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NUCLEAR REGULATORY COMMISSION D,, '
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%v f KING OF PRUSSIA, PENNSYLVANIA 19406-1415 UUL 3 C h Docket No. 50-245 Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations Northeast Nuclear Energy Company P.O. Box 270 Hanford, Connecticut 06141-0270
Dear Mr. Mroczka:
Subject:
Millstone Unit 1 Inspection 91-12 This refers to the routine safety inspection conducted by Mr. D. Dempsey and others of this of6ce on May 21 through June 29,1991, at Millstone Unit 1 in Waterford, Connecticut. The preliminary findings were discussed with Mr. H. Haynes and others of your staff at the conclusion of the inspection.
Areas ext. mined during the inspection are described in the enclosed report. Within these areas, the inspection focused on issues important to public health and safety, and consisted of performance observations of ongoing activities, independent verincation of safety system status and design con 6guration. interviews with personnel, and review of records.
Your prompt correction of design deficiencies identified in response to post-accident containment temperature response analyses and high energy line break reviews indicate programmatic strengths in the areas of engineering and safety assessment. You are encouraged to continue to utilize effectively the insights provided by your design basis reconstruction program in on-going i
efforts to identify and resolve safety significant issues at Millstone 1. Also, scheduling of outage activities such that at least one source of emergency power remained operable at all times demonstrated a sensitivity to shutdown risk concerns. Your conservative response to solve significant design challenges identified during the refueling outage demonstrated a good regard for the safe operation of Millstone 1.
i The two violations of NRC requirements identified in the enclosed Notice of Violation have been categorized by severity level in accordance with the " General Policy Statement and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (Enforcement Policy). You are required to respond to this letter, and in preparing your response, you should follow the instructions in the Notice of Violation.
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Your cooperation with us is appreciated.
l Sincerely,
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p Edward C. Wenzinger, Chief.
Projects Branch No. 4 Division of Reactor Projects
Enclosures:
1.
NRC Region I Inspection Report No. 50-245/91-12 cc w/encls:
W. D. Romberg, Vice President, Nuclear Operations D. O. Nordquist, Director of Quality Senices R. M. Kacich, Manager, Nuclear Licensing S. E. Scace, Nuclear Station Director, Millstone H. F. Haynes, Nuclear Unit Director, Millstone Unit 1 Gerald Garfield, Esquire Nicholas Reynolds, Esquire K. Abraham, PAO (2 copies)
Public Document Room (PDR)
Ixcal Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector State of Connecticut I
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o U.S. NUCLEAR REGULATORY COMMISSION REGION I Report /
Docket:
50-245/91-12
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License No.:
DPR-21 i
Licensee:
Nonheast Nuclear Energy Company j
P. O. Box 270 j
Hartford, CT 06141-0270 i
Facility Name:
Millstone Nuclear Power Station, Unit 1 i
i Inspection At:
Waterford, CT l
Dates:
May 21,1991 - June 29,1991 i
Inspectors:
W. J. Raymond, Senior Resident Inspector D. A. Dempsey, Resident Inspector, Unit 1
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D. H. Caphton, Reactor Engineer, DRS j
E. J. Benner, Reactor Engineer Intern, DRS D. H. Jaffe, Project Manager, NRR 1
1 Approved by:
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Eugene M. K y, Chief, Reactor Profe' cts Section 4A Date Scope:
NRC resident inspection of plant operations, radiological controls, maintenance,.
surveillance, security, outage activities, licensee self-assessment, and periodic reports.
Routine review of plant operations was conducted during periods of backshifts I
(evening shifts) and deep backshifts (weekends, holidays, and midnight shifts).
Inspection coverage was provided for 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> during backshifts and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during
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deep backshifts.
Results:
See Executive Summary i
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m-3 EXECUTIVE
SUMMARY
MILLSTONE UNIT 1 INSPECTION 91-12 l
Plant Operations Effective management of temporary modifications to plant systems was reflected by a low backlog of jumpers and bypasses. By scheduling outage activities to ensure the availability of at least or.e emergency power source at all times, an awareness of and sensitivity to shutdown risk issues was demonstrated.
1 An equipment tagging error which resulted in an unplanned initiation of the standby gas treatment system was not considered to indicate a programmatic weakness in this area. One violation was identified by the licensee involving errors by two persons concerning the misloading of a fuel bundle into the reactor core. A similar violation was reported in Unit 3 Inspection 91-04.
Radiolocical Controls There were no noteworthy findings in this area.
t Maintenance / Surveillance Good performance was noted during core spray pump motor replacement and emergency diesel generator rewind activities. Strong management oversight of emergency service water repair activities was evident. Good valve maintenance practices were evidenced by the successful l
completion of the containment integrated leak rate test. However, continued effort to identify and resolve the source of excessive leakage of containment atmosphere control system valves is i
warranted. The containment integrated leak rate test was professionally performed using a high quality procedure. Test supervision and quality assurance coverage were good.
Encineerinc and Technical Support The identification and resolution of several significant design discrepancies exemplified an excellent safety perspective. Noteworthy were efforts directed toward resolution of spurious 4
Group IV isolation signals, core spray pump motor replacements, and high energy line break
~i issues. Plant design change requests were of high quality.
