ML18082A189
| ML18082A189 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/20/1980 |
| From: | Dante Johnson, Keimig R, Norholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18082A188 | List: |
| References | |
| 50-272-79-32, 50-311-79-37, NUDOCS 8004170293 | |
| Download: ML18082A189 (20) | |
See also: IR 05000272/1979032
Text
U. S. NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
50-272/79-32
Report Nos.
50~311/79-37
50-272
Docket Nos.
50-311
REGION I
License Nos.
CPPR-53
Priority -----
Category
c
81
Licensee:
Public Service Electric and Gas Company
80 Park Place
Newark, New Jersey
07101
Facility Name:
Salem Nuclear Generating Station - Units land 2
Inspection At:
Hancocks Bridge, New Jersey
I'nspection Conducted:* November 18 - December 15, 1979 and Januar 9, 1980
Inspectors:
Approved by:
Section
Inspection Summary:
~- z" .. F°t:J
date
~- 21J- r~
date
date
2-~t)-/"4
date
Inspections on November rn*.:.*oecember*15~ *1979 and.January 9, 1980 (Combined .
Report Nos. *50.:.272179.:.32 and*so.:.311;79.:.37)
Unit l Areas Inspected: *Routine inspections by the resident*insp~ctor of plant
operations including:
tours of the facility; log and record reviews; review of
licensee events; IE Bulletins and Circulars; implementation of licensing commit-
ments; and, followup on previous inspection items.
The inspections involved 49.
inspector-hours by the NRC resident inspector and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> by a regional based
inspector.
- Results£
Ohe item of noncompliance was identified. (Infraction - failure to
follow procedures, Details 7).
Region I Form 12
(Rev. April 77)
'.t
.. ..
Unit 2. Areas Inspected:
Routine inspections by the resident inspector of plant
preoperational testing including:
tours of the facility; IE Bulletins and
Circulars; reportable items under 10 CFR 50.55(e); followup on previous
inspection items; and, preparedness for issuance of an operating license.
The inspections involved 15 inspector-hours by the NRC resident inspector.
Results.:
No items of* noncompliance were identified .
l.
2 .
"DETAILS
Persons Contacted
S. LaBruna, Maintenance Engineer .
A. Meyer, Site QA Engineer
E. Meyer, Project QA Engineer
.
H. Midura, Manager - Salem Generating Station
P. Mo ell er, Associate Engineer
W. Reuther, Site QAD
F. Schna~r, Station Operating Engineer
R. Silverio, Assistant to the Manager
J. S~illman, Station QA Engineer
J. Zupko, Chief Engineer
The inspectors also interviewed and talked with other licensee personnel
during the course of the inspections including management, clerical,
maintenance, operations, performance, quality assurance, and construction
personnel.
Status of Previous Inspection Items
(Closed) Unresolved Item (272/77-24-02):
Placement of fire hose in
outdoor hose houses.
The inspector verified through observation that
hose houses were adequately equipped with fire fighting gear and were
fitted with breakaway locks.
(Closed) Follow Item (272/78-29-03): Resolution of DR PD-0607.
The
inspector reviewed documentation tracking repair of a Hagan Comparator
Module.
This module had been the cause of an inoperable pressurizer
pressure channel.
Resolution of the DR included repair of a cold solder
joint and a design change to improve voltage regulation.
The inspector
had no further questions.
(Closed)" Unresolved. Item (272/79-18-05): Nuclear Review Board Charter
rev1s1on.
The inspector reviewed NRB Charter, Revision 5, dated September
25, 1979 and noted that it was now consistent w.ith Technical Specifications
Section 6 in the area of audit coverage and membership for the NRB.
The
inspector had no further questions.
(Closed) Follow Item (272/79-22-04): Evaluation of insulation weight
omission in seismic analysis calculations. The inspector reviewed calcula-
tions made on the pipe section irr question with and without insulation
weight taken into consideration. The stress values obtained were not
significantly different. It was further noted that the insulation had
been added for personnel protection due to the proximity of the pipe to a
walkway.
The inspector.had no further questions on this item.
4
(Closed) Open Itein (311/79-03-13):
Environmental preservation controls
ror records storage in accordance with ANSI N45. 2 .* 9.
The licensee had
addressed this concern by storing duplicate records in a records storage
facility at Iron Mountain.
Transmittal on a routine basis was verified.
- . The inspector had* no furthe.r questibns.
.
.
(Open) Unresolved Item (272/78""29-0l):
Evaluation of several vital
inverter failures during 1978.
