IR 05000443/1987016
| ML20236C031 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/21/1987 |
| From: | Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236B959 | List: |
| References | |
| 50-443-87-16, GL-87-09, GL-87-9, IEB-87-001, IEB-87-1, IEIN-86-106, IEIN-87-021, IEIN-87-21, NUDOCS 8710260462 | |
| Download: ML20236C031 (23) | |
Text
..
.
l I
U.'S. NUCLEAR REGULATORY COMMISSION
Report No.
50-443/87-16 Docket No.
50-443 i
License No.
NPF-56 i
Permit No.
CPPR-135 Priority --
Category B/C o
Licensee:
Public Service Company of New Hampshire
'
l 1000 Elm Street l
Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit 1 Inpsection at: Seabrook, New Hampshire Inspection conducted: July 7 - September 8, 1987 Inspectors :
A. C. Cerne, Senior Resident Inspector D. G. Ruscitto, Resident Inspector C. A. Carpenter, Reactor Engineer A. G. Krasopoulos, Reactor Engineer i
/o/2/ //
Approved by:
/
.
.
o T. C. Elsas ief, Reactor Projects Section 3C Dat6 j
Inspection Summary:
Inspection on July 7 - September 8,1987 (Report No.
,
50-443/87-16)
!l Areas Inspected:
Routine inspection by two resident inspectors and two j
region-based inspectors of work activities, procedt.res, and records relative j
,
l to design control, modifications, testing, maintenai.ce, surveillance, train-i ing, and plant operations during cold shutdown.
The inspectors.also reviewed licensee action on previously identified items, including licensee event reports (LERs), a construction deficiency report (CDR) and an NRC Bulletin i
and performed plant inspection-tours.
The inspection involved 270 inspection
hours by four NRC inspectors.
This report also documents the results of'a
.j plant inspection conducted by the Regional Administrator of NRC Region I
,
I on September 2, 1987.
Results: One violation was identified concerning the control building air handling (CBA) system.
This licensee-identified noncompliance was the direct result of inadequate corrective action taken to address a previous violation of NRC regulations. The applicable enforcement criteria of 10CFR2, Appendix C, 8710260462 871021-PDR ADOCK 05000443 G
.i
.]
l.
i
..
.
I I
l
- i l
were discussed with the licensee with respect to the policy on when self-identified problems qualify for the issuance of a Notice of Violation.
Two additional unresolved items were opened.
The first dealt with the overall j
conduct of a weld repair to the
"A" train residual heat removal (RHR) system j
and the second involved questions concerning diesel generator testing and l
loading.
j l
l t
!
l i
l
!
l l
,
I
.
1 s
!
l l
i e
l i
,
i i
!
l j
i
_ _ _. _ _ _ _ _
m._
__
___..__:.__.._m.___--__._m m-,
q l
w
.,
DETAILS
!
1.
Persons Contacted W. B. Derrickson, Senior Vice President, New Hampshire Yankee (NHY)
T. C. Feigenbaum, Vice President, Engineering and Quality Programs W. J. Hall, Regulatory Services Manager D. E. Moody, Station Manager P. M. Richardson, Training Manager G. S. Thomas, Vice President, Nuclear Production J. M. Vargas, Manager ~of Engineering
,
l J. J. Warnock, Nuclear Quality Manager i
2.
Plant Status During this reporting period, the plant remained in operational Mode 5, cold shutdown, with primary temperature about 115 degrees F and depressurized.
I a.
On July 15, 1987 smoke was reported in the lube oil conditioner room.
l The lube oil conditioner is a centrifugal purifier manufactured by l
I DeLaval. The fire brigade responded rapidly and professionally.
The i
inspector observed initial entry-into the room, fire brigade inspec-'
tion and smoke clearing efforts.
Operators rapidly de-energized both the conditioner and its associated heater.
Licensee ~ investigation j
l has revealed a failure of the spindle bearings. The inspector
!
I reviewed Station Information Report (SIR)87-063 noting ' adequate
corrective action plans for the conditioner and its associated maintenance procedures.
",
A courtesy notification was made to the-NRC Operations Center.
The inspector noted that the control room operators, firefighters, security officers and cognizant station staff personnel responded promptly and appropriately to this event.
No violations were identified.
l b.
On July 9,1987 while in Mode 5 (cold shutdown), a feedwater l
'
isolation signal (FWIS) was received.
The P-14 (steam generator l
level Hi-Hi) interlock was activated on "B" steam generator (SG)
I while switching SG recirculation from "A" SG to "B" SG.
The level l
rise was due to water in the isolated and pressurized feedwater header flowing to "B" SG as the piping depressurized during l
i switchover.
The inspector agreed with the licensee analysis that this was not a reportable event under 10CFR50.72.
The licensee evaluation of root cause was not complete at the time this report was issued.
The inspector will follow-up this data in the routine review of SIR 87-62.
l
x
'
...
- .
4'
c.
On August 13, 1987 an inadvertent safety injection signal (SIS) was generated on "B" train.
The cause of the SIS was " prick punching!'
on the main control board (MCB) adjacent to the "B" train low pressurizer (PZR) pressure ' safety injection (SI) block / reset switch.
This hammering ~ caused the contacts in the switch to bounce, resetting the. signal. With a low pressure condition in existence, as is normal in Mode 5, multiple engineered safety.
features (ESF) were actuated upon the instantaneous' reset. All plant equipment functioned as designed.
The "B" centrifugal charging pump (CCP) and emergency diesel generator (EDG) started, Train "B" high and low head SI valves a
realigned, a Train
"B" Phase A containment isolation occurred, a Train "B" containment-ventilation isolation signal was generated, the control building Train
"B" emergency air cleanup fan started, and a main feedwater isolation occurred.
Approximately 10,000 i
gallons of borated water was injected into the reactor coolant I
system (RCS),
'
The inspector observed recovery actions from the control room and reviewed the data logger records and chart recorders in verifying j
plant response. Operator actions were determined to be correct and-i timely in restoring the plant to a normal shutdown' lineup. Addi-tional discussion of this event is contained in paragraph 5 of this
,
report under LERs87-012 and 87-015.
I d.
On September 5,1987 a failure occurred on a 345KV SF6 bus duct causing activation of protective relays.
Licensee investigation of this incident is ongoing and will be followed up in a future NRC inspection report.
3.
