IR 05000443/1987002
| ML20206L669 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 04/09/1987 |
| From: | Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20206L567 | List: |
| References | |
| 50-443-87-02, 50-443-87-2, IEB-76-02, IEB-76-2, IEB-84-02, IEB-84-2, IEIN-86-106, IEIN-87-008, IEIN-87-012, IEIN-87-12, IEIN-87-8, NUDOCS 8704170198 | |
| Download: ML20206L669 (20) | |
Text
_
.t
..
..
q:
%.
'
<
.
a
.I
,
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
'
Report No.
50-443/87-02 Docket No.
50-443 License No.
NPF-5C Permit No.
CPPR-135 Priority Category B/C
--
Licensee:
Public Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit 1 Inspection at: Seabrook, New Hampshire
'
,
Inspection conducted:
January 3 - March 9, 1987
9
Inspectors:
A. C. Cerne, Sr. Resident Inspector D. G. Ruscitto e ident Inspector J. G. Hun-ctor Engineer
,
Approved by:
/
'
T.C.Elsass y tef, Reactor Projects Section 3C Uatd Inspection Summary: Insrection on January 3 - March 9, 1987 (Report No.
50-443/87-02)
i Areas Inspected:
Routine inspection by twi residen'1! inspectors and one region-based inspector of work activities,' procedures, and records relative to startup testing and license issuance; post < core loading, heat up and hot functional testing; maintenance, surveillance and plant operations; and licensee event reports. The inspectors also reviewed licensee action on previously identified items, including
.
a 10 CFR 50.55(e) report and licensee actions in response to I&E Information Notices and performed plant inspection-tours. The inspection involved 432 inspec-tion-hours by three NRC inspectors.
Results: An apparent violation was identified concerning the inplementation of the design control program with respect to the control building makeup air subsys-tem (Section 4). An unresolved item concerning the IE inverter was identified for futu e follow-up.
With respect to several previously identified open items, including a 10 CFR
.50.55(e) report and responses to I&E Information Notices, licensee actions and corrective measures were verified to be either complete or satisfactorily in' pro-gress.
8704170198 870410 PDR ADOCK 05000443 Q
).
'
..
..
/
DETAILS 1.
Persons Contacted T. C. Feigenbaum, Vice-President, Engineering and Quality Programs W. T. Hall, Regulatory Services Manager a
T. L. Harpster, Director of Emergency Preparedness
,0. G. McLain, Technical Support Manager D. E. Moody, Station Manager G. S. Thomas, Vice-President, Nuclear Production L. A. Walsh, Operations Manager Interviews and discussions with other members of licensee and contractor man-agement, and with their staffs, were also conducted relative to the inspection of items documented in this report.
2.
Plant Status During this report period, the plant changed operational modes from cold shutdown to hot standby for post core load hot functional and pre-critical testing. Mode 4 was entered on February 10, 1987. On February 15, 1987, the
~*
plant entered Mode 3 and normal operating temperature (NOT) and pressure (NOP)
were achieved the following day.
Several events of minor safety significance, based upon the current status of the plant, but either reportable in accordance with 10 CFR 50.72, 50.73, or having testing / schedular impact occurred during this inspection period.
.These events are documented below:
a.
January 8,1987 - Inadvertent control room ventilation isolation.
Refer to Region I Inspection Report (IR) 443/86-54 and further discussion of Licensee Event Report (LER) 87-001-00 submitted as a result of this event in Detail 5.b of this report.
b.
January 17, 1987 - Engineered safety features actuation upon loss of offsite power to an essential switchgear bus.
Refer to further discus-sion and closure of LER 87-002-00 in Detail 5.a of this report.
c.
February 11, 1987 - Containment equipment hatch airlock found to be in-
.l
' operable. At 2:10 p.m., the licensee entered the action statement of S
Technical Specification (TS) 3.0.3 as a result of the identified failure
'
of the linkages which operate te two pressure equalizing valves on each containment equipment hatch air lock door. With both doors inoperable,
,
the Limiting Condition for Operation (LCO) of the applicable TS 3.6.1.3
'
was not met, nor could it be met by the e.ction statements of the TS.
Hence, TS 3.0.3 was applicable and required initiation of shutdown to cold shutdown within one hour.
Repair of the linkages and valves was in progress but not completed by 3:10 p.m. and the cooldown commenced.
At 3:25 p.m. the cooldown was halted because satisfactory completion of a leak test on the outer door rendered that door operable and placed the P
m
..
..
plant in TS 3.6.1.3 with one inoperable air lock door.
Repair work con-tinued on the linkage and equalizing valves on the inner door with test-ing confirming operability of that equipment hatch airlock door at 5:30 p.m.
During the fifteen-minute interval between the initiation of cooldown and the halt of the cooldown process, the plant was determined to have been in a condition described as an Unusual Event by the licensee's Emergency Plan. This condition was declared at 3:30 p.m. and immediately declassified since the criteria bearing on the classification of this event, i.e., the cooldown, was no longer applicable at that time.
A detailed inspection of licensee activities and the conditions surround-ing the classification and reporting of this Unusual Event was conducted by two Region I Emergency Preparedness specialists on February 19-20, 1987. The results of that inspection are documented in IR 443/87-07.
Further NRC follow-up on this event will be directed to the open inspec-tion issues discussed in that report.
d.
February 19, 1987 - Engineered Safety Features (ESF) actuation upon loss of the uninterruptible power supply (UPS) to a vital instrument panel.
