ML20214G986
| ML20214G986 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/16/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214G943 | List: |
| References | |
| 50-423-85-98, NUDOCS 8705270232 | |
| Download: ML20214G986 (59) | |
See also: IR 05000423/1985098
Text
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ENCLOSURE
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE
INSPECTION REPORT NUMBER 50-423/85-98
MILLSTONE NUCLEAR STATION, UNIT 3
ASSESSMENT PERIOD: SEPTEMBER 1, 1985 - FEBRUARY 28, 1987
BOARD MEETING DATE: APRIL 16, 1987
8705270232 870514l
DR ADOCK 050004 3
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TABLE OF CONTENTS
PAGE
I. INTRODUCTION......................................................... 1
II. CRITERIA............................................................. 2
III. SUMMARY OF RESULTS................................................... 3
A. Overall Summary................................................. 3
B. Background...................................................... 4
1. Licensee Activities........................................ 4
2. Inspection Activities...................................... 4
C. Facility Performance Analysis Summary........................... 5
D. Plant Trips and Unplanned Shutdowns............................. 6
IV. PERFORMANCE ANALYSIS................................................. 10
A. Plant Operations................................................ 10
B. Radiological Controls........................................... 14
C. Maintenance..................................................... 18
D. Survei11ance.................................................... 21
E. Emergency Preparedness.......................................... 23
F. Security and Safeguards......................................... 25
G. Outage Management............................................... 28
H. Licensing Activities............................................ 30
I. Engineering Support............................................. 32
J. Training and Qualification Effectiveness........................ 35
K. Assurance of Quality............................................ 37
V. SUPPORTING DATA AND SUMMARIES........................................ 40
A. Investigation and Allegation Review............................. 40
B. Escalated Enforcement Actions................................... 40
C. Management Conferences.......................................... 40
D. Licensee Event Reports.......................................... 40
E. Licensing Activities............................................ 42
TABLES
Table 1 - Inspection Report Activities
Table 2 - Inspection Hours Summary
Table 3 - Enforcement Summary
Table 4 - Licensee Event Reports
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I. INTRODUCTION
The Systematic Assessment of Licensee Performance (SALP).is a periodic, inte-
grated NRC staff effort to collect observations and data and evaluate licensee. ;
performance. SALP supplements the normal regulatory processes used to ensure
compliance with NRC regulations. SALP is' intended to be sufficiently diagnos-
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tic to provide.a rational basis for allocating NRC resources and to provide ~
meaningful guidance to licensee management to promote quality and safety of ,
plant. operation. .
The NRC SALP Board met on April IC, 1987 to review performance observations
and data in accordance with the guidance in NRC Manual Chapter 0516, " System- '
atic Assessment of Licensee Performance". A summary of the guidance and
evaluation criteria is provided in Section II of this report. ,
This report addresses performance at the Millstone Nuclear Power Station, Unit
3 from September 1, 1985 through February 28, 1987. The_ findings'and data
reflect an 18 month assessment period. Although this includes activities
during construction and initial fuel loading, the evaluation of licensee per-
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formance has emphasized the period of power operation from January 23, 1986
through February 28, 1987.
The SALP Board was composed of the following:
Chairman:
W. F. Kane, Utrector, Division of Reactor Projects (DRP)
Members:
W. V. Johnston, Director, Division of Reactor Safety (DRS)
S. J. Collins, Deputy Director, DRP
E. C. Wenzinger, Chief, Projects Branch No. 3, DRP
R. R. Bellamy, Chief, Emergency Preparedness and Radiological Protection ,
Branch, Division of Radiation Safety and Safeguards (DRSS) (Part-Time) '
E. C. McCabe, Chief, Reactor Projects Section No. 38, DRP !
E. L. Doolittle, Project Manager, PWR Project Directorate No. 5, Division
of PWR Licensing-A, NRR (Part-Time) L
J. T. Shedlosky, Senior Resident Inspector .
Other Attendees (non-voting):
N. J. Blumberg, Acting Chief, Operational Projects Section, DRS (Part-Time)
E. L.- Conner, Project Engineer, DRP (Part-Time)
,
M. C. Kray, Reactor Engineer, DRP 3
W. J. Madden, Physical Security Inspector, DRSS (Part-Time) '
W. J.~Pasciak, Chief, Effluents Radiation Protection Section, DRSS.(Part-Time) i
J. A. Schumacher, Senior Emergency Preparedness Specialist, DRSS (Part-Time)
A. A. Weadock, Radiation Specialist, DRSS (Part-Time) '
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,. II. CRITERIA
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Licensee performance is assessed in selected functional areas. These areas
are significant to nuclear. safety and the environment, and are normal pro-
grammatic areas. The following criteria were used as appropriate to assess
[
l each area.
1. Management involvement and control in assuring quality.
-2. -Approach to resolution of technical issues from a safety standpoint.
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3. Responsiveness to NRC initiatives. !
4. Enforcement history.
5. _ Reporting and analysis of reportable events.
6. . Staffing (including management). i
7. Training effectiveness and qualification.
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Based upon the SALP Board assessment, each functional area is classified into
one of three performance categories. These are: i
. Category l'. Reduced NRC attention may be appropriate. Licensee management
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i attention and involvement are aggressive and oriented toward nuclear safety;.
licensee resources are ample and effectively used so that a high level of
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performance with respect to operational safety is being achieved.
j Category 2. NRC attention should be maintained at normal levels. Licensee
i- management attention and involvement are evident and concerned with nuclear
~ safety; licensee resources are adequate and reasonable effective such that
satisfactory performance with respect to operational safety is being achieved.
, Category 3. Both NRC and licensee attention should be increased. Licensee
- management attention or involvement is acceptable and considers nuclear safety,
but weaknesses are evident; licensee resources appear strained'or not effec-
- tively used such that minimally satisfactory performance with respect to
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operational safety is being achieved.
. .
[ The SALP Board has also categorized the performance trend over the course of
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the SALP assessment period. The SALP. trend categories are:
j Improving: Licensee performance was determined to be improving near the close
of the assessment period.
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Declining: Licensee performance was determined to be declining near the close
of the assessment period.
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A trend is assigned only when a definite trend of performance is discernible,
and the SALP Board believes that continuation of the trend may result in a
change of performance level.
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III. SUMMARY OF RESULTS
A. Overall Summary
l~ This assessment found a well-staffed licensee with strong and visibly
involved managers. Strengths were observed in self-identification of.
'. problems, in response to problems, and in searching for root causes.
There was diligent attention to proper performance at all levels, and
performance improved as the SALP period progressed.
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Operations phase programs, procedures, and management controls were in '
place and fundamentally sound. Minor setpoint errors and procedure in-
adequacies in the surveillance area were one of the few weaknesses found.
Program implementation was good in all areas.
There were four reactor scrams before initial criticality and 15 more 1
- during the subsequent year. While this is a high number,-it is consist-
l ent with the performance of similar plants during initial operation.
t Also, licensee responsiveness to these events resulted in a scram fre-
) quency decrease of about a factor of two during successive four month
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operating periods. -
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Review of scrams and other operating events points to personnel error,
equipment characteristics, and component failures (in that order) as the
main factors. In many cases, scrams were caused by a combination of
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I personnel error and the high degree of difficulty of steam generator
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level control. In addition to steam generator level control equipment
I performance improvement, this SALP identified a need for reducing the
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number of illuminated control room annunciators and improving the per-
l formance of equipment such as the Power Operated Relief Valves. A strong '
i program to reduce personnel errors and improve' equipment performance is
j needed. Improvements in scram and feedwater transient reduction and in
i the number of lighted annunciators indicate that the licensee's correc-
- tive action approach is working.
I
i Licensee command and control was notably good. Activities were carefully
1
' planned and conducted, with outages being a noteworthy example. Managers
were actively involved and inserted themselves into decision-making and
} activity direction at appropriate levels. Operating supervisors and
plant personnel were knowledgeable and alert, with strong corrective
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- action evident when a discrepancy in performance and supervision occurred.
l Overall, this SALP reflects careful and safe performance of initial plant
- operation.-
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B. Background
1. Licensee Activities
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Millstone 3 received a low power license (NPF-44) on November 25,
! 1985. Initial criticality was on January 23, 1986. A full power
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license (NPF-49) was issued on January 31, 1986. Power operation
- was first attained on February 3. The plant reached 100% power on
April 17, was declared commercial on April 23, and completed the
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100-Hour Warranty run on April 29.
There were four reactor scrams before initial criticality and 15
reactor scrams during the first year of operation. The major factor
i in the scrams was difficulty with steam generator level control,
a ' which contributed to 10 scrams. There were also two unplanned and ,
three planned outages to correct equipment deficiencies and perform
i surveillances. These outages and the reactor scrams are tabulated
in Section IIID (Page 6) of this SALP. The plant achieved an 86%
capacity factor for the commercial operating period beginning on
April 23, 1986 until the end of the SALP period on February 28, 1987.
2. Inspection Activities
l Two NRC resident inspectors were assigned to the site during the
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entire 18-month assessment period. The NRC inspections are summar-
4
ized in Table 1 and re
} (4790 hours0.0554 days <br />1.331 hours <br />0.00792 weeks <br />0.00182 months <br /> per year),present an inspection
distributed as shown in effort of 2.
Table 7130 hours0.0825 days <br />1.981 hours <br />0.0118 weeks <br />0.00271 months <br />' ;
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Special team inspections were made of operational readi. ness (April
14-24,1986); as-built pipe and supports, electrical, and instrument
and controls (September 9-20, 1985); and the site emergency exercise
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(November 19,1986.)
This report also assesses " Training and Qualification Effectiveness"
and " Assurance of Quality" as separate' areas. .These separate areas
! provide a synopsis of these topics, which.are also incorporated in
other functional areas through their use as evaluation criteria.
!. For example, assurance of quality was assessed on a day-to-day basis
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' by the resident inspectors and as an integral part of all specialist
' inspections. Although the management tools for measuring quality
include QA inspections and audits, quality work is the responsibil-
t ity of every employee. Major quality factors such as involvement
of first-line supervision, safety committees, and worker attitudes
are considered in each functional area.
! Fire protection was not addressed as a separate area during this
l SALP because 10 CFR 50,~ Appendix R implementation has not yet been
- specifically inspected onsite. Engineering support was added as
i a functional area to provide better focus on support functions which
[ were previously addressed in several functional areas.
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C. Facility Performance Analysis Summary
Last Period This period
(9/1/84- (9/1/85- Recent
Functional Area 8/31/85) 2/28/87) Trend
A. Plant Operations 2 2 Improving
8. Radiological Controls 2 2 Improving
C. Maintenance 2 1 --
, D. Surveillance 3. 2 --
E. Emergency Preparedness 2 1 --
F. Security and Safeguards 1 1 --
G. Outage Management # 1 --
H. Licensing Activities 2 1 --
I. Engineering Support # 2 --
J. Training and Qualification # 2 --
Effectiveness
K. Assurance of Quality # 1 --
L. Preoperational Testing 1 ## ##
M. Fire Protection and 1 ## ##
Housekeeping
, N. Construction 1 ## ##
- Not previously assessed as a separate area
- Not assessed this period
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l 0. Plant Trips (Scrams) and Unplanned Shutdowns
j- Power Root Functional
j Date Level Description Cause Area
j 12/15/85 Cold Scram when improper.applica- Startup testing Operations
l Shutdown tion of a jumper during Procedure
j. testing resulted in revers- inadequacy.
j ing the steam generator- '
l 1evel logic.
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1 1/16/86 Hot Scram due to rate compensated Operator error Operations
i Standby steam line low pressure due
j. to too quick opening of atmos-
i pheric steam dump valve when
MSIVs were shut.
i 1/18/86 Hot Scram due to source range Construction Operations
Standby monitor spike due to welding personnel error-
cables in proximity to. work control
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nuclear instrumentation
i cables.
1 1/19/86 Hot Scram due to rate compensated Operator error- Operations
l Standby steam line low pressure due high T-avg. com-
! to opening of a steam pounded by misad-
i generator relief valve with justed relief valve
j misadjusted setpoint. setpoint
j [1/23/86 INITIAL CRITICALITY]
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2/4/86 15% Scram due to low steam Operator error Operations
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generator level during manual
j control of feedwater flow.
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) 2/7/86 15% Scram due to low steam Improper settings Engineering.
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generator level auxiliary on auxiliary steam Support ,
{ steam relief stuck open. relief valve and
! gain (high) of-
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- feedwater regulat-
ing bypass valve
i 2/10/86 15% Scram due to low steam Faulty design of Surveillance
i generator level. During control card test
- surveillance, the level set points, plus I&C
l point input to control card technician failure '
i faulted to ground. (Probe to follow special ,
j contacted two test points.) instructions on '
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probe'use.
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Power Root Functional
Date Level Description Cause Area
2/12/86 29% Scram due to low steam Procedure Operations
generator level - transfer inadequacy
from turbine-driven to motor-
driven feedwater pump with
only one running condensate
resulted in low suction pres-
sure trip of feedwater pump
and then feed instability.
2/13/86 15% Scram due to spurious actu- Equipment: Cause --
ation of RPS inoperability unknown (possible
protection (General warning power supply
annunciator) during problem)
surveillance.
2/21-3/5/86 Shutdown to lower steam Manufacturing or --
generator chemistry to with- construction resi-
in owners group guidelines. due, or resin in-
jection.
3/19/86 10% Scram due to low steam Procedure Operations
generator level caused by inadequacy
failure to shift control to
feedwater regulating bypass
valves following a turbine
trip. Remained shutdown
through 3/20 to clean trans-
former insulators.
4/10/86 15% Scram due to low steam Operator error Operations
generator level - level
oscillation started with
control rods manually moved
to control average RCS tem-
perature during turbine
loading.
4/23/86 7% Scram due to low steam Operator error Operations
generator level following
rapid power reduction from
60% in response to secondary
steam leak from moisture
separator reheater drain tank
manway cover. Steam Generator
level control was in manual
at the time of the scram.
(18 hour2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> outage).
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Power Root Functional
-Date Level Description Cause Area
! 5/9/86 80% Scram automatically followed Management error- Operations
- manual turbine trip caused poor planning.
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by decreasing condenser Fouling of circu-
i vacuum (92 hour-outage). .lating water intake
i screens while wash
system was out of
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service for main-
- tenance }
7/24/86 20% Scram due to low steam Defective valve --
i generator level after feed- positioners caused
i water isolation due to over- bypass valves to be i
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feeding caused by partially partially open ,
j open bypass valves. As one i
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consequence, a planned mid-
! cycle maintenance outage was i
j begun early. The plant re- !
i mained shutdown through i'
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August 17 (552 hour0.00639 days <br />0.153 hours <br />9.126984e-4 weeks <br />2.10036e-4 months <br /> outage).
8/17/86
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11% Scram due to low steam Operator error com- Operations
generator level - after - pounded by feed
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shifting to automatic control control system
! operators attention was alignment
l diverted from steam genera -
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tor while controlling others
! in manual (13 hour1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> outage).
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i 8/17/86 21% Scram due to low steam -Operator error Operations
! generator level after feed- compounded by feed _,
j water isolation due to high pump control re-
. steam generator level -scram sponse time
j occurred because of inade-
! quate coordination between
- two operators who opened a
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feedwater regulating valve
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as feed pump speed was in-
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} creased (10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> outage). t
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9/6/86 80% Scram due to low steam Random equipment --
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- generator level following failure
j spurious closure of a feed-
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water isolation valve (25
j houroutage).
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Power Root Functional
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Date Level Description Cause Area
1/13/87 100% Scram followed low vacuum Operator error Operations
turbine trip after circulat- (improper lubicat-
ing water pumps tripped due ing water lineup)
to low lubricating water flow
(31 hour3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br /> outage).
1/14/87 7% Scram due to high source Operator error Operations
range neutron flux when trip (brushed against
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block was accidentally reset panel switch)
(7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> outage).
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The below table summarizes scram performance versus time, and shows the decrease
in scram frequency as the first year of operation progressed.
AT/BELOW 15% POWER ABOVE 15% POWER lTOTAll
1 WHEN EQUIP PERS & PROCED EQUIP PERS & PROCED
Before Criticality 0 4 -- --
4
2/86 - 5/86 2 5 0 2 9
6/86 - 9/86 0 1 2 1 4
10/86 - 1/87 0 1 0 1 2
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NOTE: The root causes in these Tables are the opinion of the SALP Board based
on inspector assessments and may differ from the LERs.
