IR 05000293/1987021
| ML20235N139 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/29/1987 |
| From: | Anderson C, Paulitz F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235N098 | List: |
| References | |
| 50-293-87-21, NUDOCS 8707170450 | |
| Download: ML20235N139 (10) | |
Text
'
i
.
'iJ.S. NUCLEAR REGULATORY COMMISSION
REGION I
l Report No. 87-21 Docket No.
50-293 License No. DRP-35 Category C Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 J
Facility Name: Pilgrim Nuclear Generating Station Inspection At: Plymouth, Massachusetts l
Inspection Conducted: May 11-15, 1987
'
'
Inspector:
2(
t h
N7 Frederick Paulitz, R actor Dat4
'
'
Enginee i
Approved by:
(o f 81 Clific/rd J/ Anderson, Chief
' Date Plant Systems Section,EB Inspection Summary:
Routine unannounced inspection of plant modifications for the Safety Enhancement Program Diesel Generator, HGA relay replacement and followup of previous inspection findings by one Region based inspector.
Results:
No violations, deviations or other unacceptable conditions were identified.
8707.t70450 870707 PDR ADOCK 05000D93 O
- _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ -
._
_ _.
.__
_
. _ _ _ _ _ _ _.
_ _ - _ _ _ _ _ - - - _ - _ _ _ - - _ _ _ _
s
.
DETAILS 1.0 Persons Contacted 1.1 Boston Edison Company
- R. Bird, Senior Vice President
- K. Roberts, Nuclear Operation Manager R. Swanson, Nuclear Engineering Manager H. Brannan, Quality Assurance Manager T. Sowdon, Radiation Section Manager
- R. Grazio, Field Engineering Section Manager
- P. Hamilton, Acting Compliance Group Leader R. Sherry, Chief Maintenance Engineer J. Mattia, Quaiity Assurance Supervisor Group Leader P. Kahler, Senior Licensing Engineer J. Coughlin, Senior Electrical Engineer F. Mogolesko, Principle System & Safety Analysis Engineer W. Riggs, Senior Mechanical Engineer S. Bibo, Senior Quality Assurance Engineer J. Bryan McLaughlin, Senior Electrical Engineer J. Pawlak, Power System Group Leader M. Maquire, Station Electrical Engineer 1.2 Nuclear Energy Service W. Ringen III, Consultant 1.3 United States Nuclear Regulatory Commission
- M. McBride, Senior Resident Inspector
- J. Lyash, Resident Inspector
- L. Bettenhausen, Acting Deputy Director, Division of Reactor Safety Denotes those present at the exit meeting held on May 15, 1987.
- 2.0 Licensee Action on Previous Inspection Findings 2.1 (Closed) Unresolved Item (87-09-01) Core Spray System Valve Position During the review of IFI(85-30-16), which was closed in inspection report 87-09, the inspector noted the following discrepancies between
'the Operation Procedure No. 2.2.20 " Core Spray", revision 29, approved January 31, 1987 and the folicsing drawings:
Core Spray System P&ID drawing M242. revision Ell
Core Spray System FSAR drawing figure 7.4-8.
'<
_ _ _ _ _ _ _ _ _ _ _ _ _ _. -
_ _.
_ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _. _ _ _ _.
s
,
The procedure indicates the following valve position:
Page 8, VII Operating Procedures, A. Standby Status,
5.
M0-1400-25 A&B -Switch in Auto.
Valve position indication indicates valve open on Panel C903.
6.
MD-1400-24 A&B Switch in Auto.
Valve position indication indicates valve closed on Panel C903 Page 2.2.208-3. Valve Check List
-
M0-1400-25A normal position open
-
MO-1400-24A normal position closed Page 2.2.20B-10 Valve Check List
-
M0-1400-25B normal position open
-
MO-1400-21B normal position closea The drawing indicates that MO-1400-25A(B) are normally closed and i
MO-1400-24A(B) are normally open.
However, the procedure indicates
'
that MO-1400-25A(B) are normally open and MO-1400-24A(B) are normally closed.
The inspector determined that the licensee had received information I
from General Electric Company concerning the seismic adequacy, for
]
certain applications, for the HGA type relay..The licensee performed a safety evaluation, number 2054, dated January 29, 1987. This safety.
'
evaluation recommended temporary repositioning of the MO-1400-24A&B j
and 25A&B valves due to the potential inoperability of the M0-1400 25A&B (normally closed) valves.
