ML20056G371
| ML20056G371 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/16/1993 |
| From: | Kay L, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20056G362 | List: |
| References | |
| 50-245-93-18, 50-336-93-13, 50-423-93-14, NUDOCS 9309030052 | |
| Download: ML20056G371 (7) | |
See also: IR 05000245/1993018
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
Report Nos.
50-245/93-18
50-336/93-13
50-423/93-14
Docket Nos.
50-245
50-336
50-423
License Nos.
DPR-65
Licensee:
Northeast Nuclear Energy Company
P.O. Box 270
Hartford, Connecticut 06141-0270
Facility Name:
Millstone Nuclear Power Station
Units 1,2, and 3
Inspection At:
Berlin Office and Waterford, Connecticut
Inspection Conducted:
June 14-25,1993
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July 1,1993, Region I office
Inspectors:
T. Scarbrough, Sr. Mechanical Engineer, NRR
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Il Kay, Reactor Engineer, ES,
Engineering Branch, DRS
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Approved by:
W. Ruland, Chief, Electrical
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Section, Engr. Branch, DRS
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9309030052'930826
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Areas Insnected: The licensee's self-assessment activities for their motor-operated valve
program and unresolved item 50-336/92-27-03 regarding assumptions used in thrust
determinations for power-operated relief valve block valves 2-RC-403 and 405.
Results: The licensee's failure to adequately review and document operability evaluations
following motor-operated valve testing performed in December 1992, for Unit 2 resulted in a
violation. The test procedure used for dynamic testing failed to incorporate requirements and
acceptance limits for determining operability prior to returning the tested MOV to service.
The status of both unresolved items was updated; both items remain open.
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DETAILS
1.0
PURPOSE
The purpose of this inspection was to review and verify the adequacy of the licensee's
dynamic test data for motor-operated valves (MOVs) as a follow-up to two previously
identified unresolved items and to monitor the licensee's self-assessment activities for the
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MOV program at the Millstone Nuclear Power Station.
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2.0
OPERABILITY OF PRESSURIZER PORV BLOCK VALVES DURING
CYCLE 11 (UPDATE) UNRESOLVED ITEM NO. 50-336/92-27-03
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During a partial loss of normal power event at Millstone Unit 2 in July 1992, the licensee
identified risk-significant failure modes involving inadvertent power-operated relief valve
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(PORV) actuation when safety injection systems were unavailable. This event was
documented in NRC inspection report No. 50-336/92-22. PORV block valves 2-RC-403 and
2-RC-405 are required to isolate such a PORV actuation. Prior to this event in May 1992,
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the licensee initiated a reportability/ operability evaluation form (REF) determination based on
new MOV target thrust calculations which . suggested that the block valves would not have
been capable of isolating a PORV under worst-case design basis conditions. Licensee
evaluations of this finding performed in August and September 1992, during Cycle 11,
concluded that the valves were operable. This conclusion, however, conflicted with a
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calculation performed by Babcock and Wilcox Nuclear Services to determine the thrust-
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producing capability of the valves. This calculation indicated that the PORV block valves
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would not have developed the required closing thrust without exceeding the thrust and torque
limits of the actuator. Although the licensee determined that these valves were operable,
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modifications were performed to increase the valve actuator gear ratio and change the spring
packs to provide more thrust capability.
During NRC inspection No. 50-336/92-27, that reviewed the licensee's above operability
determination, the inspector identified an unresolved item pertaining to the assumptions made
in evaluating the operability of pressurizer PORV block valves. In particular, the unresolved
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item addressed the lack of:
justification of using a valve factor of 0.3,
reconciliation of the apparent conflict between the preliminary operability
determination and the implementation of valve modifications,
explanation of why the valve design basis, supposedly settled under the Three Mile
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Island Action Plan items between 1982 and 1988, was still undetermined in 1992, and
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resolution of the reportability of valves found inoperable by GL 89-10 standards.