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IV. PERFORMANCE ANALYSIS
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.A. Plant Operations (1365 Hours, 19%)
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1. Analysis
- The previous SALP, completed prior to initial fuel loading, rated
i " Operations Support" as Category 2, consistent. Concerns included
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control over jumpers and lifted leads, tagging, log keeping, shift
! turnover adequacy, and root cause addressal. Of these, only equip- -
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ment tagging presented a concern during the current SALP period. ;
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On March 1, 1986, a reactor coolant system-(RCS) hot leg injection ;
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valve was. tagged shut.without an effective cross-reference to the
i subsequent plant heatup. The result was accomplishment of a pro-- '
hibited change in operating mode during heatup. The licensee then
i verified proper flow in other systems and' reviewed all tagouts and
i tag clearances. Overlapping management controls were implemented t
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to prevent similar occurrences due to a failure in one management
i system. These corrective actions were comprehensive. No further -
tagout problems were observed during the remaining 12 months of the
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SALP period.
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The transition from construction and testing to power operations l
occurred smoothly, mainly due to the significant nuclear operating '
experience of Northeast Utilities. Adding to this was an early
shift, during construction, in control room activities to that of
an operating plant, to the use of a plant-specific simulator for '
operator training, and to strict adherence to. written procedures.
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Overall, the startup test program was well managed and controlled.
! Initially, there were NRC concerns about the number of persons in
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the control room, about too"many tests being done simultaneously,
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and about the number of last minute procedure changes. When the
- licensee was informed, prompt and effective action was taken. Ex-
! cess personnel were not allowed in the control room. More deliber-
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ate test conduct was observed. Test preparations and procedures
became more thorough and timely. Startup testing proceeded rapidly
and in accordance with NRC requirements. Startup personnel were-
- knowledgeable and quickly' identified and corrected testing problems.
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A good interface was evident between startup, operations, reactor
engineering, I&C, maintenance, and QA/QC. The entire startup or-
f ganization was' assessed as' extremely capable and professional ~.
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As shown in the Supplementary Table in Section IIID (Page 9) of this
j SALP, there were a high number of scrams, with improvement evident
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by the decreased scram frequency with time. Fifteen of the 19
scrams were due to personnel and procedures. Eleven of the 15 were
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at or below 15% power. System and personnel responses to the scrams '
were proper. The errors were mainly in manual control of steam
generator levels. Quick operator response to changing conditions
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- was evident. Operator performance is considered to have lowered
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the scram rate substantially because feedwater transients which
caused many of the early scrams were handled expertly later on, and
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scrams were thereby averted. Proficient operator actions which
prevented challenges to safety systems included res
l petitive losses of fourth point heater drain pumps,ponses to re-
to feedwater-
regulating valve _ failure with a simultaneous motor-driven feed pump
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' minimum flow valve _ failure, and to loss of turbine plant closed
cooling water. Operator excellence was also shown in prompt re-
i sponse to'a major steam leak from the moisture separator drain tank
and to leaks of turbine electro-hydraulic control. system fluid. -
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Overall, operating shift functioning was evaluated as smooth and
professional. Activities were conducted carefully and with suffi-
1
cient formality. The operators.themselves were strong proponents
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of control room formality. Operator attitudes were' assessed as
l positive. Alertness was routinely observed in operator performance ,
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during day'and backshift inspections. Distractions such as ex-
traneous reading material were not permitted or observed in the
{ control room. Shift turnovers were observed to be consistently
thorough and effective. Briefings for tests and infrequent evolu-
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tions were detailed and involved free exchanges of questions and
answers. Written procedures were routinely followed. Shift-logs '
j and records were found discrepancy free during frequent. review.
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A high number of main board annunciators were illuminated'during
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operation, with about 100 identified in April 1986. This was
! largely due to annunciation of conditions which did not affect
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operation but presented a potential distraction to operators. The
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licensee has since reduced the lighted annunciators to about 60.
This is acceptable progress, but continued attention to this~ aspect
i is needed. '
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Operator technical knowledge was good. During the NRC license ex-
! aminations given this SALP period, 43,of the 52 license candidates
j passed. No significant generic weaknesses were noted.' This~83%
!
pass rate is a substantial improvement over the_ previous pass rate ;
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of 52% (11 of 21). Also, the operators _have consistently exhibited
detailed and thorough knowledge of the' equipment, its status, and
associated requirements. A recent example was shift supervisor ~(SS) i
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L review of surveillance of charging pump suction valve surveillance.
The SS recognized.that the valves were interlocked such that their
cycle times should be added to determine shift-over time between
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water sources, precipitating licensee reassessment of the associated
- . engineered safety feature response times..
} A few minor training weaknesses were observed..- The operators did
- not know how to take local manual control of a feedwater regulating '
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valve upon valve positioner failure-(an'in-line valve was used as
an alternate control instead). Also, an incorrect simulation of
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plant configuration contributed to a safety injection when a main
steam isolation (MSI) signal was reset with an atmospheric dump set
below steam line pressure. In these cases, the training response
and initiation of procedure and simulator changes were evaluated
as appropriate and timely, and representative of the licensee's
overall good approach to correcting problems.
Licensee management support of training and rewarding of operator
proficiency has been evident. The facility has a modern plant-
specific simulator with a training staff that has been expanded to
about 20 individuals. Several experienced operators have been pro-
moted to the training staff. There is a six shift rotation during
power operation, with full-time training for one shift being a re-
gular part of the rotation. Station management involvement in
training was evident in their knowledgeable discussions with NRC
personnel and in their obvious interaction with the training staff
and observance of simulator training.
Management attention to operations was evident in frequent plant
superintendent control room tours and detailed weekly plant material
walkdowns by a team of Health Physics, Maintenance, and Operations
supervisory personnel.
Overall, operating procedures were satisfactory. No major procedure
inadequacies were found. Personnel routinely followed procedures
and properly proposed appropriate changes. There were many minor
changes, as is expected during initial operation. In this case,
that is considered reflective of licensee determination to eliminate
procedure inadequacies. However, procedure flaws contributed to
three operational problems. One was a reactor scram on loss of
feedwater flow when an additional condensate pump was not started
before shifting feed pumps. A second was an emergency core cooling
pump inoperability for over three days when a valve was left shut
during surveillance. The third was a feedwater isolation and reac-
tor scram while shutting down without shifting to bypass valve
feedwater control prior to tripping the main. turbine.
Operations Review Committee performance was very good. Meeting inputs
were well prepared and showed a clear understanding of issues. The
approach to problem resolution was technically sound, very thorough,
and routinely conservative. Root causes of problems were actively
pursued. During meetings and in NRC discussions with higher level
managers, there was a licensee willingness to face facts and an
atmosphere of healthy self-criticism.
Housekeeping was poor at the end of construction and into startup
testing. After cleanliness and storage problems were identified
to station management, general plant cleanliness was upgraded and
attention was placed on removing or securing heavy items near safety-
related equipment. Four cubicles in the Engineered Safeguards
Features Building were completely cleaned and painted out. Cleanup
_ _
.- .
13
efforts and correction of packing leakage diminished the number and
size of the radiologically contaminated areas which had begun ex-
panding as the plant progressed into operation. Overall, house-
keeping was satisfactory.
1
Licensee Event Reports were routinely reviewed by the inspection
staff and generally _found to be complete, accurate, timely and to
contain adequate corrective actions. A special NRC Incident Re-
sponse Branch review of ENS reporting and classification as well
as use of event cause codes in LERs found adequate reporting.
Licensee command and control of operations were strong overall.
Managers were aware of operating status and details, and actively
inserted themselves at the appropriate organizational level. Shift
management was knowledgeable and exerted positive control over
activities affecting operation. An exception was the three-day
emergency core cooling pump inoperability. This involved improper
shift supervisor staff assistant (SSSA) performance, inadequate
training of SSSAs in valve operating requirements, and an inadequate
surveillance procedure. Strong corrective actions were taken. The
procedure and training flaws were corrected. Licensee review found
a lack of potential for other similar events. The individual and
cognizant line management were reprimanded. No similar occurrences
were observed.
Licensee management is strong. Corporate and unit goals and poli-
cies are detailed and well communicated, and administrative controls
i are effectively implemented. There is a strong safety first orien-
tation at all levels in the licensee's organization. Licensed
operators were professional, knowledgeable and thorough, and their
performance became excellent as the initial operating period pro-
-
gressed. The operator errors were assessed to be largely due to
inexperience with steam generator level control characteristics and
- to the high degree of difficulty of manual control of steam genera-
tor level. Scram frequency decreased in the latter part of the SALP
period. Housekeeping also improved. _ Concerns identified in the
previous SALP were effectively addressed. Overall, operating per-
1
formance was satisfactory and improving.
2. Conclusion
i
.
Category 2, Improving.
3. Board Recommendations
l
Licensee: Reduce unnecessary annunciations and reactor scrams.
j
l
NRC: None. '
i
I
!
)
l
__ _ _. __ . . __ _ - - . _ ._ -_. _ _ _ _ . _ _ - _. __
.. .
14
8. Radiological Controls (845 Hours, 12%)
1. Analysis
The Radiological Controls Program was previously rated Category 2,
consistent. There were no major concerns identified.
1.1 Radiation Protection
An effective, well-defined organizational structure is in place
to control unit radiological work activities. Adequate levels
of supervisory and technician level personnel were available
to support routine radiological activities. NRC inspectors
observed that Radiation Protection (RP) supervision were ac-
cessible to the RP staff and exhibited a strong "in-the-field"
presence. RP and Operations supervision regularly perform
joint tours of the controlled area to identify sources of con-
tamination and potential radiological concerns, and licensee
records and files document actions being initiated as a result.
The RP staff performed aggressively in directing decontamina-
tion efforts and initiating fixes for identified contamination
sources.
The number of audits of radiological operations routinely per-
formed by the RP corporate staff was assessed as good. However,
the audits were noted review to station RP activities as a
whole without always providing an in-depth review of unit ac-
tivities. This weakness was identified by the licensee and
actions have been started to improve the audit system. Over-
all, audits were considered satisfactory and the associated
audit program corrective actions were good.
Clear procedures and policies were in place and effectively
implemented. No procedure deficiencies were found by the NRC.
Radiation Protection personnel were trained and qualified in
accordance with a good program. However, RP technician re-
qualification training in 1986 contained only a limited dis-
,
cussion of plant systems. That NRC identified aspect was found
!
to be the only significant deficiency. The licensee is re-
viewing this matter.
Several noteworthy improvements were initiated in the radio-
logical training program. Detailed mockups of the RCP seals
and of a radwaste shipping cask have been procured to enhance
job-specific radiological training. An elaborate, complete
chemistry laboratory has been constructed in the training
facility for the instruction of chenistry technicians.
3
I
i
. .
15
The program for surveying, posting, and controlling radiological
areas was found to be well implemented. An extensive and
thorough radiation survey program to evaluate shielding effec-
4
tiveness was performed by the licensee during unit startup.
NRC specialist review of the Radiation Work Permit System found
it effective in controlling radiological work activities.
A special inspection was conducted to review post-accident
sampling. The licensee was able to adequately demonstrate this
capability. There were specific concerns with equipment
operability, monitor calibration procedures, and handling and
analysis of high activity samples. These required licensee
response and improvement, but were relatively minor items.
Collective exposure during 1986 was low (approximately 27
, person rem). The monthly exposure average was typically well
below the ALARA goal of 5 person-rem / month of operation. Dur-
ing previous. review of station ALARA activities, it was noted
that sometimes conflicting exposure goals were developed
separately by the unit and corporate staffs. The exposure
goal-setting process for 1987 has been improved in that the
unit ALARA staff provided significant input to the formulation
of corporate goals.
Overall, in plant health physics was a notable licensee strength.
This is attributed to a sound program, a capable staff, and
supervisory excellence.
1. 2 Unit 3 Radioactive Waste Management
Reviews of the liquid waste, gaseous waste and ventilation
systems installation and testing found that the systems were
installed as described on the Piping and Instrumentation Dia-
grams in the FSAR and testing was completed in an orderly and
timely manner to support initial criticality, power ascension
and commercial operation. Design deficiencies in those systems
discovered during preoperational and startup testing were re-
solved with very little impact on plant startup schedules.
There was no operational radwaste management assessment because
of the low level radwaste of processing activity during in-
itial operation. (Unit 1 and 2 radwaste considerations were
not considered in this Unit 3 SALP.)
1. 3 Radiological Effluent Control and Monitoring
Area, effluent and process radiation monitoring capabilities
were demonstrated during preoperational and startup testing.
There were recurring problems with the adequacy of monitor
calibration and licensee performance of Technical Specification
!
% y.- -
. .-- . . . = . - - - . . - - . -- - -.
k
. . ,
.
16
i
i
,' action statements required by monitor inoperability. Procedural
i. inadequacies included the failure of effluent monitor surveil-
lance procedures to adequately test Technical Specification
,.
required auto-isolation and alarm annunciation features and
'
also resulted in the non-conservative calibration of contain-
ment high range area monitors. The above deficiencies resulted
in the generation of several LERs. Additionally, weaknesses
i
in the comparison of monitor and laboratory sample data and
quality control for vendor laboratories were noted, showing
inattention to technical detail in procedural development and
review. Nonetheless, the licensee's analysis of radioactive .
samples was in agreement with the NRC values. The licensee '
provided adequate technical resolution of the weaknesses.and
promptly updated and corrected the procedures.
i
, 1.4 Water Chemistry Controls '
f
The licensee demonstrated a strong commitment to water chemis-
- try control.
' Chemistry analysis was. thoroughly reviewed during ,
i
the daily management meetings, and operations were thereupon
modified to optimize chemistry conditions.
i
.
Reviews of the water chemistry control program indicated a
! generally adequate program was developed and implemented. The
j
licensee was using upgraded analytical procedures and state
! of the art instrumentation in the laboratory. The training
!
and qualification program for supervisors and technicians in-
! cluded formal classroom training, written demonstration of
proficiency and,.for technicians, participation in an intra-
,
1
laboratory spiked sample program evaluated by their supervision. i
An elaborate chemistry laboratory has been constructed for the
l
instruction of chemistry technicians. The good training and
i facilities contributed positively to performance. ,
3
' NRC review found five of twenty-two comparisons of analytical a
' results against NRC standard samples were in disagreement. '
!
The differences were due to laboratory control program weak- t
nesses including single point calibrations of instruments, lack
} of measurement control charts, and sampling errors.
t
,
4
The program for controlling water purity in the primary and
secondary coolant loops was adequate. The ifcensee provided
i
a documented management commitment to and support for the pro-
l gram and closely monitored performance. During testing of
!
plant water systems, the licensee noted and corrected condenser
l inleakage, closely monitored unusual sulfate levels in the -
i
steam generators and administrative 1y controlled contaminants
j~ i
at levels generally well below consensus guidance. However, ;
i
several occurrences were noted suggesting-inadequate design
! review and failure to note lessons from Unit 2's operating ex-
i
4
i
l
l l
l
I
_ . _ . . ,
- - . ..- . - .. . . - - . _ - - -
. ..
17
-perience. As~a probable result of inadequate delay for Nitro-
gen-16 decay, the sampling location of the reactor coolant pro-
vided high radiation levels with no evidence of failed fuel.
Resin retention filters experienced strainer failures due to
.
'
.
) design problems identified initially at and corrected in Unit
j 2. In addition, the licensee failed to monitor the feedwater
system for metallic transport which could result in excessive
, sludge buildup in the steam generators.
The licensee proceeded with caution while steam generator
'
secondary chemistry difficulties were being worked out during ~
the power ascension test program. That involved a shutdown
]
from 30% power to drain and refill the steam generators to
lower sulfate concentrations below the vendor recommended limit
of 20 parts per billion. Also, domineralizer webbing was re-
1
placed to prevent resin pass-through to the steam generators.
3 The seven-day shutdown taken in this' case is considered repre-
- sentative of the licensee's normal emphasis on safety and
- quality having priority over operation.
r- During this assessment period, the licensee implemented a generally
effective radiological controls program supporting early commercial
operation. Recurring deficiencies were noted, however, with the
adequacy of procedures for and calibration of area radiation and
effluent monitors. Overall, inasmuch as the low levels of radiation
,
and contamination encountered during initial startup and operation
i did not present a strong radiation protection challenge, a high
- performance rating was not considered appropriate. Performance was
!
! assessed as satisfactory, and improving as a result of the quality
4
and results of corrective actions.
f 2. Conclusion
[
Category 2, Improving.
.
l 3. Board Recommendation
na
f
Licensee
Improve technical oversight of radiological monitor calibration,
'
j and laboratory quality assurance / quality control activities.
NRC
None. l
l
3
1
..
4
!
!
. .
18
C. Maintenance (359 Hours, 5%)
1. Analysis
The previous SALP rated Maintenance as Category 2, consistent. It
was recommended that the licensee schedule completion and implemen-
tation of maintenance procedures and training programs. This has
since been accomplished.
During this SALP period, maintenance was reviewed during two region-
based inspections and by the resident inspectors. No scrams or
challenges to protective systems were attributed to maintenance.
Safety system readiness and inservice testing (IST) performance
evidenced the effects of good preventive and corrective maintenance.
An example was the rebuilding of two service water pumps late in
the SALP period because of IST results.
