[Federal Register Volume 62, Number 200 (Thursday, October 16, 1997)]
[Notices]
[Pages 53840-53845]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-27417]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-271]
Vermont Yankee Nuclear Power Corporation, Vermont Yankee Nuclear
Power Station; Issuance of Partial Director's Decision Under 10 CFR
2.206
Notice is hereby given that the Director, Office of Nuclear Reactor
Regulation, has taken action with regard to a Petition dated December
6, 1996, submitted by Mr. Jonathan M. Block, on behalf of the Citizens
Awareness Network, Inc. (CAN). The Petition requested evaluation of
certain Memoranda enclosed with the Petition relating to the Vermont
Yankee Nuclear Power Station operated by the Vermont Yankee Nuclear
Power Corporation (Licensee) to see if enforcement action is warranted.
The first document enclosed with the Petition is a CAN Memorandum
dated December 5, 1996, that reviews information presented by the
Licensee at an enforcement conference held on July 23, 1996, involving
the minimum flow valves in the Vermont Yankee residual heat removal
system. CAN raises a concern that the corrective action taken by the
Licensee in opening these valves may have introduced an unreviewed
safety question with regard to containment isolation.
The second document enclosed with the Petition is a CAN Memorandum
dated December 6, 1996, that contains a review of certain Licensee
Event Reports (LERs) submitted by the Licensee in the latter part of
1996. Various issues are presented, such as fire protection, tornado
protection, thermal protection for piping lines, equipment operability,
and equipment testing. On the basis of its analysis of the LERs, CAN
reaches certain conclusions regarding the performance of the Licensee
and actions that should be taken.
On the basis of these documents, CAN requests that the NRC
determine whether enforcement action is warranted pursuant to 10 CFR
2.206.
The Director of the Office of Nuclear Reactor Regulation has
granted the Petition in that the NRC staff has evaluated the majority
of issues and LERs raised in these Memoranda to see if enforcement
action is warranted based upon the information contained therein. The
conclusion of the evaluation is that no further enforcement action is
warranted for those issues and LERs that are closed. LERs which remain
open will be resolved through the normal inspection and enforcement
process and will be addressed in a Final Director's Decision after the
NRC staff has completed its evaluations. The reasons for the staff's
conclusions are provided in the ``Partial Director's Decision Pursuant
to 10 CFR 2.206'' (DD-97-25), the complete text of which follows this
notice and is available for public inspection at the Commission's
Public Document Room, the Gelman Building, 2120 L Street, NW.,
Washington, DC 20037, and at the local public document room located at
Brooks Memorial Library, 224 Main Street, Brattleboro, VT 05301. A copy
of the Decision will be filed with the Secretary of the Commission for
the Commission's review in accordance with 10 CFR 2.206(c) of the
Commission's regulations. As provided for by this regulation, the
Decision will become the final action of the Commission 25 days after
the date of issuance, unless the Commission, on its own motion,
institutes a review of the decision in that time.
Dated at Rockville, Maryland, this 8th day of October 1997.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
Partial Director's Decision Pursuant to 10 CFR 2.206
I. Introduction
On December 6, 1996, Mr. Jonathan M. Block, submitted a Petition to
the Office of the Secretary of the U.S. Nuclear Regulatory Commission
(NRC) pursuant to Section 2.206 of Title 10 of the Code of Federal
Regulations (10 CFR 2.206). The Petition was submitted on behalf of the
Citizen's Awareness Network, Inc. (CAN or Petitioner), and contained
two Memoranda from CAN. The first Memorandum enclosed with the Petition
is dated December 5, 1996. It reviews information presented by the
Vermont Yankee Nuclear Power Corporation (Licensee) at a predecisional
enforcement conference held on July 23, 1996, involving the minimum
flow valves in the residual heat removal (RHR) system at the Vermont
Yankee Nuclear Power Station (Vermont Yankee facility). CAN raises a
concern that the corrective action taken by the Licensee in opening
these valves may have introduced an unreviewed safety question with
regard to containment isolation.
