ML20151L868
| ML20151L868 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/13/1988 |
| From: | Conlon T, Ruff A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20151L852 | List: |
| References | |
| 50-416-88-10, GL-83-28, NUDOCS 8808040166 | |
| Download: ML20151L868 (23) | |
See also: IR 05000416/1988010
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, UNITED STATES
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k'JCLEAR REGULATORY COMMISSION
o '* REGION 11
j[ 101 MARIETTA ST., N.W.
e,,,, ATLANTA. GEORGIA 30323
Report No.: 50-416/88-10
Licensee: System Energy Resources, Inc.
Jackson, MS 39205
Docket No.: 50-416 License No.: NPF-29
Facility Name: Grand Gulf
Inspection Conducted: May 23-27, 1988
Inspector: d. W WJ
A. B. Ruff, Telivltfader, Region II
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Eate' Signed
Team Members:
C. Paulk, Region II Inspector
C. Smith, Region II Inspectior
L. Foster, Region II Consultant
Approved by: N A- 7//3/F9'
Date Signed
T. E. Conlon, Chief 8
Plant Systems Section
Engineering Branch
Division of Reactor Safety
SUMMARY
Scope: This special announced inspection was performed to assess the
licensee's response to Generic Letter 83-28, Required Actions Based
on Generic Implications of Salem Anticipated Transient Without Scram
(ATWS) Events. Areas inspected included post-trip review, equipment
classification and vendor interface, post-maintenance testing,
reactor trip system (RTS) reliability, lic" s 's action on IEB's,
IEN's, and previous inspection findings.
Resul ts: In the areas inspected, no violations or deviations were identified.
8808040166 880721
PDR ADOCK 05000416
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REPORT DETAILS
1. Persons Contacted
Licensee Employees
- 8. Angle, Manager 0AS
- M. D. Bakarich, Support Systems
D. Bottemiller, Simulator Training Supervisor
S. Cothran EngineeringSupervisor(SIMSProject)
8. Crow,IECForeman
- D. Cupstid, Plant Support
- L. F. Daughtery, Compliance Supervisor
- M. A. Dietrich, Supervisor Quality Audits
- J. Dimmette, Jr Manager - Plant Maintenance
- W. C. Eiff, Prilicipal Quality Engineer, NPE
T. Errington Nuclear Engineer, Technical Support
'D. Graves,15CForeman
- F. Hennington, Planning Supervisor
- C. A. Hutchinson, General Plant
B. Higginbotham, I&C Technician
G. Imes, Component Data Base Supervisor (SIMS Project)
F. Key, Instrumentation and Control (I&C) Technician
- D. L. Pace, Manager, Nuclear Design
- J. Reaves, Manager Quality Systems
- J. Roberts, PM&C, Manager
- R. F. Rogers, Manager SpecialProjects
, *W.R. Rogers,MaterialsManagementSupport
M. Rowland, QA Engineer (SIMS Project)
- P. Simpson, Licensing Engineer
J. Simpson, Material Management Superintendent
- J. Summers, Compliance Coordinator
J. Tahler, Project Manager (SIMS Proiect)
- F. Titus, Director Nuclear Plant Eng neer
R. Wells, Engineering Supervisor (SIMS Project)
D. Williams, 0)erational Analysis Section
- H. J. Wright, Manager Plant Support
- G. Zinke, Technical Support Superintendent
Other licensee employees contacted during this inspection included
craftsmen, engineers, operators, mechanics, security force members,
technicians, and administrative personnel.
NRC Resident Inspectors
- R. Butcher, Senior Resident Inspector
- J. Mathis, Resident Inspector
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- Attended exit interview
2. Background
On February 25, 1983, both of the scram circuits at Unit 1 of the Salem
Nuclear Power Plant failed to open upon an automatic reactor trip signal
from the reactor protection system. This incident was terminated manually
by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers was determined to be
related to the sticking of the undervoltage trip attachment. Prior to
this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power
Plant, an automatic trip signal was generated based on steam generator
low-low level during plant start-up. In the case, the reactor was triaped
manually by the operator almost coincidentally with the automatic tr p.
This failure of the breaker to open upon an automatic reactor trip signal
was undetected by the licensee prior to plant restart.
As a result of the problems identified with circuit breakers at Salem and
at other plants, NRC issued Generic Letter (GL) 83-28, Required Action
Based on Generic Implications of Salem ATWS Events, dated July 8,1983.
This letter required licensees of operating plants to respond to
intermediate-term actions to ensure reli bility of the Reactor Trip System
(RTS). Actions to be performed by the licensees included development of
programs to provide for post-trip review, classification of equipment,
vendor interface, post-maintenance testing, and RTS reliabil:ty
improvement.
The licensee responded to GL 83-28 in several letters dated from
November 4, 1983 to April 1, 1988. These responses described the
licensee's compliance with NRC positions in GL 83-28. This inspection was
performed to verify com)11ance to the licensee's responses and to assess
the adequacy of the 1icensee's current program, planned program
improvements and implementation of present procedures associated with
post-trip review, equi classification, vendor interface,
post-maintenance testing,pmentand RTS reliability for the Grand Gulf Nuclear
Station (GGNS). The results of the inspection are discussed in the
paragraphs that follow.
3. Post-Trip Review
The licensee was requested in GL 83-28, Required Actions Based on Generic
Implications of Salem ATWS Events, to describe their program, procedures
and data collection capabilities to assure that the causes for unscheduled
reactor scrams, as well as the response of safety-related equipment, are
fully understood and applicable corrective actions completed before plant
restart.
