Part 21 Report - 1995-132

ACCESSION #: 9503090138
                       LICENSEE EVENT REPORT (LER)

FACILITY NAME:  Salem Generating Station - Unit 1         PAGE: 1
OF 7

DOCKET NUMBER:  05000272

TITLE:  Technical Specification (TS) 3.0.3 Entry; Both Trains of
        the Solid State Protection System (SSPS) Being Inoperable

EVENT DATE:  02/01/95   LER #:  95-001-00   REPORT DATE: 
03/02/95

OTHER FACILITIES INVOLVED:                          DOCKET NO: 
05000

OPERATING MODE:  1   POWER LEVEL:  100%

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR
SECTION:

LICENSEE CONTACT FOR THIS LER:
NAME:  Frederick Wiltsee, LER Coordinator   TELEPHONE:  (609)
339-5163

COMPONENT FAILURE DESCRIPTION:
CAUSE:      SYSTEM:       COMPONENT:       MANUFACTURER:
REPORTABLE NPRDS:

SUPPLEMENTAL REPORT EXPECTED:  NO

ABSTRACT:

On 2/1/95, Unit 1 and 2 entered TS 3.0.3 when both Solid State
Protection'System (SSPS) trains were declared inoperable after
discovery
that the AC power distribution within the SSPS is susceptible to
a common
mode failure.  The NRC granted discretionary enforcement allowing
4 days
to implement design changes to modify the power distribution
arrangement-within SSPS.  During Unit 1 implementation of the
design
changes, a number of power supply related problems were
encountered.  On
2/3/95, Unit 1 initiated a shutdown in accordance with TS 3.3.3.1
for
exceeding reactor trip bypass breaker closure time in support of
SSPS
design changes.  Unit 2 entered TS 3.0.3 on 2/3/95, for both SSPS
trains
being inoperable, NRC rescinded discretionary enforcement at time
of TS
3.0.3 entry.  This occurred as a result of original design of the
SSPS.
The apparent cause of the SSPS power supply failures has been
attributed
to aged components and the lack of preventive maintenance. 
Design
changes have been completed to rewire the SSPS, power supply
leads.  New
power supplies were installed and additional power supplies
were-cleaned,
tested and returned to service.  Evaluation is on going to
determine
appropriate action concerning preventive and predictive testing
requirements for SSPS power supplies.  Event discovery followed
original
identification of this issue by Diablo Canyon.

END OF ABSTRACT

TEXT                                                         
PAGE 2 OF 7

Plant and System Identification:

Westinghouse - Pressurized Water Reactor

Energy Industry Identification System (EIIS) codes appear in the
text as
{xx}

Identification of Occurrence:

Technical Specification (TS) 3.0.3 Entry; Both Trains of the
Solid State
Protection System (SSPS) Inoperable

Event Date:    2/1/95 and 2/3/95

Report Date:   3/2/95

This report was initiated by Incident Report No. 95-066, 95-073,
and
95-075.

Conditions Prior to Occurrence:

Unit 1         Mode 1         Reactor Power 100 %

Unit Load 1151 MWe

Unit 2         Mode 2    Reactor Power 10E-08 amps

Unit Load 0 MWe

Description of Occurrence:

On February 1, 1995, Unit 1 and 2 entered TS 3.0.3 when both SSPS
trains
were declared inoperable after discovery that the AC power
distribution
within SSPS {JC} is susceptible to a common mode failure.  SSPS
input
signals which originate in the turbine building (auto stop oil,
stop
valve limit switches, and reactor coolant pump breaker position
signals)
were susceptible to failure (short circuits), due to the
consequential
effects of design basis accidents, including earthquake, and
environmental effects of pipe ruptures.  The Nuclear Regulatory
Commission (NRC) granted discretionary

TEXT                                                         
PAGE 3 OF 7

Description of Occurrence: (cont'd)

enforcement allowing 4 days to implement design changes to modify
the
power distribution arrangement within SSPS and change fuse sizes.

This concern is due to the location of equipment, wiring, and
junction
boxes with respect to high energy lines and the non-seismic
design of the
turbine building which feeds the 15 and 48 volt power supplies
and field
contacts powered from the SSPS input bays.

