IR 05000443/1988002
| ML20151G672 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 04/06/1988 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151G666 | List: |
| References | |
| 50-443-88-02, 50-443-88-2, GL-87-12, IEB-88-002, IEB-88-2, IEIN-87-023, IEIN-87-23, IEIN-88-005, IEIN-88-5, NUDOCS 8804200099 | |
| Download: ML20151G672 (21) | |
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U. S NUCLEAR REGULATORY COMMISSION
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REGION I
' Report No.:
50-443/88-02 Docket No.:
50-443 License No.:
NPF-56 Permit No.:
CPPR-135 Licensee:
Pub'.ic Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit No.1 Inspection At: Seabrook, New Hampshire, Inspection Conducted:
February 2 - March 28, 1988 Inspectors:
A. C. Cerne, Senior Resident Inspector - Seabrook Station D. G. Ruscitto, Resident Inspector - Seabrook Station E. H. Gray, Senior Reactor Engineer, Materials and Processes Section, Engineering Branch Division of Reactor Safety P. C. Wu, Corrosion Specialist, Chemical Engineering Branch, Office of Nuclear Reactor Regulation Approved By:
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f///f7 tonald R. Haverkamp(/hief /
' (Ta te Reactor Projects Section No. 3C
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Inspection Summary:
Inspection on February _2 - March 28,1938 U eport No. 50-a43/83-02)
Areas Inspected:
f.outine safety inspection during normal and backshif t periods
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I by two resident inspectors and two specialists ( 176 hours0.00204 days <br />0.0489 hours <br />2.910053e-4 weeks <br />6.6968e-5 months <br />).
The areas reviewed included operational safety, licensee action on previous inspection findings, NRC Bulletics, Information Notices end Generic Letters, licensee reports, service water system corrosion, fitness for duty and miscellaneous follow up issues.
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Results: No v'olations were identified.
With respect to the follow-up of
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Licensee Event Report 87-017, an unresolved item was identified regarding the testing of certain safety injection system valves associated with the safety injection accumulators.
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TABLE OF CONTENTS Eage
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I.
Persons Contacted......................
2.
Summary of Facility and NRC Activities I
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a,
' Resident Inspector Activities
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b.
Visiting Inspactor Activities
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c.
Plant Status......................
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3.
Operational Safety
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a.
Plant Inspection Tours (71707, 71710)*.........
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b.
Operational and Security Events (93702)
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4.
Licensee Action on Previous Inspection Findings.......
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I a.
Unresolved Item.87-10-04 (92701)
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b.
Unresolved Item 87-16-02 (92701)
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c.
Unresolved Item 87-23-01 (92701)
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d.
Unresolved Item 87-26-01 (92701)
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e.
Violation 87-02-02 (92702)
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Violation 87-16-01 (92702)
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Violation 87-20-01 (92702)
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5.
Licensee Reports
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Licensee Event Report 87-017 (92700)........
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Licensee Event Report 88-001 (92700)........
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10 CFR.21 Report 87-88-05 (92700)........
6.
NRC Bulletins, Information Notices and Generic Letters
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NRC Bulletin 83-01 (92703)......,.......
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NRC Bulletin 88-02 (92703)..............
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NRC Generic Letter 87-12 (92703)
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d.
NRC Information Notice 88-05 (92701)
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7.
Fitness for Duty Pengram (92701)
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Service Water System Corrosion (62703)...........
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Follow-up Issues - (92701)
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a.
Station Information Report 87-92
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Reactor Coolant Pump Locking Cups............
10. Open Items..........................
11. Unresolved Item.
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12. Management Meetings.....................
"The NRC Inspection manual inspection procedure that was used as inspection guidance is listed for each applicable report section, i
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OETAILS 1.
Persons Contacted
- W. A. DiProfio, Assistant Station Manager T. C. Feigenbaum, Vice President, Engineering and Quality Programs W. J. Hall, Regulatory Services Manager
- 0. E. Moody, Station Manager G. S. Thomas, Vice President, Nuclear Production J. M. Vargas, Manager of Engineering
- J. J. Warnock, Nuclear Quality Manager
- Attended exit meeting conducted on March 30, 1988.
Interviews and discussions with other members of licensee and contractor management, and with their staffs, were also conducted relative to the inspection of items documented in this report.
2.
Summary of Facility and NRC Activities a.
Resident Inspector Activities The Senior Resident Inspector attended a conference of the American Nuclear Society on Febrtary 29 - March 2,1988 in King of Prussia, Pennsylvania.
The Senior Resident Inspector participated in an inspection at the nuclear facility at the University of Lowell on March 15-16 and March 22, 1988.
The Resident Inspector attended a two week training course in advanced boiling water reactor technology at the NRC Technical Train-ing Center commencing on February 28, 1988.
b.
Visiting Inspector Activities On February 16-19, 1988, an NRC:RI emergency preparedness specialist conducted a ' routine inspection of the NHY emergency preparedness program.
During that inspection (Inspection Report 50-443/88-03)
I several NRC open items were reviewed.
On February 18-19, 1988 an NRC:RI senior reactor engineer and an NRR corrosion specialist conducted an inspection of the service water system. They attended a licensee presentation on the status of cor-rective action on February 18, 1988.
Their inspection findings are included in paragraph 8 of this report,
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On February 22-26, 1988 an NRC:RI operations specialist conducted a routine operational safety inspection.
During that inspection, (Inspection Report 50-443/88-04) NRC Temporary Instruction 2515/91 concerning Generic Letter 83-28 was reviewed.