The licensee's response to GL 83-28 provided a description of their
program pertinent to performing post-trip reviews. The inspector reviewed
their program, procedures, and interviewed key personnel to assess the
adequacy of the licensee's program for post-trip reviews. In addition,
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the evaluation and conclusion items of the NRC Safety Evaluation Report
(SER) in letters to GGNS on May 5,1985 and October 22, 1985 were verified
to be established. The inspection was formatted to verify that a post-
trip review program has been implemented and meets the following
attributes:
Procedures and equipment exist to cover post-trip review.
Safety assessments of the reactor trip are clearly delineated as part
of the post-trip review.
Post-trip review procedures are reviewed periodically by an onsite
safety review committee such as Plant Safety Review Committee (PSRC)
and upgraded in any areas that have been identified as deficient.
Plant personnel preparing and/or reviewing post-trip documentation
receive initial training and refresher training in post-trip review.
Responsibilities and authorities of plant personnel who will preform
the review and analysis of these events are clearly defined.
Criteria for determining the acceptability of authorized restart have
been established.
Criteria for comparing plant information with known or essential
plant benavior have been established.
Guidelines are established for preservation of evidence of reactor
trips.
The licensee's program with respect to each of tnese attributes is
discussed individually as follows:
a. Procedure and Equipment Exist to Cover Post-Trip Review.
Procedures that were used to verify program actions and documentation
are listed in paragraph 9. For simplicity and clarity, they will be
referred to in the the body of this report by number only.
AP-01-S-06-26 is the primary procedure for post-trip reviews.
ONEP-05-1-02-I-1 addresses immediate action for a reactor trip (or
what to do if a reactor trip should have occurred and did not) and
leads the operctor into 10I 03-1-01-4 for scram recovery. Procedure
01-S-06-5 specifies requirements to document incident reports and
requires notification of reportable events to applicable agencies.
All unplanned reactor trips require an incident report. 01-S-06-76,
addressed NRC and INP0 concerns in the area of reactor trips and is
intended, inpart, to aid in reducing the number of reactor trips. In
addition, the licensee is a member of the BWR Owners Group Scram
Frequency Reduction program.
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AP-01-S-06-26 provides the folicwing information:
(1) Purpose
(2) Responsibilities
(3) Definitions
(4) Date Collection
(5)On-shiftAnalysis
. (6) Restart Decision
(7) Off-shift Analysis
(8) Report Distribution
(9) BWR Scram Report for BWR Owners Group Scram Frequency Reduction
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The above procedure also provides for:
(1) Written statements from individuals involved in the event.
(2) Pre-trip plant status including pre-trip activities which could
(3) potential
Post-trip have
plant contributed
status to the event
(4) Plant response and verification that Reactor Protection System
(RPS) operated properly i
(5 Capture of plant transient information through charts, logs, etc
(6 Engineered safety feature actuations
(7 Cause of trip and Root cause identification
(8) Trip Classification
(9) Sequence of Event Description
(10) Restart Authorization
The reactor trip review places a reactor trip into one of four
classes:
Class I -
The cause of the trip is positively known and has
been corrected; all Safety-Related and other important
equipment functioned properly during the trip.
Class II -
The cause of the trip is positively known and has
been corrected: some Safety-Related equipment did not
function properly; however, the malfunction has been
corrected or a Technical Specification constraint does -
not prohibit'a startup
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Class III - The cause of the trip is positively known and had not
been corrected or the cause of the trip is not
positively known.
Class-IV -
Some Safety-Related and/or other important equipment
functioned in an abnormal or degraded manner during
the trip and the malfunction has not been corrected or
prevents startup due to Technical Specification
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The data collected for the event includes sequence of events from 80P
and NSSS Alarm Typers, NSSS Post trip log from NSSS Alarm Typer and
80P Post trip log from 80P on-demand typer. During plant operation
the GETARS (General Electrical Transient Analysis Recording System)
is in the Sentinel mode. It is capable of recording many parameters
(channels) for which plots can be made. GGNS presently records 263
channels of information with GETARS in Sentinel mode. The GETARS
data is recorded and saved for approximately 4.67 minutes during a
transient involving a scram ( a) proximately 47 seconds pre-trip and
233 seconds post-trip). Typical parameters monitored and capable of
being plotted from GETARS Sentinel program or other records are:
Narrow Range Level
Narrow Range Dome Pressure
Wide Range Dome Pressure
Total Steamline flow
APRM A thru H
Total Core Flow
Wide Ran e Level
Total Fe dwater Flow
The inspector observed a demonstration of the GETARS and reviewed
post trip analysis for scram numbers 48 and 49. The post trip
analysis were considered to be satisfactory.
b. Safety assessments of the reactor trip are delineated as part of the
post-trip review.
The safety assessment is clearly delineated in procedure 01-S-06-26
under "On-shift Analysis" to determine:
(1) Verification of the proper operation of safety-related systems
and other important equipment involved in the trip operated as
expected
(2) Abnormal indications or degraded trends in equipment performance
(3) Event occurring out of the normal or expected sequence
(4) Failed or abnormal response of equipment to control signals
(5) Unusual chemistry results or radiation readings
(6) Determination if a Technical Specification Safety Limit was
exceeded.
Restart decisions are all made by the General Manager, and if the
reactor trip is a Class III event, the PSRC's recommended actions
have to be resolved and/or completed.
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c. Post-trip review procedures are reviewed periedically by an on-site
safety review committee such as PSRC and upgraded in any areas that
have been identified as deficient.
Administrative Procedure 01-S-02-02 provides for biennial review of
AP-01-S-06-26. The Document Control Supervisor initiates and tracks
the review process by distributing a two year Review of Directives
form to responsible managers and section su3ervisors/superiendents
who in turn will assign a knowledgeable incividual to review the
directive / procedure. The Document Control Section retains all
returned two-year review forms as quality records,
d. Plant personnel pre 3aring and/or reviewing post-trip documentation
receive initial training and refresher training in post-trip review
procedures.