[[Page 53841]]
The second Memorandum enclosed with the Petition is dated December
6, 1996, and contains a review of certain licensee event reports (LERs)
submitted by the Licensee in the latter part of 1996. Various issues
are presented, such as fire protection, tornado protection, thermal
protection for piping lines, equipment operability, and equipment
testing. On the basis of its analysis of the LERs, CAN reaches certain
conclusions regarding Licensee performance and actions that should be
taken. In the Petition, the Petitioner requested that the NRC evaluate
these documents, pursuant to 10 CFR 2.206, to see if enforcement action
is warranted based upon the information contained therein.
On February 12, 1997, the NRC informed the Petitioner in an
acknowledgement letter that the Petition had been referred to the
Office of Nuclear Reactor Regulation for the preparation of a
Director's Decision and that action would be taken within a reasonable
time regarding the specific concerns raised in the Petition.
II. Discussion
The NRC staff evaluation of these documents follows.
A. The Residual Heat Removal System
The first document enclosed with the Petition is a CAN Memorandum
dated December 5, 1996, that reviews information presented by the
Licensee at a predecisional enforcement conference held on July 23,
1996, involving the minimum flow valves in the Vermont Yankee RHR
system.1 The Vermont Yankee RHR system consists of two
loops. Each loop has two pumps that take suction from the suppression
chamber. Each pump has a minimum flow line equipped with a minimum flow
valve that returns flow to the suppression chamber. The RHR pumps start
automatically to cool the reactor in case of a loss-of-coolant accident
(LOCA). The minimum flow valves close to prevent flow from being
diverted from the reactor core to the suppression pool when flow is
being supplied from the RHR pumps to the reactor core, and open
automatically on high pump discharge pressure to protect the RHR pumps
if other valves between the RHR pumps discharge and the reactor core
are not yet open.
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\1\ Several statements in the December 5, 1996, Memorandum are
either unclear or incorrect. A single power supply failure does not
prevent RHR minimum flow valves in both loops from operating,
contrary to the statement on page 2 of the Memorandum. Minimum flow
valves in both loops will not remain open if a single power supply
failure occurs, contrary to the statement on page 3 of the
Memorandum. Also, on page 4 of the December 5, 1996, Memorandum, CAN
questions the remote manual closure capability of the minimum flow
valves. The minimum flow valves have remote manual closure and
opening capability, but the pump protection logic will override any
remote manual closure or opening signal.
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The Licensee discovered a vulnerability to single failure which
could prevent the minimum flow valves from opening to protect the RHR
pumps during a LOCA. To resolve this concern, the Licensee changed the
normal and failed positions of these valves from CLOSED to OPEN. The
Petitioner is concerned that the corrective action taken by the
Licensee in opening these valves may have introduced an unreviewed
safety question with regard to containment isolation.2 A
pipe break outside containment would breach primary containment with an
OPEN minimum flow valve.
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\2\ The NRC staff assumes Petitioner's reference to an
``unreviewed safety question'' is in the context of the NRC's
regulation 10 CFR 50.59, ``Changes, Tests, and Experiments''.
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This issue must be addressed in terms of the Vermont Yankee
facility licensing basis. The basic design for early boiling-water
reactors, including the Vermont Yankee facility which was reviewed and
accepted by the NRC, considered the piping of the RHR and Core Spray
(CS) Systems to be a closed extension of primary containment. Failure
of the passive pressure boundary (piping) of these systems during
either the short-term (injection phase) or long-term (recirculation
phase) course of a design-basis accident (DBA) was not a design basis
assumption. As a result, the RHR and CS suction and minimum flow lines
were not provided with containment isolation valves, or if valves were
provided in these lines, they were not provided for the purpose of
meeting containment isolation requirements and thus were not classified
as containment isolation valves. In most if not all cases, the
penetrations of concern in the older plants were originally provided
with at least one valve capable of performing the containment isolation
function, and these valves are periodically tested under inservice
testing (IST) program requirements. The Vermont Yankee minimum flow
valves can be remotely closed and are periodically tested under the IST
program.
For more recent facilities, emergency core cooling system (ECCS)
closed systems outside containment are required to have at least one
recognized isolation valve at each penetration. This is not the case
for the Vermont Yankee facility.