This inoperability was due to the potential for HGA relay 14A-K20A&B "b" contact bounce during a seismic
.
event. This relay contact bounce is discussed in a memorandum to i
J. Pawlak from J. Rogers, dated January 23,1987, " Evaluation of Failure in FGA Relays 14A-K20A&B (Core Spray)".
This HGA problem and corrective action is further discussed in paragraph 2.2 of this l
report. The plant conditions required for the temporary repositioning of the above valves are:
Reactor is in cold shutdown condition.
(<212 F)
Reactor shall not be made critical.
Reactor pressure is at atmospheric condition.
Operations personnel are informed of the modified core spray lineup and procedure changes required to manually initiate core spray have been made.
The inspector reviewed the above safety evaluation.
No deficiencies were identified within the safety evaluation conclusion. The safety evaluation conclusion is that the core spray system would still meet all necessary safety functions even with the above valve changes af.d plant conditions.
The licensee's field compliance group is tracking this item to assure the following actions prior to plant restart:
l l
s
)
The operating procedure is revised to reflect the normal valve j
lineup as described in the FSAR and other design documents.
The valve lineup is changed to agree with the revised operating j
procedure, FSAR and other design documents.
t
The corrective action for the HGA relay 14A-K20A&B is complete.
This unresolved item is closed.
j 2.2 (Closed) Inspector Followup Item (86-37-09)HGA Relay Potential i
Seismic Deficiency j
This potential generic seismic deficiency was brought to the attention of Region I by another licensee who responded to General Electric (GE), Power Systems Management Business Department, Service j
Advice Letter (SAL) number 174.1.
The Senior Resident Inspector at t
the Pilgrim Nuclear Station (PNS) inquired what Boston Edison Company
'
was doing about this seismic design deficiency for the HGA relays which were installed at the facility.
The personnel at PNS were not aware of the SAL. The deficiency was confined to normally deenergized relays, which used normally closed "b" contacts, which-could repeatedly open and close during a seismic event.
The Nuclear Engineering Department (NED), Power Systems Group (PSG),
analyzed 248 HGA relays, with respect to system operability, and identified a number of relays which required replacement by a GE type HFA relay. This analysis is documented in Calculation No PS-29, Revision 0.
The NED System and Safety Analysis Group (S&SA) reviewed the above analysis and safety evaluation. The S&SA's HGA relay contact evaluation resulted in the identification of additional relays that would require corrective action than had been identified by the PSG. The S&SA group was concerned that during a seismic event the chattering contact would not allow a relay / coil to pick-up and
,
seal itself in.
The PSG performed another review based upon the following two basic classifications:
No further action was required if the contacts / logic met one or more of the following criteria:
The normally closed contacts were not being used.
-
-
The normally closed contacts were used for a alarm function.
-
The normally closed contacts were used for light indication.
The relay was found to be a spare relay.
-
The relay / contacts were not Class IE.
-
The chattering contact would.only result in an operating
-
device completing its safety function.
- An evaluation would be required,to determine further corrective
,
action if the contacts were found to not meet the above non-action criteria.
- _ _ _ _ _ _ - - _ _ _ _ _ - _ - _ - - - _ - _
_
_
_.
_
-- - - _ __ __ - _ -
_ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _.
-
__
%
.
The final number of relays that the licensee determined were to be replaced was fifteen (15).
The modification package PDC 97-27 had been initiated for the 15 relays which required corrective action.
These corrective actions are to be completed prior to plant restart.
This item is closed.
2.3 (0 pen) IFI (85-30-15) Motor Operator Valve (MOV) Overload Protection In a previous inspection the NRC determined that the licensee selection of MOV thermal overload heaters was not consistent with either the vendors recommendations nor was it consistent between MOVs having the same function and horsepower rating.
Presently the heaters provide no MOV protective function but are used for alarm only during surveillance testing or accident conditions.
The NRC had been concerned that the above inconsistencies might impact the alarm function resulting in MOV failure.
The licensee's corrective action, for the above problem, has been deferred until the overall plant electrical protective coordination study has been completed.
This study extends from the 345KV system down to the 480 volt system.
This IFI remains open pending completion of the licensee's corrective action to provide MOV protection and electrical system protection coordination.
2.4 (0 pen) Unresolved Item (87-09-02) Battery Maintenance During a previous inspection the 125 volt and 250 volt direct current (DC) system was inspected to determine the effectiveness of the licensee's preventative maintenance program.
