IR 05000293/1990009
| ML20059J663 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/20/1990 |
| From: | Chiramal M, James Trapp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20059J655 | List: |
| References | |
| 50-293-90-09, 50-293-90-9, NUDOCS 9009200106 | |
| Download: ML20059J663 (8) | |
Text
-
+
.
- ..
.
.
.
- .
.
.
d
!
!
U. S. NUCLEAR REGULATORY' COMMISSION
-'
REGION-I
Report No. 50-293/90-09-
Decket No.:50-293
,
.
License No DPR-35-L Licensee:
Boston Edison Company
800 Boylston Street I
Boston, Messachusetts 02199 i
Facility,Name:. Pilgrim Nuclear Generating Station;
Inspection At:' Plymouth; Massachusetts"
~
i
'
- Inspection Conducted: August' 6-10. 1990-I Inspector:
k 8,2p7; 90-twoo Jamesd. Trapp,Sr; Reactor fngineer-Special'
'ddte
'
,
TestC&rograms, EB, DRS~
d kfo--
Approved by:
1e Matt Chiramal, Acting Chief, Special Test dhte
'
'
Programs Section, Engineering Branch,JDRS Inspection Summary:. Inspection on August 6-10,~1990 (Inspection Report l
No. 50-293/90-09).
!
Areas Inspected:
Routine unannounced inspection.of~the licensee's control.of
-
design,- design changes, modifications, and temporary modificati_ons. _In addition,.
L the engineering and technical support organization, staffing,; communications,.
'
!
quality assurance, training,'and management support werel reviewed; JWeak areas identified during.thet previous' SALP period were: also inspected.
-Results: No violations or deviations.were: identified. 1The design and I
modifications process was found to comply with regulatory and ' station i
- administrative requirements.
a
+
,
ll w
9009200106 900829 T PDR ADOCK 05000293; G
PNU..
-
-.
'. ^ ',,
.
v.
- '.
...
!
,
.
.
DETAILS'
1.0 Persons Contacted
'
1.1 Boston Edison Company (BECO)
<
- M. Akhtar, Mod. Mgmt. Div, Mgr.
- R. Anderson, Plant Manager.
_
- J. Bellefeuille, Reactor Safety & Perf. Div.
- P. Cafarella, Principal System Eng'.
J
- R. Fairbank, Eng. Dept. Mgr.
- P_. Hamilton, Compliance Div. Mgr.
-
- K. Highfill, PNPS Director.
- J. Kel_1y, Sr, Compliance-Eng.
-
-
- V.-Oheim, Deputy Eng. Mgr.
- J.?Pawlak, Principal-Eng.
- F. Schelberger, Principal Quality: Eng.
- L.'Schmeling, Acting Plant'Mgr.-
R. Schifone, QA/QC Division Manager
- N. Simpson, Pnincipal' Eng, Tech Group
,
E. Wagner,. VP Nuclear Engineering i
1.2 U.S. Nuclear Regulatory Commission I
C. Carpenter, Resident Inspector
!
J. MacDonald, Sr. Resident: Inspector
"
W. Olsen, Resident. Inspector r
- Denotes present at the exit-meeting held August 10,'1990.
2.0 Plant Design Change (PDC) Review (IM 37700 &'37828)
.
- )
The objective _ of inspecting PDC. documentation is to ascertain _.that design
. changes and modifications to_ safety related systems _ receive-adequate
'
engineering review and that design changes are implemented.in accordance with regulatory requirements andtplant administrative procedures.
Included-J in this review-is a_ verification that adequate _ training, drawing-revisions;
'
design document update, and -testing are performed prior to declaring-
,
modified systems operable.
'To. satisfy this objective, two PDC packages were selected'for review:
'
,
j PDC 89-25, "Resize Core Spray Injection Orifice Plates,""
-
i PDC 89-05, " Replacement of the HPCI Gland Seal: Condenser-Condensate
Pump P220,"
'
PDC 89-25 was implemented to. increase-the margin between the Technical
[
-
Specification required core spray pump flow rate and discharge, pressure,-
'
and the system design.