Plant Inspection Tours The inspectors observed work' activities in progress, completed work and plant status in ieveral areas during general inspections of the plant.
They examined wo k for any obvious defects or noncompliance with regu-latory 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, noncon-forming 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, the inspector reviewed control room logs and records including night orders, shift journals, shift turnover sheets, completed Repetitive Task Sheets (RTS), 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 Specification (TS) Limiting Conditions for Operation (LCOs) and Action Statements.
_ _ _ _. _ _. _ _ _ _. _ _. _.
. _ _. _
-.
_.
_
.______________________m___.
,
a.
During various plant inspections and visits to the control room for observation of' plant operations, the inspector examined the_following components, conditions and activities, noting appropriate licensee corrective action, where required:
--
A surveillance test performance on "B" safety injection pump (SI-P-68) for vibration analysis was witnessed.
The inspector verified recirculation valve positions, local instrument readings and pump operating conditions.
He reviewed the operating procedure with the senior control room operator (SCRO) and verified that the main control board (MCB) valve lineup was proper and that procedural precautions and prerequisites were met, j
--
The inspector observed reassembly of the "A" Train primary I
component cooling water (PCCW) heat exchanger (CC-E-17A) lower -
I head supports.
He reviewed the maintenance procedure and support drawings in use and noted the presence ~of a quality-control (QC) inspector at this work station. An interview with.
.
the QC inspector confirmed his knowledge of the requirements and the applicable acceptance criteria.
--
The inspector examined the repair of a leaking safety injection (SI) accumulator _"0" (SI-TK-90) sample line valve'(SI-V-8003).
He reviewed the applicable work package. including the procedures and procurement documents.
The inspector periodically examined the reactor coolant system
--
(RCS) vent located off the pressurizer header _ piping to the power operated relief valves (PORV). The position of RC-V-468, the removal of the blind flange and the screened protection of the open vent were all checked to confirm compliance with Technical Specification 3.4.9.3 for RCS overpressure protection.
During a backshift plant inspection, the
"B" train component
--
cooling heat exchanger discharge and bypass, temperature con-trol valves were observed to have been aligned for operation with the manual operating clutches engaged locally.
The inspector interviewed the cognizant system engineer and verified that the control room was aware of this condition e
and that a work request had been generated to control the valve status in support of instrument calibration activities in progress.
Subsequent inspection revealed that the subject valves had been returned to their normal lineup for remote operation.
The inspector observed tube sleeving operations on the "A"
--
train primary component cooling heat exchanger (CC-E-17A). On the lower tube sheet, a hydraulic expansion process was being utilized, while on the top tube sheet, a mechanical expansion i
..
.
1 rolling technique was_ concurrently in progress; ' Work controls per design coordination report (DCR)87-223 and work request (WR) 87WO06172 were in evidence, as was station authority (SORC 87-172) for the contractor conduct of the work per a TECCO
procedure (site labeled ES-87-1-10).
The inspector noted_that the three inch sleeve material (70-30_Cu-Ni) had been selected
]
for consistency with the tube sheet design and for optimization i
of heat transfer characteristics.
An engineering evaluation (NHY 87-001) had been performed to~ consider the different-sleeving options. Input from the heat exchanger vendor (Joseph j
Oat) was also considered to maximize heat transfer in light of the desired flexibility for additional, future. tube plugging
,
requirements.
q
.1
--
Over the course of this inspection, random checks of several
')
valves were conducted to confirm correct valve position in accordance with the governing operational controls and to verify locked valve status of certain valves in accordance with-the " inadvertent boron dilution" requirements of the Facility Operating License (NPF-56).
Where procedural evolutions neces-sitated changes to the standard valve positions, subsequent inspections spot-checked the final valve conditions to verify proper alignment _in accordance with the applicable procedural
'
controls.
b.
Portions of the "B" train RCS, CS, RHR, SW and PCCW systems were reviewed to verify the systems were properly aligned and fully operational for the current plant conditions.
The review included verification that accessible major flow path valves were correctly-positioned, power supplies were energized, lurication 'and component cooling were proper, and components were operable based c., a visual inspection of equipment for leakage and general conditions.
No noncompliance were identified.
However,-with respect to equipment operability determinations, inspector questions were raised as discussed in paragraph 9 of this report.
c.
On September 2,1987, the NRC Region I Regional Administrator con-ducted an extensive tour of the Unit 1 facility with the resident inspectors and licensee management.
The status of the plant and current operational activities, as evidenced by the observed com-ponent conditions, were discussed, as was equipment maintenance,
'
conduct and planning.
Battery maintenance, in progress, was observed. The maintenance database.for the repair of several valve.
packing leaks was checked for completeness with respect to acted examples of boron crystallization.
Overall plant housekeeping was also observed and found to be acceptable.
The station manager issued an internal memorandum assigning responsibility in each disciplinary area for which the regional administrator's comments dictated appropriate follow-up.
No specific deficiencies or major problem areas were identified.
-
.-
..
'
Where required, further inspection was conducted'(eg:
imit switch status on valve, RH-V-36) to confirm component accept-ability with respect to the questioned field conditions.
With respect to all of the above plant inspection-tour and independent inspection items, no violations were identified.
i 4.
Licensee Action on Previously Identified Items a.
(Closed) Construction Deficiency Report (85-00-20): Failure of Fire Dampers to Close Under Flow Conditions.
This construction deficiency report (CDR) concerns the inability of certain fire:
.l dampers to close under air flow conditions.
NHY addressed this problem by making modifications which included the replacement of j
the damper closure springs with a heavier grade spring and appli-cation of additioral lubricant in the damper guide' tracks.
The licensee performed a calculation (M3-MISC-60) entitled " Closure of Dampers under Air Flow Conditions" to determine which dampers required modification or testing. The inspector reviewed this j
calculation as well as General Test Procedure GT-M-05 entitled
'
"Non-Manual Valves and Dampers".
He also reviewed miscellaneous modification and test packages to. verify the following-
>
-- All dampers have been included in the licensee evaluation.
-- Those dampers which required modification or testing had been modified or tested as appropriate.
This review did not identify any unacceptable conditions or concerns
,
with the stated licensee corrective action. This CDR is closed, b.