In view of the licensing criteria applicable to the Vermont Yankee
facility, maintaining the minimum flow valves of the RHR system in the
OPEN position is permitted and acceptable. The Vermont Yankee final
safety analysis report (FSAR) does not describe the minimum flow valves
as being in the CLOSED position, and placing these valves in the OPEN
position is not a change to the facility under the meaning of 10 CFR
50.59 and no unreviewed safety question is presented. For the above
reasons, no enforcement action is warranted with regards to this issue.
B. Licensee Event Reports
The second document enclosed with the Petition is a CAN Memorandum
dated December 6, 1996, that contains a review of several LERs
submitted by the Licensee in the latter part of 1996. Various issues
are presented, such as fire protection, tornado protection, thermal
protection for piping lines, equipment operability, and equipment
testing. On the basis of its analysis of the LERs, CAN reaches certain
conclusions regarding Licensee performance and actions that it believes
should be taken. First, CAN requests that the NRC and the Licensee
review all safety analyses conducted since initial startup of the
Vermont Yankee facility with particular attention to their role in
providing a complete and up-to-date FSAR. Second, the Licensee needs to
correct serious deficiencies in its design change control process and
should undertake a historical review of its design control
documentation to verify its accuracy. Third, the Licensee should
perform a global evaluation to determine how many modifications have
been inadequately tested since startup. Fourth, the Licensee needs to
initiate a thorough retraining program to review and emphasize the
underlying safety purposes of Technical Specifications, the FSAR,
design bases and NRC regulations in relation to routine operation of
the Vermont Yankee facility, emergency preparedness, and practical
implementation of the NRC's ``defense in depth'' philosophy. Finally,
CAN strongly recommends that the Licensee's Vermont Yankee staff
receive training on the proper use of the ``Single Failure'' criterion.
The LERs identified in the CAN Memorandum are briefly discussed
below.
(1) LER 96-13: ``Two fire suppression systems do not meet design
requirements due to personnel error on the part of [the] vendor who
designed and installed the systems''
CAN asserts that the LER did not address the cause and consequences
of
[[Page 53842]]
the foam suppression system deficiency, which is one of the two fire
suppression systems addressed in this LER. CAN is correct in that the
Licensee did not determine a precise root cause because such a long
time had elapsed since the occurrence (1978). It is not unreasonable
for a licensee to be unable to ascertain the exact root cause of a
personnel error that took place many years before (18 years in this
case). Key points that are considered in reviewing an LER are (1)
whether the specific problem is being appropriately addressed, (2)
whether the potential for a broader problem exists and, (3) if a
broader problem exists, whether it is properly addressed. In this case,
the Licensee reviewed its current design process and procedures and
determined that a similar occurrence would not be expected to occur
now, and the Licensee had two teams that were actively reviewing the
fire protection design bases and searching for the types of problems
reported in the LER. CAN is incorrect in stating that the consequences
of the foam system deficiency were not discussed in the LER. The
Licensee stated that any fire in the area would be contained and
suppressed, preventing its spread to safety-related equipment.
Because the design deficiencies addressed in this LER were
licensee-identified and corrected, they were treated as Non-Cited
Violations in Inspection Report 50-271/96-05 in accordance with Section
VII.B.1 of the NRC's Enforcement Policy,\3\ and the LER was closed in
Inspection Report 50-271/96-06. Further enforcement action is not
warranted.
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\3\ The NRC's policy and procedures for determining the
enforcement action that may be warranted for a violation are
discussed in NUREG-1600, ``General Statement of Policy and
Procedures for NRC Enforcement Actions'' (Enforcement Policy).
Because regulatory requirements have varying degrees of safety,
safeguards, or environmental significance, the first step in the
enforcement process is to evaluate the significance of the violation
and then assign a severity level to the violation. A violation is
assigned one of four severity levels. As described in Section IV of
the Enforcement Policy, Severity Level I is assigned to violations
that are the most safety significant and Severity Level IV is
assigned to violations that are the least safety significant.