While searching for a ground fault on 480-volt emergency bus E-51, a circuit breaker from motor control center (MCC) E-512 to UPS inverter IE was manually opened after the d-c power supply (Battery 1A to 125 volt d-c bus 11A) was verified as available to the inverter. An unexpected ESF actuation occurred.
Licensee analysis at that time attributed the cause of the actuation to a two-second delay between the time the a-c breaker opened and the d-c battery power picked up the load.
At the time of this event, a four-hour report to the NRC was made via the emergency notification System (ENS) in accordance with 10 CFR 50.72 and issuance of an LER is pending. The licensee originally suspected that the subject UPS inverter had not functioned as designed. However, a later attempt to duplicate this event on February 25, 1987 failed to produce the same results, since the inverter picked up its d-c power supply and continued to supply the a-c UPS loads, as designed, without tripping.
It is now believed that the ground fault itself may have caused the event coincident with the opening of the a-c supply breaker.
The licensee intends to conduct additional testing on the subject Elgar inverter in an attempt to establish the cause of the problem. This testing is currently planned in Mode 5 after cooldown from the precriti-cal, hot functional testing of the plant. This issue will be inspected further by the NRC after the issuance of the LER.
However, since satis-factory conduct of the planned inverter testing has a direct bearing on the operability of this equipment, this issue is also considered unre-solved pending completion of the testing, review of the results by the NRC, and corrective action, if necessary, on the part of the licensee to prevent reoccurence of any similar anamolous event (443/87-02-01).
..
..
,
e.
February 21,1987 - Inadvertent actuation of the emergency sirens in Rockingham County, New Hampshire.
Following maintenance on a single siren, licensee personnel requested a silent test of that siren from the Rockingham County dispatcher. The dispatcher mistakenly pressed the wrong button, actuating all sirens in New Hampshire. The sirens sounded for less than 30 seconds. The licensee is investigating possible human factors changes to the siren control panel to prevent recurrence of such an error. The overall adequacy of the public emergency alerting system sirens will be the subject of future NRC/ FEMA inspection.
f.
~ February 23, 1987 - Caustic contamination of the demineralized water storage tank (DWST) and demineralized water (DW) system. This problem was caused by improper operation of the plant water treatment (WT) system.
Although neither system is safety-related, this occurence is significant in the fact.that recovery actions delayed testing for about two weeks and lessons were learned with respect to the correct performance of manual plant operations in such areas. The inspector reviewed the pre-liminary station information report (SIR 87-021) on this subject.
During addition of caustic solution to the neutralizing tank in the WT system, both caustic pumps were in operation with the common discharge valve open. Following completion of neutralization, the auxiliary opera-tor (AO) inadvertently left one pump running and shut the discharge valve.
The positive displacement caustic pump overpressurized the discharge line causing valve seat leakage and injecting concentrated caustic into the on-line WT stream. This rapidly depleted the functional capacity of the demineralizers and contaminated the DWST and DW system which was in operation.
About 30 minutes later, the oncoming A0 noted high conductivity in the WT system and stopped the pump. His alert actions prevented an even more serious contamination problem. The flushing operations required to re-store these systems to an acceptable level of cleanliness were extensive and included draining the 200,000 gallon DWST. The shortage of de-mineralized water created by this event caused a delay in the conduct of precritical, hot functional testing, in progress at that time. No violations or safety concerns were identified regarding this matter.
g.
March 4, 1987 - Inadvertent actuation of the reactor trip system.
In preparation for a scheduled test of the reactor trip breaker (RTB) auto-shunt trip test switch, the RTBs were properly aligned with the "B" by-pass breaker racked in, to allow the "A" reactor protection system (RPS)
logic to control both breakers and permit testing of the "B" trip test switch.
This switch is located inside the back panel of the RTB cabinets.
Upon initiation of the test, however, the "A" trip test switch was mis-takenly pressed, causing a reactor trip with the signal also initiating a feedwater isolation (FWI).
.
+
-
--
.,
v.
,-
%
.,, -
. - -
-y
.. - - <
y e
,
w
..
..
Since all the control rods were in the core and feedwater was limited to. low bypass flow conditions, this event had very little impact upon plant conditions.
The operators reset the FWI signal and continued RTB testing. A four-hour report to the NRC was made via the ENS in accord-ance with 10 CFR 50.72.
The inspector discussed with operations personnel the human factors element to this event which was caused by operator error.
The subject train related auto-shunt trip test switches ("A" and "B") are both located in the back of the RTB cabinets and can be actuated only if the respective train-related cabinet door is opened. However, the "A" and
"B" color-coded markings are located only on the outside of the cabinet doors. Therefore, with the doors opened the relationship of each switch to its related train is not visibly marked. This may have contributed to the noted error and is being pursued by operations personnel as a human factors enhancement item.
Further NRC follow-up of this issue will be addressed after the issuance of the LER on this subject.
h.
March 6,1987 - Identification of a missed TS surveillance activity. The licensee discovered that the daily TS surveillance log had not been com-pleted within its allowable periodicity.
The control room operator mis-takenly selected the "At All Times" logsheet instead of the " Daily" log-sheet. As corrective action, all " Daily" logsheet parameters will be incorporated into the respective " Mode" logs and the " Daily" logsheet will be discontined. The licensee plans to submit a 30-day report to the NRC in accordance with 10 CFR 50.73. The inspectors will review this item further, as a routine inspection follow-up to the LER, when issued.
3.
Plant Inspection Tours The inspectors observed work activities in progress, completed work and plant status in several areas of the plant during general inspections of the plant.