Corrective maintenance was generally performed in strict accordance
with policies and work orders. Troubleshooting and significant
supervisory involvement led to accurate problem assessment and
formulation of proper corrective action. Work was thorough and
technically sufficient. Rework was seldom required. Only one in-
stance of poor maintenance was observed. A feedwater regulating
valve stem packing was tightened enough to retard valve motion.
It then failed to close on a Feed Water Isolation signal because
the packing was too tight (the next valve downstream did close).
Later, the same valve stuck in automatic control and then popped
open, causing a feedrate which caused reactor power to exceed 3445
MWth (101% of design). The licensee has since committed to full
stroke testing of such control valves after packing adjustments.
There were three instances of breaching or fouling of fire, control
building, and Secondary Leak Collection and Recovery System (SLCRS)
boundaries by process fluid hoses or staging. Also, there were
numerous instances of broken penetration seals, either by work in
progress or left over from construction. Increased licensee man-
agement attention was applied, and the resident inspectors noted
that such occurrences decreased in frequency.
The maintenance department was fully staffed with well trained, com-
petent and dedicated mechanics, electricians and machinists having
diverse backgrounds. Maintenance assistance was available from the
other three Northeast Utilities plants. Observations and discus-
sions showed maintenance supervisors and managers to be knowledge-
able, and active in on-scene oversight of activities. Effective
planning minimized outage and operational scheduling impacts. Co-
ordination with other departments was excellent. In fact, communi-
cation and cooperation between all departments, both at grass roots
and management levels, has been a key to timely and effective
troubleshooting and corrective maintenance on numerous occasions.
. .
19
A computerized maintenance management system (PMMS) has been in-
strumental in planning, controlling and documenting work. Its
machinery history function has been routinely used to trend equip-
ment performance for establishing corrective actions. PMMS is con-
sidered an excellent tool for managing maintenance.
Control of maintenance and testing was generally very effective.
Outages usually included between 700-900 work activities and tagouts
with minimal interference or failures in the control program. Main-
tenance and modification activities during normal plant operations
were controlled and performed within the bounds of Technical Speci-
fication Limiting Conditions for Operation. This was evident in
the routine daily performance of 6-8 preventive maintenance activi-
ties. Infrequent lack of control was observed, however: work on
main turbine stop valves commenced without a reapproved work order;
staging cross-bracing blocked operation of a Feed Pump turbine trip;
and Linear Variable Differential Transformers (LVDTs) were installed
on the Feedwater Regulating Valves without Operations Department
approval and to installation details modified after PORC approval.
These events affected equipment which is not safety-related, were
detected and corrected by the licensee, and had no operational
consequences.
Removal of a trash conveyor from its foundation created a potential
access route to the protected area. This maintenance error was
licensee-identified and promptly corrected.
Unavailability of improved replacement parts resulted in delaying
troubleshooting for all potential causes of Power Operated Relief
Valve (PORV) leakage, and effective repair of leaking PORVs was not
timely. As a result, although the valves have undergone a major
-
modification as well as two separate repairs, both PORVs were
blocked for a major part of the plant's operation. In addition,
due to either PORV and blocking valve leakage or safety valve leak-
age, the TMI action plan mandated positive indication of safety
valve status indicated open safety valves for most of the operating
period.
A significant maintenance action involved improper blowdown ring
settings on the Main Steam Safety Valves (MSSVs). In response to
an NRC information notice, the licensee spent considerable effort
verifying the ring settings for all 25 site valves, noting and cor-
recting a related problem of short ring lock pins and readjusting
the rings to a common setting. The ring readjustment was based on
documented phone conversations to the vendor. These confirmed the
technical manual setting values. In this case, maintenance
thoroughness significantly improved the assurance of proper MSSV
blowdown.
__________ - __-_ -
. .
20
The procurement program was well organized and allowed material
traceability to work orders. The warehouse was administrative 1y
well controlled and housekeeping was adequate.
In summary, licensee performance in the maintenance area has been
good overall, with the discrepancies noted being isolated and non-
representative. The maintenance program is properly established,
implemented and staffed. Plant equipment has performed with a high
degree of reliability.
2. Conclusion
Category 1.
3. Board Recommendations
Licensee: Assure thorough testing after maintenance.
NRC: Maintain current level of inspection.
.
1
4
1
I
. .
21
D. Surveillance (554 Hours, 8%)
1. Analysis
Surveillance was rated Category 3 during the previous assessment
period. A major factor was the tardy development of procedures.
This analysis is based on frequent NRC inspections by the resident
inspectors and four inspections by regional specialists.
The management program for controlling surveillances was found
especially strong in the Instrumentation and Controls (I&C) and
Maintenance Departments. Both departments used an automated system
to identify up-coming surveillances. Initially, the Operations Oe-
partment used a manual tracking system. Although it was cumbersome,
all required surveillances were completed. Operations is now also
using a computer system for tracking surveillances.
The inspectors have found that the technicians or operators conduct-
ing a test generally have a very good understanding of both system
and procedure requirements. This is particularly significant when
the complex electronic systems included in the Unit 3 design are
considered, and is a notable strength of the program.
The surveillance procedures are very detailed and form a solid basis
from which to build a successful program. Licensee personnel have
demonstrated a strong commitment to these procedures by active use
of the procedure change system. Changes were requested and drafted
by persons working with surveillance tests and processed in accord-
ance with the Technical Specification system for procedure changes.
The surveillance program has been managed conscientiously. Event
reports (LERs) documented seven missed surveillances. All were
licensee-identified. No single type of surveillance or responsible
working group was responsible for the missed tests. LERs also
identified some inadequate shift checks and compensatory actions.
Inasmuch as the lapses represent seven of several thousand surveil-
lances, and no significant safety degradation was involved, the
overall performance of required surveillances was excellent.
Surveillance caused a reactor scram from 15% power when a technician
inserted a test probe too far into a test point, contacting another
test point and grounding the level set point signal. The potential
for such an occurrence had been previously realized, and instruc-
tions had been issued to use short (non-standard) test probes for
such measurements. After this event, the licensee corrected the
basic problem by installing a barrier between the test point rows.
There were also seven instances of incorrect instrument setpoints
as the result of inadequacies in surveillance or calibration pro-
cedures. Four of these affected Reactor Protection System instru-
_ ..._ _. _ _ . _ _ . _ . - . _ _ . . =_ . _ . _ _ . _ . _ _
._.7._._._
.. , v
4
22
l
-
l
t
mentation setpoints. As a result, non-conservative settings were
j. used in over-temperature differential pressure scram setpoints, .
4
intermediate range neutron flux monitor scram setpoints,-reactor.
coolant system flow setpoints and the power range neutron flux.P-8
i interlock setpoint. Although none of these resulted in exceeding 'o
a Limiting Safety System Setting, their existence showed a potential
for such an excess. Because of these problems, the licensee re- 7
evaluated Technical Specification setpoints by comparing NSSS Vendor
,
Safety analysis documents to the Technical Specifications and the
l settings specified in surveillance and calibration procedures.
1
Recalculations were made for each setting; these contained all the.
,' conversions needed to track between plant primary parameters and
instrument electrical values. These licensee corrective actions
were assessed as comprehensive and found no additional inadequacies.
'
Five other occurrences resulted from inadequate surveillance proce-
) dures. These included isolation of service water to safeguards pump
j heat exchangers without the knowledge of shift supervision; incor-
.
rectly set throttle discharge valves in the control room pressuri- '0
,
zation system; an unnecessary safety injection during Engineered ,
j Safeguards Features (ESF) actuation relay testing; application of
I
full Reactor Coolant System (RCS) pressure to low pressure letdown
1 system piping during ESF actuation relay testing; and the failure
i to carry through a construction design change by deleting references
- to uninstalled remote shutdown panel transfer switches from the,
.
Technical Specifications and the surveillance procedures.. While 1
1
these items are minor from a safety viewpoint, they point out in- ,
j adequacies in validation of procedures prior to operational use. ..
'
(A procedure validation program is being considered by the licensee.)
'
In summary, although the program is sound overall, surveillance
! procedures have detracted from performance because'of setpoint and
i other problems. This appears to be a carry-over effect from the
tardy development of surveillance procedures. The excellence noted-
in performing prescribed surveillances indicates a potential for
a higher rating once it is demonstrated that procedure inadequacies
j have essentially been eliminated.
I '
j 2. Conclusion c .,
\ . , ,
'
Category 2. c,
,t, -
1 3. Board Recommendations 7
- .
l Licensee: Continue to emphasize procedure adequacy, and give evalu-
] ation of procedure validation priority emphasis.
,
,, 1
j NRC: None.
5
4
$
l
1 , ;
)
i
!
!
!
. - - . - , . - . - - . - , _ - , - , - - - - - . - - - - _ , , - . . - _ ~ , - . - , . . - - . - . - -
- .. . . ~. . .- - - . . - - - - - . ~ -
1 j,
'
23
.
t
!
-
s, E. Emergency Preparedness (173 Hours, 2%)
T ,
1. Analysis
.
J
During the previous SALP, this area was rated Category 2. Timely
,
resolution of NRC concerns was identified as needing improvement.
1
Emergency preparedness is a site function with conImon Emergency
' Plans, facilities, and personnel. This assessment covers the Sep-
-
tember 1, 1985 through February 28, 1987 period. It represents an
, ,
evaluation of all three Units but does not repeat applicable parts
j_ ,
of the three unit a uessment in-the Millstone 1/2 SALP for the
- period ending May 31, 1986. During the current assessment period,
there were two region-based inspections.
'
" Inspection on July 7-10, 1986 closed fifteen emergency preparedness .
- items. Two long lead time items remain open. These are a descrip--
s tion of the Offsite Facilities Information System (OFIS) and its
!
' maintenance procedure for inclusion in the Emergency Plan, and com- ;
pletion of the installation and testing of the Technical Support
Center (TSC) and Operations Support Center (OSC) hardware [0FIS,
4
i
Area Radiation Monitoring System (ARMS), Safety Parameter Display
System (SPDS), and the evacuation alarm]. Initially, a planning-
date for completion of the procedures was set for January 1986.
This is presently projected for completion in June 1987. That
- schedule is acceptable to the NRC. '
.
c i
,
r
The annual exercise was observed on November 19-20, 1986 (full par-
ticipation, including ingestion pathway). No significant deficien- j
cies were identified, but several minor weaknesses were noted. Two j
i of these were the direct result of a power failure caused by an ice '
storm. Back-up procedures and ' equipment worked satisfactorily,
is Both the Control Room and TSC staffs were knowledgeable and innova-
4
tive is solving problems presented as part of the exercise scenario.
The Control Room staff response.was prompt and conservative. They
j
quickly recognized changing plant conditions and were able to anti- '
l ' cipate possible corrective actions. The TSC staff demonstrated the
i t
ability to promptly identify and classify scenario events and make l
l protective action recommendations to offsite agencies. Emergency l
i Response Organization personnel were well trained and qualified for
!
their positions, and positive command and control of all emergency
4
response facility operations was demonstrated by the respective ,
'
i
'
facility managers. Overall licensee performance on the exercise-
was good.
Dedicated emergency response facilities are well maintained onsite
i
by the licensee. The Emergency Operations Facility (EOF) and TSC
.
i '
are common facilities for all three units. Both facilities have
adequate space and were designed to meet-the habitability require-
-
! , ments of NUREG-0696. Units 1 and 2 share a common OSC, with a
,t
l
.
J
.<
t
,v=wey-+,e g.9 . -n+.,s = wm,u.- -+-- --w.%-..-m-,--m., # .w .o e, r--o..m c , , , , , , - . , -,,,-r ,-sv, ,.,e,.,, ,=t--n-c. ,-
_
. .
24
separate OSC for Unit 3. All facilities are well equipped to func-
tion under emergency conditions. During the November 1986 exercise,
the emergency response facilities were promptly staffed and acti-
'
vated by the Emergency Response Organization personnel. Augmenta-
tion of the initial response to the emergency facilities was timely.
Contingency planning was evident when a hurricane was carefully an-
ticipated in August 1986. Severe weather preparations were imple-
mented. Shutdown planning was halted when the storm track shifted
substantially.
The Emergency Preparedness Staff at Millstone is ample, consisting
of a Senior Emergency Preparedness Coordinator and an Emergency
Preparedness Coordinator. Both have offices onsite. Additional
assistance is available from the Emergency Preparedness Supervisor
at the Corporate Headquarters in Berlin, Connecticut. Northeast
Utilities continues to maintain an excellent working relationship
with the State of Connecticut and local governmental agencies, as
evidenced by the continuing cooperation demonstrated during exer-
cises.
Overall, the licensee has a sound emergency preparedness program.
,
Management has adequately focused attention on this area as evi-
denced by good exercise performance, well-maintained emergency re-
sponse facilities, and an excellent working relationship with off-
site officials. There are few open NRC items.
2. Conclusion
Category 1.
3. Board Recommendations
None.
_ _
- - - . _ .
..
. .
25
F. Security and Safeguards (409 Hours, 6%)
1. Analysis
During the previous SALP period, no regulatory concerns were iden-
tified and the licensee's performance was assessed as Category 1.
The licensee was primarily involved in training and qualifying new
security force members and installing and testing new systems and
equipment for the integration of the Unit 3 program into the exist-
ing program for Units 1 and 2. During the current period, the lic-
ensee's staff was involved in monitoring the performance of new
security systems and equipment, evaluating the effectiveness of
training and assessing the need for changes as a result of imple-
menting the expanded security program.
In the current assessment period, a total of four preoperational
reviews, one special inspection and five routine inspections were
performed by region-based inspectors. Nine-of these inspections
involved the licensee's physical protection (security) program and
one reviewed the licensee's control of and accounting for special
nuclear material.
Corporate and plant management's involvement in and support for the
security program was very evident, resulting in the relatively
trouble-free integration of Unit 3 into the site security program.
The allocation of a sufficient number of experienced technical and
support personnel resulted in sound designs, good planning, timely
procurement, and quality installations.
An aggressive and comprehensive surveillance program was developed
to monitor the performance of new systems and equipment in their
initial period of use. The program was carried out by a team com-
posed of personnel with expertise in security, engineering, I&C,
and computers. The team approach was highly effective in accomp-
lishing this activity and was continued during the development of
routine surveillance testing and maintenance procedures.
As experience with systems, equipment and facilities was gained,
plans were developed and modifications were initiated for upgrading
existing systems, equipment and facilities. This demonstrated the
licensee's continuing attention to establishing and maintaining a
high quality and effective security program.
Staffing for the expanded security organization involved hiring -
about 150 new personnel. Due to the shortage of qualified candi-
dates in the local area, extensive recruitment efforts were required.
These efforts were successful and the necessary mannW, training,
and qualification were achieved on schedule. These efforts further
demonstrate the licensee's intent to implement a quality program.
!
l
1
. .
26
The training and qualification program was well developed with qual-
ity. lesson plans and instructional aids. It was adminisi;ered by
three full-time and experienced instructors provided by the security
contractor. The training program is effective and of high quality,
as indicated by the relatively small number of identified personnel-
errors. Training is continually upgraded as a result of feedback'
from operational experience and on-the-job performance observations.
Oversight of the training program is provided by a senior licensee
security supervisor and this is. considered by the NRC to be a major
strength of the program.
The licensee developed and implemented a comprehensive records man-
agement system. It included such things as manufacturers' specifi-
cations, acceptance criteria and testing data for the new systems
and equipment, design and construction information for new systems
and facilities, as well as the routine security program records.
The system provided for clear identification, ease of retrievability
and mandatory retention periods, and demonstrated the licensee's '
commitment to quality.
Necessary revisions to the licensee's corporate security audit pro-
gram, to reflect the integration of Unit 3 with the site security
program, were accomplished during the pre-operational phase as the
new systems and equipment were accepted for operation by the licen-
see. In this manner, the licensee was able to ensure that all new
program elements were included. ..The audit plan is comprehensive
and is maintained up-to-date in order to provide quality information
concerning the implementation of the program.
Fourteen Unit-3 related event reports, which required reporting
in accordance with 10 CFR 73.71, occurred during this' assessment
period. Seven of these events were minor problems with new equip-
ment / systems; of those, three concerned a deficiency in the new
intrusion detection system which, when located, was promptly cor-
rected by an engineering modification. Two events involved person-
nel errors by members of the security force. One of these was a
security officer leaving his post early; this individual was re-
trained. The other_was a security officer found asleep on duty;
this individual was fired. Licensee response to these two events
showed their strong insistence upon proper performance of duty.
Four events resulted from poor interface / coordination between vari-
ous plant functional groups and security. Another event resulted
from a contractor employee who-surrendered a' weapon prior to enter-
ing the plant protected area. The remaining event' involved a bomb
threat. Each of the above events was appropriately handled and
compensatory measures were promptly initiated when required. The'
event reports were clear,' concise and promptly submitted to NRC.