The inspector reviewed an engineering
evaluation dated September 18, 1979, which concludes that, while the
inverters meet design criteria, modifications or replacement to improve
reliability would be desirable. This item remains open pending licensee
imp.lementation o.f an appropriate inverter improvement program.
Unit l
3.
Shift Logs and Operating. Records
a.
The inspector reviewed the following plant procedures to determine
the licensee established requirements in this area in preparation
for a review of selected logs and records.
AP-5, Operating Practices, Revision 9, April 23,. 1979;
AP"-6, Operational Incidents, Revision 6, February 22, 1979;
AP-13, Control of Lifted Leads and Jumpers, Revision 3, February
22, 1979;
Operations Directive Manual; and,
AP-15, Tagging Rules, Revision 0, April 13, 1979.
The* inspector had. no .questions tn this area.
b.
Shift logs and operating records were reviewed to verify that:
.
.
.
Control room log sheet entries are filled out and initialled;
Auxiliary fog sheets are filled out and initialled:
Log entries involving abnormal conditions' provide sufficient
- detail to communicate equipment status, lockout status, cor-
rection* and restoration;
Log book reviews are.being conducted by the staff;
Operating orders do not conflict with Technical Specification
requirements;
5
Incident reports detail no violation of Technical Specification
LCO or reporting requirement; and,
Log's and records were maintained in accordance with Technica,l
Specifications and the procedures in 3.a above.
c.
The re.view included the fo.llowing plant shift logs and operating
records as indicated and discussed with licensee personnel:
Log No. l - Control Room Daily Log, November 16, 17, 19, 20,
25-27, 29, December 1-6,. 8-13;
- Log No. 3 - Control Console Reading Sheet, November 16, 17, 19,
20, 25-27, 29, December l-6, 8-13;
Night Orders, October 29 - December 10, 1979
4.
Plant Tour
a.
Durihg the* course of the inspections, the inspector made observations
and conducted multiple tours of:
(l) Control Room
(2) Relay Room
(3) Auxiliary Bi.J.ilding
(4) Vital Switchgear Rooms
(5} Turbine Building
(6) Yard Areas
(7) Radwaste Building
(8) Control Point
b.
The following determinations we.re made:
(l) Monitoring instrumentation:
The inspector verified that selected
instruments were functional and demonstrated parameters within
Technical Specification limits.
(2). Valve positions.
The inspector verified that selected valves
were in the position or condition required by the Technical
Spec.i fi cations for the applicable p 1 ant mode .
(3)
(4)
(5)
(6)
(7)
6
Radiatian controls. The inspector verified by observation that
control point procedures and posting requi.rements were being
followed.
Plant housekeeping conditions.
Observations relative to plant
housekeeping and fire hazards identified no notable conditions
except for Auxiliary Building elevation 122'.* A considerable
accumulation of outage-related items has collected in this
~rea. Cleanup is in progress.
Fluid leaks.
No fluid leaks were observed which had not been
identified by station personnel with corrective action initiated,
as necessary.
Piping vibration.
No excessive piping vibration was noted
during the plant tours.
Selected pipe hangers and seismic restraints were observed and
no adverse conditions were noted.
(8). Technical specifications. Through log review and observations
during tours, the tnspector verified compliance with selected
Technical Specification Limitin~ Conditions for Operation.
The
following parameters were sampled frequently:
RHR flow rate,
Boric Acid Storage Tank levels and concentration, emergency and
off s.ite power availability, source range nuclear instrumentation,
system operability verification prior to Mode Changes.
(9) Control room annunciators.
Selected lit annunicators were
discussed with control room operators ta verify that the reasons
for them were understood and corrective action, if required,
was being taken.
(10) By frequent observation through the inspection including s.hift
turnovers., the inspector verified that control room manning*
requirements of 10 CFR 50.54(k.) and the Technical Specifications
were* being met.
In addition, the inspector observed that
frequent tours were made by shift supervision.
c.
The following acceptance criteria were used for the above items.
(1) Technical Specifications
(2) Operations Directives Manual
(3)
Inspector Judgement
d.
Except as noted above, the inspector had no questions relative to
observations during plant tours.
7
5.
Licensee Event Reports CLER' s)
a.
In Office Review of Licensee Event Reports
The .inspector reviewed LERs submitted to the *NRC:RI office to verify
that details of the event were clearly reported, including* the
accuracy of the description of cause and adequacy of corrective
action.
The inspector determined whether further information was
required from. the licensee, whether generic implications were involved,
and* whether the event warranted onsite followup.
