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NUCLEAR REGULATORY COMMISSION
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e° ARLINGTON.TEXAS 76011
September 30, 1985
Docket: 50-267/CAL 85-04 Wan COUNTY CUMMIS 1U;dlfic
° CEIVE
OCT 31985
Public Service Company of Colorado
ATTN: 0. R. Lee, Vice President
Electric Production GREELEY. COLO.
P. 0. Box 840
Denver, Colorado 80201-0840
Gentlemen:
My letter of July 19, 1985, confirmed your commitments and authorized the
restart and operation of Fort St. Vrain (FSV) at a power level no greater than
15%. Your letters of August 30, September 10, 11, and 23, 1985, requested
authorization to operate FSV at a power level no greater than 8% for a period
of up to 45 days, not to extend beyond November 30, 1985.
We have reviewed and evaluated your additional submittals and we have
verified acceptable implementation of established procedures. We find it
acceptable for FSV to be operated up to 8% of Rated Thermal Power to
remove moisture from the reactor coolant system. This authorization
is based on the availability and capability of the liner cooling system to be
satisfactorily placed in service and operated to remove core decay heat
subsequent to a postulated High Energy Line Break (HELB). Our evaluation of
the acceptability of operating FSV in this mode is contained in the enclosed
Safety Evaluation.
In authorizing the restart and operation of FSV, we understand that you have
committed to the following actions:
1. With exception of the FSV plant power level , all other commitments
contained in the July 19, 1985 Confirmatory Action Letter continue to
apply.
2. Unless authorized by the NRC to proceed, PSC will restrict the FSV power
level to no greater than 8% of the licensed thermal power level .
3. Power operation of the FSV plant to remove moisture utilizing nuclear
heat will be restricted to maximum of 45 days from the date of this
letter, not to extend beyond November 30, 1985.
CERTIFIED MAIL - RETURN RECEIPT REQUESTED
851175
Public Service Company of Colorado - 2 - September 30, 1985
4. Verification that the operation of the liner cooling system can be
established totally by manual actions.
5. The required procedures have been provided for manual operation
and verification of operation of the liner cooling system and the
required operations personnel training in the use of the procedures
for manual operation and verification of the liner cooling system
has been completed prior to assuming shift duties.
If your understanding of these commitments is not the same as ours, please
contact this office immediately at (817) 860-8100.
Sincerely,
�0✓Robert D. Martin
uu Regional Administrator
Enclosure:
As stated
cc: see next page
cc:
C. K. Millen, Senior Vice President
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. 0.
Box 85608 San Diego, California 92138
Kelly, Stansfield R O'Donnell
Public Service Company Building
550 - 15th Street, Room 900
Denver, Colorado 80202
Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Executive Staff
Assistant, Electric Production
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 194
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
SAFETY EVALUATION
CONCERNING LIMITED LOW POWER OPERATION OF
FORT ST. VRAIN
DOCKET NO. 50-267
1.0 INTRODUCTION AND BACKGROUND
By Confirmatory Action Letter dated July 19, 1985, the staff authorized
the restart of Fort St. Vrain (FSV) and its operation up to 15 percent of
Rated Thermal Power. This allowed the licensee to operate the reactor
to remove moisture, but limited operation until certain equipment
qualification issues were resolved. Specifically, the licensee had to
complete aging and equipment operability time studies required for
equipment qualification under 10 CFR 50.49.
By letter dated August 20, 1985, the licensee reported to the staff on the
progress of his equipment qualification program. New problems discovered
by the licensee included:
- Equipment items subject to submergence
- Unqualified taped electrical splices
- Contaminated or rusted electrical junction boxes
- Discrepancies in equipment model numbers in comparison to test
reports
- Additional equipment potentially exposed to a harsh environment
In discussions between the licensee and the staff, the licensee agreed
not to operate the reactor until this situation could be reviewed and
approved by the staff.
