ML18086B002
| ML18086B002 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/06/1981 |
| From: | Greenman E, Hill W, Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18086A999 | List: |
| References | |
| 50-272-81-23, 50-311-81-21, NUDOCS 8111050602 | |
| Download: ML18086B002 (23) | |
See also: IR 05000272/1981023
Text
-
f
I
U. S. NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
(DCS Numbers - see
attached sheet)
Report Nos.
Docket Nos.
License Nos.
Licensee:
50-272/81-23
50-311/81-21
50-272
50-311
REGION I
Public Service Electric and Gas Company
80 Park Plaza
Newark, New Jersey
07101
F ac i 1 i ty Name : __
s_a_l e.;._m.;_,.;;N.;..;.uc.;._l..;.:.e..;.:.a.;_r ""'"'G..;.:.e..;.:.n.;;;..er..;.:.a..;.:.t...;..i .;.;.,ng~S....:;.t.;;;.a t.;;_i....:;.o..;.:.n_-_...;.U..;.:.n...:..i t.::..;s;._;;;,1-'a;;;,;.n,;..;.d;....,;;;.2 _
Inspection At: __
H_a_nc_o_c_k_s _B_r_i .....
dg..._e_..,_Ne_w_Je_r_s_e ....... Y ________ _
1981
Inspectors:
L.
- orrholm,
~m./M1)*
W. M. Hill, Jr., Resident Reactor Inspector
Approved By:
Uwa/_/lJl~
E. G. Greenman, Chief, Reactor Projects Section No. 2A,
Projects Branch No. 2, DRPI
SEP 2 8 1981
date
SEP 2 8 1981
date
_1of6if;
d'ate
Inspection* Summary:
. ';:'..
Inspections* on* August* 4 *..;*September *14; 19.81 * (Comolnea *Report* Numbers* 50.:.212/81.:..23
and 50-311/81.:..lll
Unit 1 Areas* Irisaected: Routine inspections 5y the resident inspectors*#of plant
operations incl u mg tours of the facil i'ty; conformance wi.tfl Technical Specifica-
tions and operatfng parameters; log and record review; reviews of licensee events;
and followup on prevfous inspection items.
The inspection involved 68 inspector-
flours 5y the resident NRC inspectors.
Results:
Two items of noncompliance were identified (Failure to comply with
Technical Specification requirements prior to mode changes - paragraph 6; failure
to perform periodic survei.11 ances - paragraph 3}.
Unit 2 Areas Ins~ected: Routine inspections by the resident inspectors of plant
startup testing including tours of the facility; conformance with 1 icense require- .
ments and Technical Specifications; and reviews of licensee events; and followup
on previous inspection items.
The inspection involved 84 inspector-hours by the
resident NRC inspectors.
Results: Three items of noncompliance were identified (Failure to comply with
Technical Specification requirements prior to mode changes - paragraph 6; failure
to comply with a license': condition - paragraph 7, failure to perform periodic
surveillances- paragraph 3).
( 8111050602 811007
1
1
PDR ADOCK 05000272
G
\\
I
Report Nos. 50-272/81-23 and 50-311/81-21
DCS Nos *
050272-810910
050272-810902
050272-810810
050272-810414
050272-810404
050272-810422
050272-810220
050272-810426
050272-810505
050272-810506
050272-810429
050272-810512
050272-810513
050272-810514
050272-810518
050272-810501
050272-810520
050272-810521
050272-810526
050272-810528
050272-810616
050272-810612
050272-810616
050272-810621
050272-810715
050272-810624
050272-810628
050272-810703
050272-810626
050311-810810
050311-810822
050311-810829
050311-810831
050311-810902
050311-810903
050311-810911
050311-810419
050311-810407
050311-810423
050311-810426
050311.,8l0427
050311-810501
050311-810505
050311-810430
050311-810515
050311-810517
050311-810519
050311-810520
050311-810522
050311-810523
050311-810603
050311-810526
050311-810530
050311-810601
050311-810603
050311-810616
050311-810606.
050311-810608
050311-810623.
050311-810608
050311-810609
050311-810612
050311-810618
050311-810707
050311-810423
050311-810622
050311-810623
050311-810625
050311-810626
050311-810625
050311-810627
050311-810701
' .
~
I
L
.. DETAILS
I. Persons Contacted
J. Driscoll, Chief Engineer
L. Fry, Station Operating Engineer
J. Gallagher, Assistant Maintenance Engineer
S. LaBruna, Maintenance Engineer
H. Midura, Manager - Salem Generating Station
L. Miller, Station Performance Engineer
F. Schnarr, Reactor Engineer
R. Silverio, Assistant to the Manager
J. Stillman, Station QA Engineer
R. Swetnam, Radiation Protection Engineer
The inspector also interviewed and talked with other licensee personnel
during the course of the inspections including management, clerical,
maintenance, operations, performance and quality assurance personnel.
2.
Status of Previous Inspection Items
(Closed) Follow Item (272/80-32-04).
Inclusion of IE Circular 80-18
considerations relative to design changes in EDD-1.
The inspector
reviewed EDD-1, Operational Design Change Control, Revision 3,
approved on March 30, 1981. All considerations listed in IEC
80-18 have been incorporated both by reference and by inclusion
of the Circular in the procedure.
The inspector had no further
questions on this item.
(Closed) Unresolved Item (272/81-14-01}. Technical Specifications for
mechanical snubber surveillance.
By correspondence dated August
10, 1981, the licensee requested a change to the Technical Speci-
fications regarding testing of mechanical snubbers.