The inspectors examined work for any obvious defects or noncompliance with regulatory requirements or license conditions.
Particular note was taken of the presence of quality control inspectors and quality control evidence such as inspection records, material identification, nonconforming material iden-tification, housekeeping and equipment preservation. The inspectors inter-viewed station staff personnel, craft personnel, supervision, and quality inspection personnel in their respective work areas.
a.
During random inspections of the plant, the inspector noted certain dis-
'
crepancies in the use of safety tags.
Following discussions with the licensee, a quality assurance (QA) surveillance was performed to verify compliance with the Seabrook Station Maintenance Manual (SSMA) as it relates to tagging requirements. The resultant surveillance report (87-00102) indicated similar findings to those identified by the NRC.
These concerns were brought to the attention of station staff by the QC Supervisor.
.-.
-
_ _ _ _ _.
.-
- -.
.
..
..
Another QA surveillance (report 87-00082) was performed in response to the NRC identification of superseded and uncontrolled electrical sche-matic and schedule drawings, posted on the inside doors of several 120/
240 volt a-c circuit breaker panels for vital MCC and other control cir-cuits.
The surveillance findings confirmed the existence of out-of-date circuit listings, representing unauthorized " operator aids". This con-cern was brought to the attention of the operations department, which initiated Request for Engineering Services (RES) 87-0165 to develop a system of controlled document postings for all station power panels and MCC distribution panels. The inspector reviewed a schedular plan for the drawing update and posting of the correct electrical schedules on the subject distribution panels, noting tentative completion of this corrective action in June 1987.
The inspector further questioned the status of one particular 120 volt breaker switch, found in the open position, but noted on the most recent electrical drawing (NHY-310106, SH-E40a) to represent a main control board (MCB) "A" train standby control power circuit. Operations person-nel traced the circuitry to a nonsafety-related receptacle on the MCB, which apparently was never wired or energized.
To address this apparent drawing error, another RES (87-0326) was initiated to revise those docu-ments necessary to illustrate the proper status of the spare cables and circuits on the affected drawings.
This action, in conjunction with the scheduled corrective measures planned in response to RES-87-0165, represents an acceptable approach to resolving the problems identified with the out-of-date electrical drawings.
Licensee response, by both QA and operations personnel, to the tagging and drawing concerns raised by the NRC inspector, appeared to be commensurate with the safety impact of the findings.
The inspector had no further questions on these issues at that time, but will continue to monitor both the safety tagging process and the adequacy of posted circuit drawings, used as operator aids, during future inspec-tions.
b.
Based upon problems identified with certain types of electrical equipment at other nuclear facilities, the inspector questioned the status and use of the following components at Seabrook Station:
Brown Boveri battery ground detector relay (ITE278)
--
--
GE HGA relays The licensee provided evidence that the ITE278 relay is not in use at Seabrook and that the status of HGA relays had been addressed in response to problems identified in IE Bulletins 76-02 & 84-02. Only two HGA re-lays are utilized in a safety-related application at Seabrook, both of i
i
j
- - _ _
_ - _ _
..
..
which provide only alarm functions with respect to underfrequency condi-tions on the diesel generators.
Both installed relays (81X) were re-inspected by the licensee in accordance with recent 1987 work requests.
The inspector had no further questions on this subject.
c.
Region I IR 443/86-34 documented that the installation of caps on plant piping drain lines would be the subject of future NRC inspections. The Technical Support Supervisor published a memorandum concerning station policy on drain and vent caps (SS#28417, December 2, 1986). The inspec-tor reviewed this memo and determined that it provided a comprehensive and rigorous discussion of this subject.
Subsequent plant inspections found no missing caps. The inspector concluded that the station policy has been effectively implemented and had no further questions on this matter.
d.
On January 15, 1987 with the plant in Mode 5, an improper valve lineup caused a high level in the "D" steam generator (SG) which triggered a feedwater-isolation signal. While placing the "D" SG in a component
lineup configuration for wet lay-up chemical addition, level in the SG increased to the P-14 setpoint due to gravity flow of water from the condensate storage tank (CST).
Subsequent investigation revealed that feedwater valve FW-V100 was not procedurally required to be closed; and this incorrect valve positioning caused the gravity flow and high SG level conditions. A procedure revision was subsequently effected to close FW-V100 during such operations in the future.
The inspector reviewed the shift superintendent (SS) report analysis of the event to determine if this inadvertent ESF actuation was reportable in accordance with 10 CFR 50. A comprehensive research of regulatory guidance had been performed by the SS, including reference to NUREG-1022, Supplement 1.
The inspector confirmed the SS interpretation that since FWI was not required in Mode 5, its actuation was not reportable. As a follow-up to this occurrence and a previous event where inspector in-quiry was instrumental in the licensee decision on reportability (refer-ence: IR 443/86-54), the inspector reviewed the lesson plan developed by the Regulatory Services Department to train the operations staff on NRC reporting requirements. The plan appeared to be comprehensive and it included the latest revision to the Nuclear Production Reporting Manual (NPRE).
During subsequent inspector activities and discussions with the opera-tions staff, the inspector noted an increased awareness on the part of the responsible operators of the requisite reporting criteria and re-quirements.
Follow-up of'the above FWI occurrence provided one example of the effectiveness of licensee training on reporting requirements, based upon previous lessons learned.
With regard to all of the above independent inspection activities and items reviewed as a result of plant inspection tours, no violations were identified.
.
_-
. -..
..
.
4.
Licensee Action on Previously Identified Items a.
