ML19322A920
| ML19322A920 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/17/1973 |
| From: | Murphy C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Thornburg H US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML19322A903 | List: |
| References | |
| NUDOCS 7911270685 | |
| Download: ML19322A920 (1) | |
See also: IR 05000269/1973007
Text
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, UNITED STATES
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ATOMIC ~ ENERGY COMMISSION
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DIRECTORATE OF REGUIATCRY OPEPATICNS
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REGION ll - $U4T E 818
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AT L, ANT A, GEoRos A 30303
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H. D. Thornburg, Chief, Field Support and Enforcement Branch
Directorate of Regulatory Operations, Headquarters
DUKE POWER COMPANY (OCONIE 1), LICENSE NO. DPR'38, DOCKET No.
50-269 - ENFORCEMENT CORRESPCNDENCE
As dir, cussed during a telecon with R. C. Paulus and in
accordance with Manual Chapter 0800, Section 0860.08 f, enclosed
for Headquarters' review and dispatch is the inspection report and
enforce =ent correspendence to the licensee as a result of a site
inspection on June 20-21, 1973, and a manage =ent meeting between
the Region II Director and DPC corporate management on July 19, 1973
7-
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C. E. Murp. , .ief
RO:II:CEM
Facilities Test and Startup Branch
Enclosures:
1.
Draft ltr to DPC
2.
RO Rpt. No. 50-269/73-7
cc v/ encl. 2:
RO:HQ (5)
-u =3DR Central Files
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Regulatory Standards (3)
Directorate of Licensing (13)
cc encl. 2 only:
'PDR
' Local PDR
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- NSIC
'DTIE,'OR
' State
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"To be lispatched at a later date.
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DETAILS
Prepared by:
Aa44
cc)d L
F. Jape, Reactor d'nspdctor
Date
Facilities Test and Startup
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Branch
Pfu)~lh
~7- M- 73
R. F. Warnick, Reactor Inspector
Date
Facilities Test and Startup
Branch
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C. E. Murphy,,Ch,ief
Dat'e
Facilities Test and Startup
Branch
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Dates of Inspection: June 20-21, 1973
Reviewed by:,
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7.7
C. E. Murphy, Chief
/ Date
Facilities Test and Startup
Branch
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1.
Individuals Contacted
Duke Power Company (DPC)
J. E. Smith - Plant Superintendent
J. W. Hampton - Assistant Plant Superintendent
O. S. Bradham - Instrument and Control Engineer
D. M. Thompson - Assistant Maintenance Supervisor
J. W. Sigman - Maintenance Supervisor
D. Laning - Assistant Maintenance Supervisor -
L. E. Summerlin - Staff Engineer
J. W. Cox - Assistant Plant Engineer
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2.
General
The purpose of this inspection was to examine the management
system for handling plant problems and plant modifications,
the use of miscellaneous test procedures, and the functioning
of the Station Review Cot =nittee (SRC) .
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3.
Review of Administrative Controls for Processing Desien Changes
The administrative controls for processing station design
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changes were reviewed by the inspector. The applicable
controls in effect at the station include the following:
a.
Administrative Procedure No.10, " Station Modifications,"
dated June 14, 1973, waich presents six steps to follow
when a change is to be implemented. The steps are summarized
below:
Step 1 requires the change to be submitted as a station
problem report to the SRC for review.
Step 2 requires the' proposed change to be reviewed by
the SRC.
Step 3 requires the station superintendent to review the
change and determine if installation can proceed with his
approval or if additional approval is required from the
Assistant Vice President, Steam Production and if review
by the Nuclear Safety Review Committee (NSRC) is required.
Step 4 requires approval by the organization or an
equivalent organization which was responsible for
the original design.
Step 5 requires the use of approved checklists, in-
structions, procedures, and drawings to perform the
work.
Step 6 allows the work to be completed on verbal approval
provided the change is documented within five days.
b.
Technical Specification 6.1.2.1.d (5) , " Station Review
Committee, Responsibilities," requires the SRC to review
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proposed changes or modifications to the station design.
c.