(Closed) Unresolved Item (85-06-03): Fire Loop and Fire Pump Tests Not Performed. At the time of the original inspection, the adequacy of the fire water system could not be verified because certain '
required system flow tests had not been performed. The tests had been delayed because the fire loop was not fully installed and the pumps were being overhauled and decertified by the manufacturer.
These tests included the fire loop flow test and the fire pump performance flow test. Successful test completion was required prior to receipt of new fuel on site.
The inspector reviewed the following documents to verify test performance in ' preparation for fuel receipt.
-- Non Conformance Report (NCR) 82-1396A
-- Procedure, AT-13.1, Revision 2 " Fire Pump Flow Capacity Testing"
-- Hydrostatic Pressure Test Fire Pump Certification
-- Fire Pump and Controller Specification, 9763-006-238-14
_ _ _ _ _ _ _
,
..
i
,
l
-- Engineering Change Authorization (ECA) 06/108100F,'"FPe
Water Chlorination System"
-- Procedure CP3.3, Revision 1, " Miscellaneous Systems-Chemistry Control Program"
-- Test Procedure 0X0443.11, Revisian 0, "Three Year Fire System Flew Test"
!
!
Through review of the above documents, the inspector determined that the licensee's actions satisfy NRC: concerns in that, the tests were i
performed within the prescribed time frame and that acceptance l
l criteria of the applicable industry standards and design criteria were met. This item is closed, c.
(Closed) Unresolved Item (86-27-05): Emergency Boration Piping Exits the Controlled Temperature Environment Without Heat Tracing.
The inspector reviewed responses to requests-for the engineering
services (86-RES-0400 & 0401) provided by YAEC 'and UE&C, both of l
which indicate that under worst case environmental conditions (ie:
plant shutdown in the winter), a temperature in excess of 65 degrees j
F would be maintained in the subject environmental zone, PB-13. An independent NHY Engineering Evaluation also was conducted to analyze the design area heating controls with respect to a 55 degree F reference temperature, below which four.wefght percent boric acid l
begins to precipitate out of solution. While the NHY report confirmed the adequacy of the existing PAB heating /ventilati(n design for maintenance of temperature controls-above 65 degrees F, it also concluded that the addition of a video alarm system (VAS)
low temperature alarm for area PB-13 would enhance the remote t9mperature monitoring capability for that-portion of the PAB
.
containing systems with boric acid in solution.
DCR-86-0572 is being revised to incorporate the recommended alarm
'
into the VAS.
The inspector had no further questions on either the engineering evaluations of the present design or the system enhancement provided by the recommended design change.
This unresolved item is closed.
d.
(Closed) Unresolved Item (87-09-01): Performance of 10CFR50.59 Review of Station Procedure Changes.
In this item,-the inspector expressed a concern regarding procedure changes classified as non-intent changes.
It was understood that no assessment for unreviewed safety questions was required, other than that implied by a deter-mination of "non-intent", until the SORC reviewed the change.'This could occur as late as 14 days after the change was implemented.
The I
licensee responded by revising the Station Operating Procedure Change form in a way that requires the originator and reviewers to. assess whether the change impacts the Technical Specifications, FSAR or.
license.
This provides additional assurance.that changes.are properly processe..
.
The second part of this item dealt with the inspector's concern that-some changes to procedures were dependent on SORC out of meeting reviews.
His concern was that this practice could result in some procedure changes not satisfying the~ Technical Specification quorum requirements for review.
The licensee subsequently implemented a new method to accomplish and document the SORC review of procedures, procedure revisions and changes.
" Walk-In. Presentations" will be
!
used for procedure changes.
This entails copies of the procedure change being provided to all members at the SORC meeting. A pre-sentation of the change will also be made, including a description of the procedures, the applicability of 10CFR50.59 to the change'and the background for this determination.
The inspector reviewed these corrective actions and had no additional questions or concerns. This item is closed.
l e.
(Clored) Unresolved Item (87-10-01): Fail-safe Application of the MSIV Design. A NHY letter (NYN-87-81) to the NRC on June 17,1981 deleted fail-safe testing requirements for the main steam isolation valves (MSIV) from the Inservice Testing (IST) program. Addition-
,
l ally, a NHY position paper on the MSIV failure mode designation was
'
presented to NRR reviewers to clarify the FSAR inconsistencies. The l
inspector confirmed that the previous NRR staff review and conclu-
'
sions were not materially affected by the revisions and that the FSAR changes could be administrative 1y handled during'the course of the next FSAR amendment submittal.
The inspector also reviewed an Independent Safety Engineering Group
'
(ISEG) evaluation of the MSIV design and its relationship to the ESF actuation (LER 87-009-00) documented in the NRC 87-10 IR. -The MSIV design, logic, procedures and administrative contcols were found to be both adequate and consistent with the safety functions of these valves. Also, the discrepancies identified by NRC inspection were
-
determined to neither involve any unresolved safety issues nor I
affect any other valves in the plant.
l An additional NRC question was raised regarding the potential impact of failure of the "B" train MSIV test / reset switches upon the overall safety functions and operability of the MSIVs. A NHY Engineering Evaluation (87-010) was conducted to review the question of whether i
a switch / circuit failure could cause an MSIV to cycle open/ closed l
around its 10*s closed signal.
Both startup and power situations were. analyzed with the resulting conclusion that the hypothesized nonsafety switch failure would neither fault the Class 1E equipment in the MSIV logic cabinets, nor preclude the fast closure of.the MSIVs upon an isolation signal. Also, it was determined that no l
plant transients would occur because of the suggested MSIV cycling
'
scenario.
'
l l
<
-
p
..
.
Thus, the current MSIV design has been re-evaluated _from a " fail safe" standpoint by both NRR and internal licensee reviews and deemed acceptable for its safety-related functions.
The incon-sistencies with respect to the IST program have been corrected and a future FSAR amendment will correct the other documentation discrepancies.
The inspector has no further questions on the MSIV design or safety-related operating characteristics. The questions
raised by LER 87-009 have been answered and this issue is hereby'
l closed.
)
f.
(Closed) NRC Bulletin 87-01: Thinning of Pipe Walls in Nuclear power l
Plants.
This NRC Bulletin requested licensees to submit information concerning their programs for monitoring'the thickness of pipe walls in certain carbon steel piping systems.
NHY responded to NRC Region I on September 11, 1987 in letter NYN-87106. Other related inspec-
.
tion of this subject area was conducted in following up Information
,
Notice 86-106 and is documented in NRC Region I IR 87-02. This
l bulletin is closed.