(Closed) Construction Deficiency Report (CDR) 85-00-07: Heating, Venti-lating and Air Conditioning (HVAC) Design Problems. This CDR dealt with an inadequate HVAC design discovered during the Seabrook Equipment En-vironmental Qualifications Review Program. The licensee determined that additional HVAC capacity was required and four specific modifications were performed:
(1) Equipment vault stairways and electrical area above.
Added chiller units which supply gir conditioners in three areas.
(2) Control building train A switchgear room and train B tunnel.
In-creased ventilation air flow by modifications to existing fans.
(3) Mechanical penetration area. Modified duct work to direct ventila-tion to two equipment areas.
(4) Primary auxiliary building (PAB). Added cooling system to the ven-tilation system for operation when outside air is above 75 degrees F and added duct work for distribution of air to equipment.
New Hampshire Yankee (NHY) submitted the final 10 CFR 50.55(e) report to NRC Region I on June 6, 1986 (SBN-1097).
The inspector reviewed the various engineering change authorizations (ECAs) implementing the design changes listed above and performed field equipment walkdowns to verify as-built system configurations.
Minor drawing discrepancies were noted on certain NHY piping and instru-mentation drawings (P&ID) and were discussed with licensee operations staff personnel. The test results of an acceptance test (1-AT-68) for the PAB chiller system was reviewed, as was the operating procedure, 051023.68, for the residual heat removal vault stairway and the elec-trical area portion of the enclosure air handling system. The inspector questioned the licensee's policy and convention as to specifying normal valve and switch lineups in the operating procedures relative to the illustration of valve or other component position as shown on the P& ids.
The NHY configuration manager responded by indicating that changes to the P& ids would be made to provide consistency with the convention for normal lineups prescribed by the operating procedures. The inspector had no further questions on this issue. This CDR is closed.
b.
(Closed) Unresolved Item 85-35-01: Piping and Instrumentation Drawing Errors. The NHY configuration management department has completed its final P&ID verification program. The inspector conducted a detailed review of several drawings as follow-up to this item, as well as a review
of specific drawings to confirm the adequacy of certain design change activities. While minor typographical and drafting errors were identi-fied, the scope and volume of such errors has been significantly reduced
, - _. -
,
._
..
.
to the point that the NHY P& ids appear fully capable of performing their function. Those particular errors identified by the inspector were pro-vided to the configuration manager for correction. No errors of any safety significance were identified.
The Office of Nuclear Reactor Regulation (NRR) has accepted the NHY in-itiative to provide "B" (Basic) level drawings in the FSAR in an upcoming FSAR change submittal. Additionally, if information on the more detailed
"0" (Design) level drawings is required, the licensee intends to make such drawings available outside the scope of the FSAR. The inspectors had no further questions concerning this matter. This item is closed.
c.
(Closed) Unresolved Item 86-46-03: Containment Air Purge (CAP) System.
This item was closed in Region I IR 443/86-54. As further follow-up to this item, the inspector reviewed the final disposition to SIR 86-030 concerning the failure of containment purge valve CAP-V-3.
This SIR included vendor reports from Posi-Seal International, the valve manufacturer, and ASCO, the manufacturer of the solenoid valve which also failed.
These reports indicated that the solenoid valve failed as a result of copper particles that may have been left in the air line as a result of previous rework to properly orient the solenoid in response to an engineering resolution of an applicable Nonconformance Report (NCR)
93/1924.
The Posi-Seal analysis indicates that the actuator overhang (90 degrees rotation from vertical) did not impose sufficient additional stresses to have caused the valve failure.
It was concluded that the leaking solenoid valve caused a " hunting" condition in the main valve which, when acting against spring pressure, resulted in a large number of oscilla-tions. This then led to galling of the stem and the identified CAP-V-3 failure. The inspector noted that the licensee evaluation of this fail-ure represented an example of an in-depth follow-up to a specific equip-ment deficiency. Such comprehensive failure analysis is considered a positive attribute in the effectiveness of the maintenance / engineering organization.
d.
(0 pen) Deviation 86-54-01: Control Building Air Handling (CBA) System Design Discrepancy. This deviation, as well as several related CBA issues were identified in IR 443/86-54. The following discussion updates the overall CBA issue and includes actions and events occurring subse-quent to issuance of the deviation.
(1) Request for Additional Information (NYN-87013)
On February 9, 1987, NHY provided NRR with an expanded description of the CBA design in Attachment 1 to letter NYN-87013. Attachment 2 contained a definition of single train operation of the CBA system and Attachment 3 contained confirmatory dose calculations.
__
_ _ _, _ _ _ _ _ _ _.. _ _ _
_
_
_,
.
Discussions between NHY and NRR are continuing and the resident in-spectors are following the review precess which will include further safety evaluation and which will be documented in a supplement to the safety evaluation report on the CBA system.
Ultimate determina-tion of the acceptability of both the current system configuration for low power operation and any modified design, which may be sub-mitted prior to issuance of a. full power license, resides with NRR.
Such determination had not been reached with respect to low power license issuance as of the end of this reporting period.
(2) Response to Inspection Report 50-443/85-54 (NYN-87024)
NHY responded to the Notice of Deviation by letter (NYN-87024) dated March 2, 1987.
In this response, the licensee reiterated their position with respect to thei. plan for a future design change.
As additional corrective actions, they indicated the following:
A clear definition for single train operation of the control
--
room makeup air subsystem has been established and will be provided to each shift superintendent.
--
Abnormal operating procedure OS1223.01, " Response to Control Room Ventilation Isolation", has been revised to better de-lineate proper operator response requirements upon receipt of a CBA isolation.
