IR 05000443/1988010
| ML20155E638 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/28/1988 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20155E606 | List: |
| References | |
| 50-443-88-10, IEB-87-002, IEB-87-2, IEB-88-005, IEB-88-5, IEIN-86-050, IEIN-86-50, IEIN-88-025, IEIN-88-046, IEIN-88-25, IEIN-88-46, NUDOCS 8810120361 | |
| Download: ML20155E638 (66) | |
Text
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.:
50-443/88-10 License No.:
NPF-56 Licensee:
Public Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit No.1 Inspection At: _S_eabrook, New Hampshire inspection Conducted: July 6 - September 6,1988 and September 21, 1988 Inspectors:
A. C. Cerne, Senior Resident Inspector, Seabrook Station D. G. Ruscitto, Senior Resident Inspector, Seabrook Station E. Yachimiak, Operations Engineer (Examiner), PWR Section, Division of Reactor Safety C. J. Conklin, Senior Emergency Preparedness Specialist, Emergency Preparedness Section, Division of Radiation Safety and Safeguards Approved By*
O $ ^uo (2AlN w3 Donald R. Haverkamp, Chiaf, eactor Projects Date Section No.3C Inspection Summary:
Areas Inspected:
Routine inspection on day and backshirts by two resident inspectors and two regional specialist inspectors of actions on previous inspection findings, NRC Bulletins and Information Notices, operational safety, licensee potentially reportable occurrences and operational events, maintenance and survelliance activities, design changes, allegations, training, and electrical c(nfiguratior, control.
G810120361 881006 POR ADOCK 05000443 O
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Inspection Summary (Continued)
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Results:
1. General Conclusions.
A repetitive weakness was identified in the implementation of the tagging program involving physical removal of a section of non-safety related piping containing a valve which was caution tagged. While the non-safety nature of the equipment indicates that regulatory requirements were not violated, the recurrent nature of the incident indicates that further management attention in this area is warranted (Refer to paragraph 8.b).
A weakness was identified in the licensee's reporting system with respect to diesel generator failures (Refer to paragraph 4.k)
A weakness was identified in the calculations associated with ncn-class 1E loads powered from class 1E power sources.
Licensee evaluat',on of this problem is continui19 and is being tracked under existing unre solved item 88-06-01 (Refer to paragraph 4.j).
A licensee strength was demonstrated in the handling of testing and inspection of flanges and fittings in accordance with NRC Bulletin 88-05.
Strong participation by quality assurance and engineering personnel con-tributed to the licensee's ability to respond to this industry wide problem in a timely fashion (Refer to paragraph 6.b).
2.
Violations.
A violation was identified regarding the failure to report diesel gener-ator failures in accordance with the technical specifications (Refer to paragraph 4.k.).
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TABLE OF CONTENTS Page 1.
Persons Contacted..............................................
2.
Summa ry of Faci l i ty and NRC Acti vi ti es.........................
a.
Resident Inspector Activities.............................
I b.
Visiting Inspector Activities.............................
c.
Plant Status..............................................
3.
Operational Safety.............................................
a.
Plant Inspection Tours (71707, 71710)*....................
b.
Operational Events (93702)................................
4.
License Action on Previous Inspection Findings (92/01).........
a.
Unresolved Item 86-54-02: CBS Piping Design..............
b.
Unresolved Item 87-10-02:
RHR Valve Alignment Questions..
c.
Unresolved Item 87-16-03: Operation of the SUFP on an Emergency Bus...........................................
d.
Inspector Follow-up Item 87-22-01:
Siren Modifications...
e.
Inspector Follow-up Item 88-09-01:
TSC/E0F Technical Support.................................................
f.
Inspector Follow-up Item 88-09-02:
TSC/OSC Multiple Access Po1nts...........................................
g.
Inspector Follow-up Item 88-09-03:
Departing Shift 0osimetry...............................................
h.
Inspector Follow-up Item 88-09-04: Media Center Responses to Press Inquiries......................................
1.
Unresolved Item 88-02-01:
SI Accumulator Isolation Valve Control Circuitry.................................
j.
Open Item 88-06-01: Non-Class IE Loads Powered From Class IE Sources........................................
k.
Violation 88-06-02:
EDG Failure Reporting................
5.
' Licensee Reports (92700).......................................
a.
Construction Deficiency Report 86-00-09:
Veritrak/Tobar Transmitters............................................
b.
10 CFR 21 Report 87-88-04: Gould Relay Failures..........
c.
10 CFR 21 Report 87-88-03:
Service Water System Valve Liners and Seats........................................
d.
Station Information Reports...............................
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Table of Contents (Continued)
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Page 6.
NRC Bulletins and Information Notices (92701)..................
a.
NRC Bulletin 87-02: Fastener Testing to Determine with Applicable Material Specifications.................
b.
NRC Bulletin 88-05: Nonconforming Materials Supplied by PSI and WJM..........................................
c.
Licensee Report of Defective Circuit Breakers..............................
d.
NRC Information Notice 88-25: Minimum Edge Distance for Expansion Anchor Bo1ts..................................
e.
Inadequate Testing to Detect Failures of Safety-Related Pneumatic Components or Systems................................................
7.
Surveillance / Maintenance (61840, 61726, 62703).................
a.
OX 1456.81: Operability Test of ISI Valves...............
6.
EX 1804.044: Safety and Relief Valve Setpoint Pressure Test....................................................
c.
EX 1804.016: Diesel Generator Auxiliary Coolant System Quarterly Test
......................................
d.
IX 1680.921: SSPS Train "A" Actuation Logic Test.........
e.
EX 1804.015: Diesel Generator 1B 18-Month Operability and Engineered Safeguards Pump and Valve Response Time Testing Mode 5 Surveillance............................
f.
0X 1406.02: CBS Pump and Valve Quarterly Test and 18-Month Remote Position Indication.............................
g.
System........................
8.
Design Changes and Modi fications (37700, 37701)................
a.
Post Accident Sample System (PASS)........................
b.
Secondary Component Cooling Water (SCCW) System...........
9.
Allegation Followup (92701)....................................
10.
Training (41400, a1701)........................................
a.
General Employee Training..........
......................
b.
Operator Training.........................................
11.
Electrical Configuration Control (92701).....................
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12. Management Meetings (30703,30702).............................
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Table of Contents (Continued)
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Attachments:
A.
Meeting Attendees, Meeting conducted August 17, ik B.
Meeting Slides, Meeting conducted August 17, 1988 The NRC Inspection Manual inspection procedure that was used as inspection
guidance is listed for each applicable report section.
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DETAILS 1.
Persons Contacted - New Hampshire Yankee (NHY)
E. A. Brown, President and Chief Executive Officer
- W. A. DiProfio, Assistant Station Manager
- T. C. Feigenbaum, Vice President, Engineering, Licensing and Quality Programs W. J. Hall, Regulatory Services Manager
- D. E. Moody, Station Manager G. S. Thomas, Vice President, Nuclear Production
- J. M. Vargas, Manager of Engineering
- J. J. Warnock, Nuclear Quality Manager Attended exit meeting conducted on September 9, 1988
Attended exit meeting cond" ted on September 22, 1988 Interviews and discussions with other members of licensee and contractor management, and with their staf f s, were also conducted relative to the inspection of items documented in this report.
2.
Summary of Facility and NRC Activities a.
Resident Inspector Activities On August 8-11, 1988, the Resident Inspector attended a Resident Inspector Seminar in King of Prussia, Pennsylvania.
On August 8-19, 1988, the Senior Resident Inspector travelled to Rockville, Maryland for a temporary assignment with the NRC Office of Nuclear Reactor Regulation.
On August 17, 1988, the resident inspectors attended a management meeting between the NRC and NHY in King of Prussia, Pennsylvania.
(Refer to paragraph 13 of this report)
On September 1, 1988, the Senior Resident Inspector was reassigned to another duty station.
The Resident Inspector was assigned as Senior Resident Inspector, b.
Visiting Inspector and NRC Management Activities On July 18-22, 1988, an NRC Region I operations engineer (examiner)
conducted a routine inspection of plant operations and previously identified items.
His inspection findings are included in this report.
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On August 16, 1988, the Director, Office of Nuclear Reactor Regula-tion visited the site. He held discussions with the Resident Inspec-tor and toured the plant. The NHY inventory department staff was requested to provide information concerning the Seabrook program for material receipt inspection and identification of fraudulent or substandard parts.
On September 21, 1988, an NRC Region I senior emergency preparedness specialist conducted a routine inspection of previously identified items.
His inspection findings are included in this report, c.
Plant Status During this r~orting period, the plant remained in operational Mode 5,
cold shutJown, with primary temperature between 105 and 140 degrees F and depressurized. Major maintenance was conducted on ser-vice water cooling tower pump SW-P-110A, the reactor trip breakers, the chemical and volume control system, the control building air handling sy stem, the waste gas system, the diesel generators and switchyard circuit breakers and bus ducts.
Major 18-month surveillance was conducted on the emergency diesel generators, emergency core cooling systems, engineered safety fea-tures actuation systems and ventilation filters.
On July 19, 1988, while performing surveillance testing on the train
"A" containnent building spray system, an improper valve lineup caused approximately 5,000 gallons of water from the refueling water storage tank to flow to the suction of the operating train "A" residual heat removal pump suction and into the reactor coolant sys-tem.
Details of this event may be found in paragraph 7.f of this report.
Significant design changes were initiated on the secondary component cooling water and post a
Inspection Report
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50-443/88-09 stated,
" The Technical Support Center (TSC) and Emergency Operations
Facility (EOF) staff displayed questionable engineering judge-
ment and/or did not recognize or address technical concerns
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(50-443/83-08[9]-01)."
Several issues addressed below were cited as examples. Overall engi-
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neering judgement displayed in both the TSC and EOF was adequate,
however, the following activities were noted to be isolated areas of
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weakness which were intended to be addressed by the licensee.
