiveness will

be evaluated during SALP 7.

Maintenance personnel during the period received virtually no

l

training. In February the licensee began sending maintenance

personnel to the Bay City Skills Training Department for General

Maintenance Training that is not specific to nuclear applications.

The Training Department has not received a request for Systems

Training for Maintenance Personnel.

The Training Department during the SALP period has added to its

staff several individ'uals with extensive experience in operations,

maintenance, or instrumentation and controls. This in plant,

hands-on experience contributes to the quality of lesson plans

and presentations. Students seem to exhibit a high degree of

respect for the instructors. Management involvement was reduced

because of the frequent temporary offsite assignments of the

Training Administrator.

There were no licensing actions which provided a clear

opportunity to judge this attribute. Based on interface with

CPC's licensing and operations personnel, it appears that the

training and qualification program makes a positive contribution

to the understanding of technical issues and adherence to

procedures with few personnel errors. Based on first-hand

experiences with operations personnel, the NRR licensing

Project Manager believes, however, that some improvement

could still be achieved.

33

__ __. _ _ _ _

.

O

2. Conclusions

The licensee is rated Category 1 in this functional area.

3. Board Recommendations

None.

i

l

i

l

l

34

__ _ _ _ _ _ _ . _ _ _

.

.

V. SUPPORTING DATA AND 9JMMARIES

A. Licensee Activities

The unit engaged in routine power operation throughout most of SALP 6

except for a scheduled outage for the 20th plant refueling which began

on September 6, 1985 and was completed on November 8, 1985.

The remaining outages throughout the period are summarized below:

December 31, 1984 - January 6,1985 Scheduled outage for

surveillance on RDS

valves l

l

April 5-17, 1985 Scheduled outage to l

repair RDS valves  ;

1

May 15-19, 1985 Outage to repair

recirculating pump seal

May 25-20, 1985 Shutdown to repair leak

on heat exchanger on

recirculation pump

May 26-27, 1985 Shutdown to repair leak l

on heat exchanger on

recirculation pump

November 14-18, 1985 Vessel flange

0-ring leakage

November 19-24, 1985 Vessel flange

0 ring leakage

December 7-8, 1985 Steam leak

February 11-17 RDS valves leaking

The plant scrammed six times (four occurred while the plant was less

than 0.1% power). In 1985, the two at power scrams were manually

initiated. One was caused by a failure to manually reset a feedwater

valve prior to plant startup while the other was caused by a minor

steam leak in the recirculation pump room. The four remaining scram

signals were caused by susceptibility of the picoammeters to

electrical noise at low neutron flux levels, a known operating

characteristic of the equipment with little safety significance.

B. Inspection Activities

An emergency preparedness exercise was conducted during the SALP

period by Region III involving observations by nine NRC representatives

of key functions and locations during the exercise.

Violation data for the Big Rock Point Plant is presented in Table 1,

which includes Inspection Reports No. 84013-86006.

35

_ _ _ _ _ _

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

.

.

.

Table 1

ENFORCEMENT ACTIVITY

FUNCTIONAL N0. OF VIOLATIONS IN EACH SEVERITY LEVEL

AREA

III IV V

A. Plant Operations 1 1  !

B. Radiological Controls 1

C. Maintenance / Modifications

D. Surveillance and Inservice Testing

E. Fire Protection

F. Emergency Preparedness 2

G. Security

H. Outages 1

I. Quality Programs and

Administrative Controls

Affecting Quality

J. Licensee Activities

K. Training and Qualification

Effectiveness

TOTALS 1 1 4

C. Investigations and Allegations Review

Several allegations pertaining to alleged deficiencies with the

licensee's security program were received from a member of the public

during this evaluation period. While no immediate safety concerns

were identified the investigation and resolution of the allegations

have extended beyond the close of t,his evaluation period and will be

addressed in a future inspection report.

D. Escalated inforcement Actions

A Severity Level III violation was issued early in 1986 for two

separate incidents which occurred in 1985 resulting from a failure of

supervisory personnel and repairmen to follow procedures. No civil

penalty was issued because of prior good performance and extensive

and comprehensive corrective actions.

E. Licensee Conferences Held During Appraisal Period

1. January 29, 1985 (Glen Ellyn, Illinois) i

{

Licensee presentation on history and operation of Reactor

Depressurization System.

1

36

_ _ _ _

,

,

.

2. March 12, 1985 (Glen Ellyn, Illinois) t

Meeting to review Systematic Assessment of Licensee Performance

(SALP 5).