The inspector reviewed the single train definition and determined that, based upon NRR. acceptance of the radiological assumptions and calculations, the definition, as written, is clear and enforceable.
Change 03 to OS1223.01 (Rev.00), " Response to Control Room Ventila-tion Isolation," addressed several of the questions previously raised in discussions among the licensee, the inspectors and NRR reviewers. Specifically, the changes to OS1223.01 ensured that the non-safety related exhaust air system was isolated, provided in-structions for manual isolation of automatic dampers and added a precaution not to run both emergency filter fans simultaneously for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
No changes were made to the normal operating procedure OS1023.51 (Rev. 03) as a result of these discussions, nor were the normal operating procedures addressed in the licensee re-sponse letter to the deviation.
This failure to re-evaluate the effectiveness of the normal operat-ing procedure concurrently with the revision to the abnormal operating procedure appeared to significantly contribute to the violation identified in subparagraph (4) below.
-
- -
-
- - -
..
..
(3) Radiation Monitor Failure (RM-6506B)
On March 3, 1987 the "B" train east air intake radiation monitor failed and the licensee appropriately entered TS Action Statement 3.3.3.1 for inoperable instrumentation. The inspector questioned the shift superintendent as to why TS Action Statement 3.7.6 on the CBA system had not been entered. The inspector provided the SS with the guidance formulated in NYN-87013 and the SS agreed that entry into TS 3.7.6 was appropriate.
The SS directed that the east intake be isolated by shutting 1-CBA-V-9 and the west intake placed in service by opening 2-CBA-V-9.
Following isolation of the east intake, the inspector attempted to verify that the east intake purge valve, 1-CBA-V-4 had been opened and the west intake purge valve 1-CBA-V-2 had been shut.
However, both valves were found shut.
Discussion with the assistant opera-tions manager revealed that the SS had not initiated purge on the isolated intake because he believed that purging was only necessary in the event of radiation in the intake. This philosophy was not consistent with the FSAR section 9.4.1 description of the control room makeup air subsystem functions, as was supportable by design basis calculations available at that time. The decision not to isolate purge, based upon this incorrect philosophy, also contri-buted to the violation discussed in subparagraph (4) below.
(4) Control Room Ventilation and Air Conditioning System Operation, OS1023.51 (Rev. 03)
A non-intent change number 02 to OS1023.51 was issued on January 7, 1987. The procedural changes established a normally closed position for 2-CBA-V-9 and indicated that only one intake should be on-line at any given time to preclude exceeding TS limits on makeup air flow.
While reviewing the actuation logic diagrams for the CBA system in conjunction with ongoing system design reviews, the inspector noted that reset of a control room ventilation isolation (CRVI) signal
" seals in" when either 1-CBA-V-9 or 2-CBA-V-9 are closed. These valves isolate the east and west air intakes respectively.
Follow-ing discussion with licensee engineering personnel, the inspector determined that with 1-CBA-V-9 or 2-CBA-V-9 shut, the purge valves (1-CBA-V-2 and 1-CBA-V-4) shut and a makeup fan running, radiation could enter the idle makeup train through infiltration without auto-matic isolation. While this is an improbable scenario, the licensee agreed to leave the purge valve open to any isolated intake to en-sure a backflow of monitored air at all times. This action was taken in lieu of performing an analysis to demonstrate that this infiltration was within the limits of the FSAR. The normal valve
.
,
y
_
.__
-_,
-.,,
.,,, -
_ ~...
m
--
- -,-
..
.
lineup of OS1023.51 (Rev. 03) was modified by an intent change (No.
03) issued on February 9, 1987. This change ensured that purge valve, 1-CBA-V-2 remains open when isolation valve 2-CBA-V-9 is shut.
The necessity of maintaining a purge on the isolated makeup was assumed to address situations where no radiation existed, as well as those in which an isolation had occurred and the contaminated intake was isolated, since these situations would envelope all de-sign basis conditions.
However, the change to OS1023.51 (Revision 3, described above)
failed to consider what action to take if the on-line makeup had to be switched for some reason other than an isolation signal.
It also failed to address the specific steps to be taken when clearing sigrals with one of the isolation valves shut. The failure of the SS to initiate purge of the isolated intake in subparagraph (3)
above is in part due to a lack of procedural guidance in this area.
This failure to translate design criteria and bases, as supportable by existing calculations or analysis, into procedures which direct the operators to place systems (e.g., the subject CBA subsystem)
in a configuration which meets its functional requirements (as de-scribed in FSAR section 9.4.1) represents a violation of 10 CFR 50, Appendix B, Criterion III (443/87-02-02).
(5) Air Conditioning Compressor Failure (CBA-AC-5A)
While reviewirg the operators log in the control room on January 30, 1986, the inspector noted that TS Action Statement 3.7.6, " Con-trol Room Area Ventilation System", had been entered due to a fail-ure of CBA-AC-5A, the compressor associated with air conditioning unit CBA-AC-4A. This unit supplies the cooling coil of control room air conditioning unit CBA-AC-3A. There is no specific TS on either CBA-AC-5A or CBA-AC-4A, however, there is a system flow specifica-tion through CBA-AC-3A of 25,700 cfm +/-10%.
Additionally, the control room environment temperature must be maintained within pre-scirbed habitability and equipment qualification guidelines per TS 3.7.10.
Initial licensee action taken on this failure was to enter TS 3.7.6 while determining its applicability.
This is the conservative approach and the inspector determined that this action was accept-able.