To provide. additional margin,.thCpressurejdrop
across a discharge line flow-restricting orifice was decceased. This
'
allowed a lower pump tiischarge pressure to satisfy the Technical Specification flow requirement.
The PDC provided calculations and-instructions for increasing the bore. size of the restricting orifice.
I.
o-.
e l
L, ' ^ '
'
,
..
- .
..
-
-
{
,
_PD.C 89-05 was written to. replace the HPCI Gland Sial Condensate Pump I
~ hat wat worn beyond repair. The existing pump', a WESTCO type BR4?0,
.
t was no' longer supplied by the manufacturer, therefore, the. licensee
chose to replace it with a WESTCO type SR4R9. pump.
The type SR4R9
pump had been specified as the correct model number in the HPCI skid:
supplier's vendor manual and drawings. However, the BR410 pump had been originally supplied with-the HPCI. skid,. The puap skid supplier informed the licensee that the model BR410 pump was selected for'it.'s-
hydraulic seal, design. The supplier believed the. hydraulic seals would provide improved operation if a vacuum werelto-exist in the -
?
condensate pump suction line, Based on this information,_the licensee's-
,
original. determination that _the ' system was-'being / returned to it's _
-!
intended design became-invalid, _ Additional. safety assessment for the
type SR4R9 pump, which utilizes mechanical seals, was required..-In-addition, the SR4R9 type pump was supplied as a commercial-grade part
_
and required extensive' testing and analysis-to dedicate:it as a safety related component.
'
-In general the engineering basis for the-two PDCs was sound._ Preoperational
,
y tests conducted for these modifications were thorough and detailed.1
~
Calculations used for design and to. support safety evaluations were; detailed l
and complete, i
J Licensee administrative procedures require that training, drawing updates,
'
procedure revisions,'and preoperational testing be completed prior-to
'
-declaring the modified system operable.
Modification packages, with the q
one exception noted below, were. complete and closed out_in accordance with stahon administrative procedures.
The licensee's current policy to update
'
both category' A (frequently used drawings) and Category B'(less-frequently used drawings) prior to PDC closecut, was viewed as a positive initiative by the Nuclear Engineering Department (NED).
The following observations were made while reviewing the PDC: documentation.
>
The updating of the vendor manual _ V-0257, for PDC. 89-05, was. not l
+
l complete. The engineering staff stated that the update to'the vendor
!
h manual was somehow lost and that the manual would be properly updated.
I In addition, engineering. management stated-that the control of_ vendor'
-
I manuals would.soon be controlled by NED which is intended to eliminate:
!
these types of problems in the future.
,
BECO calculation No, M-387, referenced'in PDC 89-25, was used tolupdatei lJ
-*
l'
.-the core spray quarterly p' ump. surveillance procedure-8,5.1.1.
However,-
'
the elevation of the. discharge pressure indicator was'not considered 4
- .
L in this calculation. The calculation assumed the pressure indicator
'
was located'where the sensing line penetrated the discharge line.' In i
fact the pressure indicator was located approximate'ly 2 feet below this point.
The licensee performed additional calculations which-
,
l determined that the calculated.value was approximately 1 psi
- j
'
non-conservative.
This was evaluated and found to be within the s
'
!
!
'
)
,
I
'
=
.a
,
..
.
k 4'
,
conservatism'that existed within the calculation and, therefore,.did a
not require any corrective action.
This example indicated a need for
engineering-follow-up to assure engineering analysis was properly-incorporated in plant procedures.
.7 T mporary Modification Review
!
3.0 L
Temporary modifications were reviewed to determine if the modifications were being properly controlled and safety evaluations were' complete and j
thorough.
Temporary Modifications packages reviewed were:
l d
83-016 Spent' Resin Storage Tank Level: Indication
,89-032 Lube 011 Purification Heaters-
[
90-015 RBCCW Sampling' Points-
.