The cumulative downtime for the equipment / systems related events
was less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, indicating prompt attention to and detection
and correction of the problems. This timely support by I&C and
. _
e
- - -- . - . - . . - - . - - - .. - .
,
.~ .
27
.
Computer Services to the maintenance of the security systems and
equipment demonstrates the-licensee's commitment to a high quality
,
maintenance program.
During the assessment period, the' licensee submitted seven changes
to the Training / Qualification,^ Contingency, and Security Plans under '
the provisions of 10 CFR 50.54(p) and provided its response to the
August 4, 1986 Miscellaneous Amendments to 10 CFR 73.55, codified
by the NRC. The changes were described in' a summary transmitted 1
with each revision and referenced to plan pages that were marked .i
to facilitate review by NRC. Revisions were generally of high
'
, quality. The licensee's safeguards licensing function is adequately
staffed by experienced personnel who are knowledgeable of NRC pro-
.
gram objectives. The quality'of the submittals is further evidence
i of the licensee's commitment to a quality security program. '
2
'
The licensee's program and procedures'to control and account for
special nuclear material at Unit 3 were found to be adequate, as
,
was the licensee's: plan for the protection of new fuel.
In summary, corporate and site management-involvement in the program
1 resulted in the efficient and effective integration of Unit 3 into
,
a consolidated site' security program.- Significant management over-
sight and direction and the application of a well-planned and exe-
i
cuted team approach were largely responsible for the ease with which
i
this. evolution was accomplished. These factors remained strongly i
'
evident throughout the assessment period in monitoring personnel
I
and equipment / systems performance and in customizing the program .
to meet its expanded needs whole conforming to NRC program objec- '
tives. The only NRC concern identified was for a possible adverse
effect from future retirements of highly experienced and capable
individuals, and that does not affect the performance rating for ,
the current SALP period.
2. Conclusion
- Category-1.
'
3. Board Recommendations
!
l
' Licensee: Place emphasis on maintaining high performance during the
transitions when senior personnel retire.
'
NRC: None.
.
,
l
!
,
a
. ,r- ..~,, ., - .,-,,_..,.,..,.,y ,, yy _g. . . , . , , , .,7 -4 , ,_4 n,, , e w. -
. . . - .- . -- . .- - - - - . . - -
1 . .
28
i
G. Outage Management (127 Hours, 2%) ,
, 1. Analysis
This is the first_ time that Outage Management has been assessed for
Millstone 3. The plant is in its first operating cycle so there was
, no refueling outage. There were, however, two planned and three
unplanned outages and planning activities for the March -1987 outage.
Outage planning, mobilization, performance, restoration and restart
,
were observed as part of the routine inspection program.
Outages were planned in detail and very closely managed. . As job
needs were identified, requisite plant conditions and materials were
'
, noted with expected durations delineated by exoerienced personnel.
The data was incorporated, with the aid of critical path management
software, into a master outage schedule complete with bar charts,
i sensitivity analysis for each task that might impact the critical
path, and logical ties between tasks. The schedule received senior
'
and supervisory ranagement reviews and modifications prior to outage
' commencement. Twice daily during an outage, an expanded time base
printout of the current _three-day window, including all recent up-
,
_ dates to the master schedule, was provided to-all supervisors during
i a status meeting. These meetings were characterized by accurate
,
assessments of work in progress and resolution of conflicts. Tight
l
controls over the schedule and plant conditions were maintained and
many potential problems were avoided by early addressal. During
these meetings, NRC observers noted a strong spirit of cooperation
,
and a very positive attitude toward nuclear safety and high quality
performance.
One safety impact of good planning has been that sufficient time
was allotted to careful completion of safety related equipment
lineups and surveillance requirements.
Administrative control of maintenance and tagging was for the most'
part effective. A notable exception was a failure to clear an Auto-
mated Work Order and its associated tag during recovery from an
outage, leaving a-Safety Injection Hot. Leg Recirculation manual
isolation valve shut until in Mode 3 (hot shutdown). During that
same outage,' two work orders that resulted in the breach of Control
4
Room and Secondary Leak Collection Recovery System.(SLCRS) pressure
boundaries were approved by Shift Supervisors without realization
that these boundaries were impacted. This was during one of the
earlier outages. Corrective actions for these licensee-identified
4
losses of control appear effective, in that there have been no re- '
peat occurrences.
Management at both the unit and department levels was proficient'
at planning and scheduling, determining contingency _ strategies, and
quickly adapting to changing conditions. An example was the 24-day
1
, , - , - ..,,,,--,v, , , , - - - sm,,- -- . , . . . ,,c - , . , - ,- ,_,~,-. ,.e- -v,- , - ,
, - . . - .. . = - . _ . _ . - - . . ___ .
i .. -.-
1
29
.
outage beginning in July 1986. The licensee had a future three-week
outage in the planning stage when performance of an inaccurate sur-
veillance procedure damaged the letdown. relief valve (this over-
pressurization of letdown piping is incorporated in the Surveillance
functional area), forcing the plant down to cold shutdown for re-
pairs. As the reactor was being cooled down, planning sessions were
called to scope the full work load and determine the critical path
,
for accomplishment of all-' commitments up to February _1987 (the ten-
l tative date for a mid-cycle outage). Many vendor and material ar-
- rangements were expedited to meet the new schedule. What started
l_ as an unplanned shutdown grew into a successful 24-day outage during
( which much corrective and preventive maintenance, and major techni- '
s
' cal specification surveillances, were completed. .This effective
utilization of forced plant conditions eliminated the need for an
additional outage during the SALP period.
'
,
In summary, the controi of outage activities was a significant man- '
agement strength, based on the quality evident in the successful
l completion of numerous complex tasks during five observed outages.
1
2. Conclusion
Category 1.
l 3. Board Recommendations
i None.
!
,
i
,
i
T
. - . - .
- . - - - - = - '
. .
30
I. Licensing Activities
1. Analysis
The previous SALP rated this area as Category 2. To assure more
timely resolution of licensing issues, increased licensee management
involvement in the licensing process was recommended.
Licensee management involvement in licensing activities was evident
during the current SALP period. An example was their extensive
comments on the staff's Station Blackout [10 CFR 50.54(f)] letter
and associated discussions with NRC reviewers and management. In
addition, licensee management was active in the NUMARC industry
group addressing this activity, thereby ensuring a high level of
review and decision making on this. issue. The licensee, Northeast
Nuclear Energy Company (NNECO), also consistently demonstrated evi-
dence of prior planning and assignment of priorities, and had well
stated, controlled, and explicit procedures for control of licensing
activities. NNEC0 worked aggressively toward completing license
conditions and commitments to the NRC, and maintained a priority
list corresponding to the NRC Licensing Action Priority list. The
Unit 3 lead licensing engineer's ability to provide schedular and
technical information on past and present licensing activities in-
dicated that licensee records were complete, well maintained, and
available.
The NNEC0 licensing staff was evaluated as well qualified, and NNECO
assigned the necessary technical people to develop complete, high
quality responses to NRC requests. For exaraple, NNECO technical
staff and managers attended four NRC staff meetings to support the
NRC review of a request for approval to operate with N-1 loops.
Requests for information were responded to in meetings, conference
calls, and correspondence. Licensee responses were usually tech-
nically sound, had appropriate management review and approval, and
were submitted on or ahead of schedule.
One expedited Technical Specification change was requested, for
extending the 18-month diesel generator surveillance schedule. The
licensee notified the NRC of the schedule well in advance, and plant
management promptly responded to a request for more information.
NNEC0 was generally responsive to the staff's concerns. They took
the initiative to resolve issues through conference calls and meet-
ings, and promptly followed-up with response submittals. For the
proposed installation of one feedwater venturi inspection port in-
stead of two, the licensee provided a drawing showing the proposed
venturi meter installation which showed the inspection port and the
effect of the inspection port opening on the accuracy of the reading.
The licensee also provided ASME paper 83-JPGC-PTC-3 which described
a similar installation at Calvert Cliffs. NNEC0 was also excep-
.
. .
31
tionally responsive during the NRC staff's review of reactor cool-
ant system flow anomalies at the Callaway and Wolf Creek plants.
When asked to repeat the RCS flow measurements taken at Callaway
and Wolf Creek, the licensee took the data and promptly provided
the results in a meeting with the NRC staff, even though flow
anomalies had not been observed at Millstone 3.
Infrequent lack of responsiveness to NRC concerns was noted. One
example was responses and submittals concerning open items on reac-
tor coolant loop stop valve interlocks in the staff's Safety Evalu-
ation Report (SER) on N-1 loop operation. Some drawing submittals
were not the latest revision available. After a staff visit to the
site to obtain the latest drawings, review indicated that further
revision was needed to eliminate additional errors.
For the eight Technical Specification change requests submitted,
the licensee advised the NRC of the need for the changes and the
submittal schedule well in advance. Seven of the eight requests -
were thorough and technically sound. The eighth was an exception
which would have allowed the Nuclear Review Board (NRB) quorum to
consist of less than a majority of the NRB members. This was found
unacceptable by the NRC. Submittal of this request represented an
apparent lack of understanding of the intent of the Westinghouse
standard technical specifications.
Initially, licensee submittals lacked details on criteria for
reaching a "no significant hazards" determination. Improvement was
shown during the SALP period. A recent submittal related to ESF
response times contained detailed information from the associated
safety analysis, providing a strong basis for the "no significant
hazards" determination.
Overall, the licensee provided effective licensing liaison with NRR
and showed a clear understanding of the issues. There was effective
centralization, with one point of contact with NRC. Timely and
acceptable resolution was thereby facilitated.
2. Conclusion
Category 1.
3. Board Recommendations
Licensee: Assure accuracy of submittals to the NRC.
NRC: None.
l
l
l
.-- -
. .
32
I. Engineering Support (262 Hours, 4%)
1. Analysis
This is a new functional area. It encompasses technical activities
in addition to thos'e provided by the operations, maintenance, and
instrumentation and controls (I&C) departments.
Northeast Utilities maintained an appropriately sized onsite engi-
neering presence in both the operating company (NNEC0) and the sup-
port company (NUSCO). The NNEC0 engineering department is currently
staffed to 28 full time employees and includes reactor, mechanical
and electrical engineering functions as well as in-service inspec-
tion (ISI). The NUSCO onsite engineering group includes mechanical,
electrical, I&C, and civil / structural / stress engineering. Each of
these four groups has a NUSCO engineer as supervisor, with the large
majority of working level engineers being contracted from the Unit
3 architect / engineer. The onsite groups report directly to central
management at the utility headquarters. Additional technical sup-
port is provided by the Production Test Group. These electrical
and electronic technicians and engineers, mainly concerned with
generation and distribution equipment, are used for complex trouble-
shooting and repair problems.
The above groups above were composed of technically knowledgeable
personnel with skillful, seasoned supervision. They exhibited per-
severance and dedication to perform tasks correctly the first time.
Examples included timely and thorough assessments of the effects
of failed snubbers on the systems they restrained and the active
and timely resolution of pipe vibration problems.
The NNEC0 Reactor Engineering and ISI sections effectively.antici-
pated plant conditions and scheduled related surveillances. Reactor
physics and core surveillances were accurate, well controlled and
timely. Numerous NRC observations of inservice pump testing found
skilled and kr.owledgeable technicians performing well-controlled
tests and questioning the results for possible trends. Technique
and measurement accuracy for Local Leak Rate Tests (LLRTs) have
, never come into question. Mechanical and electrical NNECO engineer-
ing sections performed well in supporting evolutions affecting plant
operation. An example was identifying Volume Control Tank tempera-
ture reduction as a temporary means of decreasing reactor coolant
pump seal leakage. This group also did an excellent job of origi-
nating and managing special inservice tests (ISTs) when required.
An example was the special IST of the motor-driven auxiliary feed-
water pumps, identifying the cause of low suction pressure trips ,
as a pressure oscillation. I
'I
l
i
-
. _ _ _ _ _ _ _ . - _ _
. .
33
i
i
Support from NUSCO engineering was essential and well utilized.
Numerous design deficiencies were effectively addressed. Contain-
ment Recirculation System (RSS) heat exchanger service water con-
nections and chemical addition tank seismic supports were both up-
graded prior to full power licensing. The Condensate Storage Tank l
was redesigned by NUSCO with improved overpressure relief protection.
Main Steam Valve Building (MSVB) Heating, Ventilation and Air Con-
ditioning (HVAC) design problems were addressed with interim changes.
(Permanent modifications are planned.) The Main Steam _ low pressure
trip sensing lines froze (rendering 3 of 4 channels inoperative);
alternately, the Environmental Qualification high temperature limits
were routinely exceeded. Both site engineering groups coordinated
effectively to correct such deficiencies.
Engineering and design considerations contributed to several events
during this SALP period. The reactor scrams and feedwater isola-
tions due to steam generator level transients were in part due to
the equipment design, the difficulty of manual control of steam
generator level, the need for extensive system grooming, and impro-
per equipment setpoints. Corrective actions were generally good
and performance improved, but there appears to be considerable room
for further engineering improvement of steam generator level control.
Two errors in the Reactor Protective System (RPS) Overtemperature
Differential Temperature (0 tdt) setpoint calibration procedure led
to incorrect entries into the RPS for calculation of 0 tdt. One was
an incorrect constant for the setpoint calculation, the other was
a setpoint reduction for excessive Axial Flux Difference (AFD).
Both errors were caused by failure to recognize changes in NSSS
vendor setpoint documents. The result was a slight (<1%) non-con-
servative shift in the trip setpoint. In a similar instance, due
to a change between procedure setpoints and the Technical Specifi-
cations, three loop protective interlock P-8 reset at a thermal
power higher than was allowed by Technical Specifications. These
items were discovered by licensee reactor engineers, thoroughly
analyzed, and subjected to timely'and sound corrective actions
(procedure changes and re-review of~all RPS setpoints).
In one case, immediate corrective actions were evaluated as not
conservative enough. During Startup Report review, the NSSS vendor
discovered that the reactor coolant system resistance temperature
detector (RTD) response time interpretation and acceptance criteria
were in error. The error involved late provision of information
by the vendor and licensee failure to retrofit that information~ into
the Startup Manual. When correctly evaluated, loop 2 RTDs exceeded
the acceptance criteria and required a review for impact on the
Final Safety Analysis. Between the time the vendor raised the issue
and the time a Justification of Interim Operation (JIO) was provided,
five accident analy;es were in question. The licensee did not then
trip the loop 2 bistables that provide protection for these acci-
. _ - _ _ . .
. ___ _ _ _ _ .
, ..'
34
dents. Those_three bistables, as well as.two interlock and one
permissive bistables, were tripped about eight hours later when the
vendor-supplied JIO did not satisfactorily address all five accident
analyses. Subsequently, following reallocation of some design mar-
gin and additional JIO, the licensee reset these 6 bistables.
Several engineering issues which adversely affected performance were
being acted upon. Continued licensee attention to resolution of
the following of these .is needed:
- Steam Generator Feedwater Flow oscillations.
- Elimination of illuminated. control board annunciators.
- Power Operated Relief Valve internal leakage.
- Control Building ventilation radiation monitor causing spurious
ventilation isolations.
- Main Steam Valve Building Heating and Cooling.
In summary, engineering support has been satisfactory. There was
a high workload and much competent work. Some design changes were
not carried through to modification of the technical specifications
and procedures. Further, some procedures were not changed to re-
flect technical specification changes in reactor protection set-
points. Performance would have been better if steam generator level
control difficulties had been resolved-early during initial opera-
tion, and if . unnecessary control room annunciations had been signi-
ficantly lower. However, the problems were not unusual for early
operation,-and the licensee response ~was sound.
2. _ Conclusion
, Category 2.
3. Board Recommendations
Licensee: Resolve issues requiring engineering attention.
NRC: None.
f
l
'
<
!
i
- ,-- , - . - - , , - , - . , ,
.--n , - -- ~+ - , - -
- -- - . - . . .. - - - - .
,, .-
l
35
l
l
.
.
J. Training and Qualification Effectiveness
1. Analysis
,
Training and Qualification Effectiveness.is an evaluation ~ criterion !
for each functional area. During this SALP, it also is being con- I
<
sidered as a separate area (for the first time). This area is a i
synopsis of the assessments in the other. areas. Training effective-
i ness has been measured primarily by the observed performance ofL
!
licensee personnel and, to a lesser degree, through program review. !
A strong training commitment was. evident in the investment in. staff l
j and facilities. ..The plant specific simulator was a'significant- l
i benefit in training operators and was used to train managers as well. ;
,
The licensee has built the training staff to'over twenty instructors, t
!
three quarters of whom hold operating licenses. There is a strong '
'
supervisory organization to manage the training staff.
Four reactor scrams-were assessed as having training implications.