Consistent with the recognition that violations have different
degrees of safety significance, the Enforcement Policy recognizes
that there are other violations of minor safety or environmental
concern that are below the level of significance of Severity Level
IV violations. These minor violations are not normally the subject
of formal enforcement action and are not usually described in
inspection reports. To the extent that such violations are
described, they are usually described as ``Non-Cited Violations.''
(2) LER 96-14: ``Failure to provide tornado protection for diesel
generator rooms as specified in the Final Safety Analysis Report due to
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unknown cause''
The FSAR states that large venting areas are provided to vent the
diesel generator room in the event of a tornado to provide pressure
equalization. The LER notes that the facility as constructed did not
include venting. CAN asserts that ``flaws in the FSAR cause serious,
rippling effects throughout VY's [Vermont Yankee facility's] safety
systems'' and that the Licensee ``must include assessments of the
impact of the deficient conditions upon all affected programs.''
The Licensee took immediate action to insure emergency diesel
generator (EDG) operability in the absence of the pressure relief
panels. The Licensee took immediate compensatory measures which
included blocking open the EDG room doors and posting fire and security
watches. The Licensee took additional compensatory actions for the
restoration of operability of the diesel and day tank enclosures during
cold weather months when the EDG doors had to be shut. An NRC inspector
verified that the recommended compensatory measures were properly
implemented.
The discrepancy between the actual plant design and the FSAR is a
de facto change to the facility as described in the safety analysis
report, and thus required an evaluation to meet the requirements of 10
CFR 50.59. The failure to perform such a 10 CFR 50.59 evaluation was
categorized as a Severity Level IV violation, and was dispositioned in
Inspection Report 96-11 as a Non-Cited Violation in accordance with
Section VII.B.1 of the Enforcement Policy.
Other plants have been found to have FSARs which do not properly
describe the facilities. Consequently, for this reason and as a result
of lessons learned from events at Millstone Nuclear Power Station and
Maine Yankee Atomic Power Station, on October 9, 1996, the NRC
requested information from all power reactor licensees, to verify,
among other things, that the plant FSARs properly describe the
facilities, and that the systems, structures, and components are
consistent with the design basis. In conjunction with this request for
information, and in order to encourage licensees to identify
discrepancies, the Commission approved a modification to the NRC
Enforcement Policy that allows the NRC staff to exercise enforcement
discretion for a period of 2 years for violations related to FSAR
discrepancies identified by licensees. The policy revision was
published in the Federal Register on October 18, 1996 (61 FR 54461).
In the Licensee's response to this request for information dated
February 14, 1997, the Licensee committed to complete its FSAR
verification program in 1998.
CAN raises a concern about a potential error in the Licensee's
statement in this LER of no prior occurrences, based on a James A.
Fitzpatrick Nuclear Power Plant report of a similar problem. Licensees
are only required to report prior similar occurrences at their
facility, and not at any other facility. Therefore, the Licensee was
accurately reporting that a similar event had not previously occurred
at Vermont Yankee Nuclear Power Station. This LER is closed. Further
enforcement action is not warranted. The Licensee has issued a
supplement to this LER to document the long term corrective actions to
vent the EDG room in the event of a tornado to provide pressure
equalization. This LER supplement remains open pending NRC inspection
of the Licensee's modifications to the EDG room to provide the required
pressure equalization.
(3) LER 96-15: ``Original B31.1 ANSI Code Section that Required
Overpressurization Relief for Isolated Piping Sections was not
Considered during [the] Original Design''
Certain piping sections, which would be isolated after a LOCA, were
found to lack overpressure protection contrary to code requirements.
The water in this piping could expand because of the high temperatures
accompanying a LOCA and exceed the design pressure rating of the
piping. CAN asserts that the Licensee failed to take advantage of
earlier opportunities to identify this design error when making
modifications to the six systems discussed in the LER. CAN is correct
in that the LER represented the first discovery of this problem,
although modifications had been made to the affected systems earlier.