Subsequent evaluation determined that CBA-AC-5A is not re-quired to be operable to raeet TS 3.7.6, based upon the ability to maintain control room temperature. At this time of year, cooling of the control room is unnecessary and in this respect, there was no degradation of the control room environment. The operators there-fore exited TS 3.7.6 on this basis and it was correctly indicated in the log that although Action Statement 3.7.6 was no longer in effect, CBA-AC-5A was still out of servic.
-
At a later time, both CBA-AC-5A and fan CBA-FN-14A were de-energized to repair CBA-AC-5A. This placed the unit back under TS 3.7.6 with CBA-FN-14A inoperable. The situation was identified by the inspec-tor and discussed with the Unit Shift Supervisor (USS) who entered the LC0 as of the time CBA-FN-14A was de-energized. The LCO was not violated.
(6) Independent Design Review While conducting a system walkdown in conjunction with inspection of the CBA system, the inspector noted a discrepancy between the as-built configuration and the P&ID regarding the instrument lines to CBA-PDIS-5332.
This discrepancy was discussed with the system engineer who provided a copy of ECA 98/118373A which added an addi-tional sensing line to PDIS-5332. As of the end of the report period, the P&ID had not yet been updated to reflect the design change effected by the ECA.
Verification that the affected drawing has been updated will be accomplished as a routine follow-up item during a future NRC inspection.
(7) Conclusions Three separate topics are evaluated and discussed as enumerated below:
--
Adequacy of the corrective action to the deviation.
--
Failure to enter the proper action statements for equipment failures.
--
Improper valve line-up following intake realignment.
As of the conclusion of this inspection, it is not clear that the corrective action steps described in NYN-87024 to the subject de-viation have been fully implemented with respect to ensuring that the design objectives of the system are completely understood by the operating staff.
Based on discussions with operations and regulatory services personnel, the inspector determined that during and after the radiation monitor failure on March 3, 1987, there existed confusion over purge requirements. This was evidenced by the failure to provide purge to the isolated intake.
Although a clear definition for single train operation was published in NYN-87013 in February 1987, the USS failed to enter the appro-priate action statement because he did not recognize that a single train was inoperable with a radiation monitor failur..
.
Although abnormal operating procedure OS1223.01 was revised, the engineering response to the normal closure of 2-CBA-V-9 was not translated into the normal operating procedure in such a way that the operators could switch intakes and maintain the purge.
As a result, deviation 86-54-01 shall remain open until such time as the corrective actions of NYN-87024 are complete. Additionally, the failure to open CBA-V-4 on the isolated intake on March 3, 1987 constitutes a violation, discussed in subparagraph (4) above, which represents an additional open item for which licensee response and further NRC inspection of corrective action is required.
5.
Licensee Event Reports (LER)
a.
(Closed) LER 86-002-00: Inadvertent Safety Injection. NRC Region I IR 443/86-54 described the sequence of events surrounding the inadvertent SI event of December 24, 1986. The licensee submitted LER 86-002-00 in accordance with 10 CFR 50.73 on January 23, 1987 (refer to NYN-87005).
The inspector reviewed the LER, noting it to be a routine report which documented the facts which were previously evaluated and reported in IR 443/86-54. The inspector had no further questions on this event and considered this LER closed.
b.
(Closed) LER 86-003-00 and LER 87-001-00: Control Room Ventilation Iso-lations.
These events were described in paragraph 7.d of NRC Region IR 443/86-54. The first event was reported to the NRC in letter NYN 87-009 cn January 28, 1987, while the second isolation was reported in NYN 87-015 on February 6, 1987.
The inspector reviewed both reports and determined that they were submitted within the required time frame and accurately reflected events as had been previously reported in the four-hour ENS notifications and as described in IR 443/86-54.
These LERs are closed.
c.
(Closed) LER 87-002-00: ESF Actuation Caused by a Loss of Offsite Power to an Essential Switchgear Bus. This event was reported to the NRC by letter (NYN-87017) dated February 13, 1987. At the time of the event, on January 18, 1987 the plant was in Mode 5 with the "B" diesel generator aligned and synchronized to the "B" train 4160 volt vital bus (Bus E-6)
for the purpose of testing.
The unit auxiliary transformer (UAT) breaker supplying offsite power to bus E-6 was inadvertently opened because of operator error. Since the bus was in synchronization with the diesel generator (DG), an automatic transfer of bus E-6 load from the VAT to the reserve auxiliary transfor-mer (RAT) was blocked. As designed, the bus was stripped from the DG, the. DG output breaker reclosed, and the emergency power sequencer (EPS)
stepped through its normal sequence to carry the designated loads. After
..
..
,
EPS timing completion, synchronization of the DG with offsite power was accomplished, allowing the operators to reclose the erroneously opened UAT breaker.
The incoector, at the time of this event, reviewed SIR 87-005 and dis-cussed it with the SS on shift.
FSAR sections discussing automatic transfer of power sources for 4160 volt buses and the loading sequence for the diesel generator, given a loss of offsite power situation, were reviewed. The inspector determined that plant equipment operated as designed during this event.
As follow-up corrective action, the operations manager issued a memo
-
discussing the need for additional attention to detail in switch manipu-lations on the main control board (MCB).
This memo, dated January 20, 1987, also instituted a new control room policy whereby all future MCB manipulations would be verbalized (i.e., announced by the operator) prior to the action. The inspector noted that implementation of this policy during routine observations of control room activities during the re-maining portion of this inspection period.
This LER is closed.
d.
