ML20198L179
| ML20198L179 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 12/29/1997 |
| From: | Collins S NRC (Affiliation Not Assigned) |
| To: | CITIZENS AWARENESS NETWORK |
| Shared Package | |
| ML20198L149 | List: |
| References | |
| 2.206, DD-97-26, NUDOCS 9801150191 | |
| Download: ML20198L179 (9) | |
Text
. . _ _ . . _ . _ . _ _ _ . ._
. ; H: ..
/: * ..
DD 97 UNITED STATES OF AMERICA ,
, NUCLEAR REGULATORY COMMISSION Y '0FFICE OF NUCLEAR REACTOR' REGULATION Samuel J. Collins. Director
-In the Matter of ) Docket No. 50-271
)
)
VERMONT YANKEE NUCLEAR POWER ) License No. DPR-28 CORPORATION )
(Vermont Yankee Nuclear Power '
)
Statior) ) ,
./- FINAL DIRECTOR'S DECISION PURSUANT TO 10 CFR 2.206 1
- 1. INTRODUCTION On December 6, 1996. Mr. Jonathan M Block submitted a Petition on behalf of the Citizens Awareness Network. Inc. (CAN or Petitioner), and
?
included two Memoranda from CAN. The first Memorandum, dated December 5,1996, reviews information presented by the Vermont Yankee Nuclear
, Power Corporation (Licensee) at a predecisional enforcement conference held on J
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), The second Memorandum, dated December 6, 1996, contains a review of certain licensee _ event reports (LERs) submitted by the Licensee in the latter part of 1996. The Petitioner requests that the NRC evaluate these
'O documents, pursuant to 10 CFR 2.206, to determine if enforcement action is warranted on the-basis of. information contained therein.
On February 12, 1997, the NRC informed the Petitioner in an Jacknowledgement -letter that the Petition had been referred to the Office of y
- ' Nuclear
- Reactor Regulation.(NRR) .for the preparation of a Director's Decision and that action would be taken within a' reasonable time regarding the specific 1150 g 97 $
2 concerns raised in the Petition. On October 8. 1997, the NRC issued a Partial Director's Decision that responded to the first Memorandum concerning the RHR system and all but three of the LERs listed in the second Memorandum. This Final Director's Decision addresses the NRC staff's conclusions regarding the three remaining LERs that were still being evaluated at the time the Partial Director's Decision was issued.
On November 7, 1997. CAN submitted a letter to the Director of NRR commenting on the Partial Director's Decision. CAN raised a concern that the Partial Director's Decision did not adequately address concerns raised in its Petition of December 6. 1996. In a response from the NRC staff dated November 28. 1997. CAN was informed that its letter provided no new or additional information that would warrant a review of the Partial Director's Decision. In its November 7. 1997 letter. CAN also raised a concern about asserted " systematic mismanagement" at the Vermont Yankee facility and requested certain NRC actions. The Petitioner was informed that this specific concern would be treated as a supplement to the original Petition and is addressed in this Final Director's Decision,
- 11. DISCUSSION The NRC staff's evaluation of the three remaining LERs and the Petitioner's supplemental request for action follows.
A. Licensee Event'Recorts A CAN Memorandum dated December 6. 1996 included with the Petition contains a review of several LERs submitted by the Licensee in the latter part of 1996. .On the basis of its analysis of the LERs. CAN reaches certain conclusions regarding Licensee performaace and actions that it believes should be taken. The Partial Director's Decision evaluated LERs 96-13. 96-14. 96-19.
. - -- . . =
-3 96 20, 96-21, 96-22, and 96 25 and provided a--respcr .e to CAN's overall [
-conclusions.regarding Licensee performance and reques md actions. LERs 96-15,.
% and 96 23 were stil_1 open at the time the Partial ~ Director's' Decision was issued. The staff har completed.its evaluation of these three LERs and its-conclusions 'are presented below.