The proposed
format appears consistent with Standard Technical Specifications
on this subject. The inspector had no further questions on this
item.
(Closed) Noncompliance (272/81-01-03). Failure to post in accordance with
10 CFR 19. Revision 5 to Administrative Procedure 16, Posting
of Regulatory Material, was issued on July 16, 1981. This pro-
cedure establishes responsibilities and official bulletin boards
along worker access routes for the posting of material in accor-
dance with 10 CFR 19. The inspector also confirmed that an item
of noncompliance relating to radiological health identified in
Inspection Report 50-272/81-22 had 6een appropriately posted.
The
inspector f:iad no further questions on this item.
lCl osed }_
Fo 11 ow Item (311/81-10-01}.
Review deta i 1 s of April 23, 1981
safety injection. This event, reported as LER 50-311/81-45/99X
is reviewed in this report. The inspector had no further questions *
(Closedl
(Open).
SITE
3
Noncompliance (272/80-06-02}. Exceeding steady state power limit.
The inspector reviewed Operating Memorandum 6, Guidelines For
Operation at 100 % Power, dated June 23, 1981.
Instructions for
maintaining average power and absolute limits for indicated thermal
power are provided.
The guidelines are consistent with NRC direc-
tion provided in this area.
No recurrence of this item has been
identified. The inspector had no further questions.
Noncompliance (272/80-31-01}. Failure to comply with administra-
t1ve procedures with respect to procedure periodic reviews. The
inspector noted that licensee activities to revise and upgrade
procedures are continuing with most in the final stage of prepara-
tion and review.
The inspector further noted that the requirements
of Administrative Procedure 3, Document Control Program, were not
being met. Specifically, on-the-spot changes are required, by
AP-3, to be incorporated into a revision within six months.
The
licensee stated that this is an unreasonable interval and that a
revision will be initiated to AP-3 in support of a more workable
system. Resolution of this interval is a new unresolved item
(272/81-23-07}.
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 10, May 21, 1980;
AP-6, Operational Incidents, Revision 6, February 22, 1979;
AP-13, Control of Lifted Leads and Jumpers, Revision 4, February
11, 1980;
Operations Directive Manual; and,
AP-15, Safety Tagging Program, Revision 1, November 21, 1980.
o. Shift logs and operating records were reviewed to verify that:
.
--.
~
Control room log sheet entries are filled out and initialled;
Auxiliary log sheets are filled out and initialled;
Log entries involving abnormal conditions provide sufficient
detail to communicate equipment status, lockout status, correction
and restoration;
Log book reviews are befog conducted fiy tfie staff;
4
Operating orders do not conflict with Technical Specification
requirements;
Incident reports detail no violation of Technical Specification
LCO or reporting requirements; and, -
Logs and records were maintained in accordance with Technical
Specifications and the procedures in 3.a above.
c.
__ T_he review included examination of the following plant shift logs
and operating records and discussions with licensee personnel:
Log No.
- Control Room Daily Log, August 4 - September 14, 1981
Log No.- 6 - Primary Pl ant Log, August 4 - September 14, 1981
Log No. 7 - Secondary Plant Log, August 4 - September 14, 1981
Log No. 8 - Unavailable Equipment Status Log, August 4 - September
14, 1981
Night Orders, August 4, 1981 - September 14, 1981
Lifted Lead and Jumper Log - All active
Tagging Requests - All active
Nonconformance Reports for July 1981
d.
In reviewing control room logs, the inspector noted that the control
room narrative log in both units had recorded overhead annunciator D-40,
Rod Deviation and Sequence, as inoperable for some time. Technical
Specification 4.1 .3.2.l states,
11 *** when the Rod Position Deviation Monitor
is inoperable, ... compare the demand position indication system and the
rod position indication system at least once per four hours.
11
Review of
the console reading sheets indicated that this special surveillance was
being accomplished by recording rod position every four hours, instead
of every eight as is the routine.
However, the inspector identified
several instances during one week period of review in which an interval
of eight hours had elapsed without recording rod position. During this
period, the alarm was inoperable as indicated by the narrative log.
The
following failures to conduct this surveillance were identified and the
time frame exceeded~the 25% extension allowed:
Unit l
Unit l
Unit 2
Unit 2
September 10, 1981
September 2, 1981
August 29, 1981
September l, 1981
0800 to 1600
1600 to 2400
1600 to 2400
0800 to 1600
- 5
The above failures constitute apparent noncompliance with a Technical
Specification surveillance requirement (272/81-23-01) (311/81-21-01).
The inspector had no further questions relative to logs reviewed.during
this inspection period.
4.
Pl ant Tour
a.
b.
During the course of the inspection, the inspector made observations
and
(1)
(2)
(3)
(4)
(5)
(6)
(7)
(8)
(9)
(10)
( 11)
( 12)
( 13)
The
conducted multiple tours of plant areas, including the following:
Control Room (daily)
Relay Rooms
Auxiliary Building
'.Vital Switchgear Rooms
Turbine Building
Yard Areas
Radwaste Building
Penetration Areas
Control Point
Site Perimeter
Fuel Handling Building
Containment
Guard House
following determinations were made:
Monitoring instrumentation: The inspector verified that selected
instruments were functional and demonstrated parameters within
Technical Specification limits.
Valve positions. The inspector verified that selected valves were
in the position or condition required by Technical Specifications
for the applicable plant mode.
This verification included exam-
ination of control board indication and field observation of valve
positions (Charging/Safety Injection, Auxiliary Feedwater, and
Containment Spray Systems).
6
Radiation Controls. The inspector verified by observation that
control point procedures and posting requirements were being
followed.