Several potentially deficient aspects of the preventive maintenance program were identified.
Prior to this inspection the licensee had removed the battery intercell connection corrosion from the surface.
However, the inspector noted during this inspection that there was further corrosion that could not be removed unless the connection was disassembled. This further corrective action will be conducted during some future preventative maintenance activity. Also, the inspector requested that the licensee should identify the root cause of this corrosion.
l With regard to the low specific gravity (SG) condition, of cell 15 of battery B-125V, the licensee had corrected this condition. Guidance
,
!
for this corrective action was provided by using Procedure No. 3.M.3-25 " Cleaning and Agitation of Station Batteries".
The low SG was corrected by stirring the electrolyte on February 18,1987 as documented by MR-87-46-18.
- - - _ - _ _ - _.
s
.
This unresolved issue remains open.until the licensee provides the fcllowing:
Independently identify the unknown material in the battery
electrolyte and independently verify the-effect upon the battery
<
operability.
This issue was discussed previously in inspection
. '
report 87-09.
Provide corrective action for the intercell connection corrosiori
noted above and identify the root cause of the intercell connection corrosion.
Reduce the 1 inch gap between the battery cell and the battery
rack end rail. This was identified in inspection report 87-09 as being completed based upon preliminary information from the licensee.
However, the inspector has determined by visual inspection that this deficiency had not been corrected.
3.0 Standby Emergency Diesel Generators Control Wires Terminations The licensee identified loose wires in Diesel Generator "B" control panels C102, C104A, and C104B.
These loose wires were identified during the E203 phase 3 walkdown project. Approximately 10% of the internal wire terminations were loose. These loose wires had no termination lugs.
The-licensee issued a nonconformance report (NCR), number 87-121, on March 18,1987.
The NCR Engineering Evaluation corrective action was to provide terminal lugs for all unlugged wires. To establish the root cause for the loose wires the engineering group's recommendation was to monitor
!
these terminals to determine if this condition resulted from panel vibration.
A similar condition for wire terminations (unlugged wires) existed in f
Diesel Generator "A" control panels C101, C103A, and 1038.
However, the
{
number of loose wires was less.
The licensee issued NCR 87-239 and Quality Control hold tags 87-61.14 were placed on the panel doors.
l The inspector observed the wire terminations for both diesel generator
control panels.
The unlugged wires were associated with Westinghouse BFD relays, internal wiring terminated on terminal strips, and push buttons.
Wire lugs were provided for terminations to control switches, Agastat I
relays, General Electric CR relays and external wiring on terminal blocks.
The inspector discussed with the licensee that vibration may not have been
!
l the root cause of the loose wires. The licensee is evaluating the surveillance activities, which require lifting of wire terminations, to ascertain if the lif ted wires have a higher incidence of being-loose.
The corrective action for the Diesel Generator "A" will be to provide termination lugs on all wires that are unlugged.
I
!
__ ___ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _._.
._
L
s
7
q
.
l Westinghouse Electric Corporation (WEC) was contacted by the licensee in
,
regard to the seismic qualification of the BFD relays with unlugged wiret I
WEC had not seismically qualified the BFD relay with unlugged wire terminations. Westinghouse plans to conduct this type of qualification.
However, the seismic qualification results are not currently available, q
The inspector concluded that the corrective action, taken by the licensee, for the Diesel Generator "B" and the proposed corrective action for the Diesel Generator "A" is acceptable for the following reasons:
The unlugged wires were not identified as being a direct cause of the
loose wires.
Wires.with terminal lugs would be easier to remove and reinstall
correctly during surveillance activities The BFD relays were seismically qualified with wire terminal lugs.
- The inspector had no further concerns with the diesel generator wiring.
4.0 Inspection of. Plant Modifications The purpose of this inspection was to determine if the plant i
modifications, which were determined by the licensee to not require approval by the NRC, were conducted in accordance with provisions of 10 CFR 50.59 and to determine if commitments, license conditions and requirements were being implemented properly.
These modifications are currently in progress and are at various stages of completion.
The inspector reviewed selected portio,is of the modification packages and associated documentation to determine if they were being implemented in accordance with the following commitments or requirements:
A safety evaluation (SE) had been written to assess whether the
change constitutes an unreviewed safety question or a change to the facility Technical Specification (TS).