The following LERs
were reviewed:
- --
79-36/0lT, 4 KV Vital Bus Differential Relays Seismic Deficiency
--
79-37/03L, Source Range Nuclear InstrumentationChannel N-32
- --
- --
- --.
- --*
- --
.. *--
- --
79-38/03L, Containment Air Particulate Detector Inoperable
79-39/03L, Inadvertent Entry Into Refueling Mode 6
79-41/0lT, Loss of Eddy Current Template- Plug Assembly Inside
the Reactor* Coolant System
79-42/03L, Fire Barriers Inoperable
79-43/0lT, Auxiliary Feedwater Pumps Discharge Valves Closed
During Surveillance Testing
79;...44/03L, Damaged Fue.l Assemblies
79-45/0lT, Imperfections in Component Cooling Pump Impellers
79;...46/0lT, Linear Indications in Steam Generator Feedwater
. Nozzle Welds
.
.
79-47/03L, Failed Spid~r Fingers on Six (6) Control Rods
- -- . 79;...48/0lT, Degradation of Steam Generator Tubes
- --
- lc;.._
- --
- --
79-50/0lT, Steam Generator Water Level Instrumentation Deficiency
79-52/0lT, Steam Generator Rate of Rise Restriction
79-55/0lT, Possible Malfunction of Containment Ventilation
Isolation Valve
79-56/0lT, RHR Pump Exceeds Design Runout Flow
b.
8
- --
79-58/0lT, Qualification of Control Systems for Adverse Environ-
mental Conditions
- --
79-60/03L, Loss of Safeguards Manual Initiation/Reset Functions
- --
79-6T/03L, Open Electrical Penetration Fire Barrier with No Fire
Watch
- --
79-53/03L, Loss of Fire Suppression Systems
- --
79-64/0lT, Low Pressure C02 Storage Tank Level Less Than Required
by Technical. Specifications
- --
79-65/03L, Failure to Submit Temporary Changes to Station Procedures
to SORG for Review within 14 Days
- --
79-66/03L, Meteorological Tower Instrumentation Inoperable
Onsite Licensee Ev~nt FOllowup
For those LERs selected for onsite followup (denoted by asterisks in
detail Paragraph 5.a), the inspector verified that the reporting
requirements of Technical Specifications and Regulatory Guide 1.16 had
been met, that appropriate corrective action had.been taken, that the
event was reviewed by the licensee as required.by AP-4, 6, and 7, and
that continued operation of the facility was conducted in accordance
with Technical Specification limits. The following findings relate ta*
the LERs reviewed on site*:
-.-
79-36/0lT, The initial corrective action taken consisted of
disconnecting the trip function provided by the suspect relays.
The relays have been replaced. with qualified substitutes (NRC
Inspection Report 50-272/79-25 refers).
79-39/03L, Reactor head removal is accomplished using Maintenance
nepartment Procedure* MSC,. Reactor Vessel Head and Interna 1 s
Removal and Installation. This event was caused by a failure on
the part of maintenance personnel to inform the shift supervisor
when that portion of the procedure was reached which actually
called for head detensioning and removal.
This action places the
plant in Mode 6.
The procedure has been modified (Revision ll,
dated June 25, 1979) to require a senior shift supervisor signoff
to indicate that the plant is ready to enter Mode 6.
79-41/0lT, This event resulted from failure to account for one
of 34 black plastic plugs (2" long, l" diameter) used to secure
a template inside the steam generator primary side during eddy
current testing of tubes.
The plug is assumed to have remained
inside the reactor coolant system boundary.
The inspector
9
reviewed an analysis of potential effects from continued unit
operation with the plug in the system. * The analysis concludes
that no detrimental effects will be realized from the presence
of this plug.
NRR concurs in this conclusion as documented in*
the safety analysis accompanying license Amendment 20, dated
October 30, 1979.
79"".'43/0lT, The inspector verified that the associated surveillance
test procedure had been modified to preclude isolation of
redundant pumps du.ring operability testing Of Auxiliary Feedwater
Pumps.* In addition,- this LER and a letter to NRR dated November
1, 1979, commits to locking open manually operated Auxiliary
.
-Feedwater Valves.
The inspector verified that procedure revisions
have been made to lock open the following valves:
lAFl, 11-13AF3,
ll-14AF10, 11-14AF20, 11-14AF22, and 11-14AF86.
Field verifica-
. tion of valve position and locks was also conducted by the
inspector.
The inspector had no questions.
79-44/03L, Followup on this event is documented in NRC Inspection
Repo-rts 50-272/79-15 and 79-18.