Plant housekeeping conditions. The inspector observed that house-
keeping was generally acceptable.
Any cluttered or littered area~'*
for which maintenance was not in progress, was brought to the
attention of the plant management or operating staff.
Fluid leaks.
No fluid leaks were observed which had not been
identified by station personnel and for which corrective action
had not been initiated, as necessary.
Piping vibration.
No excessive piping vibrations were observed
and no adverse conditions were noted.
Selected pipe hangers and seismic restraints were observed and
no adverse conditions were noted.
Equipment tagging. The inspector selected plant components for
which valid tagging requests were in effect and verified that the
tags were in place and the equipment in the condition specified.
By frequent observation through the inspection, the inspector
verified that control room manning requirements of 10 CFR 50.54
(kl and the Technical Specifications were being met.
In addition,
the inspector observed shift turnovers to verify that continuity
of system status was maintained. The inspector periodically
questioned shift personnel relative to plant conditions and their
knowledge of emergency procedures.
Releases.
On a sampling basis, the inspector verified that appro-
priate documentation, sampling, autnorization, and monitoring
instrumentation were provided for effluent releases.
Fire protection. The inspector verified that selected fire exting-
uishers were accessible and inspected on schedule, that fire alarm
stations were inspected on schedule, that fire alann stations were
unobstructed and that cardox systems were operable.
Technical Specifications. Through log review and direct observa-
tions during tours, the inspector verified compliance with Technical
Specifications including Limiting_ ~q_n~tt\\lor}~ _f9r_Jlp~r~J.ion_ (LCO ~s) ._
i~~e r
0 ! ~~wi~~p~~~~~~~~r~o~~~~ n~!~~;~~~p!~~-i~~~~,l~~~~*~~~~c tfii~ l p~~~:r
shutdown margin, offsite power.
In addition, the inspector conducted
periodic visual checks of protective instrumentation and inspection
of electrical switchboards to confirm availability of safeguards
equipment.
7
Security. During the course of these inspections, observations
relative to protected and vital area security were made, in-
cluding access controls, boundary integrity, search, escort,
and badging.
d.
The following acceptance criteria were used for the above items:
Technical Specifications
Operation Directives Manual
Inspector Judgement
e.
The inspector had no further questions relative to tours made during
this inspection.
5.
Review of Periodic and Special Reports
upon receipt, periodic and special reports submitted by the licensee
pursuant to Technical Specifications 6.9.1 and 6.9.2 were reviewed by
the inspector.
This review included the following considerations:
The report included the information required to be reported by
NRC requirements;
Test resu'lts and/or supporting information were consistent with
design predictions and performance specifications;
Planned corrective action was adequate for resolution of identified
problems; and,
Determination whether any information in the report should be
classified as an abnormal occurrence.
Within the scope of the above, the following periodic reports were reviewed
by the inspector:
Unit 1 Monthly Operating Report - July 1981
Unit 2 Monthly Operating Report - July 1981
No unacceptable conditions were identified *
8
6. Licensee Events
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:
UNIT 1
81-37/03L
81-38/03L
- --
81-39/0lT
81-40/03L
- --
81-41/03L
81-42/03L
- --
81-43/03L
81-44/03L
81-45/03L
81-46/03L
- --
81-4J/03L
- --
81-48/03L
- --
81-49/03L
81-50
81-51/03L
- --
81-52/03L
RHR Pump Seal Failure
Loss of Audible Fire Alarm
Service Water Leak in Containment - No. 13 CFCU
Reactor Coolant System Leakage - 1 PS 8
Reactor Trip System Instrumentation - No. 11 Pressurizer
Level Channel 3 Inoperable
Containment Air Lock 100
1 Elevation Outer Door Inoperable
Reactor Trip System Train B Inoperable During Surveillance
Test
Source Range Channel - N-31 - Inoperable
13 Containment Fan Coil Unit - Inoperable
Containment Air Lock 100 1 Elevation Outer Door Inoperable
Containment Air Lock 100 1 Elevation Outer Door Inoperable
Containment Air Lock (Elevation 100) - Cam Roller Inoperable
Not Used
Impingement of One Dead Shortnose Sturgeon, Acipenser
- *srevirostrum Lesueur on tne Circulating Water System (CWS)
12A Intake Tras~ Bars
Engineered Safety Feature Actuation System Instrumentation
18 SEC Malfunction
UNIT 1
- --
81-53/03L
- --
81-54/0lT
- --
81-55/03L
81-56/03L
- --
81-57 /03L
81-58/04L
- --
81-59/0lT
- --
81-60/99XO
81-61/03L
- --
81-62/03L
81-63/03L
- --
81-64/0lT
81-65/03L
81-66/03L
81-67/03L
81-68/03L
UNIT 2
81-06/03L
81-07/03L
81-08
81-09/03L
e
9
lB Diesel - Inoperable - Water Jacket Leak
Volume Control Tank Level Control System
Loss of #12 & 22 Station Power Transformers
Reactor Coolant System - Leakage Detection System Failure
due to Loss of IF and lG Buses
Containment Air Lock 100
1 Elevation Outer Door Inoperable
due to Misalignment
Failure to Submit 1980 Radiological Report on Time
Check Valves 11MS46 and 13MS46 - Inoperable
Fire Suppression Water System - No. 2 Diesel Fire
Suppression Pump Inoperable
Pressurizer Over Protection System - Channel 1 Inoperable
Residual Heat Removal System Valving Misalignments
Individual Rod Position Indicators - 201 and 205 -
No. 15 Containment Fan Coil Unit - Service Water Leak
No. 15 Containment Fan Coil Unit (CFCU_t - Inoperable due
to Failed Flow Transmitter
No. 13 Containment Fan Coil Unit (CFCU} - Inoperable due
to Failed Flow Transmitter
No. 13 Containment Fan Coil Unit (CFCU} - Inoperable due
to Failed Flow Control Valve
Fuel Handling Building Fire Alarm Zone - Inoperable
Pressurizer Overpressure Protection System (POPS}
Failure of No. 22 Auxiliary Feedwater Pump to Start
Not Used
Boron Injection Tank - Out of Specification
UNIT 2
81-10/03L
- --
81-ll/03L
- --
81-12/03L
81-13/03L
- --
81-14/0lT
81-15
- --
81-16/03L
- _...;.