These were the 1/16/86 scram due to too quick opening of a steam
, dump valve with the main steam isolation valves shut, the 1/18/86'
scram due to welding cables being near nuclear instrumentation
cables, the 3/19/86 scram caused by failure to' shift to feedwater
- regulating bypass valve control,.and the 4/10/86 scram due to low
steam generator level during manual control. - While better training
'
should have reduced such events, the associated training effective-
- ness is considered representative of a sound program during its
initial application to actual operation. The licensee's training
organization separately reviewed licensee event reports (LERs) and
, plant information reports-(PIRs) for training aspects, and the on-
i site safety committee (PORC) actively probed training considerations
i during its regular reviews. These feedback' loops represented man-
agement involvement and provided good corrective action inputs.
As noted in the plant operations area, operator performance ~on NRC
license examinations was good. While consistency has not yet been t
- shown in that performance, NRC concern about there being too much
of a cookbook approach to accident response and too little indivi- i
-
dual case assessment no longer exists. Also, operator performance
, on shift was excellent, with quick response to changing conditions i
evident in spite of the high number of lighted annunciators.
!
j
~
The licensee is actively pursuing accreditation.by the' Institute 1
of Nuclear Power Operations (INPO). Operator training is based.on
- Northeast Utility programs which are INPO accredited.- 1
,
.
'Non-licensed staff training was inspected and found acceptable. l
!
Plant equipment operators, maintenance, production test, and~I&C
i technicians have been observed performing normal and infrequent
1
1 !
l
,
_ _ _ - . _ - . _ _ _ . . . _ . ___ - _ _ _ - . _ _ _ _, _ , _ , _. - - , _ _ , . _ _ ,
. .
36
operations, maintenance activities and surveillances. These indi-
viduals have been found to be knowledgeable and to perform their
assigned tasks safely and competently.
The maintenance and I&C technician training program was actively
pursued. Training commitments were scheduled and strictly followed.
Senior department personnel actively assured that their juniors had
the knowledge for performing assigned tasks. NRC questioning of
in-service inspection technicians revealed excellent knowledge of
equipment, procedures, and applications.
A significant weakness in Shift Supervisor Staff Assistant (SSSA)
training was identified by the licensee. Use of the marginally
trained SSSAs for a task in excess of their training contributed
to isolation of an emergency core cooling subsystem without the
knowledge of shift supervision. This isolated incident was an ex-
ception to the generally excellent non-licensed personnel perform-
ance.
General Employee Training (GET) is common to the three Millstone
units. The program adequately addresses orientation, radiation
protection, security, emergency planning, safety, and assurance of
quality. Program content is directed by a steering committee made
up of the Unit Superintendents and other managers who determine the
emphasis of GET based on station performance goals.
In summary, the licensee's commitment to training was evident in
enhanced training staffing with a high percentage of experienced
licensed operators and expenditure of considerable resources for
training. The operators were assessed as becoming excellent per-
formers early in the initial operating period. Also, a high level
of operator and support personnel knowledge was consistently demon-
strated. Performance on NRC exams was good. Notwithstanding the
large number of reactor scrams, training was generally effective
in providing well qualified personnel who contributed positively
to safe operation.
2. Conclusion
Category 2.
3. Board Recommendations
Licensee: Continue training development to achieve accredited
training and assure consistently good operator examina-
tion results.
NRC: None.
..
. - -_ . _ . - . - . . . . - _ . - . . - - -- -. . . . - .
, , .
37
j
'
K. Assurance of Quality (424 Hours, 6%)
! 1. ' Analysis
Management involvement in assuring quality is an evaluation cri- '
] terion in each functional area. Quality assurance (QA) is an in-
4- tegral part of each functional area and the respective QA-inspection
'
hours are included in each one. This area is a synopsis of the
assessments of the assurance of quality. in other areas. During
1
the current SALP period, there were three QA inspections, inspec-
! tions by the' resident inspectors, and a readiness for operation's
j team inspection.
The related area of quality assurance was not rated during the pre-
vious SALP. Strengths were, however, noted in management's strong-
commitment to assure quality throughout the design, procurement;
construction and preoperational test phases. . No breakdowns in-
j quality programs or serious-individual quality problems were noted.
.
,
During the current SALP period, daily observations found Millstone
3 personnel to.have a standard of completing assigned work correctly
i on the first attempt. This positive attitude was repeatedly dis-
played. Shoddy workmanship or lack of attention to detail were
,
typically not tolerated by peers or supervisors. Department Heads
! were very knowledgeable of the status of work. -Plant personnel
'
exhibited a good attitude towards QA and adherence to procedures.
i The individuals closest to the work (operators, technicians,-me-
chanics, electricians, engineers, etc.) exhibited high personal
'
1
' performance standards and detailed knowledge of equipment and pro-
cedures.
'
QA/QC personnel were found knowledgeable of the tests they were
raonitoring. QC inspectors were found to be trained,' qualified and
i certified to the level of their responsibilities. Site staffing
!
levels were found adequate to support the startup test program and
i
normal operations, with headquarters and contractor personnel
available as needed. Questionable trends were investigated to de-
l termine their. root cause.
. First line supervisors provided close oversight of work activities.
Maintenance, I&C, and Production Test supervisors were generally
' knowledgeable of the plant design and station administrative re-
quirements. They were often observed'to be providing technical
, guidance and oversight to workers at'the work site. Further, Shift
Supervisors repeatedly demonstrated that they were knowledgeable
,
of plant activities and that they were effectively managing activi-
,
ties and shift personnel.
j
)
!
i
i
.
w- y =%-9,y-w-, y -,m,~w,-,,-, , v w e , e m m y--. ,---r.. --%m_., -,t t -- - w e t- -*T..r w. - , _ . . m ,
.
-. .
38
Department supervisors were also frequently observed in the plant
conducting personal inspections. NRC inspectors found them to be
knowledgeable of specific problems and active participants in prob-
lem resolution. These individuals were members of the onsite safety
committee, the Plant Operations Review Committee (PORC), and their
sound safety perspective extended into PORC activities. NRC obser-
vers continually witnessed frank, open, and knowledgeable PORC dis-
cussions of issues. PORC members clearly demonstrated sound safety
and facility knowledge, and their contribution to safety was a not-
able licensee strength. The many related examples of thorough
problem resolution include the licensee reviews upon discovery that
a reactor coolant loop hot leg injection valve, tagged shut for
maintenance while the plant was in cold shutdown, remained shut
while the plant was taken to hot shutdown. The basic operator error
was addressed. There also were two days of intensive PORC review
of operating procedures, tagout control, work activity control, work
activity control, and retest and training requirements. Procedure
improvements resulted. A design change to annunciate main steam
isolation was initiated as a side effect.
Senior plant staff were assigned as Duty Officers to act for licen-
see management on a weekly basis during operations and outages.
Management Representatives were assigned on eight-hour shifts round
the clock on site for coverage of outages. Daily staff meetings
were used to discuss plant conditions and each department was re-
quired to present the status of its work items. Issues were dis-
cussed and tracked in detailed reports which were updated and dis-
tributed daily. These controls provided excellent management of
ongoing activities.
Plant management attention was rapidly focused on problem areas by
the Plant Incident Report (PIR) system. This system has a very low
threshold for PIR origination and mandates unit superintendent re-
view as well as assignment of follow-up activity. Four hundred PIRs
were written during 1986. NRC inspectors found the PIRs to be an
excellent tool for keeping senior licensee managers informed, and '
senior managers did pay significant attention to root cause assess-
ment and corrective actions. This was routinely observed to occur
during daily management and PORC meetings.
The plant superintendent was observed making frequent control room
tours. Weekly plant walkdowns by operations, maintenance, and
health physics supervisors resulted in improved housekeeping, in
diminished size of contaminated areas, and in enhanced correction
of packing leakage and other lesser maintenance items.
Nuclear Safety Engineering (NSE), the independent safety engineering
group which is part of the corporate staff, was active in its cover-
age of Unit 3. This on-site group had ready access to the plant
staff, equipment and records. NSE assessed plant safety programs
i
!
l
__
- . . - . . -
.. -.
~i
39
and evaluated plant operating experiences.through reviews-of proce- '
dures and data including independent reviews of the resolution of
Licensee Event Reports (LERs) and PIRs. NSE made a significant
effort to participate in the Institute of Nuclear Power Operations ,
(INP0) sponsored Human Performance Evaluation Study Program (HPES).
Recommendations for corrective actions were provided from evalu-
ations of incidents or "near misses" reported to the HPES coordina-
tor. In addition to site specific corrective action, .the licensee
provided information to the INP0-HPES data base. Although no meas-
ured improvement in plant performance resulted, the fact that de-
tailed evaluations on human performance were performed is assessed
as contributing positively to root cause identification.
The Millstone Unit 3 Nuclear Review Board (NRB) was thorough in its
reviews. Its meeting agendas were extensive, the board discussions
were probing, and open issues were conscientiously tracked.
Plant. management losses have included the station superintendent,
the station services superintendent, and the unit superintendent.
The fact that no notable drop in performance resulted indicates
depth in management expertise and careful management of the transi- ,
,
tion periods involved.
The licensee's audit program was well planned. Audits were found
to be in depth and conclusive. Audit checklists were well organized
and comprehensive. Some audit findings, however, were left unre-
solved for as long as three years. Although no significant indi-
.
vidual concerns were involved, the three year delay in resolving
i findings indicates an audit response system inadequacy.
In summary, there was excellent regard for assurance of quality in
all aspects of plant operation. Management expended significant
effort to ensure that processes were controlled, that problems were
'
discovered, communicated and corrected, and that process controls
were modified to prevent problem recurrence.
2. Conclusion
Category 1.
- 3. Board Recommendations
i
None.
<
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. . . _ _ - _ . . . . , _ ,
. _ . _ _ __ . _ . . . _ . , . _ _ , _ , , . _ . . _
. . _ .
.. . _. . _ . . .
. .
- 40
l
V. SUPPORTING DATA AND SUMMARIES
A. Investigation and Allegation-Review
None.
B. Escalated Enforcement Actions
1. Civil Penalties-
None.
2. Orders
, None.
3. Confirmatory Action Letters
None.
C. Management Conferences
11/5/85 Management Meeting onsite to discuss completion status for
construction, testing, and procedure development.
1/9/86 Enforcement Conference - failure to report a construction de-
ficiency in accordance with 10 CFR 50.55(e) when an error was
detected ir, the load path for a reactor. coolant pump snubber
support. A Level III violation was ultimately issued.
3/13/86 Management Meeting to ' discuss operating experience, plant in-
cidents, and reportable events occurring during the startup~
test program.
3/27/87 Enforcement Conference to discuss the events affecting the
operability of the "B" high pressure safety injection pump '
during 11/26-30/86. A Level IV_ violation was ultimately issued.
"
D. Licensee Event Reports
. A tabulation of Licensee Event Reports (LERs)-by functional area,-and
i. an LER synopsis, is attached as. Table 4.
1. Licensee Event Reports Reviewed
LER Nos.85-001'through 85-003,86-001 through 86-059,87-001
through 87-007, and fourteen security-related event reports.
,
-
y e y, - - - , - , - _ - - - -..m- , , - , - , - , - - . ,, , ,, .---m, y -
---r-- - ,-
w -,13- +
. .. .
. .
41
2. Causal Analysis
a. Tabulation by Common Cause Factors-
Causes:
C - Communications Inadequate
Cn - Construction
D - Design Inadequacy
E - Equipment Failure
K - Lack of Knowledge (possible training inadequacy)
M - Management Planning or Control Error
Pe - Personnel Error
Pf - Procedure Not Followed
Pr - Procedure Inadequacy
l l NUMBER l l
CAUSES OF LERs LER NUMBERS
Pe & Pr 13 85-02, 86-07, 86-15, 86-19, 86-20, 86-21,
I l 186-28, 86-30, 86-56, 86-58, 87-05, 86-06,1
Security 86-13
Pe & K 10 86-01, 86-02, 86-19, 86-21, 86-26, 86-28
l l l86-30, 86-56, Security 85-32, Security- l
86-30
Pe 6 85-03, 86-04, 86-08, 86-10, 86-33, 86-44
Pe & E 5 86-13, 86-32, 86-41, 86-48, 86-49
Pe & D 5 86-03, 86-12, 86-14, 86-26, 87-02
Pe & Cn 5 86-02, 86-06, 86-36, 86-38, 86-59
Pe & M 3 86-35, 86-52, 86-56
Pe & C 1 87-01 ,
Pe & Pf 1 86-33
Note: The causes in this table are not mutually exclusive.
For example, LER 86-21 was evaluated as having personnel
error, procedure, and training causes, and was listed
under both "Pe & Pr" and "Pe & K."
b. Tabulation By Common Event Description
Failure of Safeguards Channel due to instrument line freezing: !
86-05 and 86-22.
Missed Surveillance: 86-07, 86-26, 86-33, 86-34, 87-06, and
87-07.
Safety Injection System Actuation: 86-01, 86-03, 86-19, and
86-21.
l
_
l
. . .
42
Steam General Level Transients: 86-10, 86-14, 86-15, 86-30,
86-32, and 86-48.
Supplementary Leak Collection System Boundary Problems: 86-06,
86-38, and 86-59.
High Temperatures in EEQ Monitored Areas: 86-29 and 86-50.
Security-Related Equipment Problems: 50-245/86-01, 86-02,
4
86-03, 86-04, 86-20, 86-21, and 86-31.
- E. Licensing Activities
1. NRR/ Licensee Meetings
4
a. NRC Headquarters
1/8/86 Meeting to discuss status of licensing issues in
preparation for issuing full power operating-license.
1/23/86 Meeting to discuss Millstone 3 Station Blackout.
2/19/86 Meeting to discuss NU response to NRC's 50.54(f)
letter of December 18, 1985 on Station Blackout.
7/15/86 Meeting to discuss status of licensing activities.
7/28/86 Meeting to discuss staff concerns related to 3 loop
operation.
b. Site Visits
- 5/12/86 Meeting to discuss status of licensing activi-
ties.
- 11/14/86 Meeting to review drawings of solid state pro-
tection system for 3 loop operation.
12/23-24/86 Site visit to review Plant Design change request
files.
~
2/25/87 Site visit to simulator and control room in
support of 3 loop operation.
2. Commission Briefings
1/29/86 Vote on Full Power-License Issuance for Millstone 3.
i
i
I
I
I
l
,
, y
e -+ - -- ,
.
. . .
43
3. Schedule Extensions Granted
None.
4. Reliefs Granted
None.
5. Exemptions Granted
None.
6. License Amendments Issued
1/22/86 Low Power License (NPF-44) Amendment 1 - Remote Shutdown
Instrumentation
9/9/86 Full Power License (NPF-49) Amendment 1 - Fire Protection
Audits
7. Emergency Technical Specifications Issued
None.
8. Orders Issued
None.
9. NRR/ Licensee Management Conferences
None.
i
i
i
2
- -
-
- , - -
-
- ,
. .
TABLE 1
INSPECTION REPORT ACTIVITIES
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
423/85-54 SPECIALIST 375 AS-BUILT INSPECTION OF PIPING, DUCTING,
9/9-20/85 TEAM SUPPORTS, ELECTRICAL POWER, INSTRUMENTATION
INSPECTION AND CONTROL OF SELECTED SAFETY-RELATED
SYSTEMS
423/85-55 SPECIALIST 176 OPERATING PROCEDURES, EMERGENCY PROCEDURES,
9/16-23/85 AND REVIEW 0F LICENSEE ACTIONS ON PREVIOUS
FINDINGS
423/85-56 SPECIALIST 74 REVIEW STATUS OF PREVIOUSLY IDENTIFIED
9/30-10/4/85 ITEMS, PRE 0P STATUS OF SOLID RADWASTE
SYSTEM AND TASK ITEMS IDENTIFIED IN NUREG
0737
423/85-57 SPECIALIST 205 SURVEILLANCE, CALIBRATION CONTROL, MAIN-
11/12-27/85 TENANCE PROCEDURES, EMERGENCY PROCEDURES,
OPERATING PROCEDURES, INITIAL FUEL LOAD
PROCEDURES REVIEW, PRECRITICAL TEST PRO-
CEDURES REVIEW, STARTUP TEST PROGRAM
423/85-58 SFECIALIST 32 SITE PHYSICAL SECURITY PROGRAM
9/30-10/4/85
423/85-59 SPECIALIST 131 OPERATIONAL STAFFING, OPERATIONAL STAFF
9/30-10/4/85 TRAINING, MAINTENANCE PROCEDURES
423/85-60 IE TEAM N/A ENGINEERING ASSURANCE TECHNICAL AUDIT
8/26/85-9/19/85
423/85-61 SPECIALIST 454 PREOPERATIONAL TEST PROGRAM
9/30-11/1/85
423/85-62 RESIDENT 594 REVIEW 0F PREVIOUS FINDINGS, REVIEW 0F
9/24-11/18/85 NUREG 0737 ACTION ITEMS, OBSERVATION AND
WITNESSING OF H0T FUNCTIONAL TESTING AND
RETESTING
423/85-63 SPECIALIST 31 PHYSICAL SECURITY PROGRAM
10/21-25/85
423/85-64 SPECIALIST 68 REV 0 PHYSICAL SECURITY PLAN, SAFEGUARDS
11/4-11/8/85 CONTINGENCY PLAN, TRAINING AND QUALIFICA-
TION PLAN, AND IMPLEMENTING PROCEDURES
T1-1
_
_ -
.. - _ _. .. . _ _ _ ___ __ __
. .