(Closed) LER 87-003-00: Source Range Analog Channel Operational Test.
This event was reported to the NRC by letter (NYN-87027) dated March 3, 1987 pursuant to 10 CFR 50.73. The analog operational test of each of the two source range (SR) channels is required by TS 4.3.1.1 to be per-formed every 92 days, on a staggered test basis. Although the 92-day interval was met, the staggered test basis was not. The inspector re-viewed the corrective action as described in the report and found it to be acceptable.
This LER is closed.
6.
Follow-up of Licensee Action on IE Information Notices (IEN)
a.
IEN 86-106: Feedwater Line Break. This Information Notice alerted the licensees of all nuclear power reactor facilities holding an operating license or construction permit to a potentially generic problem with feedwater pipe thinning and other problems related to the feedwater line break at Surry, Unit 2 on December 9, 1986. As a result of continuing NRC concerns related to steam, feed and condensate system piping integ-rity, the inspector discussed with the licensee their planned actions with respect to these technical concerns.
The program support manager indicated that an evaluation is underway to establish a long-term moni-toring program on this issue. Since this phenomenon would only occur over long periods of time and sustained operation for the Seabrook feed and condensate systems, the conditions identified in IEN 86-106 are not expected to occur at Seabrook in the near future, if at all. The in-spector determined that the licensee actions, to date, in this area have been appropriate. This completes the Region I planned inspection acti-
..
-
vities on this issue at Seabrook at this time. The pipe wall thinning issue will be reinspected in the future, when the licensee's planned program to address the technical concerns raised by IEN 86-106 is established.
b.
IEN 87-08: Degraded Motor Leads in Limitorque DC Motor Operators. The inspector reviewed this IEN and determined, in discussion with the lic-ensee, that no such motors are used at Seabrook Station.
c.
IEN 87-12: Potential Problems with Metal Clad Circuit Breakers, General Electric Type AKF-2-25. Because several failures of the subject breakers have occurred at nuclear power plants in recent years, the inspector in-quired into their use and application at Seabrook.
It was determined that no similar breakers are being utilized at Seabrook Station.
7.
Startup Testing a.
Containment Closeout Prior to final closeout of the containment, the inspector reviewed OS1015.19, the containment closeout procedure, as well as the licensee's closeout punch list.
The procedure included sign-offs for each depart-ment supervisor, independent verification checklists and special require-ments. The inspector also reviewed the housekeeping report, relative to the licensee's inspection conducted as corrective action to violation 86-46-01, and then conducted a preliminary inspection of the containment.
The inspector provided a list of noted discrepancies to the utilities /
radwaste supervisor in charge of last minute closecut items. None of the discrepancies had any safety significance.
b.
ST-7, Rod Drop Time Measurement The inspector reviewed the test procedure and witnessed the shift brief-ing and portions of the first rod drops.
Initial conditions, precautions and limitations were met.
Portions of a subsequent retest for one par-ticular control rod were witnessed both from the control room and from the switchgear room, where test personnel and apparatus (e.g., visicorder)
were located. No violations or safety concerns were identified.
c.
ST-10, RTD Bypass Flow Verification The inspector reviewed this procedure in preparation for test accomplish-ment. No concerns were identified.
d.
ST-12, RCS Flow Coastdown Test The inspector reviewed Revision 1 of test procedure 1-ST-12, witnessed the test conduct, verified initial conditions, prerequisites and special precautions, and discussed the jumper connections and data retrieval
__.
. _
.
points with the cognizant I&C engineer. After the first unsuccessful attempt to trip the reactor coolant pumps (RCP), the licensee evaluated the situation and determined that the ground line jumper should be con-nected to plant ground via instrument ground. This change was made and the test was successfully completed.
e.
ST-53, Turbine Driven Emergency Feedwater Start Verification Test The testing of the turbine driven emergency feedwater pump has received continuing inspection coverage from both resident and region-based in-spectors since the commencement of testing.
Inspection reports 443/87-01, 443/87-05 and 443/87-09 provide additional information of this testing.
Many test starts were witnessed by two or three inspectors simultaneously at various locations including the control room, main steam /feedwater pipe chase, piping tunnel and EFW pumphouse.
The inspector reviewed test procedure 1-ST-53, Revision 1, and subse-quently witnessed the initial runs of the pump on February 19 and Febru-ary 20,1987.
Following heatup of the reactor to normal operating tem-perature (557 degrees F) the normally closed steam admission valves off both the "A" and "B" steam generators were leaking, allowing heating of the downstream piping. The maximum allowable metal temperature for a cold start is 150 degrees F which could not be achieved with the leaking valves. The licensee thereforc decided to close the normally open up-stream isolation valves. MS-V127 is located between MS-V393 and the
"A" steam generator, while MS-V128 is located between MS-V394 and the
"B" steam generator. Their intention was to initiate pump start by opening MS-V127 and/or MS-V128 as appropriate just prior to initiating the pump start sequence.
This would place main steam pressure at the inlet of the steam admission valve (MS-V393 and/or MS-V394) prior to their opening but prevent downstream heating due to leakage.
During this initial sequence of EFW pump runs, the inspector witnessed all three pump starts, verified compliance with the test procedure and discussed the test method and results with the system engineers. The acceptability of opening MS-V127 and MS-V128 just prior to pump start was questioned since these valves are normally open. The licensee indi-cated that it was their position that data from additional runs could be evaluated to determine the effect of this procedural deviation. As of the conclusion of this inspection, EFW testing was still in progress.
NRC evaluation of the stated licensee position with respect to the de-viation from normal valve lineup configuration will be evaluated during the next inspection period.