This potential overpressurization problem has been identified at other
plants, as evidenced by the issuance of NRC Information Notice (IN) 96-
49 on August 20, 1996, and NRC Generic Letter (GL) 96-06 on September
30, 1996. The Licensee did maintain an awareness of events in this area
and identified this issue at its site before the generic communications
referred to above were issued. The NRC staff encourages licensee
initiatives to identify and correct safety problems that may be generic
to the industry in advance of generic NRC staff communications to the
industry. The Licensee's corrective actions included a design change
which provided the required overpressure protection for the
[[Page 53843]]
affected lines. The change was completed in the 1996 refueling outage.
This LER remains open. Responses from power reactor licensees to GL
96-06 were received by the NRC staff in February 1997 and are
undergoing review to assure that the overpressure protection issue is
being adequately addressed and resolved. Following this generic review,
a determination will be made of whether enforcement action is warranted
for specific plants. Information regarding the completion of this
activity and any enforcement action taken will be publicly available in
the plant specific Inspection Reports. This LER will be further
discussed in a Final Director's Decision when the LER is closed.
(4) LER 96-18: ``Inadequate Installation and Inspection of Fire
Protection Wrap Results in Plant Operation Outside of Its Design Basis,
A Single Fire Would Impact Multiple Trains of Safety-Related
Equipment''
CAN asserts that this deficiency had significant adverse safety
implications. The reported deficiency consisted of a small gap in the
fire barrier installed on a cable tray support. The cable tray
contained wiring to support operation of the ECCS. The NRC staff does
not consider CAN's claim, that a fire could have rendered both
divisions of the ECCS inoperable, credible. The Licensee's evaluation
found that existing fire protection analyses were very conservative,
and that, with the combustible loading and fire detection and
suppression equipment in the area, no credible fire threat could
challenge the functionality of the ``as found,'' wrapped cable. The
Licensee has acted appropriately to correct the fire barrier deficiency
and to prevent similar problems in the future. With the combustible
loading, fire detection, and suppression equipment in the area, the NRC
staff conceptually agrees with the Licensee's conclusion that no
credible fire threat could challenge the functionality of the ``as
found'' wrapped cable. Inspection activities were performed the week of
August 18, 1997 to verify the Licensee's conclusion.
This LER remains open. Results of the inspection and any
enforcement action as a result of this inspection activity will be made
publicly available through plant specific Inspection Reports. This LER
will be further discussed in a Final Director's Decision when the LER
is closed.
(5) LER 96-19: ``Half scram and group III containment isolation caused
by loose Reactor Protection System breaker termination''
The NRC staff agrees with CAN that this event presented no
significant risk to public health and safety. This LER is closed. No
violation was involved, therefore the NRC staff concludes that
enforcement action is not warranted.
(6) LER 96-20: ``Inadequate vender [sic] design activity and Licensee
design verification result in inability to demonstrate Fire Suppression
System Operability''
This LER involved the inability of the carbon dioxide fire
suppression system to fully extinguish a deep-seated fire, as required.
The Licensee stated in the LER that this event had no safety
significance. The NRC staff considered this LER to have little apparent
actual or potential safety significance. This conclusion was based on
the Licensee's analysis that although the carbon dioxide suppression
systems might not fully extinguish a deep-seated fire, the suppression
and detection systems would function. Fire detection would alert the
fire brigade, and because the carbon dioxide fire suppression system
had reduced the fire, the fire brigade could extinguish the fire more
easily. The NRC staff closed this LER in Inspection Report 96-11.
Pending inspector review of the Licensee's corrective actions, the
unresolved item initiated for this issue in Inspection Report 96-08
(URI 96-08-01) was left open. As documented in Inspection Report 97-05,
unresolved item 96-08-01 was closed and a Non-Cited Violation was
issued, consistent with Section VII.B.1 of the NRC Enforcement Policy.
Further enforcement action is not warranted.