I 5.
Licensee Event Reports Two of the licensee event reports (LER) discussed below deal with.
inadvertent safety injection actuation due to contact problems with l
Westinghouse OT2 type switches.
The August 1,1987 event occurred during installation of the design change necessitated by the April 16, i
1987 event. The third LER was issued after the conclusion'of this
'
i report period, but addresses an event which occurred during this inspection.
'
l a.
(Closed) LER 87-012: Inadvertent Safety Injection.
NRC Region I l
IR 443/87-13 indicated that this LER would remain open pending l
review of the design change implementation.
The inspector reviewed
!
design coordination report (DCR)87-185.
He addressed minor comments to the cognizant implementing engineer and-later verified licensee incorporation of these changes.
Specific attention was paid to the retest requirements associated with DCR 87-185.
No l
concerns were noted. He also reviewed the licensee's LER submittal to NRC (NYN-87065 dated May 15,1987) and had no questions.
This LER is closed.
b.
(Closed) LER 87-015: Inadvertent Safety Injection.
This event is partially described in paragraph 2 of.this report. The event was reported on September 11,1987 in a letter (NYN-87107) to NRC Region
'
l I.
The inspector noted the report to be particularly comprehensive
',
and noted the correlation of this event with other events of'similar nature, but not considered previous occurrences.
It is expected that during the continued work activities to complete DCR 87-185, placing the solid state protection system (SSPS) in'the test mode will preclude further spurious reset signals. Thus, " prick punching" or similar shock / impact activities in the MCB will not adversely affect operations.
This LER is closed, i
._
~h
,.
-
s c.
(Closed) LER 87m016: Control Room Makeup Ventilation System Isolation Logic Inoperable.
On August 19,1987, during the conduct of the technical specification (TS) surveillance (4.3.3.1) of the radiation monitoring instrumentation associated with the Control Building Air (CBA) intake isolations, the licensee identified that the emergency isolation functions of both trains of this system were -
inoperable..The NRC resident. inspector was notified; deportability in accordance with 10CFR50.72 was evaluated; and subsequently e
'icensee Event Report (LER 87-016) was submitted to comply with the
_
requirements of 10CFR50.73.
The inspector discussed this event with the duty shift superintendent (SS) shortly after the occurrence and later with engineering support personnel.
He reviewed both the Station Information Report,(SIR) and the subject LER and determined that this event could have been prevented from occurring if adequate corrective action for previous NRC enforcement action had been effected by the licensee.
The identification by the licensee of the CBA radiation monitors' inability to perform an intended safety function, in conjunction with the ineffectiveness of licensee corrective action to previous CBA inspection findings., constitutes a violation of 10CFR50, Appendix B.
Licensee management personnel were notified of this enforcement action at an NRC exit meeting on September 11,1987, at which time the criteria of 10CFR2, Appendix C.
V.A were discussed relative to the decision to issue a Notice of Violation.
'
In NRC inspection report (IR) 443/86-54, a notice of deviation was issued relative to the capability of the CBA system to remain functional given a postulated single failure.
In IR 443/87-02 a notice of violation was issued, also relative to the design capa-bility of the CBA system, as it was translated into procedural controls.
In the responses to these two enforcement actions', the licensee highlighted as corrective action their evaluation of the.
generic controls being utilized for translating design bases into procedures and specifically their revision to the operating pro-cedures with regard to the CBA isolation capabilities.
These corrective actions, while addresssing the specific identified concerns, placed an additional burden on the operating staff for procedural adherence.
No positive means was available to the oper-ators to alert them to the fact that the CBA radiation monitors were
" unarmed" to provide the ESF function of control room ventile' tion isolation. Also, this current event of August 19, 1987 pointed to the fact that these monitors may have been " unarmed" since the previous TS 4.3.3.1 surveillance on July 20,1987.
LER 87-016 specifically attributes the root cause of this event to system hardware design and procedural inadequacies.
In analyzing the problem, the licensee determined that the isolation signal control logic had not been properly aligned to arm the east air intake radiation monitors to provide CBA isolation capability when required.
NRC review revealed that the west air intake radiation
_ _ _ - _ _ _ _ _ _ ~ _ _ _ _ _ _ _ _ _ _
_ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - - _
. _ _.
'
.
.
/
monitors were also affected by a similar control logic problem which had specifically been the subject of the previous CBA enforcement actions.
Therefore, in accordance with the TS requirements and the licensee's own definition of CBA single train operation, both trains of CBA
!
l were found inoperable on August 19,1987.
This fact related directly to the ineffectiveness of licensee. corrective action to previous NRC identified problems with the logic for CBA isolation.
The licensee has since implemented a temporary modification to provide positive indicatior to the operators of the armed status of the CBA isolation
capability. Other corrective actions, including procedura1' revisions l
and directions, are in progress. While it is noted that a future design change is planned to significantly modify the CSA system for
'
future operations at power (ie: Mode 1), the identified failure of the licensee corrective action to address previous problems with the.
CBA system in the current modes of operation represents a violation
~
i of 10CFR50, Appendix B, Criterion XVI (87-16-01).
LER 87-016 is closed because licensee corrective action to this specific event will be followed up and tracked as a result of and
,
with respect to the violation.
l i
6.
Main Steam / Steam Generating System l
l a.
Steam Generator Wet Layup Sub-system (1)
Bac_k round. While touring the main steam and feedwater J
(MS/FW) pipe chases, the inspector noted incomplete feedwater piping which was supported using temporary hangers.
Further inspection indicated that this was unfinished construction of an engineering change to modify the SG wet layup recirculation system.
The inspector addressed three concerns to the cogni-zant licensee managers:
-- Seismic adequacy of the temporary support configuration.
-- Missing and damaged safety tags on work area boundary valves.
-- Programmatic controls regarding the ECA/DCR interface.
(2) Seismic Supports. The inspector noted that the'SG wet layup system is designated NNS-1 which indicates that seismic design criteria must be applied to a non-nuclear safety system.
The adequacy of the temporary hangers was questioned and the licensee performed an engineering calculation (C-S-1-E-0003)
demonstrating the adequacy of the piping to meet ASME code standards up to and including safe shutdown earthquake (SSE)
l conditions.