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To address the NRC concerns, the licensee initiated a new REF (No. 93-30) to reassess the
operability of these valves during Cycle 11. The minimum required thrust for 2-RC-403 and
405 was 5,507 pounds. This minimum required thrust was calculated assu ning a valve
factor of 0.38 cmd the valve port seat diameter. A stem friction coefficier . (SFC) of 0.15
was used to calculate the torque developed by the actuator. Also included was a diagnostic
test equipment inaccuracy of 9.2% and a margin for torque switch repeatability of 5%. The
available thrust for 2-RC-403 and 2-RC-405 were 6,304 and 6,324 pounds, respectively.
The inspectors discussed the bases for the valve factors used. Licensee internal
documentation indicated that the valve factor for these two-inch Velan gate valves varied
from 0.2 to 0.9. The licensee's justification for the use of a 0.38 valve factor was not
documented using available industry test data. This item remains open pending the licensee's
demonstration of the adequacy of this valve factor based on industry data for similar valves.
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The inspectors reviewed the PORV block valve design basis to determine the worst-case
accident conditions for flow and pressure. The inspectors concluded that the flow and
pressure used for determining thrust capabilities of these valves enveloped Three Mile Island
Action Plan requirements. With the exception of thejustification for the use of 0.38 valve
factor, the inspector found the licensee' evaluation provided a reasonable assurance for the
operability of the PORV block valves during Cycle 11. The inspectors also concluded that
the licensee established reasonable assurance to demonstrate the operability of the PORV
block valves in the currently installed configuration.
In a subsequent telephone conversation of July 1,1993, the licensee informed the inspectors
that a new evaluation again concluded that these valves were operable during Cycle 11. The
licensee stated that an informational Licensee Event Report (LER) No. 50-336, 93-17-00,
was being prepared to document their experience with these block valves and to inform the
industry. The licensee further stated that if future testing indicates that the assumed valve
factors were nonconservative, hardware modifications to improve the capabilities of these
valves would be initiated. The inspectors concluded that these actions were appropriate.
3.0
EVALUATION OF UNIT 2 MOV DYNAMIC TEST RESULTS
Also incorporated into the Unresolved item No. 50-336/92-27-03 was a previous inspection
finding, as documented in Inspection Report No. 50-336/92-36, that addressed a test failure
of auxiliary feedwater (AFW) pump discharge header cross-tie vaive 2-FW-44. On
December 12,1992,2-FW-44 failed to close completely. The licensee issued
Nonconformance Report No. 2-9-1099 to resolve this failure. In addition, the licensee
recalculated the new required thrust, and raised the torque switch setting from 3.50 to 4.25
to provide a closing thrust of 25,082 pounds. The valve was tested satisfactorily at a flow of
600 gallons per minute (gpm), the flow required in the safety analysis. Subsequent to this
test, the licensee reevaluated the safety function of this valve. The safety function of this
valve was determined to isolate AFW to the steam generator during a main steamline break
accident at 1650 gpm and 1194 psig. The licensee stated that they would verify these new
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requirements during a test scheduled for the next refueling outage. The licensee's review of
the initial failure was not completed at the time of this inspection.
Additionally, the inspectors noted that the licensee evaluation to address the above valve
failure had not been documented until June 1993. This valve and sixteen others had been
tested during the last Unit 2 refueling outage, completed in December 1992. The inspectors
reviewed the evaluation of the test results for the remaining sixteen valves dynamically tested
during the last outage. The inspectors noted that the results for these sixteen valves had not
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been reviewed as of this inspection. Additionally, the licensee's schedule was to complete
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these reviews by the end of August 1993.
The inspectors requested the licensee to provide the post-test opeish determinations for
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these seventeen valves and their rationale for returning each of these ,alves to service.
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While the licensee was able to articulate the bases for determining operability, such bases
were not formally documented. On June 16,1993, the inspectors met with licensee
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management to discuss the untimely test results review for these seventeen valves. The
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inspectors stated that the untimely evaluation and a lack of documented bases for
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demonstrating operability of the tested valves are contrary to the test control requirements of
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10 CFR Part 50, Appendix B, Criterion XI, and was a violation (50-336/93-13-01).
