[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]


-----------------------------------------------------------------------

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.<SUP>1</SUP> 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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    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.<SUP>2</SUP> A 
pipe break outside containment would breach primary containment with an 
OPEN minimum flow valve.
---------------------------------------------------------------------------

    \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''.
---------------------------------------------------------------------------

    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.
---------------------------------------------------------------------------

    \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 
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    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]
BILLING CODE 7590-01-P