On February 29 - March 4,1938, an NRC:RI physical security specialist conducted a routine inspection of the NHY physical security program.
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During that inspection (Inspection Report 50-443/83-05) several NRC open items were reviewed, c.
Plant Status i
During this reporting period, the plant remained in operational Mode 5, cold shutdown, with primary temperature between 110 and 140 degrees F and depressurized. Major maintenance was conducted on the circulating water (CW), service water (SW), residual heat removal (RHR) and diesel generator (DG) systems.
Passivation procedures for layup of secondary systems were conducted. Replacement of Gould type J-10 relays continued. One operational event occurring as a result of this activity is described in paragraph 5 concerning Licensee Event Report 88-001.
0 eratienal Safety 3.
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plant Inspection Tours The inspectors observed work activities in progress, completed work and plant status in several areas during general inspections of the plant. The inspectors examined work for any apparent defects or non-compliance with regulatory requirements or license conditions. Par-ticular note was taken of the presence of quality control inspectors and quality control evidence such as inspecticn records, material identification, nonconforming material identification, housekeeping and equipment preservation.
The inspectors interviewed station staff, craft, quality inspection and supervisory personnel as such personnel were available in the work areas.
During control room observation periods, during both normal working hours and on backshif ts, the inspector reviewed control room logs and records including night orders, shift journals, shift turnover sheets, completed repetitive task sheets, the temporary modifications log, weekly surveillance schedules and control board indications.
Specific note was taken of equipment in "pull-to-lock" conditions, equipment tagged, alarm status and adherence to technical specifica-tion limiting conditions for operation (LCO) and action statements.
Also, boron samples, taken from the reactor coolant system and con-nected water supplies, were spot-checked for concentration, sample frequency and documentation in accordance with specified zero power license (NPF-56) conditions.
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The inspector verified the proper position, in accordance with oper-ational procedure or tag-out controls, of specific valves during system walk-downs and checked the valve status in the control room.
Similarly, temporary modifications and component tagging, maintenance work, and design change implementation activities, as observed during plant inspection tours, were evaluated for evidence of both proper field controls and coordination of the subject work activity with the control room and operations personnel on shift.
In certain cases, the operability of specific components and the applicability of the observed work to the TS requirements were discussed with the operators.
Specifically, RHR train operability and valve lineups, steam genera-tor blowdown solenoid valve conditions and status, safety accumulator root valve. replacement activities, and the full stroke testing and results for the main steam atmospheric relief valves were inspected and discussed with operations, maintenance and technical support engineering personnel during the conduct of plant tours.
Also, general tours of the service water pump house, cooling tower, con-tainment enclosure areas adjacent to the primary auxiliary building, and the protected area fence line were conducted by the inspectors.
No violations were identified, b.
Operational and Security Eveny Several events of minor safety significance and/or reportable in accordance with 10 CFR 50 or 10 CFR 73 occurred during this inspec-tion period.
These events are documented below:
(1) On February 11, 1938 the licensee determined that the supply breaker for inverter EO-I-28 supplied from unit substation EO-US-51 had not been tested in accordance with Technical Spec-ification 4.8.4.2.
Further review of this event will be con-ducted in the review of Licensee Event Report 88-002.
(2) On March 7, 1933 NHY discovered a potential avenue of access from one vital area into another.
The access levels of the two areas were not identical and as a result, the area with a lower access level assigned was upgraded.
There were no adverse con-sequences noted as a result of this incident.
A one-hou'; non-emergency notification was made to the NRC Operations Center at 4:05 a.m. on March 7,1938 in accordance with 10 CFR 73. This event vill be the subject of NRC security specialist follow-up in review of the forthcoming Security Event Repor _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
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(3) On March 12, 1983 between 7:00-7:30 p.m.,
two fires were
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reported within one mile of Seabrook Station. One fire to the
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south of the site at an abandoned lumber warehouse presented no l
concern.
The other fire, approximately one mile northwest of the plant, involved a chemical distribution warehouse, which stored toxic chemicals and, as a general licensee, was reported to have stored small quantities of radioactive materials.
New Hampshire is an "Agreement State" and the State Radiation Health Program Manager had been notified about the fire, lhe licensee responded to a request from local authorities for assistance by providing three health physics personnel for radiological monitoring and also provided meteorological data from the site tower.
All radiation measurements taken of the air, fire-fighting water, equipment, and personnel revealed no levels above normal background levels.
At about 11:00 p.m., the local authorities evacuated a 1"ailer park near the chemical fire. At 11:45 p.m.,
FEMA contacted the NRC Headquarters Operations Officer (H00) regarding the situa-tion.
The NRC Senior Resident Inspector was dispatched to the
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plant. His review of the situation and discussion with both the
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Contrni Room Operator and offsite authorities indicated that at no time was the plant adversely impacted or threatened by the fire.
The control room ventilation system remained aligned with a fresh air intake, although the capability did exist to isolate outside air and maintain recirculation flow to the control room.
The inspector verified that licensee management had been apprised of the offsite conditions by the operators on shift and
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that operational decisions were based upon monitoring wind direction and other pertinent information.
A member of the licensee's non nuclear security guard force, stationed at the main gate to the owner controlled property, was temporarily relocated to avoid the smoke from the fire. The pro-tected area, further distanced from the fire, was not af fected.
i At approximately 2:15 a.m. on March 13, 1988, local residents of the trailer park were allowed to return home. The licensee Unit Shift Supervisor notified the NRC H00 via the ENS of the situa-tion returning to normal.