CAN asserts that this LER reveals a serious deficiency in the
Licensee's design change control process, and that the Licensee should
determine how many other modifications have been inadequately tested
since startup. The NRC staff agrees that this event demonstrated a
weakness in the Licensee's modification and testing programs associated
with fire protection. As noted under the discussion regarding LER 96-
13, the Licensee has initiated reviews of the fire protection design
bases to search for these types of problems, and believes that the
current design process and procedures are adequate to prevent similar
problems. As discussed earlier, by letter dated October 9, 1996, the
NRC staff requested information from all licensees, to verify, among
other things, the adequacy of the design change control process and to
determine the rationale for concluding that design-basis requirements
are properly translated into operating, maintenance, and testing
procedures. The Licensee responded by letter dated February 14, 1997.
(7) LER 96-21: ``Inadequate procedural controls of MOV Limit Switch
Settings result in a potential common cause failure mode with the
capacity to affect multiple safety significant components''
This LER involved two limit switches on shutdown cooling suction
motor-operated valve (MOV) to the ``D'' RHR pump. The switches measure
valve travel towards the open position. One open limit switch permits
the pump motor to start after the valve position is sufficiently open,
and the other limit switch stops valve travel so that the motor doesn't
drive the valve too far and damage the valve. The Licensee identified
that a modification to the valve's motor operator resulted in the
improper setting of these two limit switches.
Inspector follow-up, as documented in Inspection Report 97-05, led
to the conclusion that this error was of low safety significance. The
failed start of the ``D'' RHR pump because of this limit switch error
on the shutdown cooling suction valve affected only the shutdown
cooling mode of operation of the RHR system. The failure did not impact
the other modes of RHR system operation and the safety design bases
functions of the RHR system. Further, prompt Licensee action was taken
to check the other recently modified MOVs. Their limit switches were
found to be properly set and therefore their safety functions were
unaffected. This licensee-identified and corrected violation resulted
in the issuance of a Non-Cited Violation, consistent with Section
VII.B.1 of the NRC Enforcement Policy. This LER is closed. Further
enforcement action is not warranted.
(8) LER 96-22: ``Combination of poor man-machine interface, an
inadequate procedure, inadequate Operating Experience Review results in
a common cause failure mechanism, and an Emergency Diesel Generator to
exceed Tech Spec [sic] outage time''
The output breaker for one emergency diesel generator (EDG) was
found to be incapable of closing because of a missing cotter pin which
was necessary for a mechanical linkage. As a result of the absence of
this cotter pin, the breaker closing springs failed to recharge,
rendering the breaker incapable of being closed from the control room.
The only indication that the closing springs had failed to recharge was
a mechanical flag indicator located behind the breaker cubicle door.
[[Page 53844]]
No licensee procedures required verification of the closing spring
status. The closing springs were apparently in an uncharged condition
for over three weeks without discovery. Because the periodic
surveillance interval for the breaker is greater than the EDG limiting
condition for operation (LCO), the Licensee unknowingly operated in
violation of its Technical Specification (TS) governing diesel
generator operability. After reviewing the Licensee's root cause
analysis of this event, the NRC staff determined that the missing
cotter pin would not reasonably have been expected to be detected by
the Licensee's existing quality assurance program or through other
related control measures. \4\ The Licensee identified the EDG
inoperability, investigated to determine when the problem arose, and
reported that the LCO time was exceeded. The Licensee responded to the
inoperable equipment when the inoperability was discovered. The
Licensee did not intentionally exceed an LCO. Rather, the Licensee
discovered an equipment problem caused by a malfunction beyond its
control which meant that, in hindsight, an LCO had been exceeded. The
Licensee is designing a modification for this and other circuit
breakers of similar design to allow monitoring of the charging status
of the closing springs without having to open the circuit breaker
cubicle door.
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\4\ CAN asserts that the Licensee misconstrues the purposes of
TS Limiting Conditions for Operation (LCOs) as part of a ``chronic
pattern of misunderstanding'' of TS, FSAR design bases, and NRC
regulations. For the reasons described herein, LER 96-22 does not
provide a basis for this assertion.
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Because the EDG inoperability was not avoidable by reasonable
Licensee quality assurance measures or management controls, the NRC did
not issue a Notice of Violation for this issue. This is consistent with
Section VI.A of the Enforcement Policy. This LER is closed. The NRC
staff concludes that further enforcement action is not warranted.