Fuel assembly grid strap
-damage is also addressed in the safety evaluation accompanying
Ticens~ Amendment 20, dated October 30, 1979 .
79-46/0lG, As a result of NOE indications found in carbon steel
component cooling pump impellers, the licensee has elected to
replace the impellers with stainless steel. This has been
accomplished on Unit 1.
The inspector had no further questions.
79-46/0lT, Cracking in the feedwater nozzle to piping weld in
all four steam generators was identified. All have been repaired.
Details of findings, repa.irs, and further evaluation are discussed
in the licensee 1 s responses to IE Bulletin 79-13 dated July 12,
August 24, and November 15, 1979 and. in NRC Inspection Report
50-272/79-24.
79-47/03L, Fo.llowup on this event is discussed* in NRC Inspection
Report 50-272/79-18.
The significance of control rod finger
separation is also addressed in the NRC safety analysis accorn-
panyi.ng Amendment 20 to the facility operating license.
--
79-48/0lT, Visible wear marks were observed on steam generator
tubes adjacent to the tube lane blocking device on steam generator
No.s. 11 , 13, and 14.
Eddy current testing confirmed wa 11
thicknes.s degradation in five tubes.
Since motion of the tube
lane blocking device appeared to have been caused by improper
installation, reassembly was done under the direction of vendor
personnel. A 11 adjacent tubes in the affected steam generators
were plugged.
10
79-50/0lT, The inspector verified that the steam generator
level trip has been raised to 11% to account for level errors
induced by an a*cci dent environment.
The potenti a 1 for 1eve1
errors has been brought to the attention of operators via
memorandum, however, no procedural change has been made to
incorporate this concern. This item is unresolved pending a
modification to operating procedures (272/79-32-01).
79-52/0lT, The licensee raised a concern that a license re-
striction on feedwater flow rate, based on water hammer consid-
erations, limited the flow to a rate less than that assumed in
. the safety analysis for auxiliary feedwater*.
License Amendment
22, dated* November 20, 1979,. removes the license condition
related to steam generator level rate to rise. The inspector
verified that station emergency procedures have been changed to
remove any restriction on feedwater flow rate whenever loss of
secondary heat sink is threatened.
Flow rate considerations to
avoid water hammer conditions are retained in operating instruc-
tions for those situations where total loss of heat sink is not
imminent.
The inspector had no further questions *.
79-55/0lT, In responding to NRR concerns, the licensee could
not demonstrate that containment ventilation butterfly valves
would close within design time intervals with a design differential
pressure across them.
In correspondence to NRR dated November
21, 1979, the licensee confirms a commitment to keep the four
36
11 purge valves shut at all times, except when required to
perform tests pursuant to the Technical Specificattons, and to
modify the operators on the 10
11 pressure/vacuum relief valves
to meet design closure specifications. The inspector verified
that administrative controls have been applied to the large
valves to preclude operation in Modes l through 4.
In addition,
vendor modifications to the 10
11 valves have been completed to
11short--stroke
11 the valves such that the full open position is
600.
79-56/0lT, Analysis related to RHR pump NPSH indicated that RHR
pump runout conditions would be experienced if, while in the
post-accident recirculation mode, one RHR pump is used to supply
two SI pumps, two charging pumps, and two cold legs directly.
Resolution was achieved by reducing the size of a downstream
flow measuring orifice such that system flow resistance increased
enough to prevent single pump runout.
The inspector verified
that the orifice modification had. been completed and the flow
instrumentation recalibrated.
79-58/0lT, A review of environmental qualifications indicated
that the following control systems could, if subjected to the
accident environment, have an impact on systems with protective
11
functions; pressurizer power operated relief valve controls,
steam generator power operated relief valve controls, main
feedwater controls, and automatic rod control. In a response
-dated October 4, 1979, to an NRR *request for information under
l 0 C_FR 50. 54( f), the licensee concludes that no modi fi cation
due to the above concerns is required, with the exception of
procedural cautions for the operator relating to steam generator
power operated relief valve misoperation which may occur on a
postulated high energy line break in a penetration area.
At
the conclusion of this inspection period, these procedure
changes had not been made..
This item is unresolved (272/79-32-02).
79*50/03L; The inspector verified by direct inspection and
review of records that a design change has been made to increase
the trip setpoint of circuit breakers providing safeguards
- reset power.
The inspector had no further questions.