Event discovery followed original identification of this issue by
Diablo
Canyon.

During Unit 1 implementation of the design changes, a number of
power
supply related problems were encountered.  On February 3, 1995 at
1100
hours (hrs), a shutdown was initiated in accordance with TS
3.3.3.1
Action 1 due to exceeding time for having reactor trip bypass
breaker
closed in support of SSPS design changes.  Unit 1 entered Mode 3
at 1700
hrs, entered TS 3.0.3 both SSPS trains being inoperable, and
exited
discretionary enforcement.  On February 4, 1995 at 2230 hrs, the
Unit was
placed in Mode 5 for the completion of SSPS design changes.

Unit 2 entered TS 3.0.3 on February 3, 1995 at 1640 hrs, for both
SSPS
trains being inoperable, NRC Region I rescinded discretionary
enforcement
at time of TS 3.0.3 entry.  At 1730 hrs (same day), a Unit
shutdown from
Mode 2 was initiated and Mode 3 was entered at 1820 hrs (same
day).  On
February 4, 1995 at 0004 hrs, Mode 4 was entered and Mode 5 was
entered
on February 5, 1995, at 0431 hrs, for completion of SSPS design
changes.
The NRC was notified of the shutdown initiation, in accordance
with Code
of Federal Regulations 10CFR50.72(b)(1)(ii)(b).

Analysis of Occurrence:

A condition in which a fault in the circuitry for the turbine
stop valve
limit switches, autostop oil pressure switches or reactor coolant
pump
breaker position could possibly render the solid state protection
system
(SSPS) trains inoperable was identified at a plant of similar
design.
The review concluded that a single initiating event

TEXT                                                         
PAGE 4 OF 7

Analysis of Occurrence: (cont'd)

(e.g., main steam line break or seismic event), could possibly
render one
or both trains of SSPS inoperable.

The postulated failure can compromise the SSPS power supplies,
due to the
location of the 15 AMP fuses within the input bays.  This would
result in
a loss of power to the SSPS power supplies and subsequent loss of
power
to the SSPS logic and master relays (note that loss of the 48
volt power
supply will cause a reactor trip).

Electrical terminal boxes contain two SSPS instrument channels
for
turbine stop valve position indication while another terminal box
contains the auto stop oil input to SSPS.  These channels are
non-safety
related (non 1E) inputs to SSPS and are not electrically isolated
from
the safety related (1E) portion of the SSPS.

If the steam jet from the faulted main steam line was to strike
one of
the electrical terminal boxes, or if a seismic event affected the
terminal boxes, short circuits in the non 1E inputs to the SSPS
could
result.  Since the non 1E channels are not electrically isolated
from the
SSPS, the short would cause the fuses for the associated 1E
channels to
open.  The opening of the fuses for the 1E channels would result
in the
deenergizing of the power supplies for the logic circuitry of one
or both
trains of SSPS.  Assuming the credible failure of a short to
ground, the
SSPS 15 and 48 VDC power supplies would deenergize from the short
circuit.

At 0230 hrs on February 2, 1994, the NRC granted verbal
enforcement
discretion allowing 96 hours to restore operability of both SSPS
trains.
PSE&G committed to a formal written request on February 3, 1995. 
This
allowed restoration of one train of SSPS to operable status and
termination of TS 3.0.3.  The requested duration of the
enforcement
discretion was from 0230 hrs on February 2, 1995 until 0230 hrs
on
February 6, 1995, or completion of modifications.

Throughout this period, both trains would remain functional. 
During the
design modification, only one train at a time was rendered
inoperable.
During the modifications the redundant train was maintained
operable, as
well as the reactor trip function from SSPS.  In support of this
extension the following accident initiators were considered to be

TEXT                                                         
PAGE 5 OF 7

Analysis of Occurrence: (cont'd)

applicable:    1) seismic (alone); 2) seismic event resulting in
a loss
of off-site power; 3) seismic resulting in a steam line break in
the
affected area; 4) fire; and 5) steam line break.