The Senior Resident Inspector inter-viewed the licensee health physics personnel returning to the site.
No violations or unresolved inspection issues were identified as a esults of NRC follow-up of this off site even _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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(4) On March 10, 1988, seismic response spectrum recorder. SM-XR-6707-T, located in the prima ry auxiliary building was found inoperable during technical specification (TS) surveillance.
Severe corrosion and moisture damage was noted.
Although the sensor was sealed, moisture intrusion occurred. The location of tne instrument in a moisture rich environment appears to have contributed to the failure.
Licensee investigation is continu-ing with respect to any design changes that may be required.
In accordance with TS 3.3.3.3 and 6.8.2, a written Specta, Report was submitted to NRC:RI by letter NYN-SB033 dated March 21, 1986 4.
Licensee Action on Previous Inspection Finding; a.
(Closed)
Unresolved Item (87-10-04): Startup Feedwater Pump Design Questions.
This item was updated in NRClIC1nspection Report M 4137 li7 13.
Remaining inspection effort related to completion of addt-tional modifications to the startup feedwater pump (SUFP) and its control circuits. Additionally, NRC attention was required to verify adequate testing and implementation of procedure MT3.1.
Two design coordination reports (OCR) were issued.
DCR 87-124 was approved in August,1987 ar.d DCR 87-280 was approved in July, 1987.
DCR 87-124 changed the control logic for the SUFP pre-lube pump. DCR 87-280 lowered the setpoint of pressure switches FW-PSL-4233-1 and FW-PSL-4233-2, eliminated the actuation time delay and installed a hydraulic snubber in the pressure sensing line.
Following implemen-tation, a special test procedure was prepared to verify system oper-l ability and function following the design changes Special test procedure (STP) 106 was approved on December 24, 1937, and completed on January 7, 1983.
The inspector reviewed the results of STP-106 and noted particular attention was paid to verifying that the safety related featt *es of the SUFP were verified to be unaffected by the modifications.
The inspector considers implementation of this test to be a significant improvement in licensee attention to detail in the area of post modification testing.
This item is closed, b.
(0 pen) Unresolved item 87-16-02: RHR Line Weld Failures.
This item consisted of five separate issues as listed below:
(1) Evaluation of whether Interpretation XI-1-83-85 applies to ASME repairs as well as replacements.
(2) Evaluation of whether operation of the primary component cooling water system with cross connects through the thermal barrier heat exchangers and containment air handling fan coolers i; an acceptable method of heat removal with respect to single fai N re criteria.
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(3) Operability of systems with inoperable subsystems or support systems when operational criteria are met, but design bases have
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not been addressed.
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(4) Acceptability of the repair of the train
"B" residual heat removal (RHR) recirculation line weld failure on August 7,1987 while still declaring the system operable.
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(5) Investigation into the common mode failure of these welds.
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NRC review of the first tour items is ongoing.
The licensee has
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conducted an evaluation u the mode of failure of these weld lines, Engineering evaluation 87-037, entitled "Review of Instrument Source,
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Vents & Orains Connection Design and Installation Standards" was com-
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pleted on January 29, 1938. This evaluation discussed the RHR recir-s culation line failures which it attributes to excessve vibration.
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The report identified eighteen safety related assemblies with high fluid velocity which have a potential for instrument source connec-tion failure due to flow induced vibration. Data had previously been
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taken for two assemblies and more test data are required.
The in-
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spectcr reviewed special test procedure (STP) 109 entitled "RHR Flow Orifice Vibration During Dynamic Conditions".
This procedure is for the four RHR cold leg flow orifices and for the two RHR hot' leg l
orifices.
Due to a recently discovered failure of train "B" RHR
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recirculat. ion valve, RH-FCV-611, the scheduled implementation date of STP-109 is presently indeterminate.
The inspector held discussions with the System Support Manager concerning potential alternatives available in the RH-FCV-611 repair. Since the RHR recirculation line contains both RH-FCV-611 and the previously failed welds, it is
theori:ed that the unusually high vibrations which caused the weld I
failures also damaged the disc guide for RH-FCV-611.
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will maintain close liaison with the licensee concerning this repair and performance of STp-109, with specific interest concerning the operability of the train
"B" RHR system.
This item remains open.
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(Closed)
Open Item (87-23-01):
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Actuations.
NHY formed a task force to investigate recent ESF/5SPS
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actuations.
The team consisted of members of the engineering, nuclear quality, technical support and training staffs.
The task
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force was requested to determine if root cause investigations were complete and correct and if there is a common root cause for which
generic corrective measures can be taken.
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No cornon root cause for the events was identified, however certain l
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p rog ramma t ic, weaknesses were noted in the follow-up of events.
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Several recomendations were made to improve the analysis process.
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An action plan was developed which should improve incident evalua-tion, root cause determinatien and corrective action in the future.
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The inspector had no further questions.
This item is closed.
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(Closed)
Unresolved Item (87-26-01): Control Room Pressurization Problems. This item remained open pending resolution of the foilew-ing two isses:
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(1) The potential reportability of this event based upon a loss of an engineered safety feature.
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(2) Short-term actions required of the licensee to ensure that a similar situation cannot exist undetected for any length of time until a
final design modification to the subject pressure control functions is implemented.
The safety status of the makeup air subsystem, specifically the l
ability to keep the control room pressurized, has been the subject
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of NRC/ licensee discussion in the past. While it is clear that the
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control room must remain pressurtred to achieve the dose limits of 10 CFR 50, Appendix A, Criterion 19, the Seabrook FSAR does not specifically identify the makeup air subsystem as an engineered l
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safety feature.