.
'
The three temporary modifications were completed in accordance with' station procedures. -The relevant safety evaluations were thorough.. Temporary.
modifications were' being. tracked as part of-the Plan of-the Day. (P0D) ~
-
document. The temporary modification log was being reviewed monthly in
+
accordance with station procedures'.
The majority of Temporary Modifications were'less than 3 years old. Older temporary modificat. ions were removed or i
being superseded by making them permanent PDCs.One observation was'madeL t
in this area:
.
Temporary Modification 90-15 was initia'ted in response' to identifying
an unauthorized. modification-on the Reactor Building Closed Cooling Water System (RBCCW)..Tne temporary modification was written on
'
i June 6,1990 and reviewed by the Operations _ Review Committee (ORC)-on-
!
June 13, 1990. The mod fication had not'been' authorized for installa '
tion'as of August 10, 990, the last day of;the inspection.' The' licensee i
stated that small discrepancies'in-the field < installation-caused the
.
delay. The temporary modification was authorized and' installed
August 14, 1990.
This apparent lack of, timely corrective action;to' -
reestablish configuration ' control?of the RBCCW system was.found to be
,
the exception. All other temporary modifications reviewed had been-
processed in a timely manner.
4.0 Root Cause Analysis Root cause analysis is performed primarily by;the 22 onsite BECO system-
. engineers. -Equipment failures' are reported on PNPS Failure and -Malfunction Reports (F&MR).
F&MRs are screened by system engineering for applicability"
"t of root cause analysis. ~ Root cause analysis reports typically' include-the following sections: event description, root cause, corrective action,
- recommendations' for improvement, = and; safety signi ficance,- Two.F&MRs, 89-137/138 and '90-19", were reviewed to determine if. root cause analysis was being conducted in a thorough manner and to assure that adequate-
corrective actions were being implemented.
-l
.
.
-
t
'
<
,
,
s
-
-
.:
,
.,
L:,.
.
.
_,
I
'l.'
]
'S'-
'
j
,
,
F&MRs 8,9-137 and 89-138 were performed n-response to the HPCI inlet steam valve failing' to open during the-quarterly pump surveillance, test.
Thel root causeiof the failure was attributed to the. absence of torque values
,
for the.MOV torque switch. dial pointer tightening screws.- Loose dial screws
'
had caused the torque switch to malfunction resultinglin an overthrust-
,
condition.in the MOV, which' caused the motor to burn up. 'The root cause.
'
analysis documentation for this event was thorough.. Input to' root cause
.
'.
and failure analysis was provided by NED,-QA, and, system engineers, i
Extensive corrective actions were taken, including checU ng and torquing the dial l switch pointer screws on-all MOVs 1.n the. station.
F&MR 90-192 was written in response to: leakage on the~RCIC steam line drain-
-
piping.. The root cause-was attributed to.. pipe wall thinning as' a result of material erosion from high ' velocity.' steam and water droplets' downstream
.
of the steam' traps and' bypass valve.
Discussions with the-cognizant syttem
!
engineer indicated that the actual. root' cause was.a' leaking trap bypass
'
valve and a degraded steam trap ~ system.' The systems engineer!s knowledges of the failure was det' ailed but had:not been11ncorporated into the root.
i
<cause analysis-rep' ort.; The corrective action taken did not-include:a UT
'
inspection.of the drain. piping for further evidence of wall thinning.
In
?
,.
fact, the. system engineer indicated that-a >second leak was< identified and
[
o repaired following-testing of the first repair.
Long term-repairs to replace b'
the traps and. pipe material during: Refueling 0utage 8 (RF0 #8) is an i
~
>
appropriate corrective-action for this problem.
The root cause analysis and1 corrective actions designated by-the-system,
.
engineers provide.a. strong technical support for.the maintentnce orgenization.
-The extensive root cause analysis performed for equipment-failures' by the system engineers'is considered a positive ~ effort for---improving lplantL
~
reliability and. safety.