The Shift Technical Advisor's (STA) Training program provides for
initial training on the reactor trip procedure. This includes
reading and understanding the procedure, performing on-shift post
trip analysis, determining the cause of scram and completing a post
trip analysis report (See Lesson Plans SC-STA-SIM-LP-021-00, and
LP-013-00, and OP-AD-701/Rev 2). This training is documented on the
STA Qualification Card. Other licensed personnel license
candidates, and designated plant staff members receive training
through periodic required reading in accordance with AP-01-05-04-15.
e. Responsibilities and authorities of plant personnel who will perform
the review and analysis of these events are defined.
AP-01-5-06-26 requires that an on-shift analysis be performed by the
STA or shift supervisor. And when necessary, the PSRC reviews and
make recommended actions for a Class III event. The off-shift
analysis is done within ten working days after permission for plant
restart is given by the General or Duty Manager.
This off-site analysis must be done by a senior Reactor Operator, a
certified STA or by persons having equivalent experience and
background knowledge of the plant to perform the analysis.
Subsequent analysis of the Reactor Trip review is performed by
Operational Analysis Section (0AS) in accordance with NPE 701.
f. Criteria for determining the acceptability of authorized restart have
been established.
As discussed in Section b. the safety assessment determines if all
pertinent plant parameters were within Technical Specification
limits, safety related equipment operated as expected and events
occurred in a proper sequence. All restart decisions are made by the
General or Duty Manager.
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g. Criteria for comparing plant information with known or essential
plant behavior have been established.
The trip analysis, AP-01-S-06-26, provides a step by step check list
and based on the analyzer's knceledge, operations are verified to
have occurred as expected. Plant behavior is com
values contained in the Technical Specifications. pared to limitingAny deviat
noted in the report and evaluated.
h. Guidelines are established for preservation of evidence of reactor
trips.
AP-01-S-06-26, paragraph 6.4.12, requires that the original Post-Trip
analysis package be forwarded to central records. AP-01-S-05-1,
Section 6.3, provides for lifetime (of the plant) retention of these
records in the Central Records. Copies of the report which provides
a summar
actions,y,root
verification of on-shift
cause identification analysis,
and summary
observations of corrective
/ recommendations
and the post-trip analysis (without collected data) are routed to the
General Manager, Plant Operations and Maintenance Managers, various
Superintendents NRC Resident Inspector, Operational Analysis Group
andReactorEngIneeringFile
One comment on AP-01-S-06-26 is that it does not specifically call
attention to a Technical Specification requirement (Section E.7.1.a)
to notify the NRC within one hour after determining that a safety
limit has been exceeded. The licensee stated that this was covered
by AP-01-S-06-05, but since the procedure (AP-01-S-06-26)
specifically requires the STA or Shift Suaervisor (SS) to make this
determination, they would consider adding this comment in
AP-01-S-06-26.
Within the areas examined, no violations or deviations were identified.
4. Equipment Classification
The licensee was requested in Section 2.1 of GL 83-28 to confirm that all
components of the Reactor Tria System (RTS) whose function is required to
trip the reactor are ident fied as safety-related on documents,
procedures, and information handling systems ut ed in the plant. These
safcty-related activities also include maintenance work orders and spare
parts replacement. In Section 2.2 of GL 83-28, the licensee was recuested
to describe their program for ensuring that all components of otler
safety-related systems are also identified as safety-related on documents,
procedures, and information handling systems used at the plant.
The licensee's responses to Sections 2.1 and 2.2 of GL 83-28, dated
November 4, 1983, June 28, 1985, and May 14, 1985, provided details of the
current program and the developments of new programs and procedures for
safety-related equipment classification. In these responses, the licensee
made the following statements:
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"In a BWR, the components which contribute to the reactor trip
function ate found in several systems. For example, components in
the Neutron Monitoring System, Main Steam System and Reactor
Protection System may have a role in initiating a reactor tri
Therefore, the GGNS information handling system (equipment list)p.will
be reviewed to ensure that the safety related components in the
various systems which contribute to a reactor trip are properly
classified.
The existing Q-List original 1 developed by the architect-engineer is
presently being maintained Nuclear Plant Engineering and that a
new component quality classi ication listing is under development.
GGNS has in place procedures which control and require the use of the
Q-List, provide for equipment classification for work activities not
addressed by the existing Q-List, and provides for the review of
procedures which may be used in safety related activities. A review
will be undertaken to verify that procedures controlling safety-
related activities related to components which contribute to reactor
trip functions are properly identified.
A Master Equi ment List (MEL) is being developed to provide useful
data for equ pment and components in the plant, including their
safety class fications. Procedures have been developed to help
personnel determine component classifications and whether components
affected by activities are safety-related."
a. The inspector reviewed licensee's responses, appropriate procedures,
maintenance work orders, procurement documents, and interviewed
respNsible personnel to confirm that the licensee's program for
equipment classification was being implemented and was consistent
with the responses to GL 83-28. The licensee's actions concerning
the NRC's Request for Additional Information were reviewed and
discussed with responsible personnel. They anticipate that a
response to the latest NRC letter dated May 11, 1988, will be
submitted in late June 1988.
b. Discussions revealed that the licensee is developing a new
comprehensive computer program which will include information
presently contained in the Master Equipment List (MEL) and Equipment
Index (EI). This new program, "Station Information Management
System" (SIMS), was approved for use at GGNS in July 1987, and in
January 1988, the licensee started entering data into the SIMS. The
licensee expects the SIMS to be in full operation by June 1989, and
will include approximately 40,000 items. Ti 1 inspector was informed
that the SIMS is currently in use at Waterford and Arkansas Nuclear
Plants. The licensee and contractor, re9ansible for the SIMS,
demonstrated how component data is obtained, how data is entered, and
how data is retrieved and utilized by plant maintenance, engineering,
procurement, and operating personnel. Site QA is familiar with the
program and an internal audit group have audited the program. The
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inspector requested information for four components be retrieved from
the SIMS 3rogram to verify whether the components were classified as i
safety re'ated and if other technical information was available.