81~17/03L
- --
81-18/03L
81-19/03L
81-20
81-21/03L
- --
81-22/03L
- --
81-23/0lT
81-24
81-25
81-26/03L
81-27/03L
- --
81-28/03L
81-29/03L
- --
81-30/03L
81-31/03L
81-32/03L
81-33
10
Loss of Service Water Loop (21)
Containment Air Lock 100' Elevation Outer Door Inoperable
Containment Air Lock 130' Elevation Inner Door Inoperable
Containment Type B and C leak Rate Test Results Exceeded
Technical Specifications
Incorrect Cable Routing (Fire Protection)
Not Used
2B and 2C Diesel Generators Inoperable in Mode 5
Pressurizer Overpressure Protection (POPS) - Valve
Leakage
Overpressure Protection System - Channel 2 Activation
Rod 1B4 Position Indication System - Indicator Failure
Not Used
Auxiliary Feedwater Storage Tank (AFWST) Low Level
Movable Control Assemblies - Rod 2DLI, Dropped
Main Steam Isolation Valves - Closure Time Failure
Not Used
Not Used
2C Diesel Generator Overcurrent Trip
Valve 2PR6 Open/Closed Indication Malfunction
Onsite Power Distribution Systems - Trip of 2B40'.1 Breaker
Reactor Coolant Pump Flow Channel 2 Inoperable - Loop 22
No. 21 Auxiliary Feedwater Pump - Inoperable
Reactor Coolant System - Leakage Detection System Inoperable
Boron Injection Tank - Out of Specification
Not Used
,-------
UNIT 2
- --
81-34/03L
- --
81-35/0lT
81-36/03L
81-37/03L
81-38/0lT
- --
81-39/03L
- --
81-40/03L
81-41/03L
- --
81-42/03L
- --
81-43/03L
81-44/04T
- --
81-45/9_9){
... -
81-46/03L
81-41 /03L
8l-48/03L
81-49/03L
- --
81-50/03L
81-51/03L
81-52/03L
- --
81-53/03L
11
Boric Acid Storage Tank - Out of Specification
Check Valves 21MS46 and 23MS46 - Inoperable
Pressurizer Pressure Channel I - Inoperable
Radiation Monitoring System - Containment Particulate
Channel - 2R11A - Inoperable
Service Water Leakage in Containment - No. 21 CFCU
Missed Surveillance - Channel Check of Post Accident
Monitoring Instrumentation (SP(0)4.3.3.7)
Chemical and Volume Control System Letdown Line Weld
Failure
Residual Heat Removal Suction Relief Valve Actuation and
Failure to Fully Reset
Reactor Coolant Temperature Less Than Minimum Required
For Power Operation
Overpressure Protection System Inoperable Due to Valve
Leakage
Inadvertent Contamination of On-Site Storage Area
Inadvertent Safety Injection During Unit Cooldown
Individual Rod Position Indication - 1B4 - Inoperable
Individual Rod Position Indication - 1SB2 - Inoperable
Individual Rod Position Indication - 2SA2 - Inoperable
Individual Rod Position Indications - 1SB4, 2SB2, and
2SB4 - Inoperable
2B Diesel Generator Start Time Failures
Boron Injection Tank - Out of Specification
Process Monitors 2Rl1A and 2R12A - Inoperable
Missed Surveillance - Containment Airlock
' ;
b.
Onsite Licensee Event Followup
(1) For those LERs selected for onsite followup (denoted by asterisks
in detail paragraph 6a), the inspector verified the reporting require-
ments 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 accordan~e with
Technical Specification limits. The following findings~**reM.te to the
LERs reviewed on site:
Unit 1
81-39/0lT
--
81-64/01 T
--
81-41/03L
81-43/03L
81-47/03L
81-48/03L
81-49/03L
81-57 /03L
--
81-52/03L
--
81-53/03L
These containment service water leaks resulted from
failure of a dissimilar metal weld.
As noted in LER
50-272/81-64/0lT, a design change is nearly completed
which makes piping modifications to remove dissimilar
metal welds.
The change is completed on No. 13 CFCU.
This report documents unidentified reactor coolant
system leakage in excess of 1 gpm.
Investigation iden-
tified the leakage source to be packing on instrument
valve 1 PS 8.
Having identified the source as a packing
leak the applicable limit was 10 gpm and the observed
leakage rate of 2.69 was acceptable. The packing was
subsequently replaced.
The licensee has concluded that air lock failures are
attributable to improper operation.
As already docu-
mented, a program to strengthen air lock door components
has been completed.
In addition, annual requalification
training for all station personnel who may have occasion
- to use a containment airlock includes specific guidance
for careful operation of the doors. This area will be
the subject of continuing review to ensure that licensee
actions are effective in improving air lock reliability.