In
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follow-up subsequent to the exercise with licensee technical support,
operations and emergency preparedness staff, the following additional
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information was provided.
The resolution of each sub-item of
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inspector follow-up iten 88-09-01 is dcscribed individually below.
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(1) "Efforts continued to restore the emergency feedwater pump
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(EFW) af ter a large break LOCA"
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The licensee correctly stated that the EFW pump would be
required to operate to support steam generator cooldown in
the recovery phase and continued repair efforts were pru-
dent. The inspector agrees and determined that the stated
activity did not detract from the overall recovery ef fort,
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nor did it diminish other high priority recovery action in
progross or planned, and that TSC judgments were made with
long-term recovery in mind.
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(2)
"A questionable fix for the containment building spray
(CBS) system"
The inspector met with the Technical Support Manager and a
Technical Support Engineer and discussed the rationale
behind the corrective action taker to rig an alternative
water source for the CBS system.
Although the capability
of the proposed modification to the system to reduce con-
tainment pressure was never proven due to the eventual
repair of a CBS pump, the inspector determined, based on
this additional information, that the engineering judgment
and methodology involved in the proposed system and opera-
ting procedure changes were
acceptable.
The
licensee
actions were appropriate since this fix was considered to
be a "last resort" measure aftcr all prudent and subsequent
extraordinary reasures had failed to provide containment
spray by other means due to additional scenario controller
intervention.
Additionally, the licensee had previously determined that
the composition of the present TSC engineering staff, while
adequate, could be enhanced by providing an augmented staff
roster.
NHY has committed to implement this initiative.
(3) "A lack of effort to locate and isolate the release path"
This apparent lack of effort was the resu,t of licensee
decision;
not
to
pursue
entry
into
the containment
encirsure due to high radiation levels.
Discussion with
the licensee confirmed that indirect measures, such as
remote temperature, pressure and sump level indications,
were taken in a timely fashion to provide an alternate
assessment of potential leakage paths.
The inspector was
unaware of these activities during the drill. The licensee
decision to postpone entry into the containment enclosure
was intentional, based upon other recovery ef forts associ-
ated with depressuring the containment.
Restoration of a
CBS pump was imminent and activation of this system would
have stopped the release.
CBS restoration was
subse-
quently, and repeatedly, delayed by controller intervention
so that the
operators were
prevented
from affecting
repairs.
The
licensee decisions
in
this regard were
appropriate.
(4) "No effort was noted to bisdown ste.m generators (S/G) to
lessen the heat load in containment"
This comment implied that S/G blowdown was appropriate.
The actual concern was that a step in the emergency proced-
ure required the S/G to be depressurized. This step was not
performed because the TSC staff was unsure of the integrity
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of the S/G tubes because no sample was available due to
blowdown system isolation.
This TSC staff concern was
expressed to the inspector when he questioned them during
the exercise.
The NRC position in this area is that
improved guidance to the operator may be warranted and
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should be evaluated, however the decision not to vent or
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blowdown the S/Gs without sampling appears to have been
reasonable and appropriate.
(5) "Neither the E0F or TSC staff questioned a release of
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greater than 7C30 curies per second with only clad damage
and no core uncovery"
The inspector reviewed the player and controller logs for
selected TSC, E0F and engineering support center (ESC)
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staff. These logs revealed that several staff members did
question
and/or comment on
the mismatch between the
reactor coolant activity and the release rate.
Subsequent
discussions with the TSC and EOF controllers and players
also indicated that they were aware of this mismatch.
In
actuality, the ESC staff made very accurate core damage
assessments based upon the data supplied by the TSC.
The
E0F dose assessment staff made accurate dose projections
based upon the release rate, as well as correlation of
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field data to the release rate. A review of previous drill
comments, as well as the player instruction for this exer-
cise, indicated that this level of activity is recognized
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'o be an unrealistic number, which is required to provide
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the offsite dose rates necessary to exercise the entire
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emergency planning zone.
The technical staf f s had repeat-
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edly identified and questioned these mismatches in previous
drills and were told by the controllers that this high
release rate was necessary to test the off-site plans, and
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that they should not challenge the data.
Although NRC review of the specific scenario used for the
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exercise was acceptable, the above described problem indi-
cates that the licencee should place more effort in
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developing exercise scenarios where core damage and release
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rates are consistent.
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With rispect to the above identified weaknesses, the exercise inspec-
tion confirmed that the TSC/ EOF staff possesses adequate capabil-
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ities to protect public health and safety. This open item is con-
sidered closed.
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f.
(Closed) Open Item 88-09-02:
TSC/OSC Multiple Access Points.
This
item indicated that the TSC and Operational Support Center (OSC) have
multiple entrances and exits tt t are not controlled. As a result,
contamination controls were inettective at times as personnel entered
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without frisking and it couldn't be determined if continuous account-
ability was, or could be, maintained.
The TSC has a main entrance where contamination controls and initial
and continuous accountability is established and maintained. The TSC
also has a back entrance which is not locked. Although this entrance
is not normally used, the licensee agrees that it could be used, in
effect bypassing the controls established at the main entrance. The
licensee has agreed to change ER 3.1,
Operations", to control access through this entrance as well as move
the main entrance controls.
The OSC also has multiple entrances.
However, this was a condition
that was artif f:ial to the exercise.
At the time of the exercise,
the radiological control area (RCA) had not been implemented at the
station.
The licensee procedures clearly show that when the RCA is
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implemented there will be only one entrance into the OSC from the
RCA,
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The inspector noted that the licensee established and maintained
habitability throughout the exercise. Althougn some minor contamina-
tion could have occurred in the TSC, it is clear it would have been
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prcmptly recognized and would not have adversely impacted TSC
operations.
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g.
(Closed) Open Item 83-09-02:
Departing Shift Dosimetry.
This item
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indicated that no apparent consideration was given to the departing
first shift to account for possible dose when leaving the plant
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during the release, as they were not given dosiretry.
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A subsequent review of the TSC logs, as well as discussions with TSC
and OSC staf f, indicated that consideration was given to the depart-
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ing shift.
Contamination and radiation surveys were ordered and
taken.
Results indicated all areas were below background.
Because
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of this and the current wind direction, the TSC staff elected to al-
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low the departing shift to exit the site without dosimetry.
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Based upon the above review, this item is closed.
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h.
(Closed)
Open Iten 83-09-04:
Media Center Responses to the Press
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Irlqui rie s.
This
item
concerne'd
the
licensee
representa W e 5
responses to some questions in the Media Center which were not con-
.sidered adequate.
The licensee has agreed that these questions were
not fully answered. Although the answers given were current, they did
not have enough substance.
The licensee has agreed to upgrade the
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training for the Media Center spokesperson, including more informa-
tion on the NRC Incident Response Team capabilities and roles. Addi-
tionally, during a real emergency, federal spokespersons would have
been available to provide clarification as the need arose. This item
is closed.
1.
(Closed)
Unresolved Item 88-02-01:
Acr.umul a to r Isolation Valve
Actuation Logic Questions.
In meetings with licensee operations and
engineering representatives in June and August, 1988, the resident
inspectors discussed
questions
regarding
the
"maintain CLOSE0"
switch, its function and design features.
Licensee personnel ade-
quately addressed the compliance of the current design with Institute
of Electrical and Electronic Engineers (IEEE) Standard 279 and IE
Bulletin No. 80-06 guidance.
Additionally, the inspector reviewed
system test packages for the wiring veri?ication and functional
checks (reference:
general test procedure, GT-E-21) of the subject
valve circuitry to confirm the opening of the accumulator isolation
valves upon receipt of a safety injection signal with the switch in
the "maintain CLOSE" position.
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The licensee stated that the FSAR described a valve capability for
future operational testing which, while currently available, was
prohibited from use by technical specification requirements.
The
inspector evaluated this position and determined that the governing
administrative and LCO controls were adequate to prevent safety prob-
lems during routine operation and shutdown activities. Only specific
plant transitional situations and mode changes (particularly entry
into Mode 3) represent potential problem areas.
It was noted that
the Westinghouse Owners Group is evaluating accident scenarios in
Mode 3 below 1000 psig reactor coolant system (RCS) pressure and in
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Mode 4 on a generic design basis.
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In order to address the inspector's specific concerns regarding the
adequacy of current orocedures/ drawings and of future operational
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controls if technical specification requirements are revised to allow
accumulator isolation valve closure in higher modes for testing in
accordance with FSAR provisions, the licensee implemented the fol-
lowing actions:
(1)
Issued Revision 10 to the "SI-Accumulator Isolation Valves Logic
Diagram", 1-NHY-503907, to delineate the pressure setpoint above
which an alarm is actuated if the valve is not fully open.
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(2)
Initiated resisions to the affected alarm response procedures
to correct the recommended action references relative to the
proper RCS pressure setting at the safety injection (SI) unblock
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pressure.
(3) Recommended revision to the SI system description, SC-NAH/
NCH-284, Foreign Print No. 52005, for the accumulator tank iso-
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lation valves discussing valve closure af ter resatting an SI
signal with the valve controls in a "maintain CLOSE0" positio.
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The inspector reviewed licensee engineering memoranda, including one
issued by the Yankee Atomic Electric Company, Nuclear Services
Division, on the accumulator isolation valve actuation logic and
considered the adequacy of the current Emergency Response Procedures
to the SI valve respense design, including SI signal reset.
No
problems with existing controls were identified.
The inspector determined that th9 questions on the subject system
design and controls nave been adequately addressed and that the
licensee has taken steps to ensure the continued adequacy of design
control if the technical specifications are amended to incorporate
the full accumulator desigr. features discussed in the FSAR.
This
unresolved item is considered closed.