By letter dated August 30, 1985, the licensee requested authorization from
the staff to operate FSV at low power for a limited period to continue
moisture removal . Specifically, the licensee requested NRC approval to
operate at power levels not to exceed 8 percent (of full power) for a
period not to exceed 45 days or extend beyond November 30, 1985. The
licensee supplemented this request with additional information by letters
dated September 10, 11 and 23, 1985.
2.0 EVALUATION
2. 1 Desirability of Continued Moisture Removal .
The licensee has committed to and initiated multiple efforts to minimize
moisture ingress into the FSV reactor. The licensee considers it prudent
to continue to reduce the level of moisture in the reactor vessel .
- 2 -
The staff has reviewed the question of moisture ingress into the FSV
reactor in its October 1984 Assessment Report (Reference 1) . The staff
has concluded in this report that reducing moisture ingress would improve
the reliability of overall plant operations and potentially improve the
performance of the control rod drive mechanisms. The staff required the
licensee to implement modifications to reduce the frequency and severity
of further moisture ingress events.
Hence, the staff concludes that further lowering of the reactor moisture
level is consistent with the staff's previous findings on moisture ingress,
and therefore should be continued.
2.2 Qualification Status Of Equipment For Decay Heat Removal
The licensee has stated that a number of equipment qualification problems
at FSV remain unresolved. The licensee is working to resolve these issues
as rapidly as possible. The staff has found that pending resolution of
these issues , equipment necessary to maintain forced Helium circulation
cooling can not be considered qualified. Therefore, no credit can be
taken for automatic operation of this equipment.
The licensee has stated, however, that operation at low power for a 45 day
period before November 30, 1985, would rely only on manual operation of
the liner cooling system for decay heat removal following a high energy
(steam) line break accident; and that operation of this equipment in the
manual mode is not vulnerable to an accident environment.
The liner cooling system has two redundant loops and is in continuous
operation during reactor operation. The normal supply for liner cooling
water is the Reactor Plant Cooling Water System (System 46) pumps.
These pumps are located on Level 8 in the Reactor Building and east of
the 4A wall . Therefore, with the relatively low Reactor Building
temperature profiles at 8 percent power, it is expected that these pumps
would not be affected during the steam line rupture. The remainder of
System 46 is also expected to survive the accident due to the relatively
low temperature profile.
However, even if all electrical items in System 46 fail , fire water can
be manually valved into the liner cooling system. This would be a once
through system with the water being supplied by the fire water pumps
located outside of the building. In this case, no qualification is
required of any electrical items. Portions of the liner cooling
system exposed to a harsh environment can be manually operated.
The licensee has committed to have in effect, prior to restart, emergency
procedures which reflect the dependence on the liner cooling system as
the decay heat removal path. Operators would verify operation of the
liner cooling system following a high energy line break accident. This
verification would be by direct local observation of equipment operation
and valve lineup. Operation would be periodically reverified. In the
event that the operators conclude that the liner cooling system is not
- 3 -
performing its design function, they would restore its operation through
manual actions. The licensee is not taking credit for any automatic
equipment operation in assuring operation of the liner cooling system.
Instead, the licensee is relying entirely on appropriate manual actions.
Reliance on manual operation for the liner cooling system is acceptable
because the accident analysis indicates that there is adequate time
(over 12 hours) in which to gain access to the necessary components and
take action to assure that liner cooling is restored. This time limitation
is based on maintaining the reactor vessel concrete below 400°F, the level
at which the concrete could be expected to lose a significant fraction of
its compressive strength. The licensee has stated that failure of other
equipment does not affect the independent operation of the liner cooling
system and that even in the event of equipment failures, ample time margin
is available to enable operators to perform the manual operations necessary
to realign the system or to effect any necessary repairs.
In summary, the licensee has shown that the liner cooling system can be
used to remove decay heat, and has committed to implement, prior to
restart, emergency procedures to assure that:
1) The operator will not be misled by failure of equipment, and,
2) required equipment is verified to be available and operable.