Printouts for the following components were obtained and reviewed:
Reactor Vessel Low Level Switch (Safety Related)
Control Rod Drive-Hydraulic Control Unit (Safety Related)
Technical Support Room Door (Non-Safety Related)
Reactor Vessel Level Transmitter (Safety Related)
The printouts for the above com)onents contained all types of
information such as: Vendor, mocel, serial number, Q Class, safety
class, P&ID, location in , plant, vendor manual number, design
parameters, Technical Specificatlons, associated plant procedures,
drawings, and other data. When the program is completed and in full
production in June 1989, it will include information concerning
maintenance activities, surveillance activities, design changes,
NPRDS,EQ,andotherplantactivities. In addition, the program will
gene"ate maintenance work orders and preventative maintenance orders
for the components. In order to ensure that data for each component
is complete and accurate, the program provides a Component Package
Input Tracking Form which has individual checkoff blocks to verify
completion of data collection such as: system walkdown, equipment
index MEL data, P&ID check, document research, engineering review,
dataInput,dataedit,andQAaudit.
c. The following procedures associated with the classification and
cor, trol of safety-related equipment were reviewed:
(1) Procedure 01-5-07-14, revision 4, dated October 30, 1987,
established the methods and controls for maintaining the
Equipment Index (EI). This index lists technical data on
individual safety-related components. The data includes vendor,
drawing numbers, manual number, QA type, and other pertinent
information required to ensure proper maintenance and
calibration of equipment. Information from this EI is being
transmitted to the SIMS program for use in their data base.
(2) Procedure 01-5-02-4, revision 1, dated August 29, 1985,
described the method for identifying safety-related items to
which the licensee's QA program requirements ap)1y. The
procedure also describes how to determine if i; ems not
previously classified as safety-related should be reclassified.
The responsibility for maintaining the Q-List has been assigned
to the Nuclear Plant Engineering Section. This section also
evaluates and classifies components, activities, structures, and
systems as requested by other personnel.
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Any person using a procedure or performing an activity is
responsible to ensure that the activity, component or procedure
has'been classified. Procedures require that "safety-related"
be on the cover sheet of any activity associated with
safety-related equipment, systems and structures. If the
document has not been classified the user must either classify
or refer the document to knowledgeable personnel for
classification.
If the users Section Superintendent or Supervisor cannot
determine the classification, a Material Nonconformance Report
is 3repared to ensure that the equipment or activity is not
uti:ized until properly classified. The Material Nonconformance
Report is also the official mechanism to request assistance in
evaluation from the Nuclear Plant Engineering Section.
Attachment 1 to the procedure provides a criteria for use in
determining the safety classification and contains adequate
guidelines for classifying items. The guideline also requires
that directives, procedures, etc. , that establish means ~ or
methods of performing activities associated with safety-related
items be stamped "Safety Related" on the cover sheet.
Examination of several directives, procedures, maintenance work
orders, and procurement documents revealed that they were
stamped "Safety Related."
(3) Standard No. SERI-JS-08, Instrument Q-List, dated December 31,
1987, provides information for determining 10 CFR 50,
Appendix B, Quality Assurance Requirements to the design,
procurement, and installation of instrumentation at GGNS. This
standard specifies the classification of all instrumentation and
assigns responsibilities to assure that new and replacement
instrumentation meets Seismic Category I, ASME Section III, and
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Class IE Electrical requirements.
(4) Procedure 01-502, revision 5, dated January 12, 1987, provides
the requirements for maintaining the Q-List and distribution of
the controlled copies. Revisions to the Q-List are made after
, being reviewed and approved by the Quality Engineer, Quality
Assurance, Manager of Nuclear Services, Site Director, and the
Director of Nuclear Plant Engineering,
d. The inspector reviewed the Q-List, Controlled Copy No.034,
revision No.1, dated January 30, 1987, to determine if the Q-List
was controlled, if the cover sheet showed who prepared, who reviewed,
and who approved the current revision. The Q-List was prepared by
engineering, reviewed by Quality Assurance, Nuclear Services and
Fuels, and the Site Director. The Q-List was approved by the
Director of Nuclear Plant Engineering. The Q-List delineated the
systems, structures, and components that are classified as nuclear
safety-related. It also includes services that are safety-related.
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The 0-List also contains a list of Non-Safety Related Items for which
the QA program is applicable
The Q-List did not depict individual components associated with the
4 reactor protection system (RPS), however, the entire RPS was
classified as safety-related. Upon questioning, the licensee
produced other documents which had each component in the RPS listed
as safety-related. This list was broken down by mechanical
components, instrumention, and electr! cal components. Discussions
also revealed that the components in the RPS had been entered into
the SIMS program data sheets.
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e. Activities associated with the implementation of safety-related
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classification procedures were examined as listed below.
(1) Four purchase order package were reviewed to verify that the
components had been classified as safety-related. The review
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revealed that the documents contained detailed information to
ensure that the component design, fabrication, testing, and
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shipment met the licensee's procedural requirements. Documents
reviewed were:
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Purchase Order No. M98585 dated March 24, 1987
Purchase Ordei No. MP707810 dated September 19, 1985
Purchase Order No. MP808320 dated March 25, 1988 '
Purchase Order No. MP730615 dated November 10, 1987
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Procedure 01-S-09-1 depicts how procurement documents are
prepared and reviewed to ensure proper safety clas ifications. -
(2) Several Maintenance Work Orders (MW0s) were reviewed to verify
that the components had been classified as safety-related and
that activities had been completed as specified in procedures.