Furthermore, the exhaust fan, CBA-FN-15 and its i
associated modulating damper, CBA-OP-28 are not safety related, and
do not meet single failure criteria, and the control systems are
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classified as non-1E.
j With respect to the condition existing on September 20, 1987, the
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licensee stated its position that radiation at the intakes would
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still have caused an isolation of the makeup system in accordance i
with the system design basis.
The only fan operating would then be l
the cable spreading room exhaust fan and its capability to draw a a
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vacuum on the above located control room via normal leakage paths is very small.
This situation is within the bounds of the assumptions i
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most recently accepted by NRR with respect to previously identified I
CBA design deficiencies (Reference letter NYN-87051, dated April 9,
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1987 and NUREG-09S6, Seabrook Safety Evaluation Report, Supplement l
No. 7),
Assuming that the control room operators followed normal I
operating procedures, the makeup system would be restored well within
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the one-hour assumption described in the above letter. As a result i
of this information and discussions held with the Regulatory Services
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Manager and the cognizant corporate engineer, it is concluded that
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the plant was never in an unanalyzed condition.
Since the rakeup isolation function was still operable, no engtreered safety feature f
was degraded and the incident therefore was not reportable, i
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With respect to measures taken to preclude recurrence, the licensee
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initiated design coordination report (DCR)87-379 to make two system j
modifications. The first change involved switching the differential
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pressure comparator, CBA-P0Y-5312-2 from a "greater than" unit to a i
"less than" unit.
This ensures that exhaust damper CBA-DP-28 will i
modulate to maintain minimum pressurization with respect to either
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cutside atmusphere or the cable spreading room.
Secondly, alars
points were inserted on the video alarm system to alert the operator of low differential pressure on either sensing instrument after a l
ten-minute time delay.
The time delay allows normal control room i
pressure fluctuations to occur without alarm yet is set to identify.
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a differential pressure mismatch such as that which had occurred on September 20, 1937. The inspector reviewed the OCR and verified in the field that alarms F7111 and F7043 were active.
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It is noted that the installation of this design change was the re-
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suit of NRC questioning following the event.
The rapid action and development of the DCR provided evidence of a licensee responsiveness
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to NRC concerns.
The inspector discussed with the licensee their
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plans for future action in response to similar situations, based upon i
internal initiatives, rather tMn NRC questf oning.
This item is
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closed.
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(Closed) ViolatSn (87-02-02): Centrol Building Air Handling System i
~ NKY responde3 to this violation in a May _15, 1987 l
Valve Misalignment, letter to the NfC (NYN-8704).
The inspector reviewed licensee cor-rective actions as described in the letter and verified that the pro-
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cedure changes were field complete.
A procedure change was deter-I mined to adequately address those NRC concerns regarding operation
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up to 5*4 power. On January 22, 1983, NHY described the details of l
the new control building air handling (CBA) system which has been i
redesigned as a result of certain NRC identified deviations between
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the as-described and as-built systems.
Since this new design change i
eliminates the pt.rge valves which were the subject of this violation,
no further licensee actions are required to address this specific t
i issue. The NRC will continue to follow the CBA redesign process from j
l coth a licensing and inspection perspectives.
The licensing issue l
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remains cpen in Supplement 7 of NUREG-0396, The Seabrook Safety j
Evaluation Report (paragraph 6.4, Habitability Systems). The inspec-r tion issue remains open through the tracking of the original devia-
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tion 86-54-01.
This violation is therefore closed.
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(Closed) Violation (87-16-01):
Control Building Air Handling System
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Unarmed. This violation related to the discovery dueing surveDiance
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testing that the isolation function of the control room makeup air L
subsystem was inoperable for both trains.
The circuit design had
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been previously noted by the inspector to contain a feature whereby the reset function following an isolation required the manual intake I
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valves to be cycled.
However there was no method of determining at
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any given time that the system was armed without inserting an isola-i tion signal.
Licensee actions to establish measures to ensure that i
the channels were always ar,aed were inadequate, given the reset logic
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def,ign featvres, and a violation was identified on August 19. Licen-
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see corrective actions were specified in NHY letter NYN-87132, dated l
November 17, 1987 and Licensee Event Report 87-016-00, dated i
September 18, 1937.
These actions included the following:
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(1) A temporary
'ication was installed that added a white indi-
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cator light sach of the four radiation monitor haittion circuits to ate the armed status for the isolation upal
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control lor These lights indicate when each circuit is armed.
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(2) Changes were made tc the Technical Specification surveilla ce procedures for the four radiation monitors to ensure that the system is armed upon completion of the Jrveillance.
I (3) Changes were made to procedure 0510h '
- ontrol Room ventil6-l tion and Air Conditioning System OpeNcion, to sddress damper
operation, clarify centrol logic arming sequence, and explain
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the new indicator light operati0n.
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(4) A directive was issued for eatn control room shif t to log the status of the lights in the remarks secticn of the Technical
Specification Leg.
NRC review nas deter:nined that the above corrective actions have pro-vided a positive method of ensuring that the isolation channels rematn available to fulfill their design functions.
Since a major redesign of this system is scheduled prior to exceeding 5'i power, long term follow-up of tH s issue will be handled in association with closure of deviatien C6-54 ':1 which will retain open pending final design change implementation.
This violation is closed, g.
(Closed)
Violation (87-20-01):
Equ3 sent Tagging and I sol at t or!.