Other events such as turbine building crane operation, handling
and/or
dropping of heavy loads, missile generation, and tornado were
considered
as potential initiators, however, consequences were less severe
or
comparable to that of the seismic event.

The power supply issues identified during SSPS design change
implementation were due to age related component failures (i.e.
capacitors, transistors) and the lack of preventive maintenance. 
Power
supplies were found with an excessive accumulation of dust.  In
addition,
the following were identified: a wire was shorted to the rear
mounted
heat sink, a ground had propagated from the conductive circuit
board
stand-offs (the other power supplies utilize non-conductive
standoffs).

Apparent Cause of Occurrence:

The cause of this event is "Design, Manufacturing,
Construction/Installation", as classified in Appendix B of NUREG
1022.
This occurred as a result of original design of the Solid State
Protection System (SSPS).

The apparent cause of the SSPS power supplies failures has been
attributed to aged components and the lack of preventive
maintenance.

Prior Similar Occurrence:

Review of documentation did not show similar occurrences.

Safety Significance:

This event did not affect the health and safety of the public. 
It is
reportable pursuant to

TEXT                                                         
PAGE 6 OF 7

Safety Significance: (cont'd)

10CFR50.73(a)(2)(ii)(B).  In addition, this report is intended to
satisfy
reporting requirements applicable to a potential 10CFR21 concern
involving the SSPS.

A modified hot zero power steam line break (SLB) core response
analysis
(UFSAR Section 15.4.2) was performed in which all four steam
generators
blow down to the environment without operator action or automatic
mitigation of any kind.  The evaluation shows that the current
UFSAR
licensing basis analysis of SLB core response remains bounding
and DNBR
limits are not exceeded and apply to previous operating cycles.

Long term heatup analysis indicates that even though all four
steam
generators dry out, the actuation of a single motor-driven
auxiliary
feedwater (AFW) pump is sufficient to prevent reactor coolant
system
(RCS) over-pressurization, pressurize overfill, or hot leg
boiling which
are typical bounding criteria applied to long-term events such as
loss of
normal feedwater, loss of AC power to auxiliaries, and feedline
rupture.
The evaluation shows the long term consequences of the subsequent
heatup
of the reactor coolant system (RCS) was analyzed with acceptable
results.
In addition, the frequencies of these events are relatively low.

Reactor vessel integrity for the effects of pressurized thermal
shock
(PTS) were evaluated with acceptable results.

Corrective Action:

Design changes have been completed to eliminate the identified
failure
mechanism of SSPS power supplies.

New power supplies were installed in Unit 2 and one new power
supply was
installed in Unit 1.  Three power supplies were cleaned and bench
tested
satisfactorily for return to service in Unit 1.

Evaluation is ongoing to determine appropriate action

TEXT                                                         
PAGE 7 OF 7

Corrective Action: (cont'd)

concerning preventive and predictive testing requirements for
SSPS power
supplies.

                                             J. C. Summers
                                             General Manager -
                                             Salem Operations

FHW:vs
REF:   SORC Mtg. 95-026

ATTACHMENT TO 9503090138                                     
PAGE 1 OF 1

     PSEG

Public Service Electric and Gas Company P.O. Box 236 Hancocks
Bridge, New
Jersey 08038-0236

Nuclear Business Unit

                                        March 2. 1995

U. S. Nuclear Regulatory Commission
Document Control Desk
Washington, DC 20555

Attn: Document Control Desk

SALEM GENERATING STATION
LICENSE NO. DPR-70 and DPR-75
DOCKET NO. 50-272 and 50-311
UNIT NO. 1 and 2

LICENSEE EVENT REPORT NO.  272/95-001-00

This Licensee Event Report is being submitted pursuant to

the requirements of Code of Federal Regulation

10CFR50.73(a)(2)(ii)(B).  Issuance of this report is

required within thirty (30) days of event discovery.

                                        Sincerely,

                                        J. C. Summers
                                        General Manager -
                                        Salem Operations

SORC Mtg.  95-026
FHW:vs

C    Distribution
     LER File

The power is in your hands                             95-2168
REV. 6/94

*** END OF DOCUMENT ***