MWO Nos. E72980, E72984, E72982, and E72986 associated with the
RPS Relay Tests were reviewed. These MW0s were initiated to
determine if GGNS HFA Relays were binding as reported in GE SAL
No. 188.1, NRC IEN 82-13, and 88-80-03. The licensee performed
the tests as specified in GE SAL 188.1, thus closing out IFI
50-416/87-14-02. Documents associated with the above tests were
classified as safety-related.
(3) Other MW0s reviewed were Nos. E68391. E75358, E81676, E822135,
E822448, M82336, E82168, M75730, and 173908. The inspector
found that the safety-related block on MWO No. E81676 had not
been checked. The licensee immediately inve:tigated and
presented the inspector with an acceptable explanation. This
was an isolated case as all other MW0s and associated data
sheets were classified as safety-related.
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(4) The surveillance procedures listed below were also examined to
verify that they had been classified as safety related:
06-10-1821-R-0038, RV Steam Come Pressure Electronics Time
Response
06-10-1821-R-0044, RV Water Level Calibration
06-IC-1C11-M-0003, Scram Discharge High Water Level Float
Switch Calibration
06-IC-1833-R-001, Reactor Recirc System FCV Functional Test
06-1C-1C51-0-001, APRM Single Loop /Two Locu Operation
Setpoint Adjustment
A review of several vendor manuals revealed that they were not stamped as
"Safety Related." Discussion with licensee representatives revealed that
the manuals were not actually used to perform maintenance and test
activities. Licensee stated that information in the manuals was used to
prepare plant procedures and instructions which were classified as "Safety
Related." A review of plant procedures verified the licensee's
statements.
Based on interviews and the review of procedures, maintenance work
requests, procurement documents, directives, Q-Lists, and computer
programs discussed above, the inspector confirmed that the licensee has
programs and is implementing them to classify equipment and control safety
related activities.
Within the area examined, no violations or deviations were identified.
5. Vendor Interface and Manual Control
The inspector reviewed the licensee's responses to GL 83-28, reviewed
procedures, and examined the implementation of their program associated
with vendor interface and vendor manual control. Their responses
described the following program.
. Licensee management's response transmitted in their letter AECM-
84/0508, dated November 19, 1984, stated that they maintain an
ongoing interface with General Electric Company (GE) and are a member
of the Boiling Water Reactor Owners Group (BWROG). This ongoing
interface provides them with information through GE's Service
Information Letters (SILs) concerning technical items or conditions
that may be applicable to Grand Gulf. Additionally, in accordance
with recommendations of the Vendor Equipment Technical Information
program (VETIP) they are active participants in the INP0 managed
Nuclear Plant Reliability Data System (NPRDS) and Significant Event
Evaluation and Information Network (SEE-IN) program. They concluded
by stating that a program had been established and was being
implemented for vendor technical manuals.
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Licensce's response transmitted in their letter AECM-85/0157, dated
May 14,1985, stated that in June of 1984, letters were sent to
vendors of safety. related equipment in an attempt to establish a
vendor interface. Thirty-three of the vendors contacted res>onded.
Of those responding, approximately 75 per cent transmitted acditional
technical information. Licensee management determined that the
manpower requirements to continue a program of periodically
contacting vendors was prohibitive compared to the knowledge gained.
They further stated that an avaluation had been performed to assure
proper- division of responsibility between System Energy Resources,
Inc. (SERI) and vendors who provide maintenance or testing service.
They concluded by stating that the VETIP as defined in the March 1984
NUTAC document and the SERI programs described in this letter and
letter AECM-84/0508 were valid responses to GL 83-28, Item 2.2.2.
Licensee's response transmitted in their letter AECM-88/0062, dated
April 1,1988, describes the process for receipt and acknowledgement
of GE Service Information Letters (SIls). They stated that all GE
SIls are sent to SERI addressed to the GGNS Site Director with an
attached receipt acknowledgement form. The form is completed by SERI
and returned to GE. In addition, GE sends SERI, approximately every
six months, a listing of all SIls~ issued to date. This list is used
by SERI to verify receipt of all SILs. Those that have not been '
received are requested from GE.
Procedure No. 0701, specifies requirements for evaluations of onsite
and offsite documents by the Operational Analysis Section (OAS). The
)rocess provides for an initial screening by an Investigative
Engineer to determine if the document is applicable to GGNS or
requires evaluations by reviewing P& ids, system descriptions, the
FSAR, or other pertinent plant documentation. Documents that have
been evaluated and determined not to be ap?1icable to GGNS are closed
out. For situations involving documents t1at are applicable to GGNS,
an evaluation is performed to assess possible technical and safety
concerns. The evaluation is required to be of sufficient depth to
determine root cause, safety significance, generic implications, and
necessery corrective action. Additionally, for component failures
that are applicable to GGNS, a review of NPRDS data base is
performed.
The implementing procedure for the evaluations performed by 0AS is
,
EDP-023. Operating experience information is received from several
sources. Paragraph 7.0 of this procedure saecifies the scope and
nature of the operating experience information and the source from
which they were derived. Typical examples are INP0 Significant
Operating Experience Reports and Significant Eve t Reports (SERs)
received through the mail or accessed via the Ntelear Network
respectively. The inspector verified that the scopr of the operating
experience information described was consistent wtth the program
described in licensee's letters.
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Discussions with licensee management revealed that a vendor interface
had been established with the NSSS vendor by means of contracts.