Technical Specification 6.1.2.2.a, '.' Nuclear Safety Review
Committee," states, in part , that the NSRC will ". . . review
important proposed plant changes and tests . . . ."
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d.
Paragraphs 6 and 7 of Technical Specification 6.4, " Station
Operating Procedures", describe the review and approval require-
munts for major changes to the facility.
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Criterion V of Appendix B to 10 CFR 50 and paragraph 50.59(b)
of 10 CFR 50 were used by the inspector as the basis for this
review. Criterion V requires that activities that affect quality
be accomplished in accordance with appropriate instructions, pro-
cedures, or drawings . Paragraph 50.59(b) requires a written
safety evaluation which provides the basis for the determination
that the change does not involve an unreviewed safety question.
The period of plant operation reviewed was between February 6,1973,
to June 19, 1973. During this period, the records for the following
plant changes were audited:
a.
Control Rod Drive Gas Vent Piping
This change was made as shown on Engineering Change Order No. 236,
dated March 16, 1972. The SRC concurred with this change on
March 29, 1973, over a year af ter the change order. No other
documentation was available at the site regarding this change.
The change was made contrary to Steps 1 and 3 of Administrative
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Procedure No.10 (AP-10) as relates to reviews and approvals.
The failure to follow approved instructions appears to be in
violation of the requirements of Criterion V of Appendix B to
10 CFR 50.
b.
RCP Oil Drain System
This change was made as shown on Engineering Change Order
No. 343, dated March 16, 1973, and Steam Production Depart-
ment Notice of Design Chant,e No. 47, dated March 24, 1973.
The SRC reviewed the proposed change on March 30,1973, and
recommended that the oil slinger-catcher, designed by
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Vestinghouse, be installed on all reactor coolant pumps. The
committee also recocmended that censideration be given to
installation of a large backup overflow pipe on the upper
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bearing reservoirs to deflect oil from the reactor coolant
piping in case of an oil cooler rupture.
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There is no documentation regarding action on the SRC's
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recommendation nor is there any followup in the minutes
of the SRC since March 30, 1973.
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This change was made contrary to Steps 1 and 3 of AP-10.
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Failure to follow instructions appears to be in violation
of the requirements of Criterion V of Appendix B to 10 CFR 50.
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The safety evaluation required by paragraph 50,59(b) of
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10 CFR 50 apparently was not completed. Failure to perform
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the safety evaluation appears to be in violation of paragraph
50.59(b) of 10 CFR 50.
Turbine Bypass Control Modification
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c.
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This change was made as shown on Engineering Change Order 243,
dated May 8, 1973. The change was reviewed by the SRC with
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final approval to be granted after completion of the power
escalation tests.
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This change was made contrary to Steps 1 and 3 of AP-10.
Failure
to follow instructions appears to be in violation of the require-
ments of Criterion V of Appendix B to 10 CFR 50.
The safety evaluation required by paragraph 50.59(b) of 10 CFR 50
apparently was not completed. Failure to perform the safety
evaluation appears to be in violation of paragraph 50.59(b) of
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10 CFR 50.
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Feedwater Flow-Turbine Trih!
d.
This change was rude as shown on an engineering sketch dated
June 12, 1973. SRC reviewed this proposed change on June 12,
1973, and recommended that it be made.
The change was made contrary to Steps 1 and 3 of AP-10.
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The failure to follow approved instructions appears to be in
violation of the requirements of Criterion V of Appendix B to
10 CFR 50.
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The safety evaluation required by paragraph 50.59(b) of 10 CFR 50
apparently was not completed. Failure to perform the safety
evaluatien appears to be in violation of paragraph 50.59(b)
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of 10 CFR 50.
Electrical Auxiliary Transfer 1!
e.
This change was made as shown on'an engineering sketch dated
June 12, 1973. SRC reviewed this proposed change on June 12,
1973, and recommended that it be made.
Failure
This change was made contrary to Steps 1 and 3 of AP-10.
to follow instructions appears to be in violation of the require-
ments of Criterion V of Appendix B to 10 CFR 50.
The safety evaluation required by paragraph