(9) LER 96-23: ``Inadequate Surveillance Procedure results in failure
to meet Technical Specification requirements for Radiation Monitor
Functional Testing''
The reactor building and refueling floor radiation monitor test
procedure did not verify the high alarm contact actuation as required
by TS. The NRC staff agrees with CAN that this event presented no
significant risk to public health and safety. Considering that the
monitors were verified to be fully functional, and were in the
condition required by Plant Technical Specifications, this specific
event appears to have been limited to an inadequate testing
methodology. The Licensee's corrective actions included revising the
deficient surveillance test procedure to properly test the high alarm
output contacts.
However, the LER remains open as the NRC staff has not completed
its inspection activities related to this LER. The NRC staff will look
historically to see if this is an isolated case as part of the
enforcement consideration. On January 10, 1996 the NRC issued Generic
Letter (GL) 96-01 , ``Testing of Safety-Related Logic Circuits,'' that
requested, among other things, that all power reactor licensees review
their surveillance test procedures to ensure that all portions of the
logic circuitry are being tested. The Licensee's response to GL 96-01,
due to be sent to the NRC in September 1997, will be evaluated with
respect to the Licensee's long-term corrective action for logic testing
procedures, because any associated corrective action could be
considered in determining whether enforcement action is warranted.
Information regarding any enforcement action taken will be available
publicly in plant-specific Inspection Reports. This LER will be further
discussed in a Final Director's Decision when the LER is closed.
(10) LER 96-25: ``Inadequate testing leads to misadjustment of
isolation valve mechanical stop and failure to meet Technical
Specification leak rate limits for containment purge isolation valve''
This LER involved a containment isolation valve which leaked in
excess of TS requirements. The amount of valve leakage was influenced
by the direction in which the valve was leak tested and the adjustment
of a mechanical stop. CAN's concern appears to be that the Licensee
failed to apply the single-failure criterion in assessing the
significance of the failure in its LER. section 50.73(b)(3) requires
that an LER contain an assessment of the safety consequences and
implications of the event, including the availability of other systems
or components that would have performed the safety function of the
failed system or component. In this case, the requirement is that the
assessment include the availability of a redundant component (valve)
that would have performed the safety function (torus isolation).
Petitioner's issue is thus whether the LER should have, in addition,
assessed the potential radiological consequences had a design-basis
accident (DBA) occurred with failure of the redundant isolation valve.
Compliance with section 50.73(b)(3) does not require that the
assessment consider an additional single failure beyond the failure
which forms the basis for the assessment. On the basis of required
reporting, LER 96-25 was not deficient in omitting discussion of the
potential consequences of failure of the redundant valve. Inspection
Report 50-271/96-11 dispositioned this Severity Level IV TS violation
as a Non-Cited Violation in accordance with the criteria for
enforcement discretion in Section VII.B.1 of the Enforcement Policy.
Although the event was considered to be of more than minor safety
significance, the outboard valves had successfully passed all previous
tests, and thus the demonstrated containment integrity was always
maintained for the two affected penetrations. This LER is closed. No
further enforcement action is warranted.
C. Summary
In summary, with respect to CAN's concern that an unreviewed safety
question with respect to containment isolation may have been introduced
by Licensee actions in opening the RHR minimum flow lines, the NRC
staff determined that no unreviewed safety question was introduced and,
therefore, no enforcement action is warranted. With respect to CAN's
concerns related to the LERs, the NRC staff finds that the Enforcement
Policy has been applied consistently for the LERs that have been closed
and further enforcement action is not warranted.
For those LERs which remain open the Inspection/Enforcement process
will continue until the staff has completed its investigation and
consideration of the issues involved. LER closure and enforcement
action, as appropriate, will be documented publicly as is NRC staff
practice, and will documented in a Final Director's Decision.