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(0 pen)
Open Item 88-06-01: Non-Class IE Loads Powered from Class 1E
S o u rc_e s. This item was origi ally opened to resolve the is:ue sur;
rounding the tachometer on the emergency feedwater cump (EFW) tur-
bine. Subsequently the NRC concern has been expanded to include the
entire program for design, identification and testing of non-class 1E
loads powered off of class IE sources.
(1) Background
NRC:RI Inspection Report 50-443/88-06 described a non-class 1E
circuit (EFW tachometer) which was not included in the NHY
Technical Requirements Manual (NYTR) list of devices to be
tested per technical specifications (T.S.).
The T.S. involved in this issue consists of two parts which deal
with containment penetration conductor overcurrent protective
devices and protective devices for class 1E power sources con-
nected to non-class 1E circuits.
This discussion concerns only
the class IE power s;urces connected to non-class 1E circuits.
This specification states that each protective device for class
IE power sources connected to non-class 1E circuits shall be
,
operable in Modes 1-6,
With one or more of the protective devices inoperable, the cir-
cuit mv;* be de-energized by tripping the circuit breaker or
racking out ue *emoving the inoperable device within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,
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In addition, the above status must be verified every seven days
thereafter.
The surveillance requirements necessary to declare
operability include periodic testing, inspection and preventive
maintenance of the device. The list of protective devices to be
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tested per T.S. Surveillance Requirement 4.3.4.2 were incorpor-
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ated into NYTR Table 16.3-10 (Technical Requiremer.t 15) under
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the T.S. Improvement Program.
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The NHY Systems Support Department Manager reported on May 2,
1988 that his review of the circuit indicated that the tach-
ometer for the turbine-driven emergency feedwater purrp was a
non-class 1E load connected to safety-related bus E5 via 120 vac
motor control center E515 distribution panel F.3E, circuit 4.
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Request for engineering services (RES)88-226 was w-itten on
May 6, 1988 to determi.ie wnether this circuit should be included
in Table 16.3-10 of the NYTR.
A station information report
(SIR) was initiated on July 26, 1988 to document this situation
and further clarify the reporting requirements.
Licensee event
report
(LER)88-002 and
its
supplement riocument previous
instances where other ncn-class 1E circuits were omitted from
Table 16.3-10 of the NYTR.
Additional NRC inspection of this
previous LER may be found in NRC:R1 Inspection Reports 50-443/
88-06, paragraph Sc and 50-443/88-07, paragraph 5.
Licensee evaluation of this issue was conducted as an SIR fol-
low-up.
Engineerin3 review of calculation 9763-3-E0-00-46-F,
"Failure of non-class 1E Loads on class 1E Buses" revealed
several additional loads requiring immediate resolution to en-
sure compliance with the T.S.
As of the end of this reporting
period temporary modifications had been made te nearly all of
those circuits and a permanent design change is in progress.
(2) Chronology
January 1988
Licensee review indicates that the supply breaker
to inverter 28 off of unit substation E51 is not
on the list in the NYTR.
February 1988 Following evaluation of preoperational testing
previously conducted on the breaker, it is deter-
mined that the breaker must be tested.
It fails
the test, is repaired and the system is restored
to operable status.
March 1988
LER 88-002 is submitted indicating that a review
of all unit substations reveals that the above
finding is an isolated case.
April 1988
ine inspector providos a copy of a January,1988
daily report frcm another nuclear facility about
the power supply to the auxiliary feedwater pump
tachometer which is similar to the above finding.
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May 1988
Request for engineeriig review of Seabrook EFW
pump turbine tachometer is issued by NHY (RES88-226). The licensee determines that the EFW
pump tachometer is not class 1E. The tachometer
circuit is not disconnected electrically from its
1E power source as required by the T.S. action
statement.
Licensee discovers the breakers between 2 pairs
of unit substations are also not on NYTR list.
Substation tie breakers are added to list. Sup-
plement I to LER 88-002 issued.
July 1988
Licensee review of the relevant engineering cal-
culation determines that two separate problems
exist:
(1) Coo-dination of the tie breakers in the unit
substations
(2) EPd tachometer circuit
Circuit breaker for EPd pump is opened per T.S.
after discussion with the inspector.
August 1938
Continued review of calculations indicate that
trains "A" and "B" have additional circuits which
are not analyzed and are required to be discon-
nected per
T.S.
initiated so as to be completed prior to expira-
tion of the 72-hour LCO.
A permanent design
change is in progress.
(3)
Inspection
The inspector held frequent discussions with the Technical Sup-
port Vanager and Lead Technical Support Electrical Engineer con-
cerniog progress of the analysis and installation of the tempor-
aiv.todifications. A licensee event report will be submitted.
Prwiiminary fMC review of the train "B" temporary modifications
revealed no concerns.
(4) Findings
Based on the above, the following issues remain unresolved:
(a) Adequacy of the original determination of which components
were to be incorporated into the NYTR lis _ - _ _
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(b) Licensee actions taken upon discovery of the non-class IE
EFW tachometer powered from a class 1E bus.
(c) Reportability of the above findings in accordance with 10 CFR 50.73.
An additional question that must be resolved concerning the NYTR
is whether non-class IE loads which meet seismic design criteria
may be omitted from the NYTR listing.
Licensee and NRC activ-
ities are ongoing and will be the subject of continuing evalua-
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tion.
This item under expanded scope remains open.
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k.
(Closed) Violation 88-06-02: Emergency _ Diesel Generator (EDG) Failure
!
Reporting
(1) Background. NRC:RI Inspection Report 50-443/88-06 described a
-
t?ip of the train "B" emergency diesel generator which occurred
"
on February 24, 1988. Open Item 88-06-02 was written to docu-
ment NRC questions related to the reportability of this failere.
Based upon the NRC questions, NHY conducted a comprehensive
,
review of the diesel generator logs and determined that seven
i
j
failures had occurred since issuance of the zero power license
in October 1996. The failures were analyzed and summarized in a
letter to the NRC (NYN-89102) dated July 22, 1938. The informa-
tional requirements of T.S. 4.8.1.1.3 were addressed for the
most recent failure on February 24, 19S3. Additionally, the six
previous failures were reported to bring the record up to date.
(2)
,ivirement.
The above T.S.
is applicable in Modes 5 and 6.
>
survetilance Requirement 4.8.1.2 states that the required ac
electrical power sources shall be demonstrated operable by per-
formance of Specification 4.8.1.1.3.
This surveillance specifi-
,
cation
states that all diesel generttor failures shall be
reported to the Commission in a Special Report within 30 days.
,
i
(3) Findin2s.
None of the above f ailures were reported within the
16-day time frame required by T.S. 4.8.1.1.3 and this failure to
,
report
constitutes
a
violation
of
the Saabrook Technical
,
$pecifications (SE-06-02).
'
(4) Licensee Corrective Actions.
Licensee corrective actions as a
result of this violation aid actior.5 to prevent recurrence were
~
provided to the NRC in letter NYN-83102.
hHY reporting proced-
,
.
'
ures have been revised to address EDG f ailures.
The station
information reporting system will be utilized to ensure that
appropriate post failure actions are taken.
.
Based upon the above and appropriate licensee actions initiated on
two recent diesel f ailures, the inspector considers this issue closed
and no additional resronse is required.
!
t
- -. --
,
--,- --
.
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5.
Licensee Reports
a.
(Closed)
Construr: tion Deficiency Report (COR) 86-00-09:
Veritrak/
Tobar Transmitters. NRC:RI inspection reports 50-443/87-24 and 88-06
both document the progress made in the installation of Rosemount
transmitters to c.orrect this deficiency.
Design coordination report
(OCR)86-340 was implemented to control the rework and complete the
corrective action documented in the final
10 CFR 50.55(e) report to
the NRC.
During this inspection, the inspector examined the completed field
installation of all 23 Rosemount transmitters in the Unit 1 contain-
ment building. The rework associated with change authorization No. 7
to DCR 86-349 was checked and specific installation details (e.g.,
compression fittings) were examined.
The inspector also noted that
the installed components were Rosemount Model 1154 transmitters, dif-
ferent from the Model 1153 transmitters that have exhibited manuf ac-
turing deficiencies at other nuclear power plants.
The inspector reviewed the DCR for calculations affecting instrument
setpoints and determined that certain technical specification tabular
data and limiting condition for operation setpoints require revision.
' he iiconsee submitted letters to the NRC dated May 27, July 8 and
,
August 0,
1938 (NYN-83075, NYN-88091, and NYN-88109 respectively),
which discuss the methodology used in the Rosemount setpoint analysis
and transmit the proposed techr.ical specification changes and a sup-
plemental
analysis of the relevant safety considerations.
The
inspector reviewed these documents, noting consistency with the
Westinghouse setpoint methodology (also discussed in NRC:RI inspec-
tion report 50-443/87-24) and with the values calculated in DCR
86-349.
The inspector's review of the proposed technical specifica-
tion revision > vere discussed with NRR project and technical reviewer
personnel.
The inspector confirmed that system operability considerations will
be adequately controlled by the proposed technical specification
changes, that a license amendment has been requested and is being
processed, and that the licensee has completed all corrective actions
relevant to its final 10 CFR 50.55(e) report. Adequate consideration
of the level measurement error due to reference leg heatup for the
steam generator level reactor trip and emergency feedsater actuation
setpoints was also verified to have been included in the Rosemount
data calculations. A licensee request (NYN-88082) dated June 9,1938,
regarding the need for operator action in response to level measure-
ment errors also has been transmitted to NRR for review.
All corrective measures commitments have been completed and no fur-
ther action is req"ired cf the licensee at this time.
This CCR is
considered close _ - _ -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_
___
_ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
b.
(Closed)
10 CFR 21 Report (87-88-04):
Gould Relay Failures.
The
<
failure of Seabrook-specific modified Telemecanique J-10 relays in
April and August, 198'/ resulted in a licensee investigation into the
number and use of relays installed at Seabrook Station. NHY engi-
neering evaluation 88-001, "J-10 Relay System Evaluation", concluded
that plant operation with the defective relays in service was accept-
able during Modes 5-6, but was unacceptable during Modes 1-4.