Shortly thereafter, the licensee decided to complete the permanent supports on the line per ECA.25/118179A
~
_ - - - _
_
,
'
.
.
which was implemented with work request (WR) 87 WOO 2268 in i
March, 1987.
The inspector verified that a safety evaluation,
'
in accordance with 10CFR50.59 had been performed as part of the original ECA in August, 1986. He also reviewed the documentation associated with the WR including approved weld histories, weld
,
process sheets and other QA records.
(3) Safety Tagging. The inspector identified certain discrepancies in the safety tags used to isolate the modified piping from the~
feedwater system which was in use at that time.
The shift superintendent (SS) immediately had the four boundary valves in question chain locked closed and a new tagging orde'r was issued.
(4) Programmatic Controls. A major question was raised concerning the potential for additional unfinished construction to exist and not have been identified in the transition from the use of the UE&C Engineering Change Authorizations (ECA) to NHY Design Coordination Reports (DCR).
The NHY Design Control Manual (NYDC) chapter on construction changes specifies that when an existing UE&C ECA is revised, a safety evaluation is performed.
Those ECAs to which this requirement applies were identified on the " Approved ECA List".
'
All other design changes were implemented under full control of
.
the NHY design control program (ie: the DCR process) involving i
the requisite safety evaluation.
The inspector verified i
through discussion with the licensee that only one other ECA job was still in progress in addition to the noted wet layup system rework.
This modification involved the auxiliary steam condensate (ASC) system. A similar seismic adequacy question i
on the ASC system had been previously identified by the
'
licensee, as evidenced by a request for engineering services (RES) 87-0425 in March, 1987.
As part of the review of the system design change, the inspector reviewed the hydrostatic test package for the installation of the system isolation valves.
Minor discrepancies were identi-fied, discussed with the cognizant licensee personnel,'and addressed by the licensee.
(5) Findings. With respect to all the above, the inspector had no further concerns.
No violations were identified.
b.
Main Steam Safety Valves During performance of ST-53, the turbine driven emergency feedwater (EFW) pump startup test, two main steam safety valves (MSSV) lif ted on the "B" main steam (MS) line (reference.to NRC IR 443/87-02).
The licensee conducted a detailed investigation into the cause of this event.
Station Information Report (SIR)87-023 was completed
!
-
,
..
'I
<
J 14-
,
on February 27,1987.
It included an analysis of pressure instru-:
I ment response, MS line hydraulic characteristics-and safety _ valve j
performance.
The evaluation' concluded that the:"8" MS header,-
i having been isolated'for several days (ie: " bottled up"), was filled with water from. condensed steam.
A combination of. closed or plugged drain valves. allowed this condition to occur. When EFW actuation i
valve, MS-V-128, was opened to start,the. Terry turbine, the pressure
drop caused flashing in the MS header and'the observed' pressure
.
'
i transient. Thel system was being monitored at the time by the'GETARS I
. computer. and therefore data on steam generator (SG) level, steam pressure.and steam flow were recorded. Analysis of this. data indi-
!
cates that two,' rather than three,-MSSVs lifted during.the transient.
l The two suspect valves were sent..to Wyle' Laboratories for testing'and i
found to be within tolerance.
SIR 87-023 made several' recommendations -
l which included some instrumentation modifications, procedure changes;
?
and additional monitoring ~during the next heatup.
NRC review of licensee' actions to date has~ revealed a thorough eval-uation performed by the r'esponsible licensee engineers... Procedural modifications have been made to. manually bypass the' drain line orifices when the main steam lines have been." bottled up" (i.e.,
MSIVs closed for more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).
Instrumentation to monitor-l the status of alarm point D-5788 and main steam line movement is I
installed and will be monitored during the next-heatup along with-spring can position.
Personnel access to the MS/FW pip'e chase will-be restricted during operation. 'Several-minor design changes were not approved after licensee re-evaluation indicated that they were-not necessary.
The inspectors concur that these_ changes were not-l required but would have enhanced the system.
The status of the main steam lines and drains will be the subject of-NRC inspection during
-
the next heatup. The follov-up actions specified:in the SIR will be inspected in a future report.
i l
7.
Miscellaneous Inspection l
a.
The inspector inquired whether certain Anchor Darling ' double disc I
gate valves with Limitorque operators were_i_n use at Seabrook.
These valves had experienced problems at another nuclear plant
!
Licensee investigation revealed no such valves in use at Seabrook.
'
i b.
A potential construction deficiency (10CFR50.55(e))_was identified at another nuclear plant.
This deficiency involved'Model GP-Agastat relays not making proper contact.
NHY verified-that-although Model GP relays are in use at Seabrook in non-safety _
related applications, they are adequately braced to preclude'the reported problem.
-l l
l
.
l
'
- ,
_
.
.
c.
During replacement of SI pump and motor oil, the maintenance department' suspected that improper oil may have been previously used.
Samples were sent to the Mobil Oil Corporation for analysis.
Although the oil appeared slightly darker in color than the new oil, sample results indicated conformance to the specifications.
The licensee intends to change the oil more frequently to reduce dis-coloration.
This decision reflects an overall maintenance philosophy which emphasizes the independent judgements of maintenance personnel
,
followed up by detailed technical review. In this case, licensee l
action was noted to be conservatively directed and the inspector has no further questions on this subject.
8.
Training I
During this inspection period, the inspector reviewed the licensee's I
General Employee Training (GET) program.
The inspector observed a GET requalification class; reviewed the licensee's handout material and l
lesson plan, General and Specialty Training Manual, (Revision 4), and Training Group Management Manual (Revision 0); and interviewed the GET instructor and his supervisors.
.
The inspector verified that the handout material and lesson plan adequately covered the subjects noted in 10CFR19.12.
It was noted that i
the GET program is INPO certified.
During the GET requalification class, changes that have occurred at the site over the past year, as well as problem areas, were highlighted (eg: proper keycard punching).
Other important iS ms, such as hazardous material disposal, were stressed during the instructional lecture.
The instructor was well prepared, knowledgeab e of the subject material and interfaced well with the I
students.
l The inspec(or also reviewed the training organization, as delineated in the FSAR snd the Training Group Management Manual.
Previously the Training Manager and training organization reported to the Assistant Station Manager.
Effective September 4,1987, the training group was re-organized under the Training Center Manager (now referred to as the Training Manager) who reports to the Vice President - Nuclear Production.