,
5.0 Review Of Weaknesses Identifieu During Previous'SALP.
l 5.1 Detailed Control Room Design Review'(DCRDR)l The DCRDR project can be divided into the installation'of interim
,
enhancements and development =of the conceptual design' plan. The?
conceptual design-plan.is progressing:on schedule and will be, submitted
,
to the NRC in November 1990.- The' interim enhancements;have not-
- i progressed as well as anticipated.. Problems have been encountered'in-i panel, lighting and control panel color variations, causing the new-
-panel labeling to be ineffective. ' Delays have-also been encountered -
.
in updating station procedures to conform with new equipment' labels.
'
5.2 Design ~ Basis Reconstruction l-The design basis reconstruction program is.being actively developed l
'
L by the licensee. One engineering staff member is workingion this L
project full time, with two additionalistaff devot.ing;approximately L
half their time to this project.
The licensee is currently reviewing-
,
b
.g=
f4
'
,,
'S i,
a
.
,
J.,
~
!
'
'
.
-
- l
-
- .
,.
e
- .
r
.
,
i
!
'
' design basis documents from other utilities.and industry guidance to
,
develop guidelines on what information is to be included in the PNPS l
- design basis document.
The licensee is planning.to use a computer-t based design and configuration control system to store the design
,
basis information. The-licensee demonstrated;the design' basis,
configuration control sof tware package to the in'spector.. 'P&ID drawings i
are presently being entered into the system. - The system is unlike
traditional CAD * systems in that the database information stored can~
j
'in_teract with graphic displays.
For example,=all information relat1ng-
to.the design'of.a valve may!be: accessed by-selecting the data-base i
.
information of f of the P&ID graphic dif play.
The licensee has near_ly-completed the installation of all P& ids into this-system. Upon completion, this system will lead to a readily accessible and
..
user-friendly design basis document.
The licensee stated their planning of design basis content - should: be, completed by' the 'end of this -year,-
- with a pilot system design basic reconstruction scheduled to begin,
-
next year.
- 6.0 Organization / Staffing / Management Su'pport
'
The Nuclear Engineering Department (NED) -is comprised of the.' design, analysis, ano project managers sections.
The section, managers. report to;the NED=
,
- manager. The-NED : manager' reports to the Director Nuclear Engineering -
Organization, who reports. to the senior VP, Nuclear.
Each engineering section is divided into a n' umber of'di'visionse each with'
a division manage'.
The NED is' staffed by approximately 87. engineers,.
r
.
with nearly half holding Masters degrees in Science. The engineering.
organization staffing is -very stable with an average of!9 years of BECO:
.
and 15 years of industry experience.
t The engineering staff has reduced the number of open Engineering Service Request (ESR) from 952 in September 1989,. to 435 in' June 1990 The goal'
I
.
for 1990 is to further reduce the open ESRs to:300 by-the end of 1990.
,
In addition to the reduction in backlogged ESRs, engineeringL hassaiso made
'
l a si r ficant reduction in the number of priority.B drawings which require upda W g, Priority B drawings are drawings not frequently used by.the operators.
Frequently used priority A drawings 1are maintained updated.
'l
,
The number of priority B~ drawings requiring revisions.has been. reduced
from~approximately.6000 to 1200. The new policy to update priority B
- drawings prior to declaring the modified. system operable, will prevent future problems with updating priority:B drawings.
~
'
,
NED has made enhancements in support of station activities.
The Design Section Manager, from the Braintree office, attends the Plan of the Day
.
!;
(POD) meetings held at the. station.
The Braintree engineering office is l
L kept apprised of station activities and requirements'by daily assembling l
engineering managers at Braintree and having a data and~ communications L
link which displays the POD schedule on a large screen.
This'allowsMthe
'
engineers in Braintree to participate'in the plant POD meetings.
This-effort has improved communications between the site and the engineering 1 i;
staff.
,
,
i
m
'
.
.
$j,
.o
.