Of the 112 J-10 relays which were found to be in service in the
!
plant, 57 were installed in safety-related applications. These were
replaced in accordance with DCR-87-390.
.
Because of the unique voltage requirements specified for the original
!
relays, Telemecanique was unable to ensure a qualified 4r y3ar opera-
i
tional design life for the replacement relays. Analysis showed that
i
a design life of only 4.3 years could be guaranteed. This reduction
'
in design life resulted in the generation of maintenance procedure
j
MS0514.17, "Telemecanique J-10 Relay Magnet Block Replacement". This
procedure provides the instructions necessary to change out all
j
safety-related J-10 relays prior to the end of their design life.
'
To verify that these changes were made, the inspector conducted a
,
field walkdown of selected replaced relays with the cognizant tech-
nical support engineer.
This sampling included the following relays:
,
System
Relav
Work Package
i
CBA
E42/9a-3-3
87W003095
CBA
E42/9a-3-4
87W003096
RYY-2192-1L, 2L, 3L
87 WOO 3132, 8133, 8134
RYY-2292-1L, 2L, 3L
87W003135, 8136, 8137
,
EAH
E3E/3-R1
$7 WOO 3112
.
l
EAH
E3F/Sa-R2
87 WOO 3113
'
RBC7a
87 WOO 3114
All of the above listed relays were verified to have been replaced.
A document review of the above listed work packages was performed.
!
No discrepancies were identified.
The inspector has nc further
!
questions in this area and considers this item to be closed.
,
f
C.
(Closed)
10 CFR 21 Report (87-88-03):
Service '(ate _r_ System Valve
Liners and Seats.
A generic problem was icentified with the cil-
covery in May, 1987 of the premature deterioration of the liner / seats
of certain butterfly valves supplied by Fischer Contiels.
The sub-
!
ject valves, installed in the service water system, had been modified
l
previously as corrective action in.:ccrdance with a 10 CFR 50.55(e)
report (85-00-13) in which liner detachment problems wera noted. The
i
i
root cause of the most recent deterioration problem was attributed to
l
inadequacies in the modifi d seat design and in the elastomer liner
e
bonding process applied to correct the original detachment proble ________________
,
This issue was first opened in NRC:RI inspection report 50-443/87-13
and was reviewed by an NRC:RI specialist inspector, as discussed in
report 50-443/87-18.
The licensee submitted a 10 CFR 21 report
(NYN-87091) to Region I on July 28, 1987. The inspector reviewed the
,
licensee's "Summary Report on Service Water System Valves", dated
July 29,1987, noting discussion of both short term and long term
corrective action programs.
With respect to the short term, NRC
inspectors, over the past year, have witnessed licensee implemanta-
tion of a repair and test program for the subject valves.
Twenty-
eight valves were modified with an improved valve liner / seat design
,
t
which has increased the liner thickness to preclude deterioration
(reference: DCR 87-249). Also, the instClation of design modifica-
tions (D R's87-315 and 87-401) to the piping downstream of certain
of the valves was inspected.
These changes alloweet for the subject
valves, previously utiliied in throttling applications, to be posi-
tioned either fully opened or closed, thus reducing the potential for
i
!
future deterioration.
By July, 1988, all the design changes asso-
ciated with the servica water valve rework and system redesign had
been completed.
i
Longer term corrsctive action consists primarily of a monitoring
,
program to ensure that short term corrective action has been effec-
l
tive.
Tne licensee plans to conduct an inspection of four of the
modified valves, including two that were changed from a throttling
application, during the first refueling outage. The inspector verif-
ied that this activity has been formally noted in the licensee's
integrated ccmi tment t ra c k i.9 g system (action no. RED 2082).
The
l
l
inspector also reviewed scheduled raintenance data sheets which pre-
'
i
scribe the insp2ction of two additional codified valves for seal /
liner damage.
Such checks will occur each time the servics water
strain 1rs in proximity to the valves are removed for cleanirg, at a
-
frw uency of about every two months or whenever differential pressure
indications dictate.
Also the licensee has fabricated test coupons
of the modified elasto er liner material burded to valve-like metal.
l
These test coupons have been immersed in the circulating nater pump
house basin to ecnitor the effect of seawater on both the elastomer
and the bonding process.
The inspector examined two work requests
describing the removal Of the test coupons to be conduc' ed in the
latter part of 1953 for transmittal to the elastomer manufacturer,
,
Belzona Molecular Laboratory, for pull testing.
'
The insepctor noted that both the. nrt and long term corrective
!
actions taken or planned by the licensee in response to this design
l
,
'
)
deficiency were consistent with the 10 CFR 21 report submitted to the
NRC and with the discussion of the deficiency documented in NRC:RI
inspection report 50-443/87-18.
Short term corrective actions have
been co pleted and icng term corrective actions are scheduled and
j
being tracked.
The inspector has no further questions at this tire
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- - -
_____
._
_-
-
-
- -
. _
--_
, _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
__ ______
,
.
.
with respect to the licensee evaluation of the problem, the testing
conducted tv effect a workable design solution, the actual repairs or
the plans for future monitoring of the valves to check for liner
deterioration. The licensee's overall approach to this problem from
a technical standpoint has been methodical and comprehensive. The NRC
has been kept informed of new developments and licensee plans.
This
10 CFR 21 Report is considered closed.
d.
Station Information Resorts.
Licensee station information reports
(ilR) are used to internalTy report and evaluate operational events
that may require further investigation, notification to a regulatory
agency or require root cause analysis. Licensee Event Reports and 10 CFR 21 reports normally originate with an SIR. The reports discussed
below were reviewed for compliance with the implementing instruction.
Supervisory, regulatory. services, r anagement and SORC reviews were
verified. Also examined were the technical evaluation of each event,
root cause analysis and recommendation.
(1)
>IR 88-01_0: On January 15, 1988 the train "A" amergency diesel
generator (EDG) was unloaded and shutdown during a post mainten-
ante test because of a lif ting relief valve in the auxiliary
cooling water system.
As a result of this SIR several minor
design changes were instituted to improve engine reliability and
performance.
Tne inspectors discussed these modifications with
the Systems Support Manager and the cognizant Lcad Systems
Engineer.
(2)
SIR 88-054:
This SIR was initiated to investigate the root
cause of a mispositioned circuit breaker in the service water
system.
The licensee evaluation revealed minor administrative
work control defic;encies and some human factors improvements
which should be made in the labeling of the af fected motor con-
trol centers.
6,
NRC Bulletins and Information Notices
a.
(Closed) NRC Bulletin 87-02, Supplements 1 and 2:
Fastener Testing
to Determine Conformance with Applicable Material Speci ficati on s.
As documented in NRCTRI inspection report 50-443/37-26. Bul~1etin
~
87-02 was closed based upon the conduct of testing and submittal of
test results by the licenset to the NRC. The inspector assessed all
the actions taken by the licensee in response to this bulletin and
determined that they were both complete and adequat. _ _.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ - _
. _ _ _ _ _ _ _ _ _ _ _ _ _
.
i
.
i
Subsequently, the NRC issued Supplements 1 and 2 to NRC Bulletin 87-02, requesting, and then clarifying the request for, additional
information on the suppliers and nianufacturers from which the subject
fasteners may have been purchased.
On July 21, 1983, the licensee
i
responded to the supplemental requests by letter (NYN-88099) to the
NRC.
Enclosed with the letter were a list of approved vendors who
supplied or may have supplied ferrous fasteners suitable for safety-
related applications and a list of vendors who supplied commercial
I
grade f asteners.
The licensee response also discussed the basis for
compilation of the lists and a committrent to notify the NRC of any
additional suppliers or manufacturers identified by on going procure-
ment record reviews.
The inspector reviewed the information submitted in response to Sup-
j
plements 1 and 2 to NRC Bulletin 87-02.
No questions or concerns
regarding this submittal were identified.
This bulletin remains
,
closed for inspection purposes,
b.
(Closed) NRC Bulletin 83-05, with Supplements 1 and 2: knconform-
ino Materials Su
and West Mrsey_pplied by _Pipino_ Supplies._ Inc. at Folsom, New Jersey
Manufacturin_g g mpany at Willianstown, New Jersey.
NHY responded to NhC BuiTe~ tin 88-05_byletter (NYN-88114) on August
'
25, 1933.
This letter included the detailed results of the licensee
effort to determine the impact of suspect materials at Seabroek. The
NHY program consisted of the following:
Identification of af fected materials in safety related systems
-
I
Verifying
acceptability
of
installed
materials
-
Reporting to the NRC in accordance with the requirements of the
-
l
bulletin
A total of 369 flanges ard fittings were identified in safety related
j
i
systems.
A test program was developed to measure the hardness of
carbon steel items and ferite content in stainless steel items.
Licensee representatives participated in an Electric Power Research
j
Institute workshop on the use of the Equotip test equipment.
o.lity control (QC) inspectors performed the fiold testing of each
.
flange and fitting. The data sheets were evaluated by the cognizant
J
quality assurance (QA) engireer.
Om July 15, 1988 in the service
)
water cooling tower, the inspe: tor observad field hardness tasting of
the service water system flanges.
The testing was corducted in ac-
!
cordance with procedure NHY-EHT-1, "Equotip Hardness Testing" (Revis-
ion 01, Change 01).
The inspector reviewed the procedure and work
request SSW3339 and verified that licensee OC personnel were know-
ledgeable concerning both the procedure and test equipment.
l
.
I
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
_ _ _ _
.
i
t
'
,
Independent measurements were also performed on separate pieces of
!
'
suspect material by J.
Dirats and Co. and Bechtel Corporation to
,
confirm the Equotip test results. Additionally, test results were
'
sent to the Nuclear Mant.gement and Resources Council (NUMARC) for
generic industry data compilation and analysis. Of the 369 flanges
and fitting tested at Seabrook, 30 were found to be below the minimum
Brinell hardness value of 137.