The inspector confirmed with the_ licensee the need for an FSAR change in this area, which will be submitted with a future FSAR amendment.
l The inspector also inquired about the interface between the Security Department and the Training Group as it affected implementation of the GET program. While the Training Group handles the. actual general employee training and grading of the exams, the Security Gepartment has the overall responsibility for site access and therefore, performs the administrative duties related to GET.
Review of this security / training interface identified no conflicts or problem areas.
!
l
1
____._____m
____ _ _ _ _ _ _ _ _ _____
- _ - _ _ _
--_
.
.
During the course of the inspection, two incidents in which there was j
the appearance of cheating relative to the GET examination process were-brought to the inspector's attention.
In the first incident, two workers were observed talking during a GET exam.
Both workers were administra-tively failed.
They later retook the GET exam and successfully passed. In the second incident, a worker was observed with a copy of an exam prior to the class. This exam was destroyed and the student was subsequently given a different examination.
Corrective action on the part of the licensee in response to these two incidents included further direction to the instruc-tors on how to actively proctor the exam and better control the test process.
Each test is now in individual folders and is numbered and provided accountability.
The inspector also discussed with management the licensee's policy on GET failure and cheating. The licensee has a written policy on GET exam failures (ie: two failures of the exam and the worker is not permitted unescorted access to the protected area).
However, no written policy on cheating exists.
The inspector indicated that this lack of a written policy appears to place the burden for the prevention of cheating on the instructors, not the students.
The licensee will review their policy in
this regard and plans to develop a definitive statement on cheating.
i Based upon this commitment, the inspector had no further concerns in this area.
No violations were identified.
9.
RHR System Recirculation Line Weld Failures a.
Background: On August 7,1987 a cracked weld was discovered by the licensee in a 3/4" test connection on the "B" train RHR recirculation
line at flow orifice RH-FE-2474.
The crack was precisely located at the field weld (F0316) connecting a 45 degree elbow to the line off the orifice downstream flange tap. At that time, "B" train RHR was in service per TS and "A" train RHR, PCCW and SW were out of service for maintenance.
This portion of the RHR system is designated safety Class 2 in accordance with the ASME Boiler and Pressure Vessel Code, i
Section III.
l The cracked weld was cut out and the downstream piping and valve removed.
The open pipe end was plugged, seal welded and leak tested while the sytem remained in operation.
On August 27, 1987 another weld failure was found in the other flange of RH-FE-2474.
This time a crack had developed in the 3/4" i
insert to flange weld (F0104) on the upstream flange of the same recirculation line.
This weld was ground out and repaired after the system was taken out of service.
To accomplish this evolution, "A" train RHR was restarted, tested and declared operable.
This declaration was made after licensee evaluation that the inoperable
"A" train PCCW and SW systems were not required to support opera-bility of the "A" train RHR system.
,
,
.
I l
b.
Inspection: Initial NRC interest was generated following review of
)
the operator's log for August 27, 1987.
The log' entries on this
date described the sequence of events which took place to swap trains of RHR following identification of the second 3/4" weld failure. The inspector conducted a detailed review of-all out-l standing work requests on the RHR system,' verifying that-no j
maintenance in progress could adversely affect an operability determination on "A" train. The inspector reviewed the applicable.
work permits for both repairs (87W006260, 87W006501).
He noted i
that the August 27,1987 repair was conducted under the NHY.ASME l
Section XI repair program, whereas the August 7,1987. repair was
,
i not. After discussion with licensee representatives and a review i
i of the NHY Modifications and Testing Manual (NPMT), the inspector met with the cognizant NHY Engineering supervisors.
They provided an interpretation of the ASME Code Section XI which indicated that
!
i l
certain replacements of lines less than one inch nominal diameter j
do not require an ASME repair program.
I It was noted that "A" train PCCW and SW.were still inoperable by TS definition although "A" train PCCW was filled, vented and in operation.
No heat sink was avai'lable'via the "A" PCCW heat exchanger (CC-E-17A) since "A" train SW Fischer valve rework was ongoing. The inspectors had previously questioned the-system
configuration whereby both trains of PCCW were in operation.with i
l heat transfer occurring across thermal barrier heat exchanger and-l containment air handling (CAH) interfaces.
Elevated "A" train temperatures were maintained for ' "nistry purposes by balancing pump heat and system cooling loads (heat inputs) against thermal barrier ar,d containment atmosphere temperatures (heat. outputs).
This appeared to be an unusual system configuration not evaluated in the FSAR.
The inspector questioned the methods by which heat was being trans-ferred from a limited number of "A" train loads to the ultimate heat sink thru the "B" service water system.
In the case of the thermal barrier cooling system, such a cooling mechanism required heat transfer across thermal barrier heat exchanger tubes opposite from that which would be normally experienced in routine operation.
Since flow was not reversed and design' conditions were not exceeded, this reverse heat flow was not a component problem.
The licensee confirmed this position in response to 87,-RES-1109, in which other heat changers were also evaluated with. respect to reverse heat transfer conditions.
In all cases, it was determined that if.the-components function within design temperature parameters and with the normal operating fluid flow path, no adverse impact.upon the equipment is experienced.
However, while not a component problem, the connection of two separate' trains of a single system (eg: RHR or PCCW) to fulfill the design requirements of one operational train of that system is a questionable concept that requires further engi-neering review.
Particularly where nonsafety components are
\\
._______-_-_-_________:---____
. _.
.
.
necessary to effect the full safety system function, NHY engineering personnel have agreed to review this concept for compliance with FSAR commitments and single failure design requirements.
Additional discussion with station management further~ indicated that system operation in similar-unusual configurations could be expected i
in the future since the system was basically functional in the '
l absence of decay heat'even without a heat sink.
Licensee inter-pretation of the operability definition implies that even.with'
j plant radiation and minimal decay heat, such as might be expected i
during an extended refueling outage, full support system operability l
1s not required as long as the major system can perform its design
!
'
function.
!
l
c.
Analysis: NRC concerns in this area are listed'below. The' licensee has initiated efforts to address these issues. Together they con-stitute a single unresolved item'(87-16-02) and will be the subject'
of future NRC inspection.
'
(1) Evalut. tion of whether Interpretation XI-1-83-85 -applies to ASME repairs as well as replacements.