2.3 Accident Consequences
The licensee has stated that plant operation at 8 percent power level greatly
reduces the requirements for decay heat removal . At this power level , the
average core temperature would be 640°F. Decay heat calculations would
yield a maximum temperature of 1337°F 100 hours after reactor shutdown,
without any liner cooling. With one loop of liner cooling, fuel temperatures
would reach 1281°F about 100 hours after reactor shutdown. These
temperatures are well below the 2400°F fuel temperature reached in normal
operation, and the 2900°F fuel temperature at which fission product release
begins to occur. Under these conditions, there would be no significant
fuel failure or fission product activity released. Operation of the liner
cooling system would be assured by the licensee's procedures following a
reactor shutdown.
The staff performed independent calculations concerning decay heat
removal for Fort St. Vrain when operated at the 8 percent level . The
staff's calculations find that the maximum fuel temperature reached is
about 1350°F after 3 days with only the liner cooling system in operation
(Reference 2). The staff calculated peak temperature with the liner
cooling system in operation is higher than the temperature calculated by
the licensee assuming no liner cooling because of different assumptions
about coolant flows and core temperatures in low power operation. The
staff's calculations are more conservative because the minimum allowable
flow and higher core temperatures were assumed as initial conditions. It
should also be noted that both the core heatup and the cooldown produced
by the liner cooling system take place very gradually. Hence, temporary
interruptions of the liner cooling system operation do not greatly affect
fuel temperatures.
- 4 -
The staff concludes that the liner cooling system provides adequate decay
heat removal for FSV at the 8 percent power level and that operation in
this mode has no unacceptable accident consequences.
2.4 Return To Power Operation
The licensee has stated that the need to remove moisture from the FSV reactor
will potentially delay eventual return to power operation. FSV technical
specifications (Sections LCO 4.2.10 and 4.2.11) limit the allowable moisture
levels in reactor primary coolant. The licensee states it must continue
moisture removal operation or delay the return of FSV to power operation.
Eventual startup of the plant will also require the licensee to resolve
problems with equipment qualification. Hence, if the reactor moisture
level is reduced now, and maintained in that condition, reactor power
operation can begin promptly when the other problems are resolved.
Additionally, maintaining the reactor at low moisture levels during this
period minimizes the potential for the moisture to adversely affect the
exposed reactor systems, thus delaying the plant's return to power operation.
The staff concludes that return to power operation of FSV is in the public
interest and that lower power operation to allow moisture removal should be
permitted.
3.0 CONCLUSION
The staff finds the continued low power operation of FSV to remove moisture
is desirable. This is part of the licensee's overall effort to reduce
the undesirable effects of moisture ingress on the reactor. The staff also
finds that limited operation to remove moisture now would potentially
allow Fort St. Vrain to return to power operation at an earlier date after
current equipment qualification problems are resolved. In low power
operation, the staff finds that decay heat can be safely removed by the
liner cooling system. The licensee has committed to put in place prior
to restart appropriate procedures and training to assure that the liner
cooling system can be placed in operation manually and to assure that
its operation will not be affected by unresolved problems with equipment
qualification. The NRC will verify acceptable implementation of these
procedures by inspection prior to restart. Thus, the staff concludes that
Fort St. Vrain can be safely operated for a period not to exceed 45 days
and at a power level not to exceed 8 percent of full power. However,
operation at low power levels is only acceptable for a period of 45 days
before November 30, 1985. After this period, plant operation must be in
accordance with Commission policy as stated in GL 85-15.
Date: September 27, 1985
Principal Contributors:
K. Heitner, DL
T. King, DL
A. Masciantonio, DE
References:
1. Preliminary Report Related to the Restart and Continued Operation of Fort
St. Vrain Nuclear Generating Station, October 1984.