Procedure EDP-023, paragraph 7.2.9 specified requirements for review
of SILs and initiating appropriate corrective action. The inspector
determined that the res)onsibility for evaluation of SILs had not
been adequately definec. Additionally, the evaluation of SIls
performed by 0AS was after the review and disposition of the SILs by
other plant or engineering personnel. This evaluation was in most
cases performed long after receipt of the SIL by the licensee. The
inspector identified a lack of (1) accountability for review of SIls
and (2) responsioility for initiating corrective action. Paragraph
7. 6 specifies a screening process for SILs. Additionally,
programmatic deficiencies in the absence of specific guidance for
completion of the screening process were identified. The licensee
informed the inspector that a Quality Deficiency Report, addressing
these deficiencies, had been written against this procedure. The
inspector reviewed licensee's corrective actions taken to revise
procedure EDP-023. Pursuant to this review and discussions with
licensee personnel, the program requirement described in licensee's
letter AECM-88/0062 was incorporated within the corrective action
plan. Licensee management has scheduled conpletion of these
corrective actions and implementation of revised procedure EDP-023
for June 15, 1988.
LER No. 86-026-00, Inadvertent Centrol Rnd Withdrawal was sent to
the NRC by the licensee on August 22, 1986. Informationconcerning
this possible event had previously been provided to the industry by
GE in SIL 292, dated July 1979. This SIL was never received by the
licensee, therefore, the recommended corrective action pr(vided by
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the NSSS vendor was never im)lemented. Because of this experience,
licensee management initiatec and completed an ef fort to review for
a plicability to GGNS all SIls that had been provided to the industry
b GE. To assess the adequacy of this effort, the inspector selected
t e following SIls for review.
SIL No. 452, Feedwater Flow Element Inspection and Accuracy,
dated June 8, 1987
SIL No. 292, Inadvertent Control Rod Withdrawal, dated July 1979
SIL No. 310, Stuck CR0 Collect, dated October 1979
SIL No.131, Containment Isolation Logic Channe, dated March 3,
1975
SIL No. 353, HPCI Turbine Mechanical Overspeed Trip, dated
February 18, 1981
SIL No.128 RISRI, Pravo* ,ve Maintenance for CRD Scram Pilot
Valves
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SIL No.128R152, ASCO HVA-90-405 Valves and Chemical Adhesive
Thread Lockers, dated March 2, 1984.
No deficiencies were identified in the evaluation and disposition of
the above SIls.
Licensee management has not established a formal vendor interface
with vendors of other safety related items via contractural
arrangement. However, a program has been established and is being
implemented on site for receipt, evaluation, and disposition of all
vendor informaticn. Vendor interface requirements for engineering
procurement specifications are defiaed in procedure NPEAP No.01-306,
Attachments 5. The review and evaluation of vendor information
provided during the design-engineering precass is specified in
procedure NPEAP No.01-302. Other vendor interfaces are established
through owners group. licensee
management has been As a mecber
involved of the TOI
in the Design Owners
Review / QualGroup,ity
Revalidation (DRQR) prograin for the emergency diesel generators.
Recommended changes in maintenance and surveillance of the diesel-
generator made by this group are being coordinated with the
diesel generator vendor prior to implementation on site.
Procedure Nos. E0P-008, AP-703, and 01-802 were reviewed to assess
the adequacy of the administrative controls established for receipt,
evaluation, and disposition of operating experience and/or vendor
information described in licensee s correspondence.
No deficiencies were identified in the above procedures during this
review.
SERI 03erating Manual, Policy No. 8.801, Vendor Manual Control,
specif:es requirements for the receipt, review, distr'!aution,
maintenance, revision, and use of vendor manuals. The licensee is
areseatly in the process of an Operations and Mdntenance (0&M)
Aanual Update program intended to revise the 0&M manuals for systems
and components associated with the reactor tr'p system (RTS). SERI
interoffice memorandum from R. J. Rogers III to J. E. Cross, Subject:
"G. E. Update of RTS Vendor Manuals , dat;d March 10, 1988, states
that GGNS reactor trip system operatin
be updated as part of the com)any's ponse rer,g snd maintenance
to Generic manuals will
Letter 83-28.
The 0&M manual upgrade will se perfo<med by G. E. under contract to
SERI and the scope of the initial affort includes the following
systein manuals.
Power Range Neutron Monitoring System
Startup Range Neutron Monitoring System
Traversing In-Core Probe Calibration System
Proc.ss Radiation Monitoring System
Nuclear Boiler Process Instrumentation
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Upon completion of the RTS manuals, 327 GEK's and 527 other safety
related manuals will require revalidation.
Pursuant to the above effort, changes are being made to Procedure
Number 01-S-05-4, to accomplish the following:
Implements new SERI policy on vendor manuals.
Divides vendor manuals into categories and assigns responsibility for
their review as indicated below:
Manual Category Responsibility
Reactnr Tri) Systems, I NPE & Plant Staff
Ofesels, Lu)e Oil, Maintenance
Selected Turbine Manuals,
and all Safety Related
Balance of Plant II NPE & Plant Staff
Non-safety blated Maintenance
Nonpermanent IV No review required
Plant Equipment
(Hardware / Software) V Plant Staff
Computer
Computer Related Engineering / Reactor
Engineering
Requires Document Control to track vendor manuals through the review
process.
Requires vendor manuals to be approved and issued prior to sending
them to procedure originators for coordination.
Requires procedures requiring revision because of vendor manual
changes to be revised within 60 days.
Requires Document Control to send vendor manuals to the Materials
Section for review for spare parts information and in-storage
instructions.
The inspector discussed the 0&M manual upgrade program with licensee
t management and requested information concerning the implementation date of
revised Procedure 01-5-05-4. He was informed that the procedure will be
issued for use by July 1,1988. Departmental level procedures required
i for implementation of the new SERI policy on vendor manuals are also
! scheduled for development. Review of the Vendor Manual Task Force meeting
l summary dated May 20, 1988, shows that this effort is on schedule.
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The inspector selected the following components classified as safety ,
related and associated with the RTS to verify that appropriate vendor
furnished information had been incorporated into test procedures,
maintenance procedures, or the Technical Specification.