With regard to CAN's overall conclusions based on its analysis of
the above LERs, the NRC staff has reached the following conclusions:
With respect to CAN's conclusion that the NRC and the Licensee
should review all safety analyses conducted since startup of the
Vermont Yankee facility with particular attention to their role in
providing a complete and up-to-date FSAR, the NRC staff has taken
actions as noted in the discussion above related to LER 96-14 with
respect to identifying and correcting FSAR inaccuracies. This action
was taken in a request on October 9, 1996, to all licensees, including
Vermont Yankee, to provide the requested information. In addition, the
NRC staff has implemented a series of
[[Page 53845]]
engineering design inspections, including an inspection to verify
portions of the Licensee's design control process and maintenance of
the Licensees's FSAR commitments. The results of the NRC design
inspection conducted at Vermont Yankee were reported in Inspection
Report 97-201 dated August 27, 1997.
With respect to CAN's conclusion that the Licensee needs to correct
serious deficiencies in its design change control process and should
undertake a historical review of its design control documentation to
verify its accuracy, the NRC staff has taken action as noted in the
discussion related to LER 96-20 with respect to identifying and
correcting design change control process deficiencies. In the October
9, 1996 letter to all licensees, including Vermont Yankee, the NRC
staff requested information to verify, among other things, the adequacy
of the design change control process and to determine the rationale for
concluding that design-basis requirements are properly translated into
operating, maintenance, and testing procedures. As also noted in the
discussion related to LER 96-20, the Licensee has undertaken a review
of the fire protection design bases to search for the type of problems
involved in LER 96-20, and believes that the current modification
programs are adequate to prevent similar problems.
With respect to CAN's conclusion that the Licensee should perform a
global evaluation to determine how many modifications have been
inadequately tested since startup, as noted in the discussion related
to LER 96-20, the Licensee has been required to provide verification of
the design change control process, including among other things the
rationale for concluding that design basis requirements are translated
into testing procedures.
With respect to CAN's conclusion that the Licensee needs to
initiate a thorough retraining program to review and emphasize the
underlying safety purposes of TSs, the FSAR, design bases and NRC
regulations in relation to routine operation of the Vermont Yankee
facility, emergency preparedness, and practical implementation of the
NRC's ``defense in depth'' philosophy, the NRC staff disagrees. In the
discussion related to LER 96-22, the NRC staff addresses CAN's
assertion that the Licensee misconstrues the purposes of TS LCO as part
of a ``chronic pattern of misunderstanding'' of TS, FSAR design bases
and NRC regulations. The NRC staff finds no basis to require such a
retraining program.
Finally, CAN strongly recommends that the Licensee's Vermont Yankee
staff receive training on the proper use of the ``Single Failure
Criterion.'' In the discussion related to LER 96-25, the NRC staff
addresses what seems to be the basis for CAN's recommendation: i.e. the
perception that the Licensee failed to properly apply the Single
Failure Criterion in assessing the significance of a leaking isolation
valve in LER 96-25. Compliance with Section 50.73 does not require that
the assessment consider an additional single failure. The enforcement
conference related to the minimum flow valves concerned a problem in
implementation of the Single Failure Criterion; not a misunderstanding
of the requirements of the Single Failure Criterion. Because the
Licensee did not err in the instance described in LER 96-25 and the
Petition provides no other instances in which problems were caused by a
misunderstanding of the Single Failure Criterion, the NRC staff finds
no basis for requiring additional training.
III. Conclusion
The NRC staff has reviewed the information submitted by the
Petitioner. The Petitioner's request is granted in that the NRC staff
has evaluated the majority of issues and LERs raised in the Memoranda
provided by the Petitioner to see if enforcement action is warranted
based on the information contained therein. The NRC staff has discussed
each Memorandum above and described any related enforcement action
taken for those issues and LERs which are closed. The NRC will continue
the same process and consideration for the LERs that remain open and
documentation of any inspection and/or enforcement action will be
consistent with agency practices and will also be the subject of a
Final Director's Decision.
As provided in 10 CFR 2.206(c), a copy of this Decision will be
filed with the Secretary of the Commission for the Commission's review.
This Decision will become the final action of the Commission 25 days
after issuance, unless the Commission, on its own motion, institutes
review of the Decision in that time.
Dated at Rockville, Maryland, this 8th day of October 1997.
For the Nuclear Regulatory Commission.
Original signed by
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 97-27417 Filed 10-15-97; 8:45 am]
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