This is the minimum value specified
j
in the American Society of Mechanical Engineers (ASME) material
'
specification SA-105.
The 30 fittings were individually evaluated
,
and found to exceed existing tensile strength requirements in accord-
ance with the ASME code.
The evaluation demonstrated the inherent
'
l
conservatism of the code as well as the correlation between hardness
l
and tensile strength.
NHY made seven calls to the NRC Operations
i
,
Center over the course of the testing as required by the bulletin.
!
I
These non-emergency notifications were part of the 48-hour reporting
requirements that were subsequently discontinued by the issuance of
i
Supplement 2 to the bulletin,
r
Throughout the course of the test process, the inspector maintained
close liaison with licensee OA/0C inspectors, engineers and managers.
The methodology employed in identifying, testing and analyzing the
i
suspect fittings was labor intensive.
The licensee aevoted adequate
!
researces to ensure timely completion.
The two shift testing sched-
!
ule was particularly rigorous and the total support of NHY engineer-
l
ing and quality assurance departments were in evidence.
Additional
[
NRC Headqua*ters review of this bulletin may occur as a result of
j
generic evaluation of the PSI /WJM concern.
For inspection purposes,
l
this bulletin is closed.
!
c.
NRC Information Notice 88-46 and Supplement 1:
Licensee Report of
.
I
Defective Refurbished Circuit Breakers. This Information Notice (IN)
!
l
describes discovery by another utility that certain non-safety re-
'
i
lated circuit breakers manufactured by the Square D Company were
i
actually refurbished equipment rather than new stock.
It has been
,
determined that certain suppliers were refurbishing components and
'
'
re-labeling them as new equipment.
The licensee is conducting its
I
own inspection to determine what effect, if any, this IN may have on
l
Seabrook.
During a visit to the facility on August 16,1988, the
i
D',.ector of the NRC Of fice of Nuclear Reactor Regulation discussed
l
this issue with members uf the licensee inventory and material
l
requirements departments.
^
I
t
The inspector will continue to fc' low this issue and its relationship
!
to receipt inspection of comercial grade items as well as any future
p
additional NRC correspondence such as NRC Bulletins or additional IN
,
Su,S 'ements.
For inspection purposes, this is an open item.
t
i
f
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _
__
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,
.
d.
NRC Information Notice 83-25: Minimun Edge Distance for Expansion
Anchor Bolts.
An analysis of site specific data affecting the
capacity factors of Hilti Kwik-Bolts installed at the minimum spec-
ified distance from an unsupported concrete edge revealed safety
factors greater than twice the allowable design loads, This analy-
sis, accomplished by the Yankee Atomic Electric Company (YAEC) for
the Seabrook Project, utilized conservative assumptions based upon
Seabrook design criteria, Kwik-Bolt installation specifications and
concrete compressive strength ti st data, Since no safety concern was
identified, the YAEC recommendation *^ mnnart a Nuclear Management
and Resources Council (NUMARC) initiative for generic industry-wide
action on this issue was adopted,
The inspector noted that a previous NRC unresolved item, 443/
82-03-07, had addressed consideration of the Kwik-Bolt shear cone
interaction, including the influence of the spacing of anchors at
concrete corners,
As documented in NRC:RI inspection report 50-443/
85-25, testing was conducted at the Hilti Test Facility in Tulsa,
Oklahoma to check the reduction in Kwik-Bolt capacities, in part, at
outside corners,
The results of such testing, while indicating a
reduction in ultimate capacity, were acceptable when considered with
respect to the overall expansion anchor design.
The unresolved item
was therefore closed,
The inspector noted that the past testing of the Hilti Kwik-Bolts,
while not accomplishtd specifically to address the 10 CFR 21 concerns
ra' sed in IN 83-25, has confirmed the conservatism of the design, the
acceptability of Seabrook site-specific applications and the assump-
tions made by licensee engineering personnel in calculating design
loading data,
Thus the licensee positions that Kwik-Bolt installa-
tions at Seabrook represent no immediate safety concern and that
future reviews can be adequately handled through NU,tARC appear to be
well founded,
No violations were identified.
This item is closed for inspection
purposes,
e,
IE Information Notice 86-50: Jnadequate Testing to Detect Failures
of Safety Related Pneumatic Component s_ or Systems.
The inspect 3
~
reviewed internal licensee memoranda providing evidence of engineer-
ing review and regulatory cognizance cf the subject information
notice.
The licensee ccntinues to evaluate their methods of air
system and component testing and instrument air quality sampling in
accordance with FSAR commitments.
The inspector confirmed that although no specific action is required
.by this information notice, the licensee appears to be investigating
the applicability of the relevant safety issues and tracking regula-
tory cemnitments and criteria accordingly.
No violations were
identified.
This item is closed for inspection purpose _--
_ - _ _ _ _ _ _ _ - - _ _ _ _ -
_ _ -__ _ _
____ _
__ ______-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _
.
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7.
Maintenance / Surveillance
i
a
i
a.
OX 1456.81:
Operability Test of ISI Valves.
On July 22, 1988 a
re-test of the mot".r operated suction isolation valve to the train
"B" safety injection (SI) pump, CBS-V-53, was performed in accordance
,
with surveillance procedure 0X145631,
"Operability Test of ISI
'
,
Valves". The test was completed under work request 88W2735 and con-
l'
testing (IST) valve stroke time data.
The inspector observed the
sisted of the stroking of the valve to gather the required inservice
test locally at the valve in the residual heat removal vault.
The
results of this test were an opening time of 10.69 seconds and a
closing time of 10.22 seconds. The maximum allowable stroke time was
15 seconds for each direction. No violations were identified,
,
a
!
b.
EX 1804.044:
Safety and Relief Valve Setpoint Pressure Test.
On
j
June D,1931 another nuclear facility reported problems associated
with setting main steam safety valve (MSSV) lift setpoints using
j
When these valves were subsequently lift tested with
j
steam, setpoint drift was noted. The inspector reviewed surveillance
l
procedure EX1804.044, "Safety and Relief Valve Setpoint Pressure
Test" and verified that Seabrook MSSV's are presently tested in place
l
with system pressure 15-25% below valve set pressure.
An assist
)
motor is used to provide the additional test pressure. Therefore the
!
above described problems can not occur at Seabrook.
I
i
c.
EX 1804.016:
Diesel Generator Auxiliary C'ool ant System Quarterly
l
T e_s t. On May 13, 1958 the train "B" emergency diesef ger,erator (EDG)
-
was returned to service following maintenance.
Operability of the
,
l
ED3 is normally verified by four separate surveillance tests; engine
i
start, fuel oil transfer pump performance, cooling water and air
start valve performance and auxiliary coolant performance. An admin-
j
ist"tive error resulted in declaring the EDG operable on May 16,
j
193b prior to completion of the test un the auxiliary cooling system
(EX 1804.016). Station information report (SIR)88-048 was initiated
l
because of this occurrence. The SIR indicated that the root cause of
the problem was inadequate scheduling because of an error in the
j
Specification Appraisal computer program,
The inspector reviewed
i
licensee corrective ar.tions which included adjustment of tr,e program
{
model and had no further questions.
d.
IX 1630.921:
SSPS Train
"A" Actuation ~Looic Test.
On August 19.
j
1988 the inspector witnessed portions of IW0epartment Surveillance
i
Procedure IX 1680.921, SSPS Train "A" Actuation Logic Test. The pur-
i
pose of the test is to functionally test the train
"A" solid state
!
protection system (SSPS) in accordance with technical specification
i
4.3.1.1 and 4.3.2.1.
The inspector witnessed selected steps concern-
i
ing reactor trip breaker operation locilly in the essential switch-
!
gear room. Th3 inspector noted effective communications established
with the control room, the presence of a knowledgeable electrical
quality control inspector and proper control exercised over the pro-
'
cedure by the control room personnel.
No violations were identifie _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
______
_ ___
-
_ _ _ _ _ - _ _ _ _
.
.
e.
EX 1804.015: Diesel Generator 1B 18-Month Operability and Engi-
neered Safeguards Pump and Valve Response time Testing Mode 5 Sur-
~
veillance. This is a seven-event surveillance test which satisfies
several train
"B" Mode 5 technical specification
surveillance re-
quirements. The inspector observed portions of event three and event
'
six.
Event three involved an emergency diesel generator (EDG) start
initiated by resetting the train
"B"
low steamline pressure safety
injection ("S") actuation signal from the main control board.
The
-
inspector witnessed the diesel start to a standby idling condition
and the starting of the train
"B"
(ECCS) pumps as well as feedwater isolation and main steam line iso-
,
lation.
The test was run twice because of high speed recorder prob-
lems which were eventually corrected.
In all
cases the plant
,
responded as designed.
Event six followed the 24-hour run of the
i
train "B" EDG and tested the ability of EDG 18 to start and lead upon
concurrent loss of of f site power and an "S" signal and to verify that
bus E6 s's Js its load.
ECCS pump and valve response times were
obtained and the EDG's ability to accept a cooling to*.<er actuation
("TA") s.gnal while loaded with auto connected loads was also ver-
ified,
following successful service water system.ransfer to the
cooling tower, the EDG's ability to accepe a large load rejection was
'
tested by simultaneous 1-, tripping the cooling tower pump and charging
'
pump.
The inspector noted that the control room operators and test
director were intimately familiar with the procedure and expedit-
iously performed the critical post safety injection steps required by
i
procedure.
The equipment also was verified to properly perform its
intended function. No violations were identified.
.
f.