79-61/03L, The inspector verified that maintenance contractor
personnel had again been notified of Maintenance Procedure M3Y
and Technical Specifications requirements relative to opened
fire barriers. Based on routine inspection observations, it is
concluded that the requirements are known to maintenance and
ope rat ions personnel.
No recurrence of this i tern has been
identified by the inspector;
79-63/03L, This is a repeated occurrence.
Administrative
controls imposed on the fire protecti~n cross-connect valve
with the Hope Creek site proved inadequate to prevent opening
of the valve and consequent drai ni.ng of the system volume below
Technical Specification limits. The inspector verified that
corrective action taken after this later event included instal-
lation of a positive locking arm and lock over the valve access
tube.
Direct knowledge of shift supervision and permission of
the Chief Engineer are now required to open valve 1FP30.
The
inspector had no further questions.
79-64/0lT, The inspector verified that changes to operating
logs have been made to raise the minimum CARDOX level to 75%.
Continual log reviews and observations relating to this specific
item have identified no recurrence.
Further concerns relative
to this item are documented in NRC Inspection Report 50-272/79-27.
79-65/03L, The inspector verified that the supervisor involved
had been reinstructed in on-the-spot change procedures.
Based
on the number of successfully executed on-the-spot changes,
this* appears to be an isolated case.
12
79-66/03L, To maintain continuity of meteorolog.ical tower
instrumentation the licensee i~ formulating plans to install*
surge protection aga.inst lightning strikes.
Pending implemen-
tation of an appropriate design change to improve meteorological
instrument .reliability, this item is unresolved (272/79-32-03).
c.
The following Licensee Event Reports required corrective action
pursuant to the license or Technical Specifications:
79-39/03L
.79-42/031
79-43/03L
79-61/03L
79*64/0lT
79-65/03L
The inspector had no further questions relative to the above LERs *
6.
Other Items
a.
As a result. of staff reviews, concerns were raised relative* to RHR
pump NPSH requirements in.the post-accident recirculation mode.
At
meetings with the staff on November 16, 1979 and November 21, 1979,.
it was determined that both Salem units would increase the number of
flow*holes in the containment sump cover, install anti-vortex baffles
in the sump, and raise the minimum NPSH level setpoint. All of
these modi fi cat ions were complete at the end of this report peri.od,
and before significant power* history had been added to the refueled
core.- on Unit l.
Thefospector had no questions re.lative to the above.
b.
On December* 10, 1979, the inspector participated in an audit conducted
by- the NRR Bulletins and Orders Task Force on site.
The subject of
this audit was implementation of NRR-approved Westinghouse guidelines
_to be used in developing small break loss of coolant accident procedures.
The audit included detailed procedures reviews and operator interviews.
The licensee had substantially followed the vendor guidelines in
modifying accident procedures. A number of comments were provided
to the licensee relative to omissions of some recommended steps ih
the procedure.
These are being evaluated by the licensee.
Some of
the comments have been incorporated. Others wi 11 not be, due to
c.
13
uniqueness in design, accident analysis considerations, or in response
to an effort to omit superfluous information from immediate action
steps in emergency instructions.
Some weaknesses were identified in operator knowledge of the TMI
accident scenario and the dynamics of saturated systems.
These will
be addressed in the licensee's continuing training program.
The inspector had no further questions in this area at this time.
The NSSS vendor, Westinghouse, had identified to the licensee and
the NRC staff a concern relative to rod control system response to a
dropped rod. * The postulated event involves a dropped rod in the
vicinity of a power range nuclear instrument detector.
Simultaneous
fa.ilure of the power auctioneering circuit could result in rod
withdrawal in response to the indicated drop in reactor power.
This
could result in an over power condition. It was agreed that automatic
rod control would be used only at less than 90% power.
Above 90%
power, automatic rod control would not be used unless Bank D rods
were at least at 215 steps.
The inspector verified that the licensee had incorporated the above*
requirements into operating procedures before returning Unit 1 to
critical operation.
The inspector had no further questions in this area.
7.
Unauthorized Tag Out of Diesel Generator (Reference: LER 79-71/03L)
.In preparation for adding lube oil to the A, B and C emergency diesels,
the automatic initiation lockout switches for the_co, Fire Suppression
System for the three diesel areas are tagged out as a safety precaution.
However, the diesel lockout switches were also tagged out at the same
time, rendering all three diesel generators inoperable.
This condition
existed for a period of 16 minutes on October 31, 1979.
Technical Specification 3.8.12 requires that with less thahthe minimum
AC electrical power sources operable, suspend all operations involving
core alterations or positive reactivity changes until the minimum required
AC electrical power sources are returned to an operable status.