I'. LER 96-15i " Original B31.1 ANSI Code Section That Required Overpressurization Relief for Isolated Piping Sections-Was ~
Not Considered During-[tha] Original Design" Certain' piping sections which would be isolated after a loss Of coolant accident (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 documented the first discovery of this problem, although modifications had been made to the affected systems earlier. 'This potential overpressurization problem has been:1dentified at other plants. as evidenced by the issuance of NRC Information Notice 96-49 on~ August 20. 1996, and NRC Generic Letter (GL) 96-06 on September 30, 1996. The Licensee was aware of events in this area and identified this issue'at its site before the generic communications previously ,
referred to.were issued. The Licensee's corrective actions included a design change that provided the required-overpressure protection for the affected lines. 'The change was completed in the 1996 refueling outage conducted during the period of September ~6. 1996 to October 30. 1996.
Because the' Licensee identified the design deficiency described in this LER by other than routine quality assurance or surveillance activities and has
. a
fN
@ n ,
p,:*- , _
]
. 5 4 i
iglemented ' appropriate l corrective. actions to reso,1ve the discrepancygthis?
old design issue" was not cited in accordance with NRC Enforcement Policy, Section: VII .B.'3,1 -The LER was closed in Inspection Report'50-271/97-11.
- 2. LER 96-18: Inadequate Installation and Inspection of Fire' _
Protection Wrap Results.in Plant Operatiot)
-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, emergency core cooling system (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. Tha staff agrees with the Licensee's analysis as documented in the LER and has found that the Licensed acted appropriately to correct the fire barrier deficiency and to prevent simil6r problems in the future.
The NRC staff- found that the deficiency described in this LER was a
~
[ viol tian of NRC requirements of--10 CFR Part 50. Appendix R.Section III.G.
However..in accordance with the provisions of NRC Enforcement Policy.Section VII.B.4. no notice of violation was. issued in this case because the deficiency (1) was identified by the Licensee as part of .the corrective actions for a '
General Statement'of-Policy and Procedures for NRC Enforcement ,
C
. Actions. NUREG-1600'(Enforcement Policy).
-e
.N
,, . $,w- .. - .,,m , . - . _ - . .- . . - sa, - . . . - - _ , - -
A ,4- 4 L .sa
}
--- *1- A-,
,y :l
.? a 5
previous issue _related to Appeadix R .(2) had the same. root'cause as the
-previous issue. (3) did not sLbstantially change the safety significance or-q= _the character of the regulatory concern arising out of the initial action, and (4) the def' '?ncy was corrected within a reasonable time following
. identificatun. The LER was closed in Inspection Report 50-271/97-80.
- 3. 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 the Vermont Yanke's Technical Specifications. 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 'ully functional and were in the condition required by plant Technical Specificetons. 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.
Because the deficiency identified in this LER was of minor safety' significance and was identified and corrected by the Licensee it was treated
, as a non-cited violation in accordance with NRC Enforcement Policy.Section VII.B.1. The LER was closed in Inspection Report 50-271/97-08.
B. Sucolemental Reauest for Action On November 7,1997 CAN submitted a letter-which raised a concern-about
' assertedT"systematicmIsmanagement"attheVermontYankeefacilityand requested that three actions be taken. In its respcnse to the Petitioner, the s
=
g ,f 7
. /
a
~6 I
LNRC staff indicated that this concern would be considered as a' supplement to:
the Petition.
LThe requested actions, along with the NRC staff's' evaluation.- are-discussed below.
- 1. "An NRC team in conjunction with an outside contractor conduct a review of a second system.t the ventilation system."
From May 5 through June 13. 1997, the NRC staff performed'a detailed-design inspeQon of the low-pressure coolant injection and RHR service water systems at the Vermont Yankee facility. The inspection team consisted of a team leader from the NRC and five contractor engineers from Stone & Webster Engineering Corporation. The systems were chosen on the basis of their
.importance in mitigating design-basis accidents at Vermont Yankee. The purpose of the inspection was_ to evaluate the-capability of the selected systems to perform the safety functions required by the design bases and the consistency of the as built configuration and system operations with the Final Safety Analysis Report (FSAR). Overall, the inspection team concluded that the two systems were capable of performing their intended safety functions.