OX 1406.02:
CBS Puma and Valve Ouarterly Test and 18 Month Remote
I
P~osition Indication. on~JW19, 1988 while perf orming surveiflance
procedure OX 1406.02, "CBS Pump and Valve Quarterly Test and 18
l
Month Remote Position Indication", about 5000 gallons of water was
{
inadvertently transferred from the refueling water storage tank
(RWST) to the reactor coolant system (RCS) via the residual heat
l
removal (RHR) system. The event occurred because valve CBS-V-2, the
train
"A"
RWST to RHR ! solation valve was opened with RH-V-22 and
RH-V-23, the train "A" RCS to RHR suction valves still opened.
The
operator immediately realized that the lineup was incorrect and
'
re-closed CBS-V-2.
i
t
NRC:RI Inspection Report 86-54 (paragraph 4.a) described a previous
similar event which occurred on September 5,1986 and describes the
i
!
design bases for the system.
Also addressed was the standard
i
Westinghouse design for interlocks in these valves and the NHY posi-
tion on how certain design features (alarms) would be added to pre-
vent recurrence of the September 5, 1936 event.
>
,
______.___________ ____ ____ _
__
_ _ _ _ _ _ _ _ _ _ _ _ _ _
!
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The inspector met with the Assistant Operations Manager and discussed
several issues related to this event, The licensee's ongoing correc-
!.
tive actions will be observed during a subsequent NRC inspection.
g.
Residual Heat Removal (RHR1 System
NRC Region I Inspection Report 50-443/87-24 described a discrepancy
,
in the dimensional gap between the train "B"
RHR pump casing and
impeller.
The licensee subsequently disassembled the train "A"
pump and found a similar problem.
The dimensional gaps were found
to be 0.0235 inches and 0.025 inches for the train "B"
and "A"
pumps
respectively.
The manufacturer (Ingersoll-Rand) specifies a dia-
tretrical clearance between 0.030 to 0.036 inches. Both pumps wearing
)
rings were machined within specification and the pumps restored to
i
service,
j
On March 13, 1938 the inspector observed the clearance measurements
made on the Unit 2 RHR pumps.
These cumps were never installed in
.
Unit 2 and were transported from storage to the Unit I turbine
,
J
building for disassembly.
The inspector noted appropriate quality
'
control hold points in the procedure.
Both quality control and
!
maintenance personnel were considered to be knowledgeble f n their
!
tasks.
The Unit 2 clearances as measured were found to be within
specification.
The 'icensee conducted an evaluation of this technical issue pursuant
!
to 10 CFR 21. Engineering evaluation E3-016 concluded tFat given the
"as found" dimensions under design thermal and seismic conditions,
i
!
pump damage would not have occurred and therefore, a substantial
t
safety hazard did not exist.
This condit* a was therefore not
reportable under 10 CFR 21.
The licensee conducted a detai'.ed review of all relevant documents
to determine whether the wearing rings were modified in some way
,
during the construction or startup phases.
The NHY effort consisted
of a review of installation and work records and a review of spare
i
part receipt and inventory records.
'ngersoll-Rand documents indi-
!
cated that the clearances were within.pecification when shipped from
their facility.
Construction and maintenance records revealed no
modifications or replacements were ever performed on the wearing
j
rings.
The cause of the out of tolerance condition could oot be
identified even though the records check was extremely detailed and
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the quality cf the records was found to be acceptable.
The licensee
concluded that all available prudent action had been taken and ttere-
fore considers the issue closed. The inspector discussed the results
]
of the engineering evaluation with the Manager of Engineering an3 the
Lead Mechanical Engineer and had no further questions.
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8.
Design Changes and Modifications
Post Accident Samplin
ments oGOGlRif37?g _ System (PASS).In order to meet the require-
a.
iMI Action Plan Requirements", (Item II.B.3), a
i
]
PASS was installed at Seabrook. During hot functional testing, dif-
i
ficulty was experienced in obtaining consistent sample results be-
cause of 1.9 dequate sample temperature control. As a result, design
coordination report (DCR)88-081 was generated to add an additional
sample cooler to the system. The inspector reviewed DCR 88-081, as
well as its DCR implementatica plan, and made frequent field inspec-
-
tions of work in progress with special emphasis in the piping sup-
'
ports in the primary auxiliary building (PAB). Although the primary
component cooling water lines which cool the new heat exchanger are
not safety related, they are constructed to seismic criteria due to
!
the design requirenents of the PAB.
The inspector had discussions
i
with the Systems Engineering Supervisor concerning the identification
of seismic /non-seismic class breaks in relation to licensee commit-
i
cents documented in NRC:RI Inspection Report 50-443/86-14.
Field
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inspection of piping and pipe supports revealed no violations of NRC
i
requirements.
Completion of pre-operational testing on the PASS
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requires the plant to be hot and is scheduled for accomplishment in
the heatup prior to initial criticality. Actual testing of the PAS $
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will be ths subject of future NRC inspection to close out TMI Item
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II.B.3.
!
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b.
Sacondary._ Component Cooling Water System
(1)
_B_a cig round.
The secondary component cooling water (SCCW) system
provides cooling water to non-safety related secondary loads in
the turbine building.
Typical cooling loads are the air com-
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pressors and condensate pu.mp air and oil coolers.
The system
includes three 50% capacity each contrifugal pumps and two 100%
capacity each large horizontal heat exchangers. The heat ex-
changer shells and tube sheets are clad with 90-10 copper
!
nickel. All other carbon steel inner substances are lined with
!
neoprene.
The tubes are 90-10 copper nickel.
These heat
I
enhangers are cooled by a non-safety related leg of the service
)
water (SW) system.
!
System inspections in 1986 and 1937 revealed significant tube
!
corrosion due to low fluid velocities at low flow.
(2) Licensee Evaluation and Corrective Action.
The N4Y engineering
department prepared engineering evaluation 88-04 in February,
j
1933 which proposed several solutions including installation of
j
low flow heat exchangers for use during low heat load cond4 tion,
j
This would allow the main heat exchangers to be placed in layup
when not in use. Design coordination report (DCR) 8B-033 was
!
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.________ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _
.
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initiated to add two additional low flow heat exchange's to the
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SW/SCCW systems.
The heat exchangers were procured from exist-
,
ing stock as they are the original Unit 2
air removal heat
exhangers. Once the new auxiliary heat exchangers (SCC-E-185A,
i
B) are installed, the main heat exchangers (SCC-E-29A,B) may be
removed and reworked or replaced with the Unit 2 coolers.
!
'
(3)
Inspection.
Despite the fact that this system is not safety
related, this design change is of general NRC interest because
j
of its relationship to heat exchanger degradation in primary
i
systems as well as aeneral workmanship and work control through-
out the plant. Th
inspartor reviewed engineering evaluation
88-04 and DCR 88-088 and maoe frequent inspections of the work-
i
site.
On July 22, 1988, the inspector identified a section of drain
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piping which had been lut off the main SCCW line in preparation
i
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for weldolet installation.
The lina contained valve SCC-V-344
'
and a tubing connection for chemistry corrosion monitoring.
The
above valve was still caution tagged and the tubing fittings
were identifisd as "Temporary Modification #10-Other".
The
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inspector discussed this activity with the shift operators and
Assistant Operations Manager. The inspector stated that removal
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of a caution tagged valve and temporarily nodified assembly
appeared to violate station procedures concerning equipment
.
tagging and temporary modification s.
Saintenance Procedure MA
!
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4.2, Revision 7, "Equipment Tagging and Isolation" states, "No
person shall physically remove any equiprnent that is tagged
'
,
t
"DANGER / CAUTION".
Maintenance Procedure MA
4.3,
Revision 7
"Temporary Modifications" indicates that changes to temporary
modifications be re-routed with appropriate notations, initial-
!
led and dated by all reviewers or a new temporary modification
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be prepared. In light of the nan-safety related nature of this
modification activity, no vioi ltion of NRC regulations existed,
i
however, it is noted that corrective action for violation
'
87-20-01 that occurred in July, 1957, did not prevent recurrence
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of a similar although significantly less serious situation.
It
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15 also noted that anothar related occurrence was reported in
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station information report $7-108 in November,1937,
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(4) Conclusiens.
It appears that additional attention is warranted
in this area especially with respect to temporary modification
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control. These modifications are clearly identified and removal
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or modification requires similar procedural controls as instal-
lation. This area will be +.he subject of continuing NRC inspec-
!
tion with respect to routine plant operations as well as readi-
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ress for initial criticality.
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9.
Allegation Review
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As documented in NRC:RI inspection report 50-443/88-07, a written response
,
on the licensee's investigation by its Employee Allegation Resolution
}
(EAR) program personnel of five separate allegations was requested.
By
!
letter (NYN-88116) dated August 29, 1988, the licensee responded with the
determination that the subject allegations are either inaccurate or relate
'
to issues which were identified and dispositioned through internal quality
'
programs. An enclosure to the licensee letter summarized each concern,
q
its review and the licensee conclusions.
.
The inspector reviewed the above letter, its enclosure and additional EAR
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files and documents relating to the investigation of each allegation. As
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was documented in the 88-07 inspection report, the inspector had pre-
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viously conducted preliminary reviews of each allegation and performed
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both field inspection and records research where appropriate. During this
inspection, the results of the licensee investigation were evaluated not
only with regard to completeness and substantiating avidence, but also
I
with respect to the inspection data independently collected and checked by
j
the NRC.
The following represent the conclusions reached for each of the
five open allegations.
'
(a) U_ncertified pioing material supplied _by Boston Pipe.
The inspector reviewed UE&C audit and nonconformance reports (NCR)
covering the Boston Pipe & Fittings Co. of Cambridge, Massachusetts
-
and the material supplied by this company for Seabrook Station.
At
!
lea,t one of the NCR's documented the receipt of fittings on site
without certification.
Additionally, a Pullman Power Products NCR
,
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was founw to have identified certain refrigeration system and support
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material whicn lacked the appropriate documentation,
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Each case of a nonconforming condition resulting frcm incomplete
]
certification appeared to be properly dispositioned with evidence of
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completed corrective action and reinspection by quality assurance
(QA) persennel.