Component Vendor Manual
Reactor Modes Switch G.E. Instruction GEH-20380,
G.E. Control and Transfer
Switch Type SBN
Reactor Vessel High GEK-73698, Volume, IV,
Water Level (Tab 4)
(821-LT-N080A-D)
Scram Discharge Gould Instruction Manual
Volume High Water PD/PDH3018 Series
Level Trip
(IC11-LT-N012A&B)
Scram Discharge Magnetrol International
Volume High Water Bulletin 46-612, Effective
Level Trip date April 1976.
(IC-11-LS-N013A-D)
Comparison of the requirements delineated in surveillance test procedures, -
maintenance procedures and the Technical Specification revealed no
deficiency.
Within this area, no violations or deviations were identified.
6. Surveillance / Post-Maintenance Testing
a. Surveillance Testing
The inspector observed the performance of Surveillance Test Procedure
Nos. 06-lC-1E31-M-0023, 06-IC-lC71-M-001, and OC-IC-M-1103. A
tabletop walkthrough of Loop Calibration Instructions 07-5-53-C11-21
and 07-5-53-C11-22 was also performed. The procedures, in general,
were very well written and easy to follow. One area of concern
identified during a previous inspection was noted. This item '
(IFI-87-26-01) deals with verifying continuity by use of digital
voltmeter (DVM). The procedure calls for a reading of "approximately
0 VAC", yet the reading was approximately 20 VAC and decreasing. The
licensee plans to provide the technicians training in the short term
and implement long term training in the annual training program.
This training should ensure that an understanding of "approximately"
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During the review. of 01-S-07-1, it was noted that five Test Change
Notices (TCNs) were outstanding for this procedure. Procedure
01-S-02-2 requires a revision to the procedure be issued within sixty
days after the third TCN is issued, unless an 3xtension is granted by
the Plant General Manager. The licensee identified that it exceeded
the sixty day limit and issued QDR-140-88. The ins
a minor discrepancy in that TCN 45 (issued 2/26/88) pectorTCN
referenced identified
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(issued 2/29/88). The licensee -then issued QDR-163-88 identifying
the fact that -a TCN . referenced a TCN that was ~not issued at 'that-
time.
b. Post Maintenance Testing
The inspector reviewed the licensee's -maintenance and
ost-maintenance testing records. Seven Maintenance Work Orders
p(W0s) were randomly selected for review. The packages were reviewed
for compliance with 01-S-07-2, for completeness, and for ensuring
data was within the allowed ranges. The W0s audited were as
follows:
W O No. Description
1. I66250 RHR Heat Exchanger 2A/2B Outlet Temp
2. E77176 CTMT Spray Time Delay Relay Channel 6
3. 175953 Turb Control Valve Press Comp Point
4. I64301 Interface Valve Pressure
5. M66292 Outboard Isolation Valve
6. ESO426 RCIC Isolation Valve
7. M75972 FPCC to Filter Demineralizer
No discrepancies were noted on the W0s.
Within this area, no violations or deviations were identified.
7. Surveillance Testing of the Diverse Reactor Trip Functions of the Reactor
Trip System
The licensee performs on-line testing of the Reactor Protection System
during periodic functional channel tests, channel checks, and scram timing
of control rods. On-line testing of the backup scram solenoid valves is
not possible without scramming the plant; however, testing of the backup
scram valves will be performed durin,g each refueling outage. The
inspector confirmed that the licensee s Loop Calibration Instructions
07-5-53-C11-21 and 07-5-53-C11-22 will independently test each backup
scram solenoid valve once every 18 months while the plant is shutdown.
Within this area, no violations or deviations were identified.
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8. Inspection Follow-up Items
a. -(Closed) IE Bulletin 88-01, Defects in Westinghouse Circuit Breakers.
The licensee responded to IEB 88-01 by letter dated March 22, 1988,
stating .that the Westinghouse DS series breakers identified in the
bulletin were not used at GGNS. Based on this submittal, this item
is closed.
b. (Closed) IE Bulletin 88-03, Inadequate Latch Engagement in HFA TYPE
Latching Relays Manufactured by General Electric (GE) Company.
As stated in licensee letter AECM-88/0094 dated March 24, 1988, SERI
investigated the use of latching-type HFA relays and had determined
that these type of relays are not used in Class 1E application at
GGNS. Three of these type relays were found in non Class 1E
applications. One of these three has been inspected per GE Letter
No.190.1 and the other two will be inspected during the next
refueling. To ensure that this type of GE relay is not installed
without inspection, a caution statement has been added to the
Material Management Information System (MMIS). Based on the
licensee's letter, this item is closed.
c. (Closed) Inspector Follow-up Item (IFI), 416/87-14-02, Inspection
and/or Replacement of GE HFA Relays Referenced in SAL 188.1.
GE Service Advisory Letter (SAL) No.188.1 dated November 14, 1986,
reported binding problems with type HFA relays. The binding concern
only applies to HFA relays manufactured prior to October 24, 1986.
Tha licensee investigated the problem and found eight potential
effected relays installed in the Reactor Protection System. These
eight relays were checked in accordance with GE instructions during
refueling outage 2 under MW0s E72979 thru E72986 and were found to be
free of mechanical binding. Licensee's Document Review Summary Sheet
No. 8-035, dated April 4,1988, summarized the licensee's investi-
gation. Based on the above, this item is closed.
d. (Closed) IE Information Notice (IEN) 88-14, Potential Problems with
Electrical Relays
, This notice provided licensees with potential 3roblems involving HFA,
PVD 218, FVD 210, and HGA relays manufactured ay General Electric Co.
Grand Gulf Nuclear Station (GGNS) has been evaluating the potential
problems with these type relays since 1932. GGNS's Operational
Analysis Section screens and evaluates NRC Bulletins, NRC Information
Notices, GE's Service Advice Letters (SALs) and Service Information
Letters (SIls), and INP0 Network documents for applicability at GGNS.