4
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
'
423/85-65 SPECIALIST 100 CHEMISTRY AND RADI0 ACTIVE EFFLUENT CONTROL
_ -10/21-25/85 PROGRAMS
i
'
423/85-66 . SPECIALIST 27 EMERGENCY PLAN IMPLEMENTATION APPRAISAL
10/21-11/8/85
4
423/85-67 SPECIAL N/A INITIAL OPERATING LICENSE REVIEW REPORT
REPORT
423/85-68 SPECIALIST 24 FIRE PROTECTION / PREVENTION PROGRAM
11/4-6/85
423/85-69 SPECIALIST 326 PRE 0PERATIONAL TEST PROGRAM
11/12-27/85
423/85-70 SPECIALIST 10 NUCLEAR MATERIAL CONTROL AND ACCOUNTING
11/12-15/85
-423/85-71 SPECIALIST 216 SURVEILLANCE PROCEDURES
11/11-22/85
i'
423/85-72 SPECIALIST 18 PHYSICAL SECURITY INCLUDING: PHYSICAL BAR-
12/16-19/85 RIERS, COMPENSATORY MEASURES, ASSESSMENT
i
AIDS, ACCESS CONTROL, DETECTION AIDS, ALARM
STATIONS, COMMUNICATIONS, PERSONNEL TRNG
'
423/85-73 SPECIALIST N/A MANAGEMENT MEETING - COMPLETION STATUS FOR
11/5/85 CONSTRUCTION, TESTING, AND PROCEDURES
423/85-74 RESIDENT 417 NUREG 0737, WITNESSING OF SYSTEM AND COM-
11/19/85- PONENT TESTING,~0BSERVATION OF CORE LOAD, '
1/6/86 SURVEILLANCE, MAINTENANCE, AND PHYSICAL
PROTECTION
423/85-75 SPECIALIST 69 PRE 0P TESTING
,
12/9-13/85
423/85-76 SPECIALIST 130 PRE 0P TESTING
12/12-20/85
423/86-01 SPECIALIST 181 STARTUP PROGRAM REVIEW, POST CORE HOT
1/6-17/86 FUNCTIONAL TESTING PROC. REV. , SURVEILLANCE
!
t TEST REVIEW AND WITNESSING
,
'
423/86-02 RESIDENT 470 PLANT EVENTS AND NON ROUTINE REPORTS, NUREG
1/7-2/24/86_ 0737 ITEMS, POST CORE HOT FUNCTIONAL TEST-
ING, APPROACH TO CRITICALITY, LOW POWER
PHYSICS TEST
?
!
- T1-2
.
-. . . - .-.- .-. . ,-- _
-.- -- - . . - .
_ _ _ _ _ _ - -
. .
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
423/86-03 SPECIALIST 36 SITE SECURITY PROGRAM
3/24-27/86
423/86-04 SPECIALIST 68 REVIEW OF PREVIOUSLY IDENTIFIED SIGNIFICANT
1/6-10/86 DEFICIENCIES
423/86-05 MEETING N/A MEETING REPORT: DISCUSSI0fl 0F PLANT EVENTS
3/13/86 REPORT
423/86-06 SPECIALIST 37 CHEMISTRY AND RADI0 ACTIVE EFFLUENT CONTROL
1/27-31/86 PROGRAMS
423/86-07 SPECIALIST 225 STARTUP PROGRAM REVIEW, POST CORE HOT FUNC-
1/19-2/14/86 TIONAL TEST WITNESSING AND TEST RESULTS
REVIEW, INITIAL CRITICALITY AND LOW POWER
PHYSICS TESTS, POWER ASCENSION PROGRAM
423/86-08 RESIDENT 300 PLANT EVENTS, NON-ROUTINE REPORTS AND OB-
2/25-4/14/86 SERVATION OF POWER ASCENSION TESTING,
VERIFICATION OF COMPLETION OF NUREG 0737
ITEMS
423/86-09 SPECIALIST 121 STARTUP PROGRAM REVIEW, POWER ASCENSION
2/18-3/14/86 TEST PROCEDURES REVIEW, TEST RESULTS REVIEW,
TEST WITNESSING
423/86-10 OPERATOR N/A OPERATOR LICENSING EXAMINATION
3/31/-4/4/86 LICENSING
423/86-11 SPECIALIST 90 STARTUP PROGRAM REVIEW, POWER ASCENSION
3/15-4/3/86 TEST WITNESSING AND TEST RESULTS REVIEW,
PRE 0P TEST PROGRAM FINAL REVIEW
423/86-12 SPECIALIST 145 OPERATIONAL TEAM INSPECTION, INCLUDING
4/14-18/86 SURVEILLANCE, MAINTENANCE, QUALITY ASSUR-
ANCE, AND FIRE PROTECTION ACTIVITIES
423/86-13 SPECIALIST 67 NONRADI0 LOGICAL CHEMISTRY PROGRAM, LABORA-
4/7-11/86 TORY ORGANIZATION, TRAINING MEASUREMENT
CONTROL AND ANALYTICAL PROCEDURE EVALUATIONS
423/86-14 SPECIALIST 71 STARTUP TEST RESULTS REVIEW AND STARTUP
4/14-24/86 TEST WITNESSING
423/86-15 RESIDENT 167 PLANT OPERATIONS, RADIATION PROTECTION,
4/15-5/19/86 SURVEILLANCE AND MAINTENANCE
T1-3
_ _
_ _ _
. .
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
423/86-16 SPECIALIST 32 WATER CHEMISTRY CONTROL PROGRAM
5/5-9/86
423/86-17 SPECIALIST 43 STARTUP TESTING RADIATION SURVEY PROGRAM
6/2-6/86
423/86-18 RESIDENT 153 PLANT OPERATIONS, RADIATION PROTECTION,
5/20-6/23/86 PHYSICAL SECURITY, FIRE PROTECTION, IE
BULLETINS, SURVEILLANCE AND MAINTENANCE
423/86-19 SPECIALIST 16 RADI0 CHEMICAL MEASUREMENTS USING THE NRC
6/2-6/86 REGION I MOBILE LABORATORY
423/86-20 SPECIALIST 31 STARTUP TEST RESULTS REVIEW
6/16-19/86
423/86-21 RESIDENT 115 SHUTDOWN PLANNING, PLANT OPERATIONS, RADI-
6/24-8/11/86 ATION PROTECTION, SECURITY, FIRE PROTECTION,
SURVEILLANCE AND MAINTENANCE
423/86-22 SPECIALIST 18
7/7-10/86 NOTIFICATION AND COMMUNICATION EQUIPMENT
AND PROCEDUP15, OPEN EMERGENCY PREPARED-
NESS ITEMS
423/86-23 SPECIALIST 12 REVIEW 0F RADIATION PROTECTION PROGRAM-
7/7-11/86 TRAINING, EXPOSURE CONTROL, SURVEYS, AUDITS
423/86-24 SPECIALIST 26 SITE SECURITY PROGRAM'
7/14-18/86
423/86-25 SPECIALIST 36 MAINTENANCE PROGRAM PROCEDURES, CALIBRATION
7/21-25/86 CONTROL, AND QUALITY ASSURANCE INTERFACE
423/86-26 SPECIALIST 43 QUALITY ASSURANCE PROGRAMS FOR AUDITS
7/21-8/8/86
423/86-27 SPECIALIST 142 LICENSEE'S IMPLEMENTATION AND STATUS OF
8/18-22/86 TASK ACTIONS IDENTIFIED IN NUREG 0737
423/86-28 RESIDENT 203 SHUTDOWN PLANNING, PLANT OPERATIONS, RADI-
8/12-10/6/86 ATION PROTECTION, SECURITY, FIRE PROTECTION,
SURVEILLANCE AND MAINTENANCE
423/86-29 SPECIALIST 36 PROBLEM AREAS ASSOCIATED WITH SNUBBERS,
8/18-22/06 PORVS AND MAIN STEAM SAFETY VALVES
T1-4
.-
. .
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
423/86-30 SPECIALIST 35 SURVEILLANCE TESTING AND CALIBRATION CON-
9/8-12/86 TROL PROGRAM FOR I&C, PRODUCTION TEST,
OPERATIONS DEPARTMENT
423/86-31 OPERATOR N/A OPERATOR LICENSING EXAMINATION REPORT
12/15-19/86 LICENSING
423/86-32 SPECIALIST 34 QUALITY ASSURANCE AUDIT PROGRAM
9/15-19/86
423/86-33 RESIDENT 132 PLANT OPERATIONS, RADIATION PROTECTION,
10/7-11/17/86 PHYSICAL SECURITY, FIRE PROTECTION, SUR-
VEILLANCE AND MAINTENANCE
423/86-34 SPECIALIST 13 NON-LICENSED STAFF TRAINING
11/17-20/86
423/85-35 RESIDENT 154 OPERATIONAL SAFETY, MAINTENANCE, SURVEIL-
11/18/86- LANCE, LER REVIEW
1/05/87
423/86-36 SPECIALIST 40 EMERGENCY PREPAREDNESS INSPECTION AND OB-
11/19-20/86 SERVATION OF THE ANNUAL EMERGENCY EXERCISE
423/86-37 SPECIALIST 10 0FFSITE REVIEW COMMITTEE ACTIVITIES
12/1-5/86
4
423/86-38 SPECIALIST 16 DEGRADED PROTECTED AREA BARRIER AND COR-
12/11-12/86 RECTIVE ACTIONS
423/86-39 RESIDENT 44 OPERATIONS AND ENGINEERED SAFETY FEATURES
12/29/86-
01/07/87
423/87-01 SPECIALIST 45 ALARA, RADIATION SURVEYS, EXPOSURES, TRAIN-
1/5-9/87 ING
423/87-02 RESIDENT 229 MAINTENANCE, SURVEILLANCE, OPERATIONS,
1/6-2/17/87 RADIATION PROTECTION, RADCON, OUTAGE
TRAINING,QA, SECURITY
423/87-03 SPECIALIST 5 PHYSICAL SECURITY PROGRAM
1/27-29/87
423/87-04 SPECIALIST 12 PHYSICAL SECURITY PROGRAM
- 2/23-27/87
T1-5
.
. .
,
I
TABLE 2
INSPECTION HOUR SUMMARY
NORMALIZED
FUNCTIONAL AREA HOURS % OF TIME ANNUAL HOURS
PLANT OPERATIONS 1365 19.1 910
RADIOLOGICAL CONTROLS 845 11.9 560
MAINTENANCE 359 5.0 240
SURVEILLANCE 554 7.8 370
EMERGENCY PREP. 173 2.4 115
SECURITY AND SAFEGUARDS 409 5.7 270
OUTAGE MANAGEMENT 127 1.8 130
LICENSING N/A N/A N/A
ENGINEERING SUPPORT 262 3.7 175
4 TRAINING N/A N/A N/A
.'
ASSURANCE OF QUALITY 424 6.0 280
OTHER* 2612 36.6 1740
1
TOTAL 7130 100.0 4790
- Includes: construction inspections, the followup of previously identified con-
struction issues, as-built inspection of piping and supports, electrical and in-
strument and controls, preoperational test program implementation, test witnessing
and review and startup test programs, its implementation and test review.
T2-1
_
_ _
__
__ ____
. .
TABLE 3
ENFORCEMENT SUMMARY
SEVERITY LEVEL
FUNCTIONAL AREA 1 2 3 5 TOTAL
4_
,
OPERATIONS 2 1
'
3
RADIOLOGICAL CONTROLS
MAINTENANCE '
SURVEILLANCE ,
l EMERGENCY PREP.
SECURITY AND SAFEGUARDS 3 1 4
OUTAGES
LICENSING I
( TRAINING
ASSURANCE OF QUALITY
OTHER 1 1
_ _ _
2
TOTAL 0 0 1 6 2 9
INSPECTION REQUIREMENT SEVERITY AREA DESCRIPTION
! 423/85-74 ANSI N45.2.2-73 IV CONSTRUCTION EDG CRANKCASE OPENED WITHOUT
! 11/19/85- HOUSEKEEPING MAINTENANCE MATERIAL CONTROLS
1/6/86
10 CFR 50, IV OPERATIONS FAILURE TO FOLLOW EDG FUEL
APP. B.V OIL TRANSFER PROCEDURES
l
SECURITY PLAN V SECURITY FAILURE TO LOCK A VEHICLE f
, SECURITY PLAN IV SECURITY FAILURE TO ESCORT VISITORS
10 CFR III CONSTRUCTION NOT REPORTING A REACTOR
50.55(e) COOLANT PUMP SNUBBER SUPPORT
DEFICIENCY
l
423/86-09 10 CFR 50 V OPERATIONS INDOCTRINATION AND TRAINING
2/18-3/14/86 APP B, CRI 2 0F PERSONNEL {
423/86-38 SECURITY PLAN IV SECURITY FAILURE TO MAINTAIN PRO-
12/11-12/86 TECTED AREA BARRIER
423/86-39 TS 3.5.2 IV OPERATIONS SERVICE WATER T0 "B" HPSI
12/29/86- PUMP ISOLATED
1/6/87
423/87-03 10 CFR 75.21 IV SECURITY FAILURE TO LOCK SAFEGUARDS <
1/27-29/87 INFORMATION REPOSITORY
T3-1
_ - - _ _ _ _ .
__ _ . _ _ . . . __ _ _ _ _ - . . . - - _ . _ ___ _ _-_ _ _ . . . _ _ _ .
-. .- ,,
'
_ .,
/- -(.
>
TABLE 4 i
!
,
LICENSEE EVENT REPORTS
- .
~
A. LISTING OF LERs BY FUNCTIONAL AREA.
,
4
CAUSE CODES
FUNCTIONAL-AREA A B C D E X TOTAL
J
OPERATIONS 12 6 3 3 24
l RADIOLOGICAL CONTROLS 3 1 4
l MAINTENANCE 3 1
~
4 '
.
SURVEILLANCE 6 2 6 2 16
1
EMERGENCY PREP. 0 0
SECURITY AND SAFEGUARDS 5 2 7 14
J
'
OUTAGE MANAGEMENT 0
TRAINING -0
LICENSING 0
ASSURANCE OF QUALITY 0
ENGINEERING SUPPORT- 2 17 2 21
TOTAL 31 25 2 10 15 0 83
f
'
Cause Codes
,
.A - Personnel Error
B - Design / Manufacturing / Construction / Installation
C - External Caus'e .
D - Defective Procedure
E - Component Failure
X - Other
l
,'
T4-1
l
- . _ - -
_. .. _ _ - .- . . . . _ - - _ , . _ . .. _ _ . ._
_ _ _ _ . _. __
l
. .
l
l
l
B. LER SYNOPSIS
i
CAUSE
LER NUMBER EVENT DATE CODE DESCRIPTION
85-001-00 12/09/85 B EMERGENCY DIESEL GENERATOR "A" FUEL OIL
HEADER LEAK - PERFORATION IN THE RETURN
LINE TUBING
85-002-00 12/15/85 A REACTOR TRIP SIGNAL - TWO LOW-LOW LEVEL
BISTABLES ON STEAM GENERATOR "C" WHEN LEVEL
INCREASED AB0VE SETPOINT - CAUSED BY IN-
STALLED JUMPER
85-003-00 12/14/85- A 480 VOLT AC EMERGENCY-BUS, REQUIRED TO BE
OPERABLE PER TS, TAGGED OUT OF SERVICE TO
PERFORM MAINTENANCE
86-001-00 01/16/86 A REACTOR TRIP WITH SI DUE TO LOW STEAM LINE
PRESSURE
86-002-00 01/18/86 B SOURCE RANGE CHANNEL A REACTOR TRIP
86-003-00 01/19/86 B REACTOR TRIP WITH SI DUE TO LOW STEAM LINE
PRESSURE
86-004-00 01/23/86 A PLANT WENT FROM HOT STANDBY MODE TO STARTUP
MODE WITH TS ACTIONS STMT IN EFFECT WHICH
DID NOT PERMIT THIS CHANGE
86-005-00 01/25/86 B TWO CHANNELS OF STEAM GEN A STEAM LINE
PRESSURE WERE FOUND TO BE FAILED HIGH DUE
TO SENSING LINES ON PRESSURE TRANSMITTERS
BEING FR0 ZEN
86-006-00 01/25/86 B VIOLATION OF SLCRS BOUNDARY PENETRATIONS
86-007-00 02/02/86 A PLANT IN MODE 2 WITH LC0 ACTION STATMT FOR
TS 3.8.4.1 NOT MET FOR VERIFICATION OF
CONTAINMENT ELECTRICAL PENETRATION ISOLA-
TION BREAKER POSITION
86-008-00 02/02/86 A
12 HOUR GRAB SAMPLES REQUIRED BY TS 3.3.3.10 WERE NOT BEING TAKEN WITH PLANT
AT 3% POWER
86-009-00 02/04/86 E FWI OCCURRED DUE TO HIGH LEVELS IN STEAM
GEN. 1 AND 4 '
I
i
!