This event is discussed in NRC Inspection Report 50-272/
81-12, detail 13.b. Licensee actions to correct SEC
voltage sensitivity will be followed as unresolved item
81-12-07.
This report details failure of a diesel jacket water
piping nipple due to improper use of the piping as
support by an individual. The report does not address
corrective action for the apparent cause identified.
Tfte licensee stated that a supplemental report will be
submitted when effective corrective action is identified *
This item is unresolved (272/81-23-02).
--
81::.::54/01 T
--
81-55/03L
--
81-59/01 T
--
81-60/99){
--
81-62/03L
~---------
l
13
This report details a potential for damage to centri-
fugal charging pumps under a postulated failure of
the VCT level control system.
Such failure has been
identified by Westinghouse as not meeting the require-
The licensee has taken
action to ensure that operators are aware of this type
of failure and are alerted to other instrumentation
available to confinn VCT level. The inspector had no
questions relative to action taken by the licensee to
preclude failure of the centrifugal pumps.
However,
the question of compliance with regulatory requirements
relative to instrumentation and control design is un-
resolved pending further review (272/81-23-03).
This event is discussed in NRC Inspection Report 50-
272/81-12. The inspector confirmed that licensee
preventive measures have been taken to preclude testing
of the Unit 3 breaker in a condition which would jeopar-
dize the 13 KV ring bus. These measures included a
caution tag on the breaker and modification of Operating
Memo-5 which relates to operation and testing of the
Unit 3 output breaker. The inspector had no questions
with respect to interim measures taken, noting that
DCR 1 SC-0577 has been initiated to precipitate a
permanent design solution.
This event is detailed in NRC Inspection Report 50-272/
81-14.
The licensee is conducting monthly radiography
to confinn the integrity of the check valve internals
while seeking a permanent solution to the apparantly
high wear rate on the valves. This item is unresolved
pending completion of the licensee's analysis and sub-
mittal of a supplementary report (272/81-23-04).
No. 2 Fire Pump failed to start during a surveillance
test due to failure of the diesel engine.
The engine
was sent to the vendor for repair and overhaul, and
was returned to service in late August.
As committed
in the LER, the licensee will submit a supplemental
report detailing the nature of the failure. This i tern
is unresolved (272/81-23-051.
Following inspection of No. 11 RHR Pump and Heat Ex-
changer for loose camera parts on June 20, 1981 (refer-
ence NRC Inspection Report 50-272/81-14} the system
was aligned for return to service and an ASME Section
XI pump test performed at approximately 8:45 p.m.
\\
14
8l-62/03L
The results of the test were acceptable but, when re-
(continued)
viewed later, appeared to have resulted in lower than
expected differential pressure. At 1:48 a.m. on June
21, -the unit was heated from Mode 4 to Mode 3 and went
critical at 9:45 a.m.
At noon on June 21, the pump
test was repeated to resolve the question of low dif-
ferential pressure and at that time the heat exchanger
inlet valve, 11 RH 14, was found closed. With this
valve in the closed position, the No. 11 train of
Unit 2
81-ll/03L
--
81-12/03L
--
81-14/0lT
--
81-16/03L
_,,,_
81-JJ /03L
81-43/03L
RHR was rendered inoperable. The valve was immediately
opened and a satisfactory pump test conducted.
The
LER notes that the verification valve lineup required '
by procedure had not been conducted.
The inspector
further noted that mode changes with less than the
required available ECCS subsystems are prohibited by
Technical Specification 3.0.4. This event contributes
to an apparent item of noncompliance (272/81-23-06).
These reports detail failures of Unit 2 airlocks.
Corrective action described for Unit l LER 81-43/03L
above involves station-wide training and should be
effective in reducing the number of failures in both
units.
This event is discussed in NRC Inspection Report 50-
311/81-11. Correct cable routing was verified by the
inspector shortly after tne error was discovered and
corrected.
A stuck: relay in the diesel generator field flash
circuitry resulted in two inoperable diesels while in
Mode 5.
Tne inspector noted tnat there appeared to be
no periodic surveillance of relays to preclude this
type of occurrence.
Tfie licensee stated that a program
of relay cleaning will be initiated. This item is
unresolved (311/81-21-021..
These reports detail inoperability of the Pressurizer
Overpressure Protection (POPS) system while below 312°F
due to leakage through valve 2 PR 2 or 2 PR 47.
Despite
repairs during the most recent cooldown, the POPS and/or
PORV continued to leak and have 5een isolated. The
licensee establishes the vent path required by Technical
Specifications within the time limits stated.
LER 81-
43 states that repairs will be made during the next re-
fueling outage. Supplements to both LER
1s will be
submitted following repair. This item is uriresol_ved
pending va_lve _r!=!pair and receipt.,,and)re;vJew of'* the sup-
p 1Ein:iental reports" ( 3ll /81-23-03}:
- .
. . .
15
81-18/03L
This inadvertent actuation of the POPS system was
caused by miscommunication between an instrument
foreman and the shift operators. Based on review
of procedures and discussions with personnel, the
inspector detennined that adequate controls to prevent
recurrence were in place.
81-22/03L
This event is discussed in NRC Inspection Report 50-
311/81-11. The inspector had no further questions.
81-23/0lT
Testing of main steam isolation valves revealed that
under the condition of a hydraulic opening in progress,
rapid shutting of the valves could not be accomplished
in less than 5 seconds as required by Technical Speci-
fications. Design changes were made in both units to
ensure that closure under isolation conditions could be
accomplished within the required time under all circum-
stances.
81-28/03L
During a safeguards loading sequence (Safety Injection
on June 3, 19811, starting of the second CFCU on the
B vital bus caused the 2B4D feeder breaker to tr1p.