2. Letter to T. L. King, NRC, from S. J. Ball , ORNL, on ORECA Analysis of
FSV 8% Power Severe Accident Sequences dated September 9, 1985
SSINS No. : 6835
IN 85-78
UNITED STATES
NUCLEAR REGULATORY COMMISSION As
OFFICE OF INSPECTION AND ENFORCEME lg11,gb^
WASHINGTON, D.C. 20555 F
i, ;rte
September 23, 1985 4,7,4O ���{/
0 J "/
IE INFORMATION NOTICE NO. 85-78: EVENT NOTIFICATION 4.4x4y n
<er /
c /
Pao.
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purposes:
This notice is being issued to revise guidance to power reactor licensees
regarding specific event notification information that should be provided to
the NRC Operations Center when reporting events in accordance with
10 CFR 50.72. This guidance supercedes that provided in IN 83-34 dated May 26,
1983. Suggestions contained in this information notice do not constitute NRC
requirements; therefore, no specific action or written response is required.
Description of Circumstances:
Significant events reported to the NRC Operations Center receive prompt
evaluation by NRC technical staff at headquarters and regional offices. To
assist in obtaining adequate information for evaluation, the worksheet has been
revised for the Operations Officers manning the NRC Operations Center. The
event notification worksheet has been provided for your information as an
attachment to this information notice. We recognize that this list is not all
inclusive, nor are all of its items applicable to each event. Rather, it lists
certain key items on the basis of past experience that are required for most
notifications. The checklist is intended to provide the licensee with the type
of information that should be provided to the NRC Operations Center when
reporting events.
It is suggested that copies of the worksheet be made available to employees
responsible for reporting event related information to the NRC to replace
copies of the worksheet provided in IN 83-34.
To prevent misunderstandings and reduce callbacks by the Operations Officer
as a result of incomplete information, the description should provide sufficient
detail for the Operations Officer to understand the event, including all system
interactions.
The NRC welcomes any recommendations for improvements to the event reporting
process.
8509190432
IE 85-78
September 23, 1985
Page 2 of 2
No specific action or written response is required by this information notice.
If you have any questions regarding this matter, please contact the Regional
Administrator of the appropriate NRC regional office or the NRC Headquarters
incident response contacts listed in this notice.
wakE Jordan, Director
Divisi of Emergency Preparedness
and gineering Response
Office of Inspection and Enforcement
Technical Contacts: Don Marksberry, IE
(301) 492-4156
Ray Priebe, IE
(301) 492-4333
Attachments:
1. Event Notification Worksheet
2. List of Recently Issued IE Information Notices
Attachment I
IN 85-78
S. NUCLEAR REGULATORY COMMISSION OPERATIONS CENTER
EVENT NOTIFICATION WORKSHEET September 23, 1985
Paae 1 of 2
TELEPHONE NUMdLR (For c.2.1 back)
vortrICATIGr TIME FACILITY OR ORGANIZATION UNIT CALLER'S NAME
CVENT CLASSIFICATION Y N EVENT CATEGORY INITIATION SIGNAL CAUSE OF FAILURE
GENERAL EMERGENCY REACTOR TRIP/SCRAM MECNNICAL
SITE AREA EMERGENCY ESF ACTUATION ELECTRICAL
ALTRT ECCS ACTUATION PERSONNEL ERROR
UNUSUAL EVENT SATETY INJECTION mow PROCEDURE INADE UACY
•r 50.72 NON-EMERGENCY LCO ACTION STATEMENT OTHER
)••r SLGURITY/SATECUARDS OTHER
•
TAaNSPORTATfON EVENT I SYSTEM: EVENT TIME ZONE EVENT MOTH DAY
DATE
OTHER: •
COMPONENT:
EVENT DESCRIPTION
•
•
•
p wER ep.OR 'O EVENT (a): Did all systems function as required ? YES I: NO, Explain aDcve.
Y-=R_OR NODE Anything 'unusual' or not understood ? NO I: yE$, Explain above.