T4-2
- . -
- - . . . .. .
. -
. .
CAUSE
LER NUMBER EVENT DATE CODE DESCRIPTION
86-010-00 02/04/86 A REACTOR TRIP AT 15% POWER DUE TO LEVEL
DEVIATION IN STEAM GENERATOR 2
86-011-00 02/05/86 B CBI SIGNAL GENERATED DUE TO NOISE SPIKE
IN ONE OF THE INSTRUMENT LOOPS
86-012-00 02/06/86 B FWI SIGNAL FROM HIGH-HIGH WATER LEVEL IN
STEAM GENERATOR "C"
86-013-00 02/07/86 B FEEDWATER ISOLATION WITH REACTOR TRIP DUE
TO STEAM GENERATOR WATER LEVEL TRANSIENT
86-014-00 02/10/86 B REACTOR TRIP DUE TO STEAM GENERATOR WATER
LEVEL TRANSIENT-IMPROPERLY DESIGNED LEAD
BEING USED
86-015-00 02/12/86 D REACTOR TRIP DUE TO LOW STEAM GENERATOR
LEVEL-ERROR IN PROCEDURE COVERING OPERATION
OF MAIN FEEDWATER PUMPS
86-016-01 02/08/86 B PRESSURIZER CUBICLE REACHED A TEMPERATURE
OF 121.2 DEGREES FAHRENHEIT AND PLANT
ENTERED ACTION STATEMENT
86-017-00 02/13/86 E REACTOR TRIP DUE TO SSPS GENERAL WARNING
86-018-00 02/14/86 D FWI ON OPENING THE "A" MAIN STEAM ISOLATION
VALVE
86-019-00 02/28/86 A SAFETY INJECTION DUE TO LOW STEAM LINE
PRESSURE
86-020-00 03/01/86 A WITH PLANT IN MODE 3, THE RCS LOOP 2 HOT
LEG INJECTION VALVE WAS FOUND TO BE DANGER
TAGGED SHUT INSTEAD OF LOCKED OPEN AS
REQ. BY MODE
86-021-00 03/01/86 A SI DUE TO LOW STEAM LINE PRESSURE
86-022-00 03/08/86 B FAILURE OF SAFEGUARDS CHANNEL DUE TO
FREEZING
86-023-00 03/11/86 0 DEFECTIVE PROCEDURE FOR MIS-CALIBRATION
OF AREA RADIATION MONITORS IN CONTAINMENT
BUILDING
T4-3
-. . _
.
. . - - . .- .-
. .
,
.
CAUSE
'
LER NUMBER EVENT DATE CODE DESCRIPTION
,
86-024-00 03/15/86 D P-8 PROTECTIVE INTERLOCK SETPOINT HIGH
86-025-00 03/15/86 B CONTROL BUILDING INLET VENTILATION RADI--
ATION MONITOR-INOPERABILITY
'86-026-00 03/01/86 A FAILURE TO MONITOR AFD
86-027-00 03/19/86 B TRAIN "A" EMERGENCY GENERATOR LOAD
SEQUENCER SAFETY INJECTION SIGNAL-
I
86-028-00 03/19/86 D FEEDWATER ISOLATION AND REACTOR TRIP DUE
TO STEAM GENERATOR WATER LEVEL TRANSIENT
,
'
86-029-00 03/29/86 B AREA ES-07 REACHED A HIGH TEMPERATURE OF
121.2 DEGREES FAHRENHEIT
86-030-00 04/10/86 A REACTOR TRIP DUE TO LEVEL DEVIATION IN
86-031-00 04/19/86 8 CBI' SIGNAL DUE'TO CHLORINE DETECTOR FAILURE
'
86-032-00 04/23/86 A REACTOR TRIP ON LOW STEAM GENERATOR WATER
LEVEL
4
86-033-00 04/29/86 A DISCHARGE OF THE LOW LEVEL WASTE DRAIN TANK
WAS PERFORMED WITH THE RADIATION MONITOR
SAMPLE PUMP DE-ENERGIZED
- 86-034-00 05/07/86 D SURVEILLANCE OF ESF BUILDING VENTILATION
4
RADIATION MONITOR SAMPLER FLOW' RATE MONITOR
WAS NOT INCLUDED IN MONITOR SURVEILLANCE
l PROCEDURES
i
86-035-00 05/09/86 A REACTOR TRIP RES8JLTANT FROM TURBINE TRIP
' DUE TO LOW CONDENSER VACUUM SCREEN WASH
REMOVED FOR MAINTENANCE-
'
86-036-00 05/19/86 A PLANT OPERATING IN ACTION' STATEMENT IN THAT
BATTERY BANK 301A-2 WAS NOT OPERABLE DUE' :
TO AN UNPERFORMED MODIFICATION TO CHARGER l
86-037-00 05/10/86 B CBI SIGNAL DUE TO CHLORINE DETECTOR
FAILURE
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CAUSE
LER NUMBER EVENT DATE CODE DESCRIPTION
86-038-00 06/05/86 B PRESSURE BOUNDARY VIOLATION WITHOUT PROPER
NOTIFICATION
86-039-00 06/25/86 8 CBI SIGNAL DUE TO CHLORINE DETECTOR
FAILURE
86-040-00 07/21/86 B CBI SIGNAL DUE TO CHLORINE DETECTOR FAILURE
86-041-00 07/24/86 E RX TRIP CAUSED BY LOW LOW STEAM GENERATOR
LEVEL DUE TO HIGH LEVEL FEEDWATER ISOLATION
86-042-00 07/25/86 B SAFETY INJECTION ACTUATION CAUSED BY IN-
TERMITTENT RESETTING OF PRESSURIZER LOW
PRESSURE SI BLOCK
86-043-00 07/29/86 B INCORRECT MAIN STEAM SAFETY RELIEF VALVE
BLOWDOWN RING SETTINGS
86-044-00 07/31/86 A BYPASSED LIQUID DISCHARGE VALVE WITHOUT
DOUBLE VALVE LINEUP VERIFICATION
86-045-00 07/31/86 E CONTAINMENT LOCAL LEAK RATES EXCEEDED
86-046-00 08/01/86 E FAILURE OF B TRAIN EMERGENCY DIESEL
GENERATOR DUE TO UNKNOWN CAUSES
86-047-00 08/15/86 D OVERTEMPERATURE DELTA T SETPOINT HIGH DUE
TO ADMINISTRATIVE ERROR
86-048-00 08/17/86 A REACTOR TRIP DUE TO STEAM GENERATOR WATER
LEVEL TRANSIENT CAUSED BY OPERATOR ERROR
86-049-00 08/17/86 A FEEDWATER ISOLATION AND REACTOR TRIP DUE
TO STEAM GENERATOR WATER LEVEL TRANSIENT
CAUSED BY OPERATOR ERROR
86-050-01 09/02/86 B AREA TEMPERATURE MONITORING MS-01 -
86-051-00 09/06/86 E REACTOR TRIP DUE TO LOW STEAM GENERATOR
LEVEL CAUSED BY FAILED FEEDWATER ISOLATION
VALVE
86-052-00 09/18/86 A MISSED FIRE PROTECTION SURVEILLANCE
T4-5
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CAUSE
LER NUMBER EVENT DATE CODE DESCRIPTION
86-053-00 10/15/86 D INCORRECT INTERMEDIATE RANGE DETECTOR
SETPOINTS
86-054-00 10/30/86 B FIRE WATCH NOT ESTABLISHED IN REACTOR CON-
TAINMENT WITHIN ALLOTTED TIME
86-055-00 11/06/86 E FAILURE OF B EMERGENCY DIESEL GENERATOR
TO START IN LESS THAN 10 SECONDS
86-056-00 11/30/86 A INOPERABILITY OF "B" TRAIN SAFETY INJECTION
PUMP COOLER
86-057-00 12/16/86 D INCORRECT REACTOR COOLANT SYSTEM FLOW
SETPOINTS DUE TO ADMINISTRATIVE ERROR
86-058-00 12/17/86 A INADEQUATE RAD MONITOR SURVEILLANCES DUE
TO INADEQUATE TS REVIEW
86-059-00 12/27/86 B UNSEALED SLCRS PRESSURE BOUNDARY
87-001-00 01/13/87 A REACTOR TRIP AS A RESULT OF CIRCULATING
WATER PUMP DUE TO PERSONNEL ERROR
87-002-00 01/14/87 A REACTOR TRIP DUE TO ACCIDENTAL RESET OF
SOURCE RANGE CHANNEL BLOCK
87-003-00 01/14/87 E FAILURE OF "B" EMERGENCY DIESEL GENERATOR
TO START IN LESS THAN 10 SECONDS
87-004-00 01/29/87 B MOTOR DRIVEN AUXILIARY FEEDWATER PUMP TRIPS
DUE TO LOW SUCTION PRESSURE TRIPS
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87-005-00 02/01/87 D CONTROL ROOM PRESSURIZATION SURVEILLANCE
FAILURE CAUSED BY MISPOSITIONED THROTTLE
VALVE.
87-006-00 02/01/87 A MISSED AREA TEMPERATURE MONITORING SUR-
VEILLANCE DUE TO PERSONNEL ERROR AND PRO-
CEDURE INADEQUACY
87-007-00 02/11/87 A MISSED TECHNICAL SPECIFICATION ON CONTAIN-
MENT DRAIN SUMP INVENTORY DUE TO OPERATOR
ERROR
T4-6
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SECURITY RELATED LERs:
CAUSE
LER NUMBER EVENT DATE CODE DESCRIPTION
50-245/85-32 12/28/85 A SECURITY OFFICER LEFT POST PREMATURELY-
50-245/86-01 1/5/86 E INTELLIGENT 000R CONTROLLER FAILURE
50-245/86-02 1/18/86 E INTELLIGENT DOOR CONTROLLER FAILURE
50-245/86-03 1/25/86 E INTELLIGENT 000R CONTROLLER FAILURE
'
50-245/86-04 2/4/86 E INTELLIGENT DOOR CONTROLLER FAILURE
50-245/86-12 4/12/86 A BREACH OF PROTECTED AREA BARRIER
50-245/86-13 4/14/86 A VITAL AREA D0OR DISARMED DURING A
SURVEILLANCE
50-245/86-14 4/21/86 C B0MB THREAT H0AX
50-245/86-16 5/1/86 A SECURITY OFFICER ASLEEP ON DUTY
50-245/86-20 8/12/86 E LOSS OF POWER TO SECURITY SYSTEM
50-245/86-21 9/11/86 E . INTELLIGENT DOOR CONTROLLER FAILURE
50-245/86-24 11/14/86 C CONTRACTOR VIOLATES SITE FIREARMS
RESTRICTION (FIREARM DID NOT ENTER
THE PROTECTED AREA). ,
50-245/86-30 12/11/86 A BREACH OF PROTECTED AREA BARRIER
50-245/86-31 12/23/86 E LOSS OF POWER TO SECURITY SYSTEM
>
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General Offices e Selden Street, Berlin, Connecticut
v,e co cicur uo ' e oma co--
msvm. ss.c mnsasevaccow- P.O. box 270
xvt* waaao"*"" "
HARTFORD, CONNECTICUT 06141-0270
L L J [,* ,,",' ['C,c"o*,,," (203) 665-5000
February 18,1987
Docket No. 50-423
B12186
U. S. Nuclear Regulatory Commission
Attn: Document Control Desk
Washington, D.C. 20555
References: (1) T. E. Murley letter to 3. F. Opeka, Systematic Assessment
of Licensee Performance (SALP) Report No. 50-423/85-99,
dated December 27,1985.
(2) 3. F. Opeka letter to T. E. Murley, Response to SALP
Report 50-423/85-99, dated February 11,1986.
(3) T. E. Murley letter to 3. F. Opeka, Systematic Assessment
of Licensee Performance (SALP) Report No. 50-423/85-99,
dated March 21,1986.
Gentlemen:
Millstone Nuclear Power Station, Unit No. 3
Systematic Assessment of Licensee Performance (SALP)
The purpose of this letter is to inform you of the status of corrective actions
taken as a result of the SALP Board's recommendations that were provided to us
in the last SALP review period. In addition to providing you with the status of
corrective actions, we would also like to take this opportunity to provide some
information concerning our performance over the past year which we believe will
be useful to the SALP board in their next assessment of Millstone Unit 3. In
Reference (1), the NRC issued the Millstone Unit 3 SALP report for the
twelve month period ending August 31, 1985. In Reference (2), Northeast
Nuclear Energy Company (NNECO) provided its responses and comments on
SALP Report No. 50-423/85-99. In Reference (3), the NRC provided its l
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comments on NNECO's Reference (2) submittal. '
At the time of our last submittal (Reference 2), a number of corrective actions
had been completed and they were addressed in that letter. This submittal will
provide information, which is contained in Attachment 1, on the additional
corrective actions taken since then.
Additionally, Attachment 2 provides a summary of some of the key
accomplishments on Millstone Unit 3 over the past year as well as some
examples of Northeast Utilities (NU) productive participation in industry
activities and positive involvement in the regulatory process.
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We believe that you will find the-actions outlined herein that address the Board's '-
recommendations satisfactory and that you may find the additional.information
on our positive involvement in the regulatory process to be of 'v'alue in your next-
~SALP assessment- of Millstone Unit 3. .Please feel freeito contact us if you
- require any additional information.
Very truly yours,
NORTHEAST NUCLEAR ENERGY COMPANY
'
E.JTfoczka - () .
SeniorVice President-
.
cc: Dr. Thomas E. Murley, Regional Administrator, Region 1 -
E. L. Doolittle, Licensing Project Manager, NRR ..
3. T. Shediosky, Senior Resident Inspector, Millstone Unit No. 3
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Docket No. 50-423-
B12136
Attachment 1
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Northeast Nuclear Energy Company
Millstone Unit No. 3
Update to SALP Report 50-423/85-99 Recommendations
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February,1987
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Functional Areat OPERATIONS SUPPORT
Board Recommendation:
Review control of ' and training ' for jumpers - and lif ted leads, . tagging, log
keeping, and shift turnover requirements to assure controls are adequate for
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power operation.
Status: -
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. NNECO has implemented Millstone. Station Administrative Control Procedures
ACP-QA-2.06.A, B & C, which cover control of bypass jumpers, lifted leads,_
i- and tagging; ACP 6.12, shif t turnover; and ACP 10.05, log keeping requirements
- for- all plants at -the Millstone Station. The controls delineated in these
. procedures were initially implemented at Millstone' Units 1 and 2 and have
4
proven to.be very effective. In or' der to. assure these controls were appropriate
for power operation .and to familiarize Millstone . Unit 3 operating -personnel
with these procedural requirements prior to power operation, these procedures
were instituted during startup testing, far in advance of power operation.
In April,1986, the NRC conducted an operations _ audit . on Millstone Unit 3
- (Audit No. 8612). No weaknesses in the area of bypass jumper and lif ted lead
, - control were identified.
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Functional Area: RADIATION CONTROL
Board Recommendation:
Assure the FSAR accurately describes the solid radwaste system.
Status:
At the present time, the FSAR reflects the as built configuration of the solid
radwaste system. The FSAR will be updated in accordance with 10CFR50.71 to
reflect any modifications made to the system in the future.
We would also like to provide some information concerning our performance
over the past year in the area of radiological controls at Millstone Unit 3. As
noted in the Millstone Unit 3 SALP report (Reference (2)), the radiological
controls implemented at Millstone Unit 3 are identical to those which have been
used at Millstone Units 1 and 2. As a result of SALP Board recommendations on -
Millstone Units 1 and 2, NNECO has strengthened radiological controls in
several areas, namely radiation worker training, radiation exposure reduction,
and radwaste handling and shipping.
Radiation Exposure Reduction
With regard to radiation exposure, the cumulative exposures at Millstone Unit 3
have been extremely low. The 1986 total was about 27 person rem.
Corpora tely, NU has recently undertaken a program to lower collective
exposures for all of our plants to meet INPO goals. The program is
investigating methods of reducing dose rates and work scope in high radiation
areas, and improving worker efficiency at all plants.