It was later determined tnat a relay modification, al-
ready installed on Unit 1, to raise the overcurrent
setting had not been made on Unit 2. The modification
was subsequently made to the 2B and 2C bus feeder
oreakers. Failure to detect this problem earlier can
also be attributed to a prior practice of testing the
accident load"ing sequence witfiout all equipment avail-
aole. Remaining equipment was tested later as it was
returned to service. As a result, the breaker over-
current setting may not fiave oeen challenged prior to
this event. The licensee stated that such testing
practices will 5e prohibited during future surveillance
testing of this type. The inspector had no further
questions on tfiis item.
81-30/03L
Th-is event is detailed in NRC Inspection Reports 50-311/
81-11 and 81-13.
81-34/03L
During a re-fill of the Boron Injection Tank (BIT)
following safety injection, failure to completely drain
tfie Bii resulted in dilution of the Boric Acid Storage
Tank (BASTI to less than 20,000 ppm.
The procedure did
not include a method for verifying the BIT completely
drained prior to placing it on recirculation with the
_BAST.
The inspector reviewed Revision 1 to OI II-4!3,05,
Flastiing, Draining, and Filling the Boron Injection Tank,
approved oy SORC on August 31, 1981. The procedure now
calls for use of a local sample point (ISJ6) to confinn
that the BIT is empty.
The inspector had no further
questions.
16
--
81-35/0lT Corrective action for this item is similar to that
described for Unit 1 LER 81-59/0lT above and will be
evaluated concurrently.
--
81-39/03L Technical Specification 4.3.3.7 requires a monthly
channel check of post-accident monitoring instrumen-
tation. The licensee had prepared a surveillance
procedure to accomplish this activity but had not
scheduled the test in the Inspection Order system. *
As a result, the test conducted on April 22, 1981
became overdue on May 30, 1981.
The oversight was
discovered on June 6, 1981 and reported as a missed
surveillance in this LER.
During the period June 2-4,
1981, several plant startups from Mode 3 to Mode 1
were conducted.
In accordance with Technical Speci-
fication 4.0.3, failure to conduct required surveillance
constitutes inoperaoility of the applicable component.
As a result, tnesemode changes contribute to apparent
noncompliance with Technical Specification 3.0.4
(311/81-21-04}.
--
81-40/03L The failed CVCS vent line was capped.
Not stated in
the report, but confirmed by the inspector, was the
nondestructive examination of several similar config-
urations in tne system.
No other prob*1 ems were identi-
fied.
--
81-42/03L During heavy feeding of steam generators at approximately
6 percent power, Tave decreased below the minimum for
criticality (541oFl.
The excursion lasted for 13 minutes,
less than the 15 minutes permitted by the Technical
Specification Action s*tatement.
The inspector confirmed
oy ooservation during several startups that operators
remain acutely aware of temperature 1 imits and take
prompt action to maintain $pecifi ed parameters.
--
81-45/99X This safety injection was caused by failure to closely
monitor steam generator differential pressures during
a cooldown.
Tne inspector confirmed that procedures for
cooldown using atmospheric relief valves (MS 10) caution
tne operator to maintain differential pressures less than
JOO psid.
In ooserving cool downs, the inspector has
ooserved tfiat differential pressure information is readily
avalla6le and dlligently monitored-';. by operators. The
inspector had no further questions.
--
81-50/03L
--
81-53/03L
17
This report details a series of failures of 28 Diesel
Generator to start within the 10 seconds required by
Technical Specifications. Fol lowi.ng arnumber of other
repairs, it was determined that one of the turbo-boost
air supply solenoids had failed and its redundant
counterpart was wired incorrectly. A subsequent check
of all diesel solenoids has been conducted and no
wiring errors discovered.
The licensee is reviewing
all diesel failures to start and will submit a supple-
mental report. This item is unresolved (311/81-21-05).
Technical Specification 4.6.1.3.b requires a containment
air lock pressure test prior to establishing containment
integrity if it has 6een broken. This requirement had
not been incorporated into procedures for Unit 2 and
does not exist for Unit 1. On June 21-22, 1981 the unit
changed from Mode 5 to Mode 1 following a maintenance
outage during wni'di containment work required that con-
tainment integrity oe 6rol<en.
The air lock test was
not conducted prior to mode change.
When the omission
was discovered on July 1, 1981 a successful test was
conducted. Failure to demonstrate operaoility of the
air lock in accordance with the schedule specified in
Technical Specifications contributes to apparent non-
compliance with Technical Specification 3.0.4 (311/81-
21-041.
c. Tfie following Unit 1 LERs, reviewed above, were submitted later than the
ti!ne required 5y Technical Specifications; 81-40/03L, 81-45/03L, and
81-52/03L. Late reporting was tne saoject of an item of noncompliance
identified in NRC Region I correspondence to the licensee dated June 24,
1981.
Tne events listed aoove occurred prior to July 15, 1981, the date
oy wFLidi tfie 1 icensee committed to nave corrective action in pl ace to
preclude late reporting. Effectiveness of licensee actions in this regard
will be evaluated during a subsequent inspection.
Tfie inspector had no further questions with respect to LERs reviewed.
J. * Tull *Power * L itense *Condi tfons (Un it* 2}
On January14, 1981, April 28, 1981, and May19, 1981, the NRC staff, in-
cluding the Senior Resident Inspector, 5riefed the Commission on the status
of Salem Unit 2 and the proposed licensing action to authorizec:'operation in
excess of 5 percent rated thermal power.