During
the 16 minutes the diesels were inoperable, the plant was being maintained
in Mode 5 (cold shutdown) with no core alterations or reactivity changes
involved.
The cause of this occurrence was personnel error in that an individual
not authorized or familiar with Technical Specification requirements,
approved a tagging request without the knowledge of the responsible Shift
Supervisor.
14
The above actions were in violatio.n of Technical Speci.fication 6.8.1
requirements in that administrative controls established by Administrative
Procedure AP-2, "Station Organization", Section 5~14, AP-15, "Tagging
Rules
11
, Section6.0 and Operating Department Memo (OM-15) were not followed.
Thi~ is an item of noncompliance~ Infraction level.
Unit 2
7. * Plant Tour
a .. * .. The inspector conducted periodic tours of all accessible areas in
the plant. During these tours, the following specific items were
evaluated:
(1)
(2)
(3)
(4)
(5)
Hot Work.
Adequacy of fire prevention/protection measures
used.
Fire Equipment.
Operability and evidence of periodic inspection
of fire suppression equipment.
-
Housekeeping.
Minimal accumulations of debris and maintenance
of requ.i red cleanness levels in systems under or f o 11 owing
testing.
Equipment Preservation.
Maintenance of special preservative
measures for installed equipment as applicable.
Component Tagging.
Implementation and observance of equipment
tagging for safety or equipment protection. Authorized logs
for 5 components were selected for review.
(6) Maintenance.
Corrective maintenance in accordance with esta-
blished procedures.
(7) Instrumentation.
Adequate protection for installed instrumen"."
tat ion.
(8) .Cable Pulling.
Adequate measures taken to protect cable from
damage while being pulled~
(9) Communication.
Effectiveness of public address system in all
areas of the site.
(10) Equipment Controls.
Effectiveness of jurisdictional controls
in precluding unauthorized work on systems in test or which
have been tested .
15
(ll) Logs.
Completeness of logs maintained and resolution of identi-
fied problems.
(12) Foreign Material Exclusion.
Maintenance of controls to assure
systems which have been cleaned and flushed are not r*eopened to
admi't foreign material.
(13) Security .. Implementation of security provisions.
Particular
.attention to. maintenance of Unit 1 protected area boundary.
(14) Testing.
Spot-checks of testing in progress were made.
b.
. The following comment applies to observations made during the tours
of the plant.
During one tour, it was noted that the floor plates at elevations
100 1 and 84 1 in the Auxiliary Building had been removed.
This
provides access from a Unit 2 (unprotected) location on elevation
84
1 to Unit 1 vital areas on elevation 100 1 and 64 1 *
The
inspector questioned the effectiveness of a single armed guard
placed at elevation 100 1 to monitor the perimeter* opening.
A
guard was subsequently placed at elevation 84 1 , providing more
positive access control at the point of entry into Unit 1
areas.
The inspector* had no further questions in this area.
8.
Reports of Significant D~ficiencies
Reports of significant deficiencies, made by the applicant pursuant to 10
-CFR 50.55(e) are revfewed upon receipt, or v.erbal notification, to determine
whether reporting requirements have been met, whether sufficient detail
has been provided to assess the significance of the event, whether the
cause appears adequately defined, whether corrective action appears
appropriate, and whether generic applicability can be ascertained.
These
reports are followed lip by site inspection to verify accuracy and to
ensure that corrective actions have been implemented.
Detailed below are several reports made by the applicant for which the
above reviews were made.
a.
By correspondence dated May 2, 1979, the applicant confirmed an
April 11, 1979 telephone report to NRC Region I relating to seismically
unqualified 4 KV Bus Differential Voltage relays. This item was
also reported for Unit 1 as LER 79-36.
Corrective action consisted
of replacement with qualified relays, and this action has been
completed (NRC Inspection Report 50-311/79-27 and 79-34 refer)..
The
inspector had no further questions on this item.
16
b.
By correspondence dated July 13, ,1979, the applicant confirmed a
June*l5, 1979 telephone report to NRC Region I relating to linear
indications. identified in installed component cooling pump impellers
manufactured by Gould Pumps.
Similar findings on Unit l were reported
as LER 79-45.
Based on evaluatiOn by the vendor and applicant, all
impellers will be replaced with stainless steel. This action has
been completed on Unit l. Replacements for Unit 2 have not been
. received..
Replacement of component cooling pump impellers is an
unresolved item (311/79~37-02).
c~
By correspondence dated October 5, 1979, the applicant confirmed an
August 30, 19.79 telephone report to NRC Region I relati-ng to potential
steam generator level errors which could result from reference leg
heatup in an accident environment.