Radwaste
Several changes have occurred during the past year which are expected to yield
significant improvements in the implementation of the Millstone Station
radwaste management program. Examples of these are:
o The Millstone radwaste handling group has been expanded in size and
reorganized under a separate supervisor who is responsible solely for
implementation of the radwaste management program.
o increased training is being given to radwaste handling and quality
control personnel to expand their knowledge of radwaste manifest
preparation, shipping, and burial regulations,
o Nuclear Engineering and Operations Procedure 6.07 " Quality
Assurance and Quality Control in Station Radioactive Material
Processing, Classification, Packaging, and Transportation" was issued
which defines the quality related aspects of the radwaste shipping
process.
o A NU corporate radwaste engineering group has been approved for
implementation in 1987. Staffing for this group, which will provide
engineering expertise in all areas of radwaste processing, is currently
underway.
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Radiation Worker Training -
Radiation worker training is administered as part of our. General Employee
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Training Program and is updated annually,to include lessons learned from the -
. previous year, as .well .as . NRC, INPO and NU .significant . findings from the
- previous year. ,
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Additionally, supervisors have been reminded of their responsibilities in assuring .
worker radiation protection. This includes providing all .the equipment, training
and controls' necessary to ensure :that their workers: perform their. jobs both .
safely and efficiently.
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in summary, we believe that the above actions illustrate NU's commitment to
~
maintaining proper radiological controls at Millstone Station.
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Functional Areat MAINTENANCE
Board Recommendation:
Establish ~a schedule for the completion and implementation of maintenance
related procedures and training programs.
Status:
All maintenance procedures necessary to support operation of the unit have
been approved and implemented.
Please refer to the TRAINING 'AND QUALIFICATION. EFFECTIVENESS
functional area for a discussion of the training programs related to
- maintenance.
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Functional Area: SURVEILLANCE
Board Recommendation:
Assure surveillance procedures support future planned testing and operations.
Particular emphasis should be placed on orderly development and review of
procedures.
Status:
All surveillance procedures have been developed, reviewed and implemented to
meet the requirements of the lechnical Specifications.
Beginning in early 1985, a significant effort was expended in the development
of the surveillance testing program. Many of the tests were incorporated into
the startup test program which eliminated duplicate testing and permitted
operational experience to be gained and factored into the surveillance test
procedures.
A number of procedures for tests which are conducted during refueling outages
or less frequently are still under development. These procedures are being
developed on a schedule which will permit adequate review and training prior to
conduct of the tests.
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Functional Area: TRAINING AND QUALIFICATION EFFECTIVENESS
This functional area was not evaluated during the previous Millstone 3 SALP.
However, we feel it is important to inform you of our progress in the areas of
training programs.
Technical Training
On October 1,1986, NU submitted the Millstone Unit 3 technical training
program to INPO for accreditation (approximately 1 1/2 years ahead of
schedule). The decision to expedite the implementation of the accreditation
process was based on NU's continued commitment to excellence. In addition,
the Nuclear Training Department has commenced development of training
programs in the Radioactive Waste Worker and Quality Assurance / Quality
Centrol disciplines to the same accreditation standards. This decision was
predicated on the belief that even though the latter two programs are not part
of the INPO accreditation effort, the critical nature of these job functions in
the day-to-day operation of the unit dictate no less a quality commitment.
The Technical Training Branch is presently staffed with nine full-time technical
instructors who are exclusively committed to supporting the technical training
requirements of Millstone Unit 3. In addition, recognizing the invaluable
benefits of practical hands-on training, NU has established a fully equipped
laboratory for each journeyman discipline. During 1986, 20 % of the entire
training program was presented to approximately 20 % of the student
population. Our 1987 plans call for each mechanic, electrician, and technician
to participate in approximately five weeks of technical training. The
curriculum chosen for the 1987 schedule was guided by the plant supervisory
staff of Millstone Unit 3 based upon their operational requirements.
In a continuing effort to estab!!sh a lead position in the industry through
innovative training techniques, NU is in the process of piloting programs in the
fields of team training, diagnostic training, and such practical hands-on courses
as Reactor Coolant Pump Seal Overhaul. In the case of the latter, the RCP
Seal course is being presented six times prior to the Millstone 3 March,1987
mid-cycle outage. This course incorporates the use of a full scale mockup of
the seat assembly mounted in a bell housing. The team training process involves
Mechanics, Quality Control Engineers, Reliability Engineers, Safety Engineers,
Health Physics Technicians and ALARA Engineers all simultaneously attending
these courses, each offering their expertise to the training process. As a result
of this multi-discipline approach, several modifications to the existing
maintenance procedures have been incorporated that should reduce radiation
exposure and radwaste production, while at the same time improving the overall
human safety aspects of conducting the job.
Operator Training
Many significant improvements have been made in the area of Operator
Training.
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The organization and staffing'of this branch has been strengthened to provide
one supervisor for each nuclear unit with (2) assistant supervisors reporting to
him. The authorized staffing level has been increased to fourteen (14)
instructors per nuclear unit.
To ensure that the Operator Training Branch can attract the talented personnel
necessary to perform this critical function, position grade levels have been
upgraded such that many experienced plant operating personnel have been
attracted to a career in the Nuclear Training Department. It is noteworthy
that this action received corporate and station support, thus illustrating _the
recognition of the importance of the training functions.
During the past year, the Millstone Unit 3 operator training programs
completed cold license training, with 42 of 45 candidates receiving NRC
operator licenses. The Licensed Operator Requalification Training program was
successfully completed by all licensed personnel, and the first training program
for replacement operators was completed with 12 of 12 candidates receiving
NRC operator licenses.
The training program for the Millstone Unit 3 Operations Shif t Advisors was
successfully completed in February,1986.
The Millstone Unit 3 plant specific simulator had an availability of greater than
98% for 1986 bringing the capability for training nuclear plant operators to the
highest possible level.
A job and task analysis has been completed for all operator job positions in
preparation for INPO accreditation. Formal learning objectives are being
developed to support operator training programs, and are being incorporated
into all on-going programs as the development activity proceeds. INPO
accreditation activities are firmly on track, and the Accreditation Self
Evaluation Report will be submitted to INPO by November 1,1987.
General Nuclear Training
In October,1986 a new organization was announced for_ the General Nuclear
Training Branch. The changes primarily affected the personnel that are
supporting general training activities at the nuclear stations and should result in
.
improved efficiency in training station engineering personnel, emergency
response training, radiation worker, fire brigade, production maintenance
management and medic first-aid safety training.
The Branch now consists of three sections, two of which are located at the
Millstone Training Center, and one at the NU corporate office. The Millstone-
based staff supports the training discussed above at both the Millstone and
Haddam Neck sites and the corporate section provides corporate nuclear
training for offsite engineering personnel. The corporate staff is also
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responsible for - managing the Shif t Technical Advisor college program at
Thames Valley State Technical College.
The General' Nuclear Training Branch's priority goal at the present time is to
achieve INPO accreditation of the Haddam Neck and Millstone Technical Staff
and Manager (TSM) Training program, 'a goal that we feel confident about
meeting. The TSM Accreditation Self Evaluation Report (ASER) was submitted -
to INPO 'on October 1,1986 and course work refinements and teaching the -
approximately fifty new courses to plant engineering personnel has begun. We
are hopeful that the INPO Accreditation Team will visit in the latter part of
1987 and ultimately grant NU this important certification.
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Functional Areat LICENSING ACTIVITIES
Board Recommendations:
Increase management involvement in the licensing review process in order to
assure more timely resolution of licensing issues.
Status:
Senior NU management is routinely and actively involved in the management of
licensing issues. This is acknowledged and documented by the NRC in recent
SALP reports issued on our other Millstone plants as well as Millstone Unit
3.(1)(2) NU hss in the past and will continue to utilize all of the experience
gained from its other nuclear plants to develop consistent and technically sound
resolutions to safety issues. A high level of management review and approval
of all correspondence with the NRC is procedurally required at NU to ensure a
consistently clear licensee understanding and responsiveness to NRC initiatives.
Additionally, we have undertaken severalinitiatives to ensure that management
remains fully cognizant and involved in unresolved licensing issues. Examples
of these are discussed below,
o We have designated a Millstone Unit 3 lead licensing engineer to
facilitate communications with the NRC Project Manager.
o Our lead licensing engineer has worked closely with the NRC
Project Manager to establish a prioritization system containing all
key outstanding licensing items. This information is updated
frequently and assures appropriate priority focus and timely
resolution.
o Periodic meetings have been held between NU management and
NRC project management to assess the status of outstanding items
and thus assure that adequate resources are committed to achieve
timely resolution.
o High levels of NU management have been extensively involved in
industry groups that support NRC initiatives. NUMARC, AIF, INPO
and EEI are representative examples.
We believe that the above actions have contributed to maintaining clear
communications between the NRC and NU on outstanding information requests
and other licensing actions thereby allowing timely decisions to be made to
resolve outstanding issues.
(1) T. E. Murley letter to J. F. Opeka, "SALP Report Nos. 50-2t 5/85-98 (Pg.
32 and 33) and 50-336/85-98 (Pg. 31)," dated August 29,1986.
(2) T. E. Murley letter to J. F. Opeka, "SALP Report No. 50 /1 23/85-99" (Pg.
28), dated December 27,1985.
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1-10
During the past year NU has continued to be very responsive to NRC staff
requests for information. NU has provided information required to satisfy the
following 8 of 11 license conditions requiring submittal of additional
information.
-
2.C.4 - 3 Loop Operation (July 1,1986)
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2.C.5 - Inservice Inspection Program (May 22, 1986)
-
2.C.6 - Instrumentation for Monitoring Post Accident Conditions R.G.
1.97 Revision 2 Requirements (December 9,1985)
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2.C.9 - Operating Staff Experience Requirements (July 3,1986)
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2.C.10 - Changes to Initial Test Program (February 12, February 20,
March 12, March 24, May 2, May 6, May 19, and July 18, 1986)
-
2.C.ll - Revised Small Break LOCA Methods to Show Compliance
with 10 CFR 50.46, TMI Stem II.K.3.31 (June 9,1986)
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2.C.13 - Detailed Control Room Design Review (May 20, 1986)
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2.C.14 - Salem ATWS Events Generic Letter 83-28 (May 13,1986)
We have continually strived to provide comprehensive, thorough, and
technically sound submittals. In cases where the NRC staff has required
additional information, we have been quick to respond to the request with
follow up telephone conference calls, meetings or additional written submittals.
We believe a prime example of this has been our pursuit of NRC approval for 3
loop operation. NU is unique in the nuclear industry in its request for approval
to operate Millstone Unit 3 with one reactor coolant loop isolated. We have
expended substantial resources to ensure that our request was founded on a firm
technical base. We have consistently demonstrated diligence in our follow-up
of NRC staff questions and concerns by providing additional information in
meetings, telephone conference calls and written correspondence. In each case,
NU was able to provide the NRC with the necessary information "on-the-spot"
or was able to obtain a clear understanding of what was required to resolve the
concern in a timely manner. It is our understanding that we have provided all
of the information necessary for the NRC to complete its review of this issue
and we are awaiting the staff's final safety evaluation and approval. We have
had a very cooperative working relationship with the NRC on this unique
licensing application.
Another area which we feel exemplifies our responsiveness to the NRC is
updating the Millstone Unit 3 FSAR. Three FSAR updates were submitted
within the first year following license issuance whereas 10CFR50.71 does not
require submittal of the first update until two years. NU has committed
substantial resources to enable us to exceed regulatory requirements in this
regard.
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1-11
We continue to maintain a knowledgable and highly motivated licensing staff.
Millstone Unit 3 licensing personnel have received training both in-house and
outside in areas such as:
- Quality Assurance
- The Nuclear Safety Ethic
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Nuclear Engineering and Operations procedures affecting licensing
(technical specification changes, license amendments, safety evaluations,
FSAR updates)
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Millstone Unit 3 Systems
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NRC Unresolved Safety issues
Additionally, Millstone Unit 3 licensing personnel are participating on various
subcommittees of the Westinghouse Owners Group.
In summary, we feel that the licensing activities associated with Millstone Unit
3 continue to demonstrate that NU management is firmly committed to
providing the proper resources and direction necessary to effectively resolve all
issues which have the potential to affect the safety of the plant.
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Docket No. 50-423
D14156
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Attachment 2
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Northeast Nuclear Energy Company __
Millstone Unit No. 3
Examples of NU Performance During Current SALP Period
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The following is a summary of various meetings, letters, or other activities that
occurred during the period January 1,1986 to January 31,1987 which we feel are
relevant to the Millstone Unit 3 SALP evaluation.
o The following plant startup milestones were achieved:
- January 23,1986 - Initial criticality.
- January 31, 1986 - Issuance of Millstone Unit 3 operating .
license NPF-49 authorizing full power operation.
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April 21,1986 - Completion of the startup test program.
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April 23,1986 - Start of commercial operation.
o January 8,1986 - A meeting was held between NU management and .
-NRC/NRR to discuss the status of remaining licensing issues prior to
issuance of the full power operating Ilcense.
o January 9,1986 - A meeting was held between NU management and
NRC Region 1 to discuss the status of remaining licensing issues prior
to issuance of the full power operating license.
o January 23, 1986 and February 19, 1986 - Meetings were held
between representatives of NU and the NRC to discuss the issue of
station blackout with respect to Millstone Unit 3.
o March 18, 1986 -
NU submitted a letter providing additional
information on station blackout for Millstone Unit 3.
o May 12,1986 - A meeting was held between representatives of NU
and the NRC Licensing Project Manager at the Millstone Station to
discuss the status of licensing activities.
o June 18, 1986 - NU provided comments on the proposed station
blackout rule. NU has been an active member of the industry effort
to resolve the USI-A-44, Station Blackout issue. In this regard, the
industry, via the Nuclear Utility Management and Resource
Committee (NUMARC) and the Nuclear Utility Group on Station
Blackout, has been working with the Staff towards a mutually
agreeable resolution to this issue. NU personnel have lead roles in
these committee initiatives.
o June 18,1986 - NU submitted a letter proposing to extend the use of
Integrated Safety Assessment Program methodology to Millstone
Units 2 and 3.
o June 25,1986 - NU submitted Revision I to the Millstone Unit 3
Inservice Test Program for pumps and valves.
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o July- 3,1986 - NU submitted a letter providing information regarding
actions taken by NU in response to IE Information Notice = 86-47,
Erratic Behavior of Static "0" Ring Differential Pressure Switches.
Although a response to this Information Notice was not required,' NU
felt it was appropriate to inform the NRC of our followup on this'
issue because Millstone was specifically mentioned in the Information
Notice as having received the subject switches.
o July 13,' 1986 - A meeting was held between representatives of NU
and NRC- project management to discuss the status of licensing
activities.
o July 22,1986 - NU submitted the Millstone Unit 3 startup report.
o July 28,1986 - A meeting was held between representatives of NU
and the NRC to discuss NRC staff concerns related to 3-loop
operation of Millstone Unit 3.
o On September 17 and 18,1986, NU hosted a Region I Fire Protection
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Organization seminar. The seminar was attended by NRC-
representatives from NRR and Region I as well as numerous utility
representatives. The seminar was well received by all in attendance
with recommendations that similar seminars be held in the future.
o In September,1986, NU Implemented an emergency preparedness
surveillance tracking system at the Millstone Station to ensure that
facilities and equipment are maintained operational. -
o On October 1,1986, NU provided comments on a draft report written
by Brookhaven National Laboratory entitled " Evaluation of
Reliability Technology Applicable to' LWR Operational Safety."- NU
has undertaken numerous initiatives aimed at maintaining high safety
system availability, such as development and use of living PRAs and
implementation of a Safety System Unavailability -Monitoring
Program.
o On November 19, 1986, a full participation emergency exercise was
successfully conducted at the Millstone Station. The exercise, which
involved Connecticut, Rhode Island, and local Emergency Planning
Zone communities, was evaluated by both FEMA and the NRC. No
major findings of deficiencies were identified.
o On January 13, 1987, Millstone Unit 3 completed 128 days of
continuous operation and estabilshed a plant record for continuous
service.
o in an effort to improve the timeliness of providing site access to
NRC inspectors, NU developed and implemented a " Read and Sign"
training program. On October 10, 1986, NU transmitted a letter to
the NRC Region I describing the program and our plans for
implementing it.
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Noteworthy
the NU QA /QC programs include the following: changes which have occurr
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The Operations QA staff has been relocated from the corporate
offices to the Millstone site. This action -is expected to
increase the effectiveness of the quality organization by
maintaining a full-time presence on site. This will allow
improved communication between the plant operating staff and
QA staff and will expand the QA department's knowledge and
evaluation of plant problems by allowing increased observation
of on-going plant activities.
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A standardized corporate QC manual has been issued which will
result in the Haddam Neck, Millstone, and Betterment
Construction QC organizations working to the same set of
procedures. This will assure consistent application of all QC
activities and will allow better utilization of personnel because
allinspectors will be trained and qualified to the same program.
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