This violation involved a failure to properly implement the require-cents of the equipment tagging program with respect to work on the service water (SW) system. NHY responded to the Notice of Violation in letter NYN-87111, dated September 18, 1937.
This response included irrediate corrective action to place the SW system in a proper configuration and to audit all active tagging orders in order to ensure that the identified problem was an isolated case.
The tagging procedure, MA4.2, was revised to incorporate lessons learned f rom this event with respect to the "stand-alone" nature of tagouts and modifications to tagouts involving boundary valves.
The indi-viduals cogni: ant of the specific tagouts in question were counseled on the above items.
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i Long term corrective action included training for all individuals identified on the Switching and Tagging List.
This training was to ensure that the requirements set forth in MA4.2 were understood.
The Independent Safety Engineering Group (ISEG) made several recom-mendations concerning tagging practices in their December 10, 1987 report to the Vice-President, Nuclear Production.
ISEG involvement in st ' issues demonstrates the NHY desire to independently evaluate the et
- tiveness of licensee programs.
Since tr, s violation occurred, several minor tagging discrepancies have been noted by the inspector; however, licensee tagging proced-ures and their implementation have demonstrated general improvement.
NHY quality assurance involvement in the tagging program continues to identify areas where additional improvements could be made and tHs self-assessment is considered to be a valuable effort in program enhancement. NRC routine inspection will continue to monitor the NHY tagging program and verify an improving trend. Based upon the above licensee cction and the improving results, this violation is closed.
5.
Licen;ee Reports a.
(Closed)
Licensee Event Report (LER) 87-017:
Technical Specifica-tion Surseillance Requirement Not Satisfied Due to Procedural Inadequacy.
This event and subsequent licensee corrective action were reviewed and documented in NRC:RI Inspection Report 50-443/
87-24.
Since procedural changes were being processed, this item remained open pending the review of the applicable procedure revis-ions and a check for the completeness of licensee corrective measures.
During this inspection, thirteen surveillance procedures were spot-checked to determine whether the required engineered safety feature (ESF) slave relay actuativ signal and device continuity tests had been included as part of <s quarterly operability verification of the different engineered safeguards circuits. While some minor reference and recording errors were noted by the inspector and brought to the attention of the cognizant licensee personnel, the procedural revis-ions generally corrected the problem identified in the subject LER by prescribing the proper criteria and performance requirements for the functional checks of the ESF slave relays. The inspector examined a sample of the applicable electrical schematic diagrams, reviewed the FSAR and system descriptions to check the required component posi-tions upon ESF actuation, and discussed procedural implementation with operations and I&C personnel. Station Information Report, (SIR)87-101, was also reviewed to determine the scope of the licensee ESF anaiysis.
The inspector noted that the licensee had conducted a complete review of all events discussed in Section 15 of the FSAR and had selected for evaluation those events categorized as American Nuclear Society (ANS) Condition III and IV events.
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The inspector had no questions regarding the scope or completeness of licensee corrective actions in response to LER 87-017 and considers this item closed.
However, during the review of the procedural changes associated with the slave relay testing for the safety injec-tion accumulator isolation valves, a system design question was raised as discussed below.
Both the electrical schematic details and logic diagram NHY-503907, for the accumulator isolation valves illustrate a design feature in the control circuitry which has been effectively rendered nonfunc-tional by the Technical Specification (TS) requirements. This design feature described in section 7.6.4 of the FSAR, provides a "maintain CLOSE" control switch position to manually block automatic opening of the accumulator isolation valves above the safety injection unblock pressure. However, TS 3.5.1.1 and 3.5.1.2 require that the isolation valves be open with power removed in MODES 1,2 and 3 and be closed with power removed in MODES 4 and 5.
Therefore, except for mode changes requiring the repositioning of the valves or valve testing, the accumulator isolation valve movement is precluded and the
"maintain CLOSE" switch position performs no useful function.
Not only is the FSAR description of this accumulator isolation valve control feature inconsistent with the current TS requirements, but also certain other problems were identified with the design as noted below. With the switch in "maintain CLOSE" and power provided to the accumulator isolation valves;
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(1) The guidance of IEEE Standard 279-1971, a commitment as provided in the FSAR, is not followed with respect to the automatic removal of blocks of the protection system.
(2) The reset of a safety injection signal would cause the accumu-
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lator isolation valves to stroke closed, contrary to the NRC
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position stated in IE Bulletin No.80-06 that upon the reset of an ESF actuation signal, all safety-related equipment should remain in its emergency mode.
(3) The preoperational test, 1 oT(I)-38, for ESF Integrated Actua-tion had only tested the opening of the subject valves upon receipt of a safety irjection signal with the control switch in
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I the "AUT0" position; not in "maintain CLOSE".
The inspector discussed the above inconsistencies with engineering, regulatory services and station staf f personnel.
It was noted that, while complying with the Technical Specifications, valve switches incorrectly placed in the "maintain CLOSE" position will present no (
problem because of the requirement that power be removed from the valves.
Also, switch position is checked on monthly surveillances.
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However, it was not clear that the licensee had evaluated the possi-bility of an accident scenario during mode changes or valve testing with power available to the valves, and switches incorrectly posi-tioned.
This situation, coupled with the need to review an FSAR revision and evaluate the completeness of the preoperational testing of the subject valves, indicates that further action and/or analysis is required by the licensee.
Pending additional justification from the licensee as to why a "maintain CLOSE" design feature for the accumulator isolation valves is acceptable or the initiation by the license of a design change to modify this feature and revise the FSAR, this issue remains unresolved. (50-443/88-02-01)
6.