This information is being computerized for easy retrievability. Upon
request by the inspector, the licensee retrieved Document Review
Summary Sheets which gave the background and summary of finding /
action taken on these relays. Results of licensee's actions revealed
that they had closed out the GE HFA relay problems in 1982 per PMI
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82/1152, SER 6-82 dated April 15, 1982. Action o- PVD 21 relays were
closed out in documents dated April 26, 1983, Nr< ember 17, 1983, and
March 8, 1986. GE Century Series Relays were c hsed out in documents
dated December 23, 1982, and March 4,1986. Li:ensee action on GE
SAL No.188.1 concerning HFA relays was completed and closed out by
Document Review Summary Sheet No. B-035.
Licensee action en HGA II and HGA III relays was documented in PMI
83/12746 1983. GGNS's HGA II and I:1 re'ays all
have coilsdated
whichNovember
are normal 15,ly energized; therefore, do not fall into
the 2 millisecond contact opening criteria. GE letter MP GE-83/20
dated September 20, 1983, also stated that the 2 millisecond criteria
was not applicable at GGNS. Based on review of documents and
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interviews with cognizant licensee personnel,-IEN 88-14 is considered
closed.
e. (Closed) NRC Information Notice No. 88-19, Questionable Certification
of Class 1E Components.
,
Licensee management performed an IE Information Notice Review in
response to the above. This review dated May 13, 1988, stated that
the information notice is not applicable to GGNS because Planned
Maintenance System (PMS) was never approved as a quality supalier and
was never added to the Qualified Supplier List (QSL). The inspector
asked if the licensee had verified that PMS was not a subcontractor
to a qualified supplier on the QSL. In response to this question,
-licensee manacement verified that Telemecanique war ~ the only vendor
who had supp1'ed Class IE fuses for use at GGNS. They further added
that Telemecanique had never qualified PMS as a supplier and they had
never procured Class IE fuses from PMS for use at GGNS. Based on the
above licensee actions, this item is closed,
f. (Closed) Deviation 87-27-01, Offsite Power System Monitoring and
Surveillance
The licensee had a UFSAR commitment to perform functional checks of
relay and control equipment for the offsite power system on a two
year or less interval. It was identified that the last completed
functional check on the protection relay system for the 500 KV system
was performed in June 1983.
The licentee has reviewed its License Commitment Tracking (LCTS)
database to assure that SERI has been assigned as the responsible
organization, where appropriate, and revised the database where
necessary. Chapter 3 of the UFSAR was re-reviewed to ensure all
commitments have been pre;,erly entered into LCTS. Procedure
01-5-15-8 was issued to govern the tracking of new commitments.
Procedure NLAP-2.7 is in the review and approval chain and will
govern the entire database, including updating and changing existing
commitments, assigning responsibilities for meeting the commitment,
and assigning intermediate due dates to ensure the commitment date
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is not exceeded. Provisions are made to keep management informed
when problems arise in meeting the due dates so that action may be -
taken at the appropriate level.
Based on the corrective action taken and the response in letter dated
February 10, 1988, this item is closed. l
9. Procedures Reviewed
The following procedures were used and/or reviewed for this GL 83-28
inspection:
a.
b.
ONEP 05-1-02-I-1, Reactor Scram
IOP 03-01-4, Scram Recovery
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c. AP 01-5-06-26, Post-Trip Analysis
d. I0I 03-1-01-1, Cold Shutdown to Generator Carrying Minimum Load '
e. AP 01-5-02-2, Control and Distribution of the GGNS Operations Manual
f. AP 01-S-05-1, Nuclear Records Procedure
g. NPE AP 710, OAS Document Evaluation
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h. AP 01-S-04-15, Required Reading Program
i. 10-S-01, Activation of the Emergency Plan i
j. AP 01-S-06-05, Incident Reports / Reportable Events !
k. 06-IC-IE31-M-0023, Rev. 25, TCN 17, RCIC/RHR and RCIC Steam Line Hi
Flow / Flow (RCIC Isol) :
FunctionalTest !
1. 06-IC-IC71-M-0001, Rev. 25, Drywell High Pressure Functional Test
(RCP/PCIS) i
m. 06-IC-IC71-M1003, Rev. 23, Turbine Control Valve Fast Closure '
(RPS/EOC RPT) Functional Test
n. 07-5-53-CII-21, Rev. O, Backup Scram Valve A Functional Test ;
o. 07-5-53-CII-22, Rev. O Backup Scram Valve B Functional Test
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p. NLAP-2.7, Rev. O, Licen,see Commitment Tracking (LCTS)
q. 01-S-15-8, Rev. O, Licensee Commitment Tracking (LCTS) for Plant
Staff
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r. EDP-023, OAS Review Handling and Dis 30sition of Operating Experience i
s. NPE AP-01-306, Engineering Document Requirements
t. EDP-008, Nuclear Plant Reliability Data System Data Submission and
Retrieval
u. AP 703, See-in-Utilization
v. A?01-802, Processing and Evaluation of 10 CFR 21 Potential Defects
w. NPE AP 01-302, Review of Supplier Documents
x. 01-S-07-1, Control of Work on Plant Eculpment and Facility t
g y, 01-5-07-14, Control and Use of the GGh5 Equipment Index
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z. 01-S-09-1, Procurement of Materials and Services .
aa. 01-S-02-4, Rev.1 Adm. Proc. Determination of Safety / Quality !
Classifications, Safety Related !
bb.01-502, Q-List Control
ec. Standard No. SERI-JS-08, Instrument Q-List !
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10. Exit Interview
The inspection scope and results were summarized on May 24, 1988, with
those persons indicated in paragraph 1. The inspectors described the
areas inspected and discussed in detail the inspection results listed in
the above paragraphs. Although reviewed during this inspection
proprietary information is not contained in this report. Dissenting
comments were not received from the licensee.