IR 05000293/1990009

Insp Rept 50-293/90-09 on 900806-10.No Violations or Deviations Noted.Major Areas Inspected:Control of Design, Design Changes,Mods,Engineering & Technical Support Organization,Staffing,Communications,Qa & Training

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: IR 05000293/1990009 ML20059J652
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- 1 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; 4 Inspection iAt:' Plymouth; Massachusetts"

~

'

- Inspection Conducted: August' 6-10. 1990- I Inspector: k twoo Jamesd. Trapp,Sr; Reactor fngineer- Special'

8,2p7; 90-

'

, 'ddte TestC&rograms, EB, DRS~

Approved by: 1e d kfo--

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 inspecte Results: No violations or deviations.were: identified. 1The design and I modifications process was found to comply with regulatory and ' station i

administrative requirement a

+

,

ll

!

w 9009200106 900829 T ADOCK 05000293; PDR G PNU ..

-

-.

' :. ^ ' , ,

. ... ; '.

!

,

. .

DETAILS'

1.0 Persons Contacted '

1.1 Boston Edison Company (BECO) *

<

• M. Akhtar, Mod. Mgmt. Div, Mg * R. Anderson, Plant Manage * J. Bellefeuille, Reactor Safety & Perf. Di _

• P. Cafarella, Principal System Eng'. J

• R. Fairbank, Eng. Dept. Mg * -

P_. Hamilton, Compliance Div. Mg * K. Highfill, PNPS Directo * J. Kel_1y, Sr, Compliance-En * V.-Oheim, Deputy Eng. Mg ;

• J.?Pawlak, Principal-En * F. Schelberger, Principal Quality: En * L.'Schmeling, Acting Plant'Mgr.-

R. Schifone, QA/QC Division Manager

• N. Simpson, Pnincipal' Eng, Tech Group ,

E. Wagner,. VP Nuclear Engineering i 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,'199 .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. o-. :

e

.. L, ' ^ ' ,

'

l

.. ;.

3  ;

-

{

-

,

• _PD.C 89-05 was written to. replace the HPCI Gland Sial Condensate Pump I

~t hat wat worn beyond repair. The existing pump', a WESTCO type BR4?0, .

was no' longer supplied by the manufacturer, therefore, the . licensee :

chose to replace it with a WESTCO type SR4R9. pum The type SR4R9 4 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 componen '

-In general the engineering basis for the- two PDCs was sound._ Preoperational ,

y tests conducted for these modifications were thorough and detaile ~

Calculations used for design and to. support safety evaluations were; detailed l and complete, i Licensee administrative procedures require that training, drawing updates, J '

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: documentatio >

+

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 futur ,

lJ -*

BECO calculation No, M-387, referenced'in PDC 89-25, was used tolupdatei 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 poin 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 actio This example indicated a need for 4 engineering-follow-up to assure engineering analysis was properly-incorporated in plant procedure .7 TLmporary Modification Review  !

Temporary modifications were reviewed to determine if the modifications were being properly controlled and safety evaluations were' complete and j thoroug 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.. Temporar 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- 4 processed in a timely manne .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 implemente 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 caus .

'

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. statio 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 roo 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 proble The root cause analysis and1 corrective actions designated by-the-system, .

engineers provide.a. strong technical support for.the maintentnce orgenizatio The extensive root cause analysis performed for equipment- failures' by the

~

system engineers'is considered a positive ~ effort for---improving lplantL reliability and. safet ,

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' label .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

. 61 ,

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 4 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 yea *

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, Nuclea Each engineering section is divided into a n' umber of'di'visionse each with'

a division manage'. r The NED is' staffed by approximately 87. engineers,. .

with nearly half holding Masters degrees in Science. The engineerin organization staffing is -very stable with an average of!9 years of BECO: .

and 15 years of industry experienc ;

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 199 '

l In addition to the reduction in backlogged ESRs, engineeringL hassaiso made 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

,

operator Frequently used priority A drawings 1are maintained update '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 drawing ~ '

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 1 m '

.

.

.o $j, v .

A

=

i 17;

'

.

.,

!

.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 basi ,!

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 o 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_PDC 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 desig '

'

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'

-

3 9 pipe stress analysis. The NED has installed softwareJand can perform pipe'

stress analysis inhouse, which was previously' performed by outside *

contractor This? increased . technical: capability, of. NED was viewed- as a -

-

positive engineering department initiativ .

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 performanc .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 perio .

.:'

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= l conducting high quality audit l

,.

1 l

,

._ 4

. , , g . , . . . ..

gy 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 weaknes .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 restriction .,

}- .

I E t L

..

!

J l

[i

' ,

.

e 1.

l I.

l

.

'

jft ,

s ,l

'

. , ,