(Closed)
Licensee Event Report (88-001):
Inadvertent Actuation of Control Building Emergency Cleanup System.
On January 29, 1988 an
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inadvertent actuation of the control building emergency cleanup sys-tem, a subsystem of the control building air handling (CBA) system, occurred. In addition a train
"A" control room ventilation isolation signal (CRVI) was generated. These portions of the CBA system repre-sent an engineered safety feature (ESF).
As part of the corrective action for LER 87-019, certain Gould/
Telemecanique relays required replacement. At the time of the event, a technician was working on a J-10 relay associated with the makeup isolation and emergency filtration features of the CBA system (Relay E42/9a3-4).
In removing the still energized contact block from the previously de-energized coil assembly, he depressed the contact plunger causing the contacts to momentarily change state from the de-energized condition to the energized condition.
These contacts perform the following functions:
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De-energizes the solenoid device associated with CBA-DP-53B causing the damper to close which trips CBA-FN-278.
This fea-ture was known and had been anticipated.
The licensee had de-energized this equipment in advance.
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Seals in power to relay 3-4 to lock in the CRVI signal.
In this case however, power to this circuit was tagged out so the coil of relay 3-4 remained de-energized.
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De-energizes the main contacts for CBA-FN-27A causing the fan to trip. This occurred.
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Energizes the alarm circuit for the train "A" makeup air isola-tion alarm on the video alarm system. This occurred.
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De-energizes the solenoid device associated with CBA-DP-27A.
Upon the opening of DP-27A, the train "A" emergency cleanup fan CBA-FN-16A starts.
This occurred.
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The plant responded as designed and the operators restored the system to a normal configuration once the maintenance workers had suspended-work on the 3-4 relay.
Following the event, a technical evaluation of the relay replacement in the CBA system was conducted by a joint operations / technical sup-port committee.
The evaluation concluded that follow-on replacement work should be accomplished on both trains simultaneously in oroer to minimize the total time that the CBA system would be out of service.
The estimated outage time was well within that allowed by Technical Specification (TS) and under these circumstances, the inspector agrees that voluntary entry into the action statements of TS 3.7.6 was appropriate (Refer to NRC:RI Inspection Report 50-443/87-16, paragraph 3.b).
The inspector reviewed the preliminary station information report (SIR)88-011, the video alarm log, tagging order 88-228,.and LER 88-001 which was transmitted to NRC:RI on February 29, 1988 (NYN-88027). NRC inspection of this LER involves two activities; review of the LER for completeness and accuracy in accordance with 10 CFR 50.73 and a technical review of the event for root cause and lessons learned.
The inspector discussed LER 88-001 with the cognizant regulatory services engineer.
It was determined that the LER met the minimum criteria for reporting and content. The inspector noted however that the LER text could have been more detailed in describing the makeup isolation since the isolation function is considered to be a report-able engineered safety feature (ESF) per the NHY Reporting Manual, Figure 3-6-1.
The inspector conducted a technical review of the actuation by com-paring the video alarm logs with circuit schematics and relay development data.
He discussed with NHY personnel the maintenance activity in progress at the time of the actuation and verified that all systems responded as expected.
The inspector noted that this is the second event which has occurred recently involving incomplete pre-test or pre-maintenance component / actuation feature isolation.
The particular event is attributed to personnel error in that the technical support engineers failed to adequately review the system schematics when generating the implementation package for the J-10 relay design change.
The licensee recognizes that future similar events with safety significance may indicate a programmatic weakness requiring additional attention.
With respect to all the above inspection activity, no violation was identified.
This LER is close (
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c.
(0 pen)
10 CFR 21 Report (87-88-05): Westinghouse Type W-2 Cell Switch Deficiencies.
This issue was initially discussed in NRC:RI Inspection Report 50-443/87-24.
On December 7,1987 the NRC issued Informatfor Notice 87-61 on this subject.
In a subsequent letter from Westinghouse to NHY (NAH-3316, dated November 5,1987), poten-tial concerns regarding the failure mode of P-4 auxiliary contacts were discussed with respect to the "turbine trip on reactor trip" feature.
Westinghouse recommended that testing requirements be reviewed to determine if this potential failure would be detected.
Licensee action on this item is not yet complete.
On October 21, 1987 the licensee performed inspections of all switches per work request 87W7488. Although visual switch inspection revealed no problems, it was decided to replace the switches upon receipt of replacements.
This will be accomplished under work request 87W7743.
This item remains open pending licensee review of procedures and ccmpletion of switch replacements.
6.
NRC Bulletins, Information Notices and Generic Letters a.
(0 pen) NRC Bulletin 88-01:
Defects in Westinghouse Circuit Breakers.
Several industry failures of Westinghouse (W) type D5-416 circuit breakers used in reactor trip circuits have been recently reported.
The problems stem primarily from pole shaft weld failures.
As a result, W issued Technical Bulletin NSID-TB-87-11 on December 1, 1987.
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NRC Bulletin 88-01 was issued on February 5, 1988 and mandated a short-term and long-term inspection program at all facilities using type 05-416 and similar circuit breakers.
Although Bulletin 88-01 added some additional inspection criteria, it essentially dictated the conduct of the inspections recommended in the W Technical Bulletin.
NHY procedure MS 88-1-1 entitled "Inspection of Reactor Trip Breakers Pole Shaf t and Alignment" was approved on February 24, 1988.