The full power license for Salem
.Unit 2 was issued on May 20, 1981 and contains several conditions to be met
prior to given dates or events. Tfle inspector reviewed a number of these
items to determine status of implementation.
The following comments apply
to tne areas *revi"ewed (Numbers refer to paragraph references in the full
power 1 icense)_:
18
2.C.(lO)(d), (f), and (h) Fire barrier and cable wrap program.
The licensee has stated that this program is complete. A special
NRC inspec.tion has been scheduled to confirm the engineering
analysis and field installation of fire barriers. During this
inspection period, a sampling of items, to be completed by July
31, 1981, in accordance with paragraph (d), was inspected. Based
on walkthrough inspections, the additional cable wrap, barrier
extensions and smoke detectors appear to be in place.
One excep-
tion was ooserved. Paragraph 2.C.(10)(9) requires the licensee
to wrap one of the redundant power cables from the diesel generators
located in the fuel oil storage tank room.
Neither of the redundant
feeds llad 15een wrapped when ooserved on September 1, 1981.
When
identified to the licensee, a fire watch was immediately established
and work initiated to wrap the B diesel power cable. Subsequent
licensee review of the design change package indicated that the work
had not oeen accomplished as directed. A review of the package by
tne licensee indicated that all other work called for by the design
cfiange llad oeen completed. Failure to wrap the diesel power cable
constitutes apparent noncompliance with a condition of facility
operating license DPR-75 (311/81-21-06).
Complete confirmation of
tfiis license condition will Be included in a subsequent inspection.
2.C.(l2lCompletion of Preoperational Testing of Circulating Water
System. Startup Procedure 34, Circulating Water System and Sodium
_Bypocfil orfte System was completed on July 24, 1981, with the excep-
tion of a data set on Circulator 228 wnile operating (taken on July
27, 1981[ and completely automatic operatfon of the sodium hypo-
clll orite system due to incompatioility of installed equipment with
current EPA standards (operating 111ainly under operator control).
Operatfon of tile Circulating Water System was verified as proper
prfor to power ascensfon aoove 50 percent. The inspector reviewed
tile completed startup procedure and discussed the test with cognizant
engtneers*.
No items of noncompliance were identified.
2.C.(141 Waterhammer test. As discussed in NRC Inspection Report
50-311/81-1~ tfie feedwater hammer test was conducted on July 23, 1981
prior to operation a5ove 90 percent rated thermal power.
No unaccep-
ta51 e piping displacement was recorded.
2.C.(241(c1liil 48 Hour endurance run of Auxiliary Feedwater Pumps.
Tfie*motor driven pumps (No. 21 and 22) were tested on August 18, 1980.
Testing of tile steam-driven pump (No. 23) was completed on August 14,
1981. Tfie inspector reviewed test results which indicated acceptable
nearing temperatures and vioration readings for the pumps.
Testing
was completed prior to operation at 100 percent power as required
oy tfie license. The licenseers 60-day report will be reviewed when
received (311/81-21-07}.
Tfie inspector had no further questions with respect to license conditions
reviewed.
19
8. Operating Events
a. Unit 1
(ll At 2:38 p.m. on August 10, a reactor/turoine trip occurred due to
a low level in No. 12 Steam Generator fo 11 owing a loss of No. 11
Main Feed Pump. Maintenance department investigation was unable
to identify any pro5lem witfi tfie pump. First out indication was
not availa5Je due to a 5lown fuse on tfie feed pump indicator panel.
Power operation resumed at 4:25 a.in. on August 11 with No. 12
"Main Feed Pump.
No. 11 Feed Pump was run for several hours to
monitor its operation and was returned to service with no apparent
prolilems.
J:i *.. Unit 2
(1L At 4:15 a.-m. on August 10, Salem Unit 2 was shut down (Mode 3) to
repat.r pacR.ing leaks on two valves in the Reactor Coolant System
11RTD
11 .oypass line. Tfie total leal<.age was 11.5 GPM as measured by
an RCS inventory oalance.
Power operation resumed at 4:25 a.m.
on August 11.
(21 The plant tripped from 100 percent power at 8:40 p.m. on August 22,
.due to low-low level in No. 24 Steam Generator caused oy loss of
No. 22 Steam Generator feedwater pump.
Tne pump tripped on indicated
low suctfon pressure attriouted to secondary plant fluctuations or speed
control oscillation. following system trou6leshooting, the unit re-
srn:ned power operation at 10:35 a.in. on August 23.
DL A *moisture carryover test using a NA-24 tracer at 100 percent power
was conducted on August 29. At 10:45 a.m. on August 29, power was
reduced to 25 percent 5ecause of steam generator cation conductivity,
res.al ting from a condenser tuoe l eaK..
Bl owdown was initiated at
1:11 p.-in. on August 29*, to restore steam generator cfiemistry. Blow-
down to tfie non-radwaste 5asin witn known quantities of radioactive
fsotope in tlie steam generator was treated as a liquid release with
tne Jilowdown radiation monitors serving as the rel ease monitors.
Estimated activity released over the period was 0.5Ci, and is con-
ststent witfi Tecfinfcal Specification Appendix B limits *
.. GI On August 31, steam generator chemistry and activity levels were
reduced suffi.ciently to support increase in power.
At 4:41 a.m. on
August 31, No. 22 Steam Generator Feed Pump (SGFP) tripped on
indicated overspeed witfi tfie plant at 85 percent power.
The resulting
transient caused a tur5ine trip/reactor trip on high level in No. 21
Steam Generator at 4:47 a.m. Trouolesfiooting was perfonned on No.