During a follow-on run, while attempting to start the pump from both MS-V393 and MS-V394, a pressure transient occurred which caused three safety valves to pop open for two to three seconds on the "B" steam generator.
Preliminary licensee evaluation determined that the transient was caused by opening MS-V128 with condensate built up in the idle "B" steam header.
The "B" steam generator main steam isolation valve and its bypass valve
_ - _ - _ _ _ _ _ _ _ _ _ _ _ _ -
- _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _
.
were both closed. The inspector examined the instrument recorder traces which confirm the sequence of events as described.
Licensee evaluation of this event is not yet completed and the inspectors will also follow up the cause and corrective actions of this anomaly in a future inspec-tion.
Revision 2 to procedure 1-ST-53 was issued on February 21,1987.
Follow-ing licensee engineering analysis, the speed acceptance criterion at 60-seconds after start initiation was revised downward to 3350 RPM. This was based on correlating time / flow data with the FSAR commitments. Sub-sequent testing revealed centinuing problems and revision 3 was issued on March 5, 1987. This revision was necessitated by a design change which modified the opening sequence of the steam admission valves. Ad-ditional pump runs were subsequently observed; however, the licensee determined that a modification to the governor orifice would then be made with assistance from the vendor representative.
1-ST-53 will be started again from the beginning following these modifications.
The inspectors I
will continue to closely monitor this test as it carries over into the next inspection report period.
While witnessing the EFW testing, the inspector noted that various suc-tion and discharge line valves of the EFW turbine-driven pump were ad-ministratively locked into position. Manual valve MS-V-95, upstream of MS-V-395, the steam supply valve at the turbine, however, was not in-cluded as a locked valve. The inspector discussed the matter with the licensee since inadvertent closing of this valve would render the EFW turbine driven pump inoperable by blocking the steam flow to the turbine.
Following the discussion, the licensee agreed to lock the valve in the open position and included it in the procedure for periodic locked valve verification.
f.
1-ST-55, Steam Dump System Test Following a review of the test procedure, the inspector witnessed the performance of section 6.1 of 1-ST-55, revision 0, which dealt with stroke testing of the condenser steam dump valves to verify acceptable opening (3 seconds) and closing (5 seconds) times. Three of the twelve dump valves were not acceptable in that: valve PV-3019 was stroked open in 3.1 seconds; valve PV-3011 did not provide open indication when stroked; and valve PV-3018 became stuck in mid position when stroked.
The licensee generated work requests to address the valve problems.
Subsequent to performing maintenance on PV-3019 and PV-3011, the valves stroked within the acceptable range of 1-ST-55.
Valve PV-3018 will be tested in accordance with 1-ST-55 to demonstrate acceptable operation upon completion of the maintenance activities.
The inspectors had no further questions at this time.
-
_ _ - _ _ _ _ _ _ _ _ _ _ _
..
-
8.
Maintenance Activities The motor driven emergency feedwater pump 1-FW-P-378 seals were found to be rubbing during performance of a technical specification surveillance.
Troubleshooting by the licensee maintenance department indicated that the manufacturer's installation instructions were inadequate to properly install the seal. The licensee repaired the pump using a step-by-step maintenance procedure. The inspector interviewed the maintenance supervisor at the job site and discussed the seal design, repair and maintenance procedure. Addi-tionally, the inspector reviewed five QA inspection reports written during the repair period covering this job. He noted that the forms accurately re-flected job conditions and that the QA findings had been addressed.
QA cri-teria verified during this job included torquing of gland stud nuts and threaded fasteners, close out cleanliness, seal replacement, and material procurement. The inspector had no further questions on the conduct and QA coverage of this maintenance activity.
9.
Design Change and Modification Program New Hampshire Yankee is responsible for conducting engineering activities to maintain tne design integrity of the plant systems at Seabrook. To fulfill their responsibilities, the NHY engineering organization maintains a design control program addressing design activities and configuration management interfaces. The NHY Design (NYDC) Control Manual defines the design control program. This program outlines the design control process from initiation through implementation, testing and final closeout. A draft form of the manual was reviewed by the NRC in IR 443/86-06 and determined to be adequate.
Subsequent to that inspection, the licensee completed their management review process and the manual was issued for implementation.
The inspector reviewed the procedures contained in Revision 5 of the NYDC manual which became effective on December 23, 1986. The licensee's design control program adequately defines the methods and responsibilities for in-itiation, review, design, evaluation, approval, implementation, post modifi-cation testing, document updating and distribution, performing safety evalu-ations and reports in accordance with 10 CFR 50.59 and ensuring that applic-able information is incorporated into the training programs.
The licensee, to this date, has not completely processed any design changes or modifications. Approximately forty changes have been implemented and categorized as " full service notification complete" with the final documenta-tion updates and signatures still outstanding.
The effectiveness of the im-plementation of the design control program will be evaluated during future NRC inspections.
10. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations.
An unresolved item disclosed during this inspection is discussed in paragraph n w
11. Management Meetings At periodic intervals during the course of this inspection, meetings were held with senior plant management to discuss the scope and findings of this in-spection. An exit meeting was conducted on March 9, 1987 to discuss the in-spection findings during the period.
During this inspection, the NRC inspec-tors received no comments from the licensee that any of their inspection items or issues contained proprietary information.
No written material was provided to the licensee during this inspection.
,
i e
i
)
__., _ _,
_ - -.
,,
.--
. - - -
. - - - - -
- - * - -
-
~
' ~ -
'