A v
=
17; i
'
.
.,
!
.The engineering organization also supports site activitie's by staffing a -
'si'e engineering of fice with one full-time-engineer and four_ engineers on
,
a' rotating basis.
On' August 7,,1990,'the NRC inspector attended a Design Review Board (DRB)'
.
convened to review modifications PDCL90-29, " Lube 011ECentrifuge. Heaters,"
'
and PDC 90-50, " Disconnect PASS Iodine: Cartridge." The DRB is made up of.
NED managers and prov' ides a multidisciplinary teviewLof PDCs prior _ to the release for station' review and. approval.
The DRB perform.ed a. detailed review of the_PDCs.
A' number of questions were raised by the committee-members on design. adequacy.
These~were addressed by the cognizant, engineer prior to DRB. approval.
The DRB' emphasis.was placed on. safety and. adequacy-of=the design.
'
'
An example of.this'was the board's discussion of; PDC. 90-50 'on the ability'
'to purge the iodineJcart' ridge. tubing following the introduction of high:
activity into this-line.. This discussion resulted inithe cognizant' engineer
-
being required to make changes to the proposed PDC. lTh'e interdisciplinary.:
review provided by the DRB was~ viewed _ as having a. significant positive impact in incorporating sound engineering judgement into PDCs.-
t The licensee stated that recent improvements have been~made'in the area'of'
-
pipe stress analysis. The NED has installed softwareJand can perform pipe'
stress analysis inhouse, which was previously' performed by outside
contractors.
This? increased. technical: capability, of. NED was viewed-as a -
-
positive engineering department initiative.
.
Engineering is present'y completing a self. assessment. 'A review of theJ
-
self assessment report Oy the NRC' inspector for the Design Section and the Analysis Section was conducted. 'The results of the assessments were':found-l to be candid. The self assessments identified perceived. strengths?and'
"
weaknesses. -The process of conducting -self assessmentsLisia' positive
-
initiative by the engineering organization to further improve.their k
performance.
7.0 QA Involvement in Engineering
'
In response to a perceived-weakness'.in the Stan' ding Plant Design Changes.
(SPDC) process, engineering management requested'that QA perform an-audit
-
.
in this area. The SPDC process is intended _to be used forfsingle discipline, minor design changes. Modifications Which: undergo the-SPDC process-' receive-less rigorous review and approval-than' standard PDCs.
A' review of QA Audit Report 90-13, " Design' Control," indicated several safety significant. findings-and recommendations for investigation and improvement.
This audit required'
an estimated 700 auditor man-hours, using.7 auditurs over 6 week period.
.
.:
In response to the audit' findings, NED initiated a: task force to di sp'o si ti on -
'
the findings.
The audit report-by the Quality Assurance Department was thorough and indicated a serious commitment of resources and effort'in=
conducting high quality audits.
,.
l
.
>
,
._
.,, g., 4..
-
.
....
y g
i
'
9.0 -Staff Training-The' licensee's t' raining program was reviewed to evaluate'the adequacylof
- training given to the' engineering staff.c Discussion with NED management
~
indicated that the engineering department has 4. " read' and sign". training
'
- program for new employees but does not have a formal training program for
.the-engineering stafi The absence'of a continuing training program for
.
"
'
the engineering.-staff is considered:to be a weakness.
10.0 Exit Meeting
~
At.the conclusion.of the site' inspection, on August-10,:.1990, an. exit interview was ' conducted with ~ theLlicensee's seniorisite representatives (denoted in Section1)-to' discuss the resultsland conclusions of;this'
>
inspection,-
. No written material was provided to the licensee by the inspector. Based on the NRC Region I re' view of this ' report and discussions held w*ithi..
,
- licensee representatives during<thisLinspection, it was determined that:
this report does not contain information subject to 10 CFR 2.790 restrictions.
.,
}-
.
I E
t L
..
!
J l
[i
'
,
.
e 1.
l I.
l
.
jft
'
,l
,
s
'
.
,
,