O_'SID= :.GENCY OR PERSONNEL CORRECTIVE ACTION(S)
. _E.D BY LICENSEE
-Elr1
¢re--rti•' f YES P•,^, ..LL SE
{` CT OFEPATION T:LL COPR£C:10N: r 7:.fA-E RESTART:
I FFE:S =' E:.SE l[ 1 ADDITIONAL INFORMA':ICN ON SACK
Attachment 1
IN 85-78 .
ADDITIONAL INFORMATION FOR RADIOLOGICAL RELEASES Septei er 23, 1935
Page 2 of 2
LIQUID RELEASE PLANNED SOURCE(S) :
GASEOUS RELEASE UNPLANNED
RELEASE RATE (Ci/sec) : EST TOTAL ACTIVITY (Ci):
R_;.-EASE DURATION EST TOTAL IODINE (Ci):
T.S. LIMITS : GRAB SAMPLE
MONITOR READING.
Areas evacuated ? Y N List below Personnel exposed/contaminated Y N Describe below
Plant Health Physics backup requested ? Y N Note: Only if T.S. exceeded or actual contamination
ADDITIONAL INFORMATION
ADDITIONAL INFORMATION FOR REACTOR COOLANT or STEAM GENERATOR TUBE LEAKS
SUDDEN OR LONG TERM DEVELOPMENT ? START DATE : START TIME :
1 LEAK RATE PRIMARY COOLANT MONITOR READINGS
ACTIVITY
LEAK VOLUME SECONDARY COOLANT CONDENSER
ACTIVITY
T.S. LIMITS:- MAIN STM. LINE :
SG BLOWDOWN
LIE OF SAFETY RELATED EQUIPMENT NOT OPERATIONAL :
SPECIAL ACTIONS TAKEN BY LICENSEE (If any) :
Attachment 2
IN 85-78
September 23, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-77 Possible Loss Of Emergency 9/20/85 All power reactor
Notification System Due To facilities holding
Loss Of AC Power an OL or CP
85-76 Recent Water Hammer Events 9/19/85 All power reactor
facilities holding
an OL or CP
85-75 Improperly Installed Instru- 8/30/85 All power reactor
mentation, Inadequate Quality facilities holding
Control And Inadequate Post- an OL or CP
modification Testing
85-74 Station Battery Problems 8/29/85 All power reactor
facilities holding
an OL or CP
84-70 Reliance On Water Level 8/26/85 All power reactor
Sup. 1 Instrumentation With A facilities holding
Common Reference Leg an OL or CP
85-73 Emergency Diesel Generator 8/23/85 All power reactor
Control Circuit Logic Design facilities holding
Error an OL or CP
85-72 Uncontrolled Leakage Of 8/22/85 All power reactor
Reactor Coolant Outside facilities holding
Containment an OL or CP
85-71 Containment Integrated Leak 8/22/85 All power reactor
Rate Tests facilities holding
an OL or CP
85-70 Teletherapy Unit Full 8/15/85 All material
Calibration And Qualified licensees
Expert Requirements (10 CFR
35. 23 And 10 CFR 35. 24)
85-69 Recent Felony Conviction For 8/15/85 All power reactor
Cheating On Reactor Operator facilities holding
Requalification Tests an OL or CP
OL = Operating License
CP = Construction Permit
SSINS No. 6835
IN 85-77
UNITED STATES
NUCLEAR REGULATORY COMMISSION CI Co"T mamma
OFFICE OF INSPECTION AND EM
20555 ; E u E V E
WASHINGTON, D. C. 20555
September 20, 1985 OCT 1''985
IE INFORMATION NOTICE NO. 85-77: POSSIBLE LOSS OF EMERGENef9WITICMIO
SYSTEM DUE TO LOSS OF AC POWER
Addressees:
All holders of a nuclear power plant operating license (OL) or a construction
permit (CP).