This was reported as a generic
item by Westinghouse, is the subject of IE Bulletin 79-21, and was
reported for Unit l as LER 79-46.
Corrective action includes raising
the low-low level setpoint to 11%, for which ECN 35397 has been
issued.
This is an unresolved item, pending verification that the
ECN has been accomplished (311/79-37-03).
Additional corrective
actions include procedure modificatians and will be verified by
resolution of unresolved item 272/79-32-01.
d.
By correspondence dated October 5, 1979, the applicant confirmed an
August 30, 1979 telephone report to NRC Region I relating to potent.i al
inability of containment ventilation valves to close .under design
containment differential pressure conditions. This item was also
reported for Unit l as LER 79-55.
Corrective actions are identical
to those outlined in Detail 5 of this report.
Modifications of the
Unit 2 10
11 valves has not been completed.
This is an unresolved
item pending completion and testing of the modifications (311/79-37-04).
e.
By correspondence dated October 4, 1979, the applicant confirmed an
August 30, 1979 telephone report to NRC Region I relating to potential
RHR pump runout conditions under certain post-accident operating
configurations. This item was also reported for Unit l as LER
79-56.. Corrective action for both units is the same, and consists
of modifying a flow orifice downstream of the pump to increase
piping flow resistance. Modification to the flow orifice is an
unresolved item (311/79-37-05).
f.
By correspondence dated November 14, 1979, the applicant confirmed a
Westinghouse Part 21 report relating to an undetectable failure in
the engineering safety features actuation system.
Specifically, the
P-4 permissive signal is derived from auxiliary contacts on the
reactor trip breakers.
The permissive allows manual reset and block
of safety injection only when the reactor trip breakers are open.
Misoperation of the permissive, detected only by status lights,
could prevent post-accident reset for realignment to recirculation
17
or could prevent unblocking of SI when the trip breakers are closed.
Westinghouse letter dated November 8, 1979 provides a test procedure
-to verify operability of the P-4 interlock.
The i.nspector verified
that the procedure had been successfully performed on Unit 1 prior
to Mode 4 operation. Verification of operability on Unit 2 is an
unresolved item (311/79-37-01).
No periodic test procedure has been
issued to provide continued assurance that the interlock remains
functional.
Issuance of an appropriate surveillance test is also
unr~solved (272/ 79-32~04).
The inspector had no* further questions relative to reports received from.
the applicant.
9.
Operational Readiness
10 CFR 50.57 states that the issuance of an operating license is, in
part, contingent upon a finding that construction of the facility has
been substantially completed, in conformity with the construction permit
and the application, as amended, the provisions of the Act, and the rules
and.regulations of the Commission.
In order to provide a basis for this finding, the inspector is conducting
a continuing review of licensee readiness to operate the facility.
This
review includes, but is not limited to, the following areas:
- --
-*--
Completion of the NRC inspection program to assess construction,
testing, and operational preparedness.
Status of facility operating procedures and personnel training.
Status of all enforcement items and unresolved matters.
Status of the- preoperational test program.
Status of construction activities_
Proposed facility Technical Specifications.
Review- of licensee outstanding items, particularly those identified
for completion or resolution after core load.
Implementation of corrective measures to Unit 2 as a result of items
identified in Unit l for Reportable Occurrences, inspection findings,
and IE Bulletin and Circulars.
Operational safety concerns arising from the above reviews will be promptly
identified to facility management for resolution prior to the inspector
reaching a finding of operational readiness.
No specific safety concerns
have been identified to date.
18
Site
10.
IE Bulletin*and Circular Followup
a..
The IE Bulletins and Circulars discussed below were reviewed to verify
that:
Licensee management forwarded copies of the response to the
bulletin to appropriate onsite management representatives.
Information discussed in the licensee's reply was supported by
- * facility records or by visual examination of the facility.
Corrective action taken was effected as described in the reply.
The licensee's reply was prompt and within th~ time period
described in the bulletin.
- The reviews included discussions with licensee personnel and obser-
vation and review of items discussed in the details below *
. b.
By correspondence dated May 3, 1979, September 14, 1979, October 11,
1979, and October 18, 1979, the licensee responded to IE Bulletin
79-07,. Seismic Stress Analysis. of Safety Related Piping.
On November 13, 1979, the licensee's Phase I program, as defined in
the October 11 correspondence had been completed, and the unit
entered operating Mode 4.