However, the team identified some issues that indicated potential programmatic concerns extending beyond the two systems that were inspected. Specifically, the team identified the following issues which indicated potential
_ programmatic concerns: (1) several examples which indicated the Licensee's correction of licensing documentation was not timely: (2) when rendering equipment inoperable for surveillance testing, the Licensee's practice
- concerning entry into the limitirg condition of operation (LCO) was not
- consistent with the guidance provided in GL 91-18. " Resolution of Degraded and Nonconforming' Conditions:"-(3) deviations from the licensing comitments made Lin response _to GL 89-13. " Service Water System Problems Affecting Safety-
jf~ '
Related Equipment:" (4) weaknesses in the development and control of calculations, and the review and approval process for calculations: and:(5).
weaknesses concerning the Licensee's translation'of design criteria and design.
bases-intoidetailed operating instructions. The results'of this inspection
- were documented in Inspection Report 50-271/97-201.
By letter dated October 27. 1997, the Licensee provided a~ schedule and detailed the plans to complete the corrective actions required to resolve the broader programmatic issues listed in the inspection report. In its letter, the Licensee listed several initiatives it has undertaken to improve its performance, These initiatives include: (1) a re-engineering of the corrective action program, (2) a large scale program to develop Design Basis Documents for the 23 most risk'significant systems. (3) initiation of-a Design
' Basis Validation Program. (4) conversion of the plant's Technical Specifications to the Standard Technical Specification format. (5) a large scale instrument setpoint calculation and verification program (6) a large scale effort to re-engineer configuration management program, and (7) creation of a System Engineering Department.
The NRC staff has concluded that the Licensee's proposed actions and schedule are acceptable and that the facility may be operated while the Licensee works to resolve these issues. The staff will continue to follow the Licensee's progress.to improve the facility's design-basis documentation and
-implement the initiatives' outlined in its October 27, 1997 letter through the normal-inspection process. A detailed design inspection by the NRC staff of an additional safety system is no'. warranted at this time.
'2 "NRC with an outside contractor and VY [ Vermont Yankee] conduct a review-
.of all backup safety l systems to assure adequacy of these systems in order to protect worker and public health and safety."
- 2 .
-g.. ,
As stated-in the reply to Item 1 above, the NRCLstaff has conducted a detailed design inspection of-two_ selected systems at the Vermont Yankee facility. .The inspection _ team found_the two systems capable of performing 4
their. Intended design functions,. As discussed in Item 1 above, the inspection:
report also documented several. issues'of programmatic concern, The NRC staff has determined that the Licensee's response to these programmatic-concerns is-acceptable andLimplementation of.the Licensee's actions will be assessed-during followup inspections. Overall, the staff finds that the detailed design _ inspection and the-followup inspection activities provide adequate assurance of public health and safety and that a design review inspection of i additional safety systems is not warranted at this time.
- 3. "Given the lack of thoroughness-by the licensee and significent flaws in '
the FSAR and design basis evaluation,' CAN questions Region I staff's competence'to effectively oversee reactors under its authority. We i therefere request that the archive of NRC's oversight failures at VY
[ Vermont Yankee].be added to the Inspector General's investigation of complicity and systematic failure to enforce NRC regulations.by NRC staff in Region I and Project Directorates."
With regard to this request, CAN's letter has been forwarded to the-Office of the Inspector General.
III. CONCLUSION ,
-The NRC staff has reviewed the information submitted by the Petitioner.
The Petitioner's request is granted in part in that the NRC staff has evaluated all-of the issues raised in the two Memoranda and the supplemental
_ letter provided by the Petitioner to see -if enforcement action is warranted on the basis of the information contained therein. 'In the Partial and the Final
? .
~; . - ;
9,' 3
- Director's Decision.- the NRC staff has discussed each Memorandum and the -
supplemental' letter;and described any related enforcement action that was L -
i taken; The Petitioner's.: supplemental: request:that the NRC in conjunction-with an'outside contractor, conduct additional' review of safety systems at the
~
Vermont-: Yankee' facility is denied, -With respect to the supplemental request-for-an investigation of NRC oversight of the Vermont Yankee facility. the-Petitioner's-supplemental letterias forwarded to the Office of the Inspector ;
General. .
I
- 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 constitute the final action of the Comission 25 days after issuance, unless the Commission, on its own motion, institutes review of the Decision in that time.
l Dated at Rockville, Maryland, this 29th day of December 1997. -
FOR THE NUCLEAR REGULATORY COMMISSION i
I mYJ s ector Office of Nuclear Reactor Regulation
,1
, _ - ,- .A-_-