The inspector also noted that contractor receiving
inspection reports required and recorded document verification anc
traceability of the subject material as a requisite part of the
inspection criteria.
Thus, while the existence of the noted NCR's
indicates that this allegation may have some basis in fact, the
identification and disposition of these problems by the licensee
also indicates that the receipt inspection process was working
effectively. The inspector found no evidence to suggest uncertified
eaterial supplied by Boston Pipe had been installed in the plant.
.
___ - __-_ _____ ___________ _ ______________ _____ ____ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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,
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(b) Uncertified electrical equipment supplied by Massachusetts Gas and
Electric.
The
inspector checked
a
sample of
purchase
orders
from
the
Massachusetts Gas & Electric Light Supply Company, noting that most
'
wire and circuit breakers were procured for general jobsite temporary
,
power and lighting.
Despite the nonsafety-related use of such mate-
,
rial, at least one NCR was issued to document the lack of proper
'
material certification.
The inspector also noted that both UE&C and
Fischbar.h, the electrical installation contractor, conducted receiv-
ing inspections which required document checks for certificates of
compliance of the inspected material in accorcance with specification
requirements.
As similarly discussed with allegation (a) above, the far.t that the
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licensee quality programs require receipt inspection checks for pro-
!
per material certification and that NCR's have been issued when com-
plete documentation was not available provides one measure of con-
firmation that the material installed meets fabrication specifica-
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tions. Even in the case of a nonsafety supplier like Massachusetts
Gas and Electric, evidence of such QA checks are available in licen-
,
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see records.
The regulatory requirements governing certificates of
'
compliance, versus material certifications like mill test reports,
i
are not in conflict with the licensee position that the manufacturer
[
provides the requisite certifying documentation.
The inspector identified no information or facts that indicated that
the Massachusetts Gas and Electric Light Supply Company had impro-
,
perly certified material or that electrical components had been
,
installed in the plant in applications for which they were unqual-
if;ed.
]
(c) Acceptable level installation of the reactor coelan_t_puyp_s.
The inspector reviewed Westinghouse and contractor records which
substantiated the licensee conclusion documented in the NYN-88116
,
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letter to the NRC.
The Westinghouse Nuclear Service Division
'
"Procedure for Setting of Major NSSS Components", Revision 2, issued
in February, 1979, delineates the level criteria for the reactor
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coolant pumps.
The inspector checked the Pullman-Higgins installa-
tion records for two reactor coolant pumps (RCP), including RCP-lC
which represented the component originally questioned in the tech-
)
nical concern addressed in NRC:RI inspection report 50-443/87-07
(reference: UE&C engineering change authori:ation 03/1557A).
For
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each pu p,
the inspector examined the "RCP-Volute Level Data Sheet -
'
After Adjustment" and independently calculated the maximum level
deviation. Although RCP-lC was slightly more of f-level than RCP-ID,
both pu.?ps were measured to be level within the Westinghouse accept-
ante criteria.
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Furthermore, the
inspector noted that a Westinghouse memorandum
issued in March, 1982 acknowledged the adjustment that was made to
the RCP support and the resulting change in tne RCP volute main
flange differential elevation.
Westinghouse engineers approved the
change at that time.
The inspector reviewed additional evaluation
of the RCP level concerns by the licensee corporate engineering
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staff to include recent Westinghouse studies on RCP "tilt" condi-
tions.
These newer studies appear to indicate that the original
Westinghouse level criteria, which the Seabrook RCP's meet, are
conservative.
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,
Therefore, with regard the question raised by this allegation, the
inspector confirmed that the reactor coolant pumps have been instal-
led and inspected to the Westinghouse design criteria and that
,
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acceptable level conditions for each RCP were verified af ter imple-
I
mentatien of the engineering change which resulted in the reposition-
)
ing of the base of one support.
(d) Weldolet in the emerynty feedwater (EFW) pump room with wrong taper
_
.
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and counterfeit identification number.
j
Visual inspection of weldolets in the EFW pump room by an NRC
i
inspector revealed no deficient or nonconferming conditions.
Tne
'
inspector also reviewed licensee nuclear quality group evaluations
of elbolets and weldolets in the EFW pump room to ensure American
.'
Society of Mechanical Engineers (ASME) code compliance, acceptable
'
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markings and traceability and weld quality and taper.
The licensee
'
evaluation included documentation reviews, visual and ultrasonic
thickness examinations, and inspection tracing of the scribed field
-
,
marks to vendor documents which verify the quality and further
'
traceability of th3 installed components.
The licensee evaluation
concluded that ASME code compliance had been confirn ed.
The inspector checked the licensee's Thickness Data Sheet resulting
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from the ultrasonic testing field examinations and reviewed a sample
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of Dravo pipe fabrication sketches, establishing traceability of
'
weldolet/elbolet field scribe marks to the heat number codes docu-
mented in the manufacturers' mill test reports.
!
The acceptability of field conditions for a number of components,
'
which might represent the subject of the stated allegation, was
!
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verified by independent NRC and li:ensee inspections.
The inspector
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concluded that this allegation could not be substantiated.
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__ __ -_- _ _ ________- _ _ _ _-____ - - __________ _____ -___.
_ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _
.
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(e) Qualification of an Authorized Nuclear Inspector (ANI) trainee
i
The inspector reviewed EAR records documenting licensee investigation
of an allegation regarding the qualification of an ANI trainee and
- , authority to conduct independent inspections.
As discussed in
..
NRC:RI Inspection report 50-443/88-07, NRC inspection of a similar
concern resulted in substantiation of certain of the facts, but in a
conclusion that neither a noncompliance with the ASME Code, nor
evidence of wrongdoing was identified.
The EAR records confirmed that the allegation previously reviewed by
the licensee involved the same ANI trainee that was the subject of
the allegation raised to the NRC.
The licensee
investigation
concluded that during the period of time from May to December,1935
when the subject ANI trainee was assigned to Seabrook, he performed
assignments in accordance with his assigned training program.
NRC
inspector review of documents dating back to the 1985 time frame
veri fied that qualified ANI's had evaluated and monitored the ANI
trainee's training, progress and inspection work..
While the facts surrounding this allegation may be true, both NRC
and licensee reviews of the stated concerns have identified no
impropriety with respect to the certification or conduct of work on
the subject ANI trainee while at Seabrook Station.
The five allegations listed as open in NRC:RI inspection report 50-443/
88-07 were addressed by the licensee in the response letter, NYN-83116.
Independent NRC inspection of these issues prior to raising the questions
with the licensee had identified no hardware problems or quality concerns.
Subsequent licensee EAR investigation of the allegations concluded that
the allegations had no substantive merit.
This inspection has included a
review of those EAR investigation results and the process by which they
were achieved. The inspector verified that licensee actions were compre-
hensive relative to the information provided in the allegations.
The
allegations generally either could not be substantiated, or represented
issues with some factual basis, but with no adverse safety impact.
These five allegation issues are considered closed.
10.
Training
a.
General Empoloyee Training
NRC:RI Inspection Report 50-443/S7-16 discussed the topic of cheating
on general employee training (GET) exams and the lack of written
policy on cheating.
During this inspection period this issue was
re visited.
The inspector reviewed the GET examiration cover sheet
which listed instructions to be read aloud by the instructor prior to
,
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the examination. These instructions specifically addressed the steps
to be taken should suspected cheating occur.
Additionally, the
f
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inspector reviewed the draft of training procedure NT-7010. "Examina-
!
tion Administration and Integrity" which also formalized the station
>
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policy on cheating. The inspector determined that licencee follow-up
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actions
this
issue
have been
appropriate and had no further
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questions.
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b.
Ojerator TraininJ
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On July 20,193S, the inspector discussed the recent Nuclear Manage-
f
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rnent and Resources Council rneeting on operator requalification ttst-
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ing, and the status of Institute of Nuclear power Operations (INPO)
accreditation with the Training Manager. In the area of INPO accred-
!
!
itation, the licensee stated that an INPO programmatic inspection is
!
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due to be performed in November of this year.
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11.
Electrical Confiouration Control
i
s
.
As documented in NRC:RI inspection report 50-443/8S-06, several engineer-
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ing discrepancies and configuration control problems identified in the
-
electrical area were resolved with the issuance of licensee engineering
!
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evaluation 88-011.
NRC open item 87-24-01 was therefore closed.
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During this inspection, the inspector identifiec certain field conditions
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for which questions of electrical detail and adequacy were raised. Spec-
}
ifically, electrical fire wrap requirements in ar.cordance with engineering
i
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change authorization 03/11295G. the protection of spared cable termina-
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tiens, the conformance of Sf6 switching station breaker alignment to the
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plant technical specifications, and the status of missing condolet covers
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were all checked and found to be either acceptable or under work request
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control.
Additionsily, the inspector reviewed a quality assurance (CA)
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assessment (reference: CAIR E8-0597) of electrical design changes where
'
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the potenLial for interface prCblems from engineering to constructien to
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startup/ operational control appeared to be high. Only minor discrepancies
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were identified as a result of this assessment.
!
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Another QA surveillance report 87-00%3 was reviewed with regard to the
ieplementation of work request activities in the cannibal 1:ation or Unit 2
'
!
equipment and spare part components, including electrical itens.
The
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Station procurement and Materials Manual (Chapter 5.5) delineates criteria
!
for the control and docutent tracking of the cannibalication crociss. lhe
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subject surveillance activity resulted in no adverse findings.
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With respect to the licensee's programs of control for electrical work
,
activities and its efforts to ensure ele:trical field configurations meet
l
design requirements, the inspector noted comprehenshe QA/QC cepartment
{
i nv ol v e'ne n t.
Based upon internal licensee assessments and NRC inspector
)
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spot-check and review, no generic prcblems or violations were identified,
.