The modifications were controlled by the licensee's administrative
procedures.
Adequate quality assurance was in effect for the above activities.
- The design change activities were conducted in accordance with the
appropriate specifications, industry codes and standards, drawings, and other requirements.
The details of the above areas of the inspecting are discussed below.
-__-____ - _______-
'
q
<
4.1 Review of the Modification Packages The inspector reviewed the following plant design change (PDC)
modification packages:
PDC No. 87-27 HGA Relay Replacement-SE No. 2136 PDC No.86-56A Station Blackout Diesel Generator-SE No. 2106
PDC NC.86-56B Station Blackout Diesel Generator-SE No, none These modifications are being implemented as major plant design
changes in accordance with the Nuclear Engineering Department
!
Procedure NED 3.02, Preparation, Review, Verification and Revision of I
Design Documents for Plant Design Changes, Revision 17.
The
{
inspector reviewed the procedure and modification packages to verify
!
that sufficient administrative controls existed for the activities
'
associated with the modification and that these controls were in I
effect.
Specifically the review was to ascertain that the following l
controls exist:
Procedures for the centrol of the modifications, including methods for initiating design cnanges; and methods to assure that the modification does not involve an unresolved safety question.
- Procedures and responsibilities for design control, including organizational responsibilities; review of design and modification program; methods for conducting safety evaluations, reviewing and approving design input requirements and performing l
independent verification; training of personnel performing I
modification activities; and auditing design activities.
'
,
l
Administrative control for design and document control, i
including interdepartmental interface; record storage; l
implementation of approved design changes in accordance with i
approved procedures; post-modification acceptance testing; and I
post modification acceptance criteria.
The safety evaluation for PDC 86-56B was part of the modification package and did not have a separate number.
The Safety Enhancement Program (SEP) Station Blackout Diesel Generator design packages were separate.
PCD 86-56A covered the site preparation and civil work.
PCD 86-56B covered the mechanical and electrical work.
The PDC 86-56B package is considered Q although the diesel generator and auxiliaries including the associated electrical components are not Category 1 seismic, ASME 3, nor Class 1E.
The inspector concluded that the interface with safety-related structures, system and components had been addressed in the design and safety evaluation.
No unacceptable concfitions were identified.
,
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
-_
_ _ _ _ _ _ _ _ _ _ _ _
-
I i
.-
i
The inspector reviewed selected portions of the modification packages including associated engineering drawings, safety evaluations and proposed changes to the FSAR.
The scope of review was to ascertain that the packages were technical adequate and complied with the related commitments and NRC requirements.
No unacceptable conditions were identified.
4.2 Implementation of the Modifications The inspector reviewed selected procurement and installation documents to verify that they compiled with the appropriate standards, specifications and drawings.
The electrical design rating of the various components were reviewed to ascertain whether they i
were compatible with the existing design.
The inspector also toured and visually inspected portions of the underground cable duct bank i
for the Station Blackout Diesel Generator modification (SBDGM).
This duct portion of the duct bank was completed under a previous modification.
The SBDGM PDC 86-56A sitework had received Onsite i
Review Committee (ORC) approval, on April 29,1987, during the ORC
'
meeting number 87-44. The licensee's tentative schedule is to have the SBDG installation tested and turned over by August 4,1987.
No unacceptable conditions were identified.
4.3 Quality Assurance The scope of inspection included a verification that the modification is being implemented in accordance with the quality assurance requirements of the following documents:
10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear
Power Plants.
Pilgrim Quality Assurance Manual, Revision 21.
- Nuclear Engineering Department Procedure NED 3.02; Preparation,
Review, Verification and Revision of Design Documents for Plant Design Changes; Revision 17.
- Other requirements and licensee commitments.
The inspector interviewed QA personnel at the NED office concerning their understanding of activities associated with plant modifications which were conducted in accordance with 10 CFR 50.59.
No unacceptable conditions were identified.
_ _ _ _ _ _ _ _ _ _ _ _.
. _ _ _ _ _ - _
_
--
,.
.
5.0 Unresolved Items Unresolved items are matters about which more information is required to ascertain whether they are acceptable items, violations or deviations.
6.0 Exit Interview The inspector met with the licensee representatives, denoted in paragraph 1 on May 15,1987.
The inspector summarized the scope and findings of the inspection at that time.
No written material was provided to the licensee by the inspector.
.
_ _ _ _ _ _ _. _ _