3. October 1, 1985 (Glen Ellyn, Illinois)

Licensee presentation on new reorganization.

4. December 5, 1985 (Glen Ellyn, Illinois)

l

Meeting to discuss the breakdown in management controls of

'

plant work activities.

l

F. Confirmation of Action Letters (CALs) l

l There were no CALs issued during this SALP assessment.

G. Review of Licensee Event Reports, Construction Deficiency Reports, I

and 10 CFR 21 Reports Submitted by the Licensee

1. Licensee Event Reports (LERs)

LERs issued during the 17 month SALP G period are presented

below:

LERs No.

84-14

85-01 through 85-09

86-01 through 86-02

Proximate Cause Code * Number During SALP 6

Personnel Error (A) 3.

Design Deficiency (8) -#

-0

l External Cause (C) 5

Defective Procedure (D) 1

Management / Quality Assurance

Deficiency (E) 0

Others (X) 1

No Cause Code Marked ** 2

Total 12

• Proximate cause is the cause assigned by the licensee

according to NUREG-1022, " Licensee Event Report System."

• • NUREG-1022 only requires a cause code for component failures.

In the SALP 5 period, the licensee issued 27 LERs in 16 months l

for an issue rate of 1.7 per month. In the SALP 6 period the l

licensee issued 12 LERs in 17 months for an issue rate of 0.7

per month. Four of the LERS were submitted for RPS activation

known as " nuisance trips" resulting from electrical noise which j

gives an upscale /downscale trip signal at less than 1% power.  ;

37

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

.,

.

,

.

The licensee submitted a request to be exempt from this reporting

'

. requirement. This request was denied because the requirement

! will be revised to address this problem. The reduction in

!

overall LERs is indicative of an improving trend.

The office for Analysis and Evaluation of Operational Data (AE00)

reviewed the LERs for this period and concluded that, in general

i the LERs are of above average quality based on the requirements

'

contained in 10 CFR 50.73. However, they identified some minor

i deficiencies. A copy of the AE0D report has been provided to the

licensee so that the specific deficiencies noted can be corrected

i in future reports.

l 2. Construction Deficiency Reports i

f

i No construction deficiency reports were submitted during the

,

assessment period.

,

3. 10 CFR 21 Reports

No 10 CFR 21 reports were submitted during the assessment

period.

j H. Licensing Activities

i

) 1. NRR/ Licensee Meetings (at NRC)

!

! SALP 5 Region III 03/12/85

i Licensing Action Prioritizations 08/14/85

Maintenance Practice Discussions 10/01/85

j Enforcement Conference 12/05/85

. Counterparts Meeting 01/30/86

l Fire Protection 03/31/86

i

!

2. NRR Site Visits / Meetings

Plant /0rientation 11/07-08/84

! Plant Orientation for PM/PD 07/07-12/85

i Licensing Action Prioritization 10/02/85

{ Fire Protection 12/20/85

i DCRDR In-Progress Audit 1/27/-31/86

i

3. Commission Meetings

l

i

! None

!

i 4. Schedule Extensions Granted

i

{ Equipment Qualification 03/27/85

i  !

I l

! l

!

1

! -

i

38

l

_

.

s

e

5. Reliefs Granted

ISI Relief Requests (Revision 3) 11/01/85

ISI Relief Requests 12/12/85

6. Exemptions Granted

Appendix R, III.G.2 03/26/85

RCS High Point Vents 07/17/85

Containment Airlocks Testing 01/08/86

ATWS Recirculation Pump Trip 03/20/86

7. License Amendments Issued

Amendment Title Date

71 Plant Review Committee Review Process 12/10/84

72 Incorporation of Byproduct License 04/18/85

73 Control Rod Drive Performance Testing

Frequency 05/01/85

l

74 Containment Isolation Valve CV-4049 06/07/85

75 Stack Gas Monitoring System 06/10/85

76 Administrative Controls 07/01/85

77 Radiological Effluent Technical

Specifications 08/26/85

78 Definition of Operability & Associated

LC0 10/02/85

79 Surveillance Frequencies 10/22/85

80 Containment Pressure & Water Level

Monitor 10/29/85

81 Reload Il Fuel MAPLHGR Limits 11/01/85

82 Plan for the Integrated Assessment 02/12/86

83 Plant Staff Reorganization and

Administrative Changes 03/10/86

8. Emergency Technical Specifications

None

9. Orders Issued

None

10. NRR/ Licensee Managment Conferences

None

39

L. .

.

.

.. . _ _ . . . . .