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NRC NEWS
U. S. NUCLEAR REGULATORY COMMISSION |
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E-mail: opa@nrc.gov |
No. 96-55 FOR IMMEDIATE RELEASE
(Thursday, April 4, 1996)
NOTE TO EDITORS:
The Nuclear Regulatory Commission staff has issued the
attached letter to Northeast Utilities (NU) requiring certain
information, under oath or affirmation, before restarting its
Millstone 3 nuclear power plant in Connecticut.
The NRC is taking this action based on preliminary
results of a special inspection that has identified design
and equipment deficiencies described in an enclosure to the
letter. The deficiencies are similar in nature to those at
Millstone 1 and 2, both of which are in shut down.
Millstone 3 is required to submit to the NRC, seven days
before restarting from its current refueling outage, actions
taken to assure future operation will be conducted in
compliance with its operating license, NRC regulations and
final safety analysis report for the plant.
Millstone 1 and 2 also are required to provide similar
information and operating assurance to the NRC before being
able to restart.
#
Attachments:
As stated
.
April 4, 1996
Mr. Robert E. Busch
President - Energy Resources Group
Northeast Utilities Service Company
P.O. Box 128
Waterford, CT 06385
Dear Mr. Busch:
On March 7, 1996, the NRC issued to Northeast Utilities (NU)
a letter describing its review of an internal NU document,
"ACR 7007 - Event Response Team Report" (7007 Report), dated
February 22, 1996 and requesting additional information on
NU's actions and plans to address the conclusions of the 7007
Report as it pertains to Millstone Unit 3 and Haddam Neck.
At the time we issued that letter, we did not have a recent
inspection history or findings with regard to Millstone Unit
3 that revealed design deficiencies similar in number and
nature to those of Millstone Units 1 and 2, but the 7007
Report identified the potential for similar configuration
management conditions at Millstone Unit 3.
Since the March 7, 1996 letter, the NRC has initiated a
special inspection at Millstone Unit 3. Based on the
preliminary results of the portion of the special inspection
conducted to date, we have identified programmatic issues and
design deficiencies at Millstone Unit 3 that are similar in
nature to those at Millstone Units 1 and 2. Examples of some
of the deficiencies found in the special inspection are
described in the enclosure to this letter. In addition, you
have recently identified design deficiencies in the auxiliary
feedwater containment isolation valves and the recirculation
spray system that have existed for more than 10 years.
These findings raise substantial questions as to whether the
Millstone Unit 3 facility is being operated and maintained in
conformance with the updated final safety analysis report
(UFSAR), license conditions, and Commission regulations.
Therefore, the NRC requires additional information to be
submitted pursuant to Section 182a of the Atomic Energy Act
of 1954, as amended, and 10 CFR 50.54(f) in writing, under
oath or affirmation, to determine whether or not the license
for Millstone Unit 3 should be suspended, modified, or
revoked. The information is to be submitted no later than 7
days prior to Millstone Unit 3 restart (prior to criticality)
from its current outage and is to describe actions taken to
ensure that future operation of Millstone Unit 3 will be
conducted in accordance with the terms and conditions of the
Millstone Unit 3 operating license, the Commission's
regulations, including 10 CFR 50.59, and the Millstone Unit 3
UFSAR. This request for information supercedes our previous
request of March 7, 1996.
Your submittal should describe actions taken to ensure that
design and configuration control deficiencies at Millstone
Unit 3 have been identified, and have been evaluated with
regard to plant operability, the existence of unreviewed
safety questions, and reportability. Your submittal should
also address your corrective actions. In particular,
seriously degraded conditions must be reported pursuant to 10
CFR 50.72(b)(2)(i) and 50.73(a)(2)(ii). Further, prior to
restart of Millstone Unit 3, you must resolve, to the NRC's
satisfaction, the issues raised in the examples described in
the enclosure, the Auxiliary Feedwater valve issues, and the
recirculation spray system matter. We also recognize that
ongoing activities by your staff as well as the NRC staff may
identify additional issues which warrant resolution prior to
restart.
In accordance with 10 CFR 2.790 of the NRC's "Rules of
Practice," a copy of this letter and your responses will be
placed in the NRC Public Document Room, the Gelman Building,
2120 L Street, NW., Washington, DC, and in the local public
document room located at the Learning Resources Center, Three
Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and in the temporary public document
room located at the Waterford Library, Waterford,
Connecticut.
Sincerely,
(original signed by W. T. Russel)
William T. Russell, Director
Office of Nuclear Reactor Regulation
Enclosure: Preliminary Inspection
Findings
. PRELIMINARY INSPECTION FINDINGS
Inoperability of the turbine-driven AFW pump during startup
and shutdown
In response to plant information report (PIR) 3-94-060, "AFW
High Energy Line Break Concerns," dated March 15, 1994, the
licensee developed a procedure to isolate the turbine-driven
auxiliary feedwater (AFW) pump whenever either of the motor-driven AFW pumps were used for steam generator water level
control (low power operations during startup or shutdown).
Technical Specification 3.7.1 requires operability of both
motor-driven and the turbine-driven auxiliary feedwater pumps
whenever the plant is in modes 1, 2, and 3.
Failure to remove plastic shipping plugs from Rosemount
transmitters
The NRC team observed two Rosemount transmitters, 3CHSFT139
and 3CCPFT12B, with plastic shipping plugs installed, and
transmitter 3GWSFT84 open at the alternate conduit entry
point, contrary to the manufacturer's instructions to
environmentally seal these openings with a stainless steel
plug. The licensee subsequently identified two safety-related and 94 non-safety related transmitters that required
corrective action based on partial inspection. The licensee
had not completed inspection of the transmitters inside
containment.
Failure to correct degraded non-safety battery
The NRC team found that the maintenance history on battery
301D-1 demonstrated marginal capacity because the monthly
surveillances were preceded by equalizing charges on eight
occasions in 1995. Although the battery is classified as
non-safety, procedure AOP-3563 cautions the operators that a
manual reactor trip is required upon the loss of the
associated DC bus. This DC bus is relied upon for the
operation of all four steam generator atmospheric dump
valves, and to provide control power for the 4160 volt
emergency bus feeder breakers.
Inadequate Control of Modification of the service water
system
A temporary modification was installed in 1990 that bypassed
the automatic service water booster pump start on high
discharge temperature in the MCC/rod control area room cooler
ducts. The safety evaluation performed was inadequate in
that it did not address the substitution of a manual action
for an automatic function and it did not address the deletion
of an automatic start feature. In addition, the licensee did
not maintain a special instruction in an alarm response
procedure required by the original bypass as a compensatory
measure.
Potential introduction of foreign material into the
containment sump
Test flange rings were installed on the recirculation system
(RSS) suction piping penetrations in the containment sump by
a modification in June 1995 (PDCR 3-94-0162). The licensee
installed bolts "snug tight" in the flange holes to prevent
thread damage and minimize the accumulation of water and
debris in the test ring bolt holes during plant operation.
This practice unnecessarily introduces material into the
containment sump that could, if the bolts are dislodged or
improperly controlled, be pulled into the suction of the RSS
pumps and result in pump damage.
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