Purpose:
This information notice is provided to alert licensees to the possibility that
modifications to plant telephone systems may result in a change in the vulnera-
bility of the Emergency Notification System (ENS) and other plant telephones to
losses of ac power. It is expected that recipients will review the information
for applicability to their facilities and consider actions, if appropriate, to
preclude similar problems occurring at their facilities. However, suggestions
contained in this information notice do not constitute NRC requirements;
therefore, no specific action or written response is required.
Description of Circumstances:
On January 8, 1984, the Palisades Nuclear Plant interrupted its offsite power
supply for maintenance activities. Subsequently all ac power from the emergen-
cy diesel generators was lost. (This event is described in detail in Informa-
tion Notice No. 84-42. ) As a result of the ac power problems, all onsite
telephones were rendered inoperable for approximately 3 hours, except for two
offsite-powered pay telephones. Loss of the ENS and normal communications
significantly hampered the notification process.
On May 7, 1985, Quad Cities Unit 1 was operating at 90% power. The Unit 2
auxiliary transformer was inadvertently shorted while the unit was shut down
with its associated emergency diesel generator out for maintenance. This
caused the loss of offsite ac power to Unit 2 and a voltage transient in Unit 1
that subsequently caused that unit to scram about 15 minutes later. Unit 1
retained offsite ac power. One division of Unit 2 was promptly powered by
autostart of the swing emergency diesel generator, and the other division was
powered within about 20 minutes by crosstie to a 4kV bus of Unit 1. When the
licensee attempted to notify the NRC Emergency Operations Center over the ENS,
the circuit repeatedly disconnected. The Quad Cities plant also was unable to
receive incoming calls from the NRC over commercial telephone lines.
8509180410
anIMT=i / 1/n(
IN 85-77
September 20, 1985
Page 2 of 3
These incidents indicate that the provisions of IE Bulletin 80-15 were not
maintained at the affected facilities at the times of the events. The bulletin
required verification that all ENS station packages that use onsite ac power
were connected to a safeguards instrumentation bus backed up by automatic trans-
fer to batteries and an inverter or an equally reliable power source. At the
time the bulletin was issued, both plants had ENS packages that were powered by
the local telephone company, making them independent of ac power sources at the
plant sites.
Discussion:
The installation of the ENS requires a station package that operates on 110
Vac. In some cases, the station package is located at the local telephone
company which supplies the required power for normal operation and emergency
power for operation during abnormal situations. However, in many cases, the
ENS package is located at the site and is served by ac power provided by the
plant.
Earlier incidents involving loss of offsite power led to losses of emergency
notification capabilities at the Davis-Besse facility on October 15, 1979, and
at the Indian Point Unit 2 on June 3, 1980. These incidents prompted the
issuance of IE Circular 80-09 and IE Bulletin 80-15. The bulletin contained a
list of those stations with ENS packages powered by the telephone company and a
list of those stations with ENS packages powered at the plant site. At that
time, both the Palisades plant and the Quad Cities plant had ENS packages with
power supplies provided by the telephone company.
Subsequent changes to provide additional circuits in the telephone system at
the Palisades plant resulted in the ENS and commercial telephone system packag-
es being powered at the plant site. Power was supplied from a bus supported by
an emergency diesel generator. However, the modification was not controlled
within the licensee' s formal modification process and was thus completed
without formal review. The modified ENS power supply was not backed by batter-
ies and an inverter, as previously provided by Bulletin 80-15, and was not
independent of the station' s commercial telephone service as reflected in the
licensee' s Emergency Plan.
During the incident at Palisades on January 8, 1984, the unit was intentionally
powered from a single emergency diesel generator on the 1C 2400-V bus to allow
isolation of a faulty switchyard breaker. The unit was defueled, and the other
diesel was inoperable due to maintenance. When the running diesel subsequently
overheated and tripped, the station was without ac power with the exception of
preferred ac. Although some other buses were repowered by offsite ac within
an hour, difficulties in closing the breakers to the 1C and 1E 2400-V buses
resulted in the extended loss of all telephones except for two pay telephones
powered by the telephone company. The 1E bus was repowered after about 3 hours
by successfully closing the breaker to the offsite source. This provided par-
tial restoration of the telephone service. However, the ENS telephones on the
1C bus were not restored for 6 hours, when they were finally jumpered to an
energized source.