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On June 17, 1993, the licensee completed preliminary evaluations and demonstrated the
operability of all seventeen tested valves. In response to this concern, corporate engineering
issued guidance to all three Millstone units and Connecticut Yankee to determine operability
of tested components prior to returning them to service. As stated in Connecticut Yankee
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Inspection Report No. 50-213/93-12, the inspectors verified that these requirements were
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adequately documented in that facility's MOV test procedures.
4.0
LICENSEE MOV PROGRAM SELF-ASSESSMENT
From June 14-25, 1993, the licensee performed a self-assessment of its MOV program,
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Following a meeting with Region I and NRR staff on May 20,1993, the licensee proposed
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their plan to perform a self-assessment of the MOV program at the Millstone Nuclear Power
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Station in a letter dated June 4,1993. Region I authorized the licensee's self-assessment in
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lieu of the NRC inspection mandated by NRC Temporary Instruction 2515/109, with NRC
monitoring of the assessment and an inspection following completion of the self-assessment
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as described in a letter to the licensee dated June 10, 1993.
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The inspectors monitored the self-assessment activities throughout the two-week period.
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Yankee Atomic Electric Company performed the self-assessment for the licensee. On
June 16 and from June 23 to 25, NRR assisted Region I in the monitoring effort. At the exit
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meeting of the self assessment, the audit team identified 29 findings, 6 unresolved items, and
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6 recommendations. The inspectors determined that the licensee's findings were equivalent
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to those that would have been identified by an NRC team.
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The inspectors concluded that the licensee's self-assessment was thorough and detailed in
their evaluation of the licensee's GL 89-10 program. The inspectors stated that the findings
were significant and emphasized the importance of dispositioning each finding adequately.
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The licensee stated that they will present a full report of the self-assessment to the NRC
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Region I staff by September 1993.
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5.0
EXIT MEETING
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The inspectors met with licensee personnel, denoted in Attachment I of this report, at the
conclusion of the inspection on June 25,1993. At that time, the scope of the inspection and
inspection results were summarized. The inspectors also requested assurance that Northeast
Utilities would establish procedure changes to require preliminary operability determinations
following dynamic testing prior to the next Millstone Unit 3 outage. The licensee agreed to
implement these changes. The licensee acknowledged the inspection findings as detailed in
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this report and had no additional comments regarding the inspection results.
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Attachment
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Persons Contacted
Herthpast NucicatEnergy Company. Corporate and Station Personng]
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'P. Austin
NUSCo, Manager, Systems Engineering
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- R. Blanchard
NNECo, Acting Engr. Supervisor
- P. lilasioli
NNECo, Unit 1 Engineering Manager
- J. DiMarzo
NUSCo, Sr. Engineer, Systems Engr.
- D. Harris
NUSCo, Licensing Engineer
- R. IIarris
NUSCo, Director, Engineering Dept.
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- 11. liaynes
NNECo, Unit 1 Director
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- S. Hodge
NUSCo, Supervisor, Systems Engr.
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- L. Loomis
NNECo, Unit 3 Engr. Supervisor
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- W. Loweth
NNECo, Sr. IST Engineer
- D. McDaniel
NNECo, Unit 3 Engineering Mgr.
- D. Nichols
NNECo, Unit 2 MOV Coordinator
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- E. Perkins
NUSCo, Nuclear Licensing
- J. Quinn
NNECo, Unit 1 Engr. Supervisor
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- J. Riley, Jr.
NNECo, Unit 2 Engr. Manager
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_UJ. Nuclear Regulatory Commission
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- R. Arrighi
Unit 3 Resident inspector
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D. Dempsey
Unit 2 Resident inspector
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- P. Eapen
Section Chief, Region I
P. Swetland
Sr. Resident Inspector, Millstone
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- Denotes present at exit meeting conducted on June 25,1993.
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