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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12. Management Meetings _
On August 17, 1988 a meeting was held in King of Prussia, Pennsylvania
with NHY senior managers at the request of the NRC.
The purpose of the
meeting was to discuss licensee plans for heatup, initial criticality and
'
low power testing.
In addition, the current status of NRC Bulletin 88-05
!
was prosented.
Both parties agreed to meet again prior to Initial
!
criticality.
A copy of the meeting handouts and attendance sheet is
!
appended to this report as Attachments A and B, respectively.
j
At periodic intervals during the course of this inspection, meetings were
held with plant managment to discuss the scope and findings of this
-
inspection. An exit meeting was conducted on September 9, 1988 to discuss
the inspection findings during the period. An additional meeting as held
'
on September 22, 1988 between the Assistant Station Manager and the Senior
Resident inspector to discuss item status not covered in the previous exit
meeting. During this inspection, the NRC inspector received no comments
.
from the licensee that any of their inspect'on items or issues contained
proprietary information. No written material was provided to the licensee
i
curing
this
inspection other than
a
listing
of minor
inspection
i
deficiencies summarized in paragraph 3.a of this report.
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ATTACHMENT A
NHY/NRC MEETING ON AUGUST 17, 1933
NRC REG'ON I, KING OF PRUSSIA, PENNSYLVANIA
Name
Title
Organi z ajt_icLn
i
W. Russell
Regional Administrator
NRC/RI
W. Kane
Director, Division of Reactor Projects
NRC/RI
W. Johnston
Director (Acting), Division Reactor
NRC/RI
Safety
J. Wiggins
Chief, Projects Branch 3
NRC/RI
R. Gallo
Chief, Operations Branch
NRC/RI
,
0. Haverkamp
Chief, Reactor Projects Section 3C
NRC/RI
M. Shanbaky
Chief, Radiation Safety Section
NRC/RI
a
A. Cerne
Senior Resident Inspector
NRC/R1
'
O. Ruscitto
Resident Inspector
NRC/RI
D. Brinkean
Project Manager
NRC/NRR
'
R. Wessman
Director, Project Directorate I-3
NRC/NRR
y
F. Brown
President
G. Thomas
Vice President, Nuclear Production
T. Feigenbaut
Vice President, Engineering, Licensing
and Quality Programs
0. Moody
Station Manager
J. Vargas
Manager of Engineering
J. Warnock
Nuclear Quality Manager
R. Sweeney
Washington Lt +nsing Representative
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ATTACHMENT B
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New Hampshire Yan<ee
)
Presentation
!
to
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JSNRC Region 1
,
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08-17-88
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_ - _ _
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AGENDA
.
Introduction
G. S. Thomas
,
.
NHY Organization
E. A. Brown
,
J
Low Power Test Program
G. S. Thomas
,
Self-Assessment
T. C. Felgenbaum
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Status of Bulletin 88-05
J. J. Warneck
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,
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G. S. Thomas
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NEW HAMPSilRE YAN (E E
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E. A. Brown
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NEW HAMPSHIRE YANKEE ORGANIZATION
Posident & CEO
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President & CEO
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Vce President
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Comptroller
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Emergency Pfarnry
Nuclear Production
Engmeering. Lhasirg
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Independent Resnew Team (IRT)
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0W POWER TEST PROGRAM
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LOW POWER TEST PROGRAM
STARTUP ORG.ANIZATION
Startie
Manager
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Shift Test
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3 Test Directas
3 Startup Engineers
3 Startup Engineers
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LOW POWER TEST PROGRAM
STARTUP ORGANIZATION
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Shift test directors and test directors (directors of test activities)
will be qualified in accordance with the requirements of
Reg Guide 1.8 as specified in the FSAR.
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All shift test directors and test directors have previously worked
in the Seabrook preoperational and startup test programs.
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Test personnel will be formed from the following organizations:
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- Technical Support
-- Engineering
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- Operator Training
- Regulatory Services
- Yankee Atomic Electric Company
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- Westinghouse Electric Company
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TECHNICAL SPECIFICATION
SURVEILLANCE TESTS
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All local leak rate tests (Type B & C) have been reperformed
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Emergency diesel generator and engineered safety features actuation
testing scheduled for the last two weeks in August
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Other surveillance testing has been incorporated into the schedule
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PREVENTIVE MAINTENANCE
Data Date 8/2/88
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ACTIVITIES PERFORMED
DEPARTMENT
1987
1988
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Mechanical
1868
1144
Electrical
2834
1558
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2634
1435
Utilities
536
207
TOTAL
7872
4344
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MAN-HOURS CONSUMED
1987
1988
TOTAL
47232
20267
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PREVENTIVE MAINTENANCE
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Data Date 8/2/88
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1987
1988
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Activity
56 %
51 %
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32 %
22 %
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RADIATION PROTECTION
TIME PRIOR TO
CRITICALITY
ACTIVITY
4 weeks
Start reissue of dosimetry to qualified rad workers
1 week
Establish Radiological Control Area for training
Just prior
Establish full Radiological Controlled Area (RCA)
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Just prior
implement full radiation protection program
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OPERATIONS
Licensed Operators
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Staff Licenses
SRO-Operations
9 SRO-Training
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Includes 16 STA - Qualified Operators
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ON-SITE EMERGENCY RESPONSE ORGANIZATION (ERO)
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Fully staffed and trained
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Demonstrated during 1986,1987 and 1988 Graded Exercises
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Fully implemented since receipt of Zero Power License
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Meets requirements of proposed change to 10CFR 50.47(d)
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S EL:-ASS ESS V E \\
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_0W POWER TESTING EVOLLTION
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T.C. Feigenbaum
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PURPOSE:
To perform a self-assessment of the preparation for and the conduct of
activities associated with the Seabrook Station low power testing evolution
in order to assess the readiness and effectiveness of personnel, programs
and equipment and to identify areas requiring immediate or long term
management attention.
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SCOPE:
The scope of the self-assessment effort will include, as a minimum, the
following topical areas:
1. Plant operations
2. Radiological controls
3. Maintenance
4. Surveillance and testing
5. Safety assessment / Quality verification
6. Control room operations
7. Effectiveness of internal problem identification and resolution
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8. Plant chemistry and health physics
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FOCUS:
Within the above topical areas, the self-assessment effort will focus on
the following organizational conduct and activities:
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1. Orgenizationalinterfaces and management effectiveness
2. Plant configuration control
3. Program / procedural adequacy and compliance
4. Communications and teamwork
5. Operational Quality Assurance effectiveness
6. Timeliness and adequacy of support of Station activities
-
7. Training program adequacy and effectiveness
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8. Timeliness and adequacy of corrective action reporting and
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follow-through
9. Adequacy of design based on Low Power Test Program elements
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SELF-ASSESSMENT TEAM ORGANIZATION:
MANAGEMENT OVERSIGHT COMMITTEE
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E. A. Brown - President and CEO
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G.S. Thomas - V.P. Nuclear Production
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T.C. Feigenbaum - V.P. Engineering, Licensing
and Quality Programs
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D.E. Moody - Station Manager
SELF-ASSESSMENT TEAM MANAGER
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N. A. Pillsbury - Independent Review Team Manager
SELF-ASSESSMENT TEAM MEMBERS *
AREAS of EXPERIENCE:
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Operations
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Maintenance
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Chemistry / Health Physics
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Training
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Engineering / Technical Support
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QA/QC
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Independent Safety Engineering Group
- Approximately 30% of each work week to be dedicated to evaluat;on activities
. _ _
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NRC INTERFACE:
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Periodic updates by Team Manager and members of the Management
Oversight Committee (bi-weekly suggested)
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Final report available to NRC Resident and Region 1 office
(approximately 6 weeks after completion of Low Power Testing)
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Normal daily contact with NRC Resident as required
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NHY/NRC critique of performance following completion of
majpr activities
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SCHEDULE of ACTIVITIES:
WEEK LOW POWER TESTING
SELF-ASSESSMENT
MGMT. OVERSIGHT COMMITTEE
Preparation
Preparation
Team Preparation
Team & Mgmt Briefing
Preparation
Self Assessment Team Start
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Preparation
Self Assessment
Team Status Report
Preparation
Self Assessment
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Heatup
Self Assessment
Team Status Report
Heatup
Self-Assessment
Team Status Report
- Precritical Concurrences Status -
Low Power Tests
Self Assessment
Team Status Report
Low Power Tests /Cooldown
Self Assessment
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Layitp
Self Assessment
Team Status Report
Layup
Self Assessment
Layup
Self Assessment
Team Status Report
Layup
Self Assessment End
layup
Draft Report
D/R Internal Distribution
Layup
Layup
Issue Final Report
Team & Mgmt Debriefing
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STA"L S 0" B J _LE
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J.J. Warnock
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BULLETIN 88-05 SUMMARY
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Falsified CMTRs - WJM/ PSI / Chews Landing
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identify Installed Fittings and Flanges (F/F) and other
material and test
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Engineering evaluation for F/F as required
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Written report to NRC
,
--
-
f.
- -
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SEABROOK APPROACH
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Documentation review
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Field walkdowns
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Procedures developed
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Testing of installed F/Fs
e
Laboratory testing of selected F/Fs
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NUMARC/EPRI support
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Engineering Evaluation
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Additional confirmations
- DRAVO
- Radnor Alloys
-- Other suppliers
- Continued NUMARC support
- -
- -
- -
... _ _ _ _ _ _ _
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. - - -
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BULLETIN 88-05 RESULTS
DOCUMENTATION REVIEW
e
Complete
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358 WJM flanges / fittings installed; 12 F/F vendor
markings not available (B31.1 only)
e
13 S/R ASME systems affected; 1 S/R B31.1
systems affected
e
Predominently carbon steel (5 stainless
steel flanges)
TEST RESULTS
e
368 tested
e
30 requiring engineering evaluation
e
No replacement anticipated
0THER
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DRAVO review consistent
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Supplier responses
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