IN 85-77
September 20, 1985
Page 3 of 3
At the Quad Cities plant, the local telephone company abandoned the copper wire
cables that were in use in 1980 and installed a fiber optics communications
system in its place. Because the fiber optics cable does not provide for
electrical power transmission, the fiber optics package at the plant had to be
provided with an onsite power source. Similarly, the site package for the ENS
had to be shifted to onsite power. The licensee powered the fiber optics
package from an instrument bus in Unit 2 and the ENS from an instrument bus in
Unit 1. These buses are supported by emergency diesel generators, but the
power supplies to the communications packages are not backed up by batteries
and an inverter in accordance with Bulletin 80-15.
During the event on May 7, 1985, when the Unit 2 instrument bus powering the
fiber optics package lost power, both the ENS and normal PBX telephones became
inoperable. The Unit 1 bus supporting the ENS package remained powered by an
offsite ac source through the switchyard, but could not communicate through the
unpowered fiber optics system. Once power was restored to the Unit 2 bus
through a crosstie, the ENS circuit repeatedly disconnected as the licensee
attempted to make emergency notifications.
These events illustrate the need for careful review of changes to plant
telephone equipment to ensure that the reliability of the ENS is not compro-
mised. In those cases where offsite communications power that is supplied by
the telephone company is replaced by an onsite power source, it is important
to consider the reliability of the power sources for all segments of the ENS
transmission path. Those plants that already supply the ENS from an onsite
safeguards instrumentation bus should be aware that the introduction of a fiber
optics connection by the local telephone company still may compromise the ENS
if the plant-end fiber optics package is not similarly powered.
No specific action or written response is required by this information notice.
If you have any questions regarding this matter, please contact the Regional
Administrator of the appropriate NRC regional office /or this office.
Edward L rdan, Director
Divisio Emergency Preparedness
and E neering Response
Office of Inspection and Enforcement
Technical Contacts: S. Long, IE
(301) 492-7159
R. Priebe, IE
(301) 492-4333
Attachment: List of Recently Issued IE Information Notices
Attachment 1
IN 85-77
September 20, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-76 Recent Water Hammer Events 9/19/85 All power reactor
facilities holding
an OL or CP
85-75 Improperly Installed Instru- 8/30/85 All power reactor
mentation, Inadequate Quality facilities holding
Control And Inadequate Post- an OL or CP
modification Testing
85-74 Station Battery Problems 8/29/85 All power reactor
facilities holding
an OL or CP
84-70 Reliance On Water Level 8/26/85 All power reactor
Sup. 1 Instrumentation With A facilities holding
Common Reference Leg an OL or CP
85-73 Emergency Diesel Generator 8/23/85 All power reactor
Control Circuit Logic Design facilities holding
Error an OL or CP
85-72 Uncontrolled Leakage Of 8/22/85 All power reactor
Reactor Coolant Outside facilities holding
Containment an OL or CP
85-71 Containment Integrated Leak 8/22/85 All power reactor
Rate Tests facilities holding
an OL or CP
85-70 Teletherapy Unit Full 8/15/85 All material
Calibration And Qualified licensees
Expert Requirements (10 CFR
35.23 And 10 CFR 35. 24)
85-69 Recent Felony Conviction For 8/15/85 All power reactor
Cheating On Reactor Operator facilities holding
Requalification Tests an OL or CP
85-68 Diesel Generator Failure At 8/14/85 All power reactor
Calvert Cliffs Nuclear facilities holding
Station Unit 1 an OL or CP
OL = Operating License
CP = Construction Permit
Hello