(2) Evaluation of whether operation of the PCCW system with cross connects through the thermal barrier heat exchangers.and CAH fan coolers is an acceptable method of heat removal with respect to single failure criteria.
(3) Operability of systems with inoperable subsystems or support systems when operational criteria are met, but design bases have not been addressed.
(4) Acceptabilityofth{repairofthe"B"trainRHRrecirculation
.
l line weld failure o6 August 7,1987 while still declaring the I
system operable.
J (5)
Investigation into the common mode failure of these welds.
10.
Operation of the Startup Feedwater Pump on an Emergency Bus The startup feedwater pump (SUFP) has a 1500 horsepower (hp) motor which
)
is rormally connected to and powered from nonsafety-related electrical
!-
bus E4.
It is also provided with an alternate connection to bus.E5, the
.
_
safety-related 4.16 kv bus which is powered by the "A" train diesel gen-
.
erator (D/G 1A) upon a loss of offsite power.
This alternate connection
'
requires manual implementation in accordance with defined procedural controls.
Based upon a concern identified by NRR of the capability of the diesel generator to power this additional 1500 hp load in conjunction,with-the normal Class 1E load requirements, the NRC staff required that the licensee demonstrate that the "A" train diesel generator is capable of'-
__---_ _ _ a
~
..
..
..
..
.
..
..
..
..
.
...
- _ _ - -
,.
.
I
l powering the SUFP while already " carrying the maximum train.A. load listed
in FSAR Table 8.3-1" (Reference: Seabrook-Safety Evaluation Report (SER)
and supplemental SERs 4 & 6).
During this inspection, the inspector reviewed the results of preopera-tional test PT-39.2 for " Loss of Offsite Power with SI" and the procedural requirements of operating procedure OX1426.02 for the "D/G 1A 18 Month Operability Surveillance".
He noted the following apparent inconsistencies with the above referenced SER discussions:
Per the results of PT-39,2, D/G 1A was carrying a load of 3600 Kw
--
(and not the 5582 Kw listed in FSAR Table 8.3-1) when the SUFP was i
.
loaded onto the diesel generator for testing.
j OX1426.02 does not duplicate the loading conditions of PT-39.2,
--
because the surveillance is run with the lesser load provided by a
!
service water pump versus a cooling tower pump on D/G 1A when.the
!
SUFP is added.
!
l
'1 l
Discussions with both the licensee and NRR on these inconsistencies I
revealed that the FSAR Table 8.3-1. load ratings were design values which could not be fully achieved during system operational testing. Addition-l I
ally, not all the equipment listed in Table 8.3-1.could. be realistically operated at the same time during the conduct of PT-39.2.
Thus, the 3600 Kw represented the maximum achievable loading at the time of test conduct.
The licensee provided background summaries of both the past D/G 1A testing
.
and the future plans for operation and testing of the SUFP on emergency j
bus E5, which served as bases for additional telephonic discussions with
NRR.
Based upon further review of this data and in consideration of their documented safety evaluations, NRR reviewers expressed concern with the adequacy of procedural controls for manual loading of the SUFP on 0/G 1A.
NRR also noted that certain wording in the surveillance requirements of TS 4.8.1.12 incorrectly implied that-only auto-connected emergency loads need be considered during the conduct of the 18 month surveillance.
-l l
The licensee has drafted a letter to NRR which agrees to changes in TS i
4.8.1.1.2f which provide consideration of all auto-connected loads during l
the affected D/G testing. Also, the licensee has committed to revise the applicable station operating procedures to ensure adequate margin on D/G 1A prior to the initiation of manual operator action to load the SUFP on that diesel generator.
Pending acceptance of these char.ges and licensee commitments by the NRC office of NRR and also pending implementation by
-
the licensee of the actions deemed acceptable by NRR, this issue regard-ing the operation of the SUFP on an emergency bus remains unresolved (87-16-03.).
!
-__-
- _ _ _ _ _ _ _ - _ _ _ - _ _ _ _..
_ _
.
,
'
'11.
Operational Activities and Operating Procedure Changes During routine control room-inspections, the inspector observed certain'
routine operations and plant status indicators which he questioned with
~
respect to either the system design or the procedural. controls being utilized for system operation or testing.
He discussed his questions-
)
with the operators on shift and later with licensee engineering personnel, i
Each issue is summarized below with a~ discussion of the NRC question, the
[
licensee response and any additional licensee action, where required.
j a.
RHR slip steam purification - This operation allows for a portion of the RHR flow to be diverted (ie: " letdown") to the chemical and volume control system (CVCS) for RCS~ purification purposes and returned directly to the suction of the RHR. pumps. The return flow path is termed the " slip stream" and provides for coolant purification while the CVCS charging system is not operating.
..
J The inspector questioned why a certain section of the slip stream piping was classified as nonASME, non safety-related.
The licensee prepared Engineering Evaluation Number 87-014 which analyzed the-slip stream piping design.
It was determined that since _the slip strecm flow of RHR can be placed in service only after the RCS is cooled to below 150 degrees F, the only safety-related consideration
,
is that the piping be seismically supported and constructed.
The inspector reviewed construction records to verify the seismic nature (NNS-1) of the slip stream piping.
The inspector also verified that operating procedures do provide the
)
requisite temperature controls for. placing the slip stream-flow path
'
into service.
However, in reviewing'these applicable c~perations procedures, he noted that certain valve' alignments for the RHR letdown to the CVCS were inconsistent with the ability to switch-from RHR "A" train letdown te "B" train letdown or vice versa.
Discussion with operations and technical support personnel led to the issuance of procedural changes (ie: OS1013.03, Revision.5 and OS1013.04, Revision 4) as they pertained to valve positions on RH-V-18 and 19.
The inspector checked that the revised procedures were successfully utilized in a subsequent RHR evolution involving the swapover from "B" to "A" train cooling.
The inspector had no further questions on RHR slip stream purification, b.
Atmospheric Steam Dump Valve'(ASDV) testing - The inspector witnessed l
stroke timing of two of the four total ASDVs (MS-PV-3001 and 3002),
reviewed the applicable surveillance procedure (EX1804.002, Revision 2) and discussed the conduct of this test with operations and tech-nical support personnel.
He questioned the actual stroke times relative to FSAR (Section 10.3.2.4) commitments, and the fact that the surveillance procedure only ' required the measurement of " closing" i
and not " opening" stroke times.