After an intervening cooldown to repair
leaks around thermocouple seals and a pressurizer manway, criticality
was achieved on December 1, 1979.
On December 3, the reactor tripped
automatically from zero power when an.instrument technician conducting
tests on one.fntermediate range nuclear instrument channel operated
a test switch on the operating channel.
The plant was again brought
critical and startup testing resumed~
.
.
During the course of plant operation and testing, seismic reanalysis
continued.
As supports in the remaining systems to be analyzed were
identified which were deficient in design using newly calculated
loads, the associated subsystems were declared inoperable and the
appropriate Technical Specification action statement imposed.
On
December 4, 1979, a number of pipe supports in diesel generator
systems were found which failed to meet design requirements.
En-
gineering review of the calculations concluded that similar conclusions
followed relating to the same hangers on all three diesel generators.
All three diesel generators were declared inoperable and the require-
ments of Technical Specification 3.0.3 invoked.
The plant was
brought to cold shutdown.
Due to the potential for similar cases of
identical redundant system problems, a decision was made to remain
.
.
in cold shutdown .until it could be determined that all such cases
had *been re-evaluated under the conditions of IE Bulletin 79-07.
Salem Unit.l remained in Mode 5 (cold shutdown) at the end of this
reporting period.
The inspector reviewed the procedures used in the seismic: recalcu-
lations. Initially, information from field-verified piping isometrics
is. used as input to the stress program, which takes into account
. SRSS load. combinations as required by the Bulletjn.
The output data
provides piping stress values and support point loadings under
. earthquake conditions. The supports are evaluated for all three
orthogonal loads and their combinations as well as torsion loads.
Pipe stresses*.have been found to be within limits for calculations
completed to date.
Support evaluations are done separately, and
have identified approximately 800 supports requiring modifications.
A significant number of these have been pipe anchors which failed to
meet design criteria for torsion capability. The inspector identified
a sample of 50 supports which had been evaluated as unacceptable.
Quality evidence available at the site confirmed that 43 of these
had been modified.
After some research at the corporate office,
evaluations were found which concluded that the remaining 7 were
acceptable as built..
The inspector expressed his concern over the
lack of a single tracking mechanism by whkh it could be concluded
that all supports in safety related systems had been found acceptable
or had been modified.
No actual failure to account for a support
was i dent i fi ed.
The 1 i censee acknowledged the inspector 1 s concerns-.
The inspector had no further questions on this Bulletin at this
time.
c.
By correspondence dated April 4, 1979, the licensee responded to IE
Bulletin 79-03, Longitudinal Weld Defects in SA-312 Type 304 Stainless
Steel Pipe Spools Manufactured by Youngstown Welding andEngineering
Company.
The response addresses both Units 1 and 2.
The licensee
concludes that none of the suspect. material. is installed in safety.
related systems at Salem. l or 2 .. This conclusion is based* on a
review. of heat numbers for material supplied by Youngstown to eight
piping fabricators and vendors who, in turn, provided material to
the Salem project.
The inspector had no further questions on this Bulletin.
d.
By correspondence dated October 24, 1979, the licensee responded to
IE Bulletin 79-24, Frozen lines, for both Units 1 and 2.
The licensee
concludes that no unprotected safety related lines were identified
in his review.
The inspector verified that unprotected lines which
had caused problems for Salem Unit l during past winters (e.g.,
turbine impulse pressure sensor) had .also been modified through an
- appropriate ECN on Unit 2.
20
The i.nspector had no further questions on this. Bulletin.
e.
IE Circular discussed below was reviewed to verify that. it had been
reviewed for applicability by cognizant management, and approriate
action i~itiated .
. The review included discussions with licensee personnel and observation
- and review of items discussed in the details below. -
.
.
. .
.
. .
.
.
.
.
IE. Circular 78-19 addressed the potential for manual override
or bypass of safety system actuation signals, particularly by
reset on non-associated systems such as radiation monitoring.
The inspector reviewed a Controls Division Safety Evaluation
(CD-SE-9) dated January 10, 1979, which documents the review
conducted by the licensee.
The analysis concludes that sufficient
annunication of SI overrides at the system level are available
and that the design provides for adequate administrative control
over manual reset actions.
The inspector had no further questions relative to this Circ.ular.
ll. Unresolved Items
Areas for which more information is required to determine acceptability
are considered unresolved.
An unresolved item is contained in Paragraph
11.c of this report.
12.
Exit 1nterview
At periodic intervals during the course of this inspection, meetings were
held with senior facility management to discuss inspection scope and
findings.