Safety Assessment /Ouality Verification A conservative regard for safe operation of Millstone 1 was demonstrated by replacement of wide range reactor vessel and torus level instruments in response to environmental qualification coreerns. A justification for continued operation regarding high energy line break issues was I
of high quality. Resolution oflow pressure coolant injection and core spray systems operability concerns related to pump motors and pipe suppons was good. The need for comprehensive i
review of all relevant design information was reinforced by these recent discoveries. Effective quality assurance was responsible for identifying incorrect Masterlock material during emergency
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service water strainer repair activities. One violation was identified concerning less than adequate supervision of outside storage of pipe material.
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4 TABLE OF CONTENTS EXECUTIVE SUMM ARY...................................... ii TABLE OF CONTENTS........................................ iii 1.0 PRINCIPALS CONTACTED................................
1 2.0 SUMM ARY OF FACILITY ACTIVITIES........................
1 3.0 PLANT OPERATIONS (IP 71707,93702,92701,92702,61701) 1 3.1 Operational Safety Verification 1
3.2 Onsite Followup of Operational Events 3
l 3.3 Previously Identified Items.............................
5 4.0 RADIOLOGICAL CONTROLS (IP 71707) 6 4.1 Posting and Control of Radiological Areas 6
l 5.0 MAINTENANCE / SURVEILLANCE (IP 61726, 62703, 37828) 6 5.1 Observation of Maintenance Activities......................
6-5.2 Observation of Surveillance Activities...................... 10 5.3 Hydrostatic Test of "B" Train Emergency Service Water System...... 14
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6.0 ENGINEERING / TECHNICAL SUPPORT (IP 37700, 37828,93702,92700) 16-l 6.1 On-Site Followup of Operational Events..................... 16 6.2 Plant Design Change Request Review - Core Spray Pump Motor Replacem en t...................................... 17 6.3 Diesel Generator Air Start Vent Valve Qualification.............. 18 l
6.4 Previously Identified Items............................. 19 7.0 SAFETY ASSESSMENT / QUALITY VERIFICATION (IP 92700,90712,92702, 3
90713) 21 l
7.1 Outside Storage of Safety Related Piping 21 i
7.2 Previously Identified Items............................. 22 7.3 Emergency Diesel Generator Qualification Test Review 22 7.4 Reactor Vessel Wide Range Level Instrument Modification 22 7.5 Emergency Diesel Generator Room Ventilation................. 23 j
7.6 High Energy Line Break Justification for Continued Operation 25
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7.~7 Post-Accident Torus Temperature Issues..................... 27 j
7.S Licensee Event Reports............................... 30 7.9 Periodic Reports................................... 31 j
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13 5.2.2 Inservice Inspection Program Weld Review The inspector reviewed NNECO documentation for three welds located in the isolation i
condenser, reactor tvater cleanup, and recirculation systems. Regarding weld number RREJ4, a weld on the "E" jet pump riser, a contracted licensee ultrasonic inspector noted an indication and determined it to be intra-granular stress corrosion cracking (IGSCC) with substantial through-wall penetration. Upon receiving the data from the ultrasonic inspector, NNECO reviewed the original radiograph of the weld to determine the existence of a longitudinal weld seam. The radiograph revealed that a longitudinal weld seam did not exist. Review of previous data showed that no indications had been reported in the suspected area.
I The licensee performed funher evaluation utilizing two ultrasonic level III inspectors qualified by the Electric Power Research Institute (EPRI) for IGSCC detection and a level II ultrasonic inspector qualified for IGSCC detection and sizing. The detection and sizing method used to evaluate the indication were recommended by EPRI. After extensive supplementary evaluation, NNECO concluded that the indication was a base material inclusion.
Based on review of NNECO evaluations, the inspector concluded that the licensee used adequate and appropriate methods to characterize the indication. and that no further licensee action was needed. However, the inspector recommended that the licensee include this weld in the next inservice inspection in order to monitor the weld.
5.2.3 Low Pressure Coolant Injection System Loop Selection Logic Test The inspector witnessed the performance of portions of refueling outage surveillance procedure SP-412K, "LPCI Containment Cooling System 1.ogic Test." The inspector verified that the procedure being utilized by NNECO test personnel was up-to-date; that initial conditions and precautions were observed; and that adequate communications were established between test technicians and control room operators. On one occasion, when the procedure could not be performed exactly as written, technicians stopped the test until an appropriately reviewed and approved procedure change was obtained.
After successful completion of the test, during system restoration, both reactor recirculation pumps tripped unexpectedly. The inspector observed ponions oflicensee troubleshooting efforts, which were directed by an instrumentation and controls department engineer. Through a walkdown of system relays, the licensee discovered the problem to have been caused when the removable handle of a test switch was inserted into the test box 180 degrees out of its normal position. The inspector noted that no mechanism existed whereby a technician could be alerted to this condition.
Despite the satisfactory test results, the licensee chose to repeat the affected portions of the procedure. All system components performed as required. The licensee modified the test switch handle to preclude future mishaps. The inspector concluded that the surveillance was performed t
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