Also, the ASDV design controls
'
_ _ - _ _ _ _ _ _ -.
. _ - _ _ _ _ _ - -
-
.
,
,,
t
l
include not only manual stroking capability for all four valves off'
the safety-related, backup air supply, nitrogen bottles, but the j
ability to stroke two of these valves off the nonsafety, instrument
{
air (IA) supply lines.
Thus, complete testing would necessarily l
include conduct of the surveillance for MS-PV-3001 and 3002 with the
instrument air supply both isolated and unisolated. This provision
!
was not provided in Revision 2 to EX1804.002.
-
l The conduct of the stroke time testing witnessed by the inspector
{
was related to the licensee's own analysis of the differences in i
'
stroke times observed on the four separate ASDVs.
The stroke time criteria for all four valves had been increased in accordance with Engineering Change Authorization (ECA 98/118583A).
As a result of this internal analysis and, in part, in response to NRC questions, the licensee issued Revision 3 to EX1804.002 and concurrently made it an operational surveillance (0X) procedure.
l The inspector reviewed this procedural change, noting that both
" opening" and " closing" times were measured and that for the two l
af fected ASDVs, the surveillance test was repeated with IA both
^
isolated and unisolated. Also, the licensee committed to revise FSAR section 10.3.2.4 where a conflict with ECA 98/118583A was noted to exist.
Based upon the above licensee actions, the inspector had no further questions regarding the observed ASDV stroke time testing.
c.
Safety Parameter Display Ssytem (SPDS) status - In response to l
specific SPDS open items documented in supplement No.6 to the SER, l
the licensee committed to certain improvements in a letter to the NRC (NYN-87026) dated March 6,1987.
During this inspection, a sample of SPDS design modifications were spot-checked.
The inspector confirmed the following licensee commitments were implemented:
The top-level critical safety function summary is continuously
--
displayed on all CRT formats.
--
The containment isolation display has been patterned and illuminated to provide a recognizable status of containment isolation from the STA console.
Also, a display screen of individual containment isolation valve position status has been added to the SPDS options available on call at the STA Console CRT.
The subcriticality and core' cooling critical safety functions
--
have been made made dependent in such a way that their displays do not unnecessarily indicate status tree challenges, when such are not applicable during normal power operations.
- _ _ _ _ _ _ _ _ _ _ _ _ _ _
.-
-___- _
l
<
e,.
I With respect to the third inspection item above, the inspector observed the subcriticality critical safety function to be " black" in Mode 5, when a " green" indication of fault tree status was i
expected.
Discussion with the. responsible reactor engineer revealed l
that this situation was abnormal and currently existed only because i
of the lack of activity within the core.
The position of the gam-mametric ex-core detectors, which provide input to the subcriticality status, was far enough removed from the in-core neutron sources so as not to bring the subject status tree under analysis.. Additionally, the licensee is reviewing further SPOS improvements.in this area such that the mcde dependency of certain critical safety functions will be i
displayed as " green" when the status tree is not challenged.
Thus, a
'
i
" black" display would only be reserved for unreliable or erroneous-l data.
The inspector reviewed.the functional descriptions for the software used to evaluate the critical safety function status trees and additional working documents describing the planned revisions.
He had no further questions on either the implemented SPDS improvements or the proposed future changes, d.
Containment sump level instrumentation surveillance activities -'In NRC IR 87-10 it was noted that the surveillance requirements for checking silicone-oil level in the containment sump level trans-a mitters (1-CBS-L-2384 & 2385) had been increased to a weekly activity. This was done to verify that the observed weeping oil from the threaded conduit joints represented no real problem.
During this inspection, the licensee changed this surveillance to a monthly periodicity.
The inspector reviewed the revised repetitive task sheets (RTS) and the basis for the change.
The licensee had determined that during the past 17 weeks in which l
weekly surveillance were performed, there had been no discernable i
loss of silicone-oil and no oil had to be added through the fill plug.
The level of silicone-oil in these transmitters relates
.l directly to their functionality and environmental qualification.
'
Prior to the observed weeping, a quarterly oil level check was l
deemed adequate.
The inspector had no questions regarding the
!
current licensee decision to conduct monthly surveillance. -In response to ASLB interest on this subject, the revised RTS have been added to the component EQ files.
12.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations.
Unresolved items disclosed during this inspection are discussed in paragraphs 9 and 10.
_-_=
.
....
\\
13. 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 September 11,1987
,
- to discuss the inspection findings during the period. ' During this -
J inspection, the NRC inspectors received no comments from the licensee that any of their inspection items or issues contained-proprietary information, l
The only written material provided to the licensee was a copy of the NRC Vendor Inspection Branch inspection report (99901082/87-01) documenting.
the observation.of fabrication and testing processes of spent fuel storage racks at U.S. Tool and Die, Incorporated.
This public document
,
was specifically given to the licensee to review for applicability of
'
the findings to Seabrook Station. While U.S. Tool and Die was not involved in the fabrication of spent fuel storage racks at Seabrook, it did supply the new fuel storage racks which are installed in the
-!
fuel storage building. Currently, no new fuel is being stored in the-subject racks.
On August 26,1987 a meeting was held with the station manager to discuss l
the NHY program and policy on personnel attentiveness.
Previously, as a
.j result of an event at another plant which is the subject of NRC Informa-tion Notice 87-21, the licensee had reviewed their personnel policy on this matter as it related to the performance of licensed operators.
The meeting with the station manager additionally addressed the licensee
{
position on attentiveness of all station staff personnel, particularly those required to be in a duty status because of regulatory requirements (eg: fire watch).
Licensee initiatives, notifications to the NRC resident inspectors and regulatory reporting requirements were all discussed.
Also, during this inspection period, the inspector confirmed with the i
station operations manager the NHY position that TS Limiting Condition I
for Operation (LCO) 3.0.3 is not intended for use as an operational convenience to permit redundant safety systems to be removed from service for a limited period of time.
Based upon problems with the i
interpretation of LCO 3.0.3 at other plants, the NRC position is that voluntary entry into LCO 3.0.3 is unacceptable.
The station operations
'
manager agreed with this NRC position and indicated that implementation at Seabrook Station was in consonance with the bases for LCO 3.0.3 issued with NRC Generic Letter 87-09.
.