Inspec-tions commenced on February 26, 1988 of all four reactor trip breakers (RTB), as well as the warehouse spare. Preliminary results indicated unacceptable pole shaf t welds on all breakers although the spare breaker was determined to be useable for up to 4000 cycles per the inspection criteria. The licensee has notified V and is awaiting technical resolution of this issue.
In the meantime, all RTBs are considered inoperabl..
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The inspector reviewed MS 88-1-1 and compared the acceptance criteria with that given in Bulletin 88-01.
He witnessed initial breaker disassembly and testing, noting licensee electrical supervision, quality ~ control inspectors and qualified non-destructive examination personnel to all be knowledgeable in their. assignments. He verified that several breaker welds did not meet the acceptance criteria and concurred in the licensee judgements with regard to operability.
No violations were identified.
The inspector determined that NHY is placing proper emphasis and effort in the resolution of this problem which will remain open until the breakers are returned to operable status. This bulletin remains open.
b.
(Closed) NRC Bulletin 88-02:
Rapidly Propagating Fatigue Cracks in Steam Generator Tubes. This bulletin was sent to Seabrook for infor-mation only since the plant uses Westinghouse Model F steam gener-ators which have stainless steel alloy support plates.
The problem addressed in this bulletin applies only to earlier models with carbon steel support plates. This bulletin is therefore closed.
c.
(0 pen) NRC Generic Letter 87-12: Loss of Residual Heat Removal While the Reactor Coolant System is Partially Filled.
NRC Information Notice 87-23 entitled "Loss of Decay Heat Removal During Low Reactor Coolant Level Operation" was issued in May, 1987 following the April, 1987 loss of decay heat removal event at another operating plant.
In that instance, lowered reactor vessel level allowed vor-texing and air entrainment to occur in the residual heat removal (RHR) system.
As a follow-up to this event, Generic Letter (GL) 87-12 was issued in July,1987. This letter requested that licensees provide the NRC with a description of the plant operations during the approach to and during operation with a partially filled reactor coolant system (RCS) condition to ensure that licensing bases are met.
The licensee responded to GL 87-12 by letter (NYN-87114) on September 21, 1937. NHY is participating in the Westinghouse Owner's Group initiative to assist in developina appropriate responses to this issue.
The most significant initiative to date is Seabrook's prcposed RHR/
RCS Mid-loop Operation Test. This special test is to be conducted to determine the maximum RHR system flowrate that may be achieved with-out cavitation during mid-loop operations.
Special test procedure (STP) 105, Revision 0, was reviewed in detail by the inspector.
He traced out system flowpaths and lineups on the applicable piping diagrams. The procedure was reviewed for consistency with GL 87-12, specifically with respect to level indications and RHR pump cavita-tion shutdown criteria.
NRC inspectors will continue to follow the planned test activities.
No violations were identifie.
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d.
(Closed)
Fire in Annunciator Control Cabinets.
This Information Notice provides information on three separate occurrences at other plants where all control room annunci-ator alarms were lost because of electrical fires.
The inspector reviewed the design of the Seabrook alarm and annunciator system and examined the emergency response procedures applicable to the classi-fication of events related to the loss of alarm capability.
The occurrence of any postulated single event which would cause the loss of all alarm capability was determined to be highly improbable because of Seabrook design features.
Only a limited number (less than 100) of hardwired annunciators are installed on the main control board. All are powered by uninterrup-tible power supply sources; some of which come from a train " A" dc bus, others from a train
"B" dc bus.
Trains
"A" and
"B" are physically independent in accordance with IEEE guidelines.
Each de bus has two 100*4 redundant, safety-related batteries available for power upon loss of normal ac power.
The video alarm system (VAS),
provides most of the plant alarms on CRT screens throughout the Con-trol Room.
Loss of specific hardwired annunciator power is itself alarmed on the VAS.
In several cases, VAS alarms are redundant to the hardwired annunciators. Loss of the VAS is handled by an abnor-mal operating procedure, "Loss of the Plant Computer", which requires increasing the frequency of several surveillances and minimizing planned plant transients.
Also, the loss of the VAS (e.g., main
- omputer and back-up f ailure) wo.:Id not af fect the hardwired annun-ciators, which provide the most vital safety information. No single component or electrical f ailure can render al the hardwired annun-ciators inoperable at the same time.
The inspector reviewed the Seabrook Emergency Response Proqram Manual and discussed control room alarm / annunciator design and applicable emergency classifications with operations personnel.
The inspector vecified the existence of appropriate emeroency action levels to address the loss of total alarm capability, depending upon plant l
conditions. Based upon the design features noted above, however, the I
loss of both the VAS and the hardwired annunciators would constitute an unlikely hypothetical situation. Also, the conditions postulated to cause such a situation would dictate the declaration of emergency
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action levels separate from the loss of alarm capability. Thus, no problems related to the generic concerns of Information Notice 88-05 were noted in either the Seabrook system design or operator response capability.
No violations were identified.
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7.
Fitness for Duty Program The inspector examined records, data and a listing of logable security incidents for information relating to the licensee's fitness for duty pro-gram.
Specific questions on the implementation of this program were asked of the licensee and their responses were provided to Region I (RI) in accordance with a RI Temporary Instruction 88-01 issued on January 21, 1988. The inspector evaluated the licensee information in accordance with the reportability recuirements of 10 CFR 73.71 and the guidance of USNRC Regulatory Guide 5.62 and NUREG-1304.
The Security Department Supervisor was interviewed regarding the site policy and the routine follow up actions taken by the licensee in response to incidents involving personnel fitness for duty.
During this inspection period, the inspector noted evidence of an active chemical screening program and examples where random testing was appro-priately utilized to eliminate any suspicion of problems.