22 SGFP controls and power operations resumed at 6:24 p.m.
20
(5)
The 100 percent generator trip test was conducted at 9:24 a.m.
on September 2, followed by the last natural circulation test,
boron mixing and cooldown. * All systems functioned normally on
the trip with no safety injection initiated. The unit returned
to power operation at 7:19 a.m. on September 3.
(6) At 4:10 p.m. on September 3, and following a review of the cali-
bration data (Startup Tests) for both channels of steam flow
indication for No. 21 steam generator, the licensee declared the
instruments inoperable and shutdown the unit. The square root
of the steam flow dp was non-linear above 92 percent power, with
moisture carryover considered to be the cause. The instruments
were recalibrated for their linear response range (below 92 per-
cent). Power operations resumed at 7:45 a.m. on September 4.
The licensee has imposed an operating limit of 90 percent power
pending an evaluation and resolution of the non-linear steam flow
indications and high moisture carryover experienced by No. 21
steam generator above 92 percent.
l7) At 7:40 p.m. on September 11, the licensee conducted a second
steam generator moisture carryover test at 100 percent power
using NA-24 tracer. Operation above 90 percent during this test
caused both channels for No. 21 steam generator steam flow in-
dications to trip their respective bistables.
The results are being evaluated by the licensee and the NSSS
vendor.
Preliminary review of the data confirm the results of
the first carryover test. At 9:50 a.m. on September 12, the test
was complete, and power was reduced to 25 percent because of high
cation concentration in the steam generators.
At 10:49 p.m. on September 13, reactor power was returned to 90
percent when the blowdown system was returned to service in order
to reduce the cation concentration in the steam generators.
The inspector had no further questions relative to operating events listed
above or the licensee's corrective action.
9.
Surveillance Activities
The inspector observed licensee's performance of the following surveillance
procedures:
a.
PD 18.1.001 Solid State Protective System - Periodic Test
b.
SP(O) 4.0.5 P - Auxiliary Feedwater System - pumps 21 and 23
c. SP(nl 4.0.5 V - Auxiliary Feedwater System - valves (partial)
(No. 2MS132)
21
The inspector confirmed that testing was performed in accordance with
adequate procedures, test instrumentation was calibrated, limiting
conditions for operations were met, removal and restoration of the
affected components were properly accomplished, and test results conform
with Technical Specifications and procedure requirement and were reviewed
by personne*1 otfler tflan the indiv'fdual directing the test. Any deficiencies
noted were reviewed and resolved by the personnel of the responsible de-
partment. The personnel performing the surveillance activities were
knowledgeable of the system and the test procedures. The inspector con-
cluded that they were qualified to perform the tests and had no further
questions regarding the performance of these surveillance activities.
10. System Operation and Review
The inspector conducted a walk down of the accessible portion of the
Auxiliary Feedwater System of Unit 2. Licensee's drawing 205336, No. 2
Unit Auxiliary Feedwater Piping Diagram, revision 6, dated April 11, 1980,
was used for this walk down.
The inspector found no unexpected conditions
with the exceptions of a two inch pipe which had been welded between the
drain line on the discharge piping of No. 23 Auxiliary Feedwater Pump and
the drain line of the corrnnon suction header from the Auxiliary Feedwater
Storage Tank.
This pipe had been installed by design change, 2 ET-0760,
to support a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> test run of the No. 23 Auxiliary Feed Pump which was
performed as part of the Startup Test Program.
The drawing in use by the
inspector had been appropriately marked,
11DCP ISSUED.
11
When the Operational
Design Change Status Notice Report, dated August 1, 1981, was reviewed,
several design change packages were indicated as affecting this drawing,
but the change package installing tfle two inch pipe and the current status
was not indicated on the status report. This item is unresolved pending
further review (311/81-21-08).
11.
Fuel Receipt
The licensee received 56 new fuel assemblies which are scheduled for in-
stallation during unit 1 refueling outage later this year. The fuel
assemblies were removed from their shipping casks, inspected, and placed
in the new fuel storage racks (dry). The inspector witnessed unloading,
transfer, and storage of 10 assemblies. The inspector confirmed that
acceptable procedures were approved and were in use. The licensee stated
that no unacceptable conditions were identified in any of the 58 fuel
assemblies.
The inspector reviewed the shipping documents and receipt
inspection documents. Daily checks of the cranes had been performed, and
weekly checks of the overload cutout had been performed within seven days
of fuel movement.
The inspector noted no unsatisfactory conditions and
had no furtfler questions regarding the receipt of new fuel.
22
12. Startup Testing
13.
14.
The inspector observed portions of startup testing conducted during this
period. Observations included the following; correct procedure in use,
crew requirements, test prerequisites, plant conditions, calibrated test
equipment, adherence to procedure, coordination, data collection, and
preliminary data review.
Portions of the following tests were observed:
SUP 82.1 - Load Swing Tests
SUP 82.2 - Large Load Reduction Tests
SUP 82.8 - NSSS Acceptance Test
SUP 82.9 - Generator Trip from 100 percent Power
Comments relative to observations of the above Startup Test are provided
in a separate NRC Inspection Report 50-311/81-22.
Operator License Examination
On September 4, 1981, the Senior Resident Inspector administered a written
partial re-examination to two applicants. The inspector was present for
the entire examination and identified no unacceptable conditions or practices.
Unresolved Items
Areas for which more information is required to determine acceptability are
considered unresolved. Unresolved items are contained in Paragraphs 3, 6,
7 and 10.
15. Exit Interview
At periodic intervals during the course of this inspection, meetings were
held with senior facility management to discuss inspection scope and
findings.