A licensee analysis of all security events logged during the fourth quarter of 1987 was reviewed and transmittal of the quarterly log to the NRC Document Control Desk was noted.
Specific incidents documented in the current quarterly log were discussed with cognizant licensee personnel to deter-mine whether appropriate licensee follow-up action was in progress.
The inspector determined that the licensee criteria for handling reportable events separate from logable events, and their policy for follow-up actions on any incidents involving fitness for duty issues were consistent with NRC guidance in this area.
No violations were identified.
8.
Service Water System Corrosion As documented in NRC:RI Inspection Report No.50-443/87-24, a corrosion problem had been identified by the licensee on the service water inlet line to the train "A" secondary component cooling water heat exchanger, SCC-E-29A.
Although the specific pipe line (identification no.
1-SW-1827-6-LI-16-2) where the problem was found represents a portion of the service water system that i s non-safety-related, evaluation of the corrosion phenomenon continued to provide additional assurance that safety-related se'ctions of the service water system were not similarly impacted by corrosion problems.
A licensee presentation of the status of their continuing evaluation of the service water system was provided on site on February 18, 1988 to the NRC inspectors and a representative from the Office of NRR.
The results of licensee inspection, to date, revealed no evidence of corrosion prob-1 ems in the safety-related portions of the service water piping. However, laboratory analysis of the non-safety-related piping where the corrosion problem was first identified revealed that a microbiologically influenced
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corresion (MIC) mechanism was in evidence and a probable contributor to the corrosion process.
This fact, while providing new information that was not available in inspection report 50-443/87-24, does not materially affect conclusion reached in that inspection; i.e., the licensee's need to continue to evaluate the potential for corrosior, in the safety-related portions of the service water system, regardless of the postulated corrosion mechanisms.
To provide additional data in the ongoing evaluation process, the licensee is planning a remote visual inspection of several hundred feet of buried safety-related service water piping.
While direct visual inspection of several areas of above ground service water piping has revealed no generic corrosion problem, the additional data provided by the planned remote inspection would not only support the planning of long-term surveillance options, but also ensure that corrosion has not adversely affected piping installed early in the construction phase and heretofore inaccessible.
As previously documented, independent NRC inspection of the open service water piping has confirmed that no significant corrosion or biofouling problems tre in evidence.
During this inspection, an inspector witnessed pipe replacement and repair activities, including Belzona lining appli-cation, in the non-safety-related piping near the secondary component cooling water heat exchanger. Piping in the service water pump house and cooling tower was also examined, including a vertical section of pipe that had been r:notely excmined by video camera. The inspectors viewed a video tape of this remote visual examination.
Inspector discussion with licensee engineering and station personnel revealed that other corrosion evaluation activities are in progress, in addition to the planning for the remote visual examination of the buried piping.
These activities include a review of the installation and QA records for the buried service water pipe welding and cement liner appli-cation; a historical investigation of the water treatment controls imple-mented for the non-safety portion of the service water sytem, versus the safety-related side; engineering analyses of the estimate of service water leakage limits that would cause operational or structural problems; and the procurement of MIC Detection Kits by the station chemistry department for the future evaluation of corrosion areas and their causal mechanisms.
In ord r to track the continued licensee plann og, review and inspection (
activities to the point where all the requi red inspection results are
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available and both the licensee and NRC evaluation processes are complete, an inspector follow up item is being opened (50-443/88-02-02). The status of this overall corrosion problem will be reviewed periodically based upon updated developments and/or schedular milestones.
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9.
Follow-up Issues
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a.
Station Information Report 87-92:
The inspector reviewed station information report 87-92 describing a reactor trip signal generated during calibration of nuclear instrument channel N-36 on October 21, 1987.
He discussed the potential reportability of this event with cognizant regulatory services personnel and concurred in their evalu-ation that this event was not reportable in accordance with 10 CFR 50.73.
b.
Reactor Coolant Pump Locking Cups: In December,1987 another nuclear facility reported a f ailure involving reactor coolant pumps similar to those used at Seabrook. The problem involved the locking cups for the turning vane dif fuser holddown bolts.
Several cups became de-tached and some were found in the reactor core. NHY corporate engi-neering department conducted an evaluation of this loose part fail-ure.
Engineering evaluation 87-035 was completed on December 9, 1987 and concluded that the Seabrook design utilizes a split cylinder design rather than the continuous cylinder design employed at the other facility. The split cylinder provides additional holding force and problems are therefore not anticipated. The inspector noted that the NHY evaluation was expeditiously performed once the problem was identified in the industry. The evaluation was comprehensive and the conclusions appropriate.
No violations were identified.
10. Open Items Open items are matters that require further review and evaluation by the inspector.
Open items are used to document, track and ensure adequate follow-up on matters of concern to the inspector.
Open items are dis-cussed in sections 4 and 8 of this report.
11.
Unresolved Items An unresolved item is a matter about which more information is required to ascertain whether it is an acceptable item, a deviation, or a viola-tion. Unresolved items are discussed in sections 4 and 6 of this report.
12. Management Meetings At periodic intervals during the course of this inspection, meetings were held with plant management to discuss the scope and findings of this inspection.
An exit meeting was conducted on March 30, 1933 to discuss the inspection findings during the period.
During this inspection, the NRC inspectors received no comments from the licensee that any of their inspection items or issues contained proprietary information. No written material was orovided to the licensee during this inspection, except for the procedural comments and inspector follow-up questions generally dis-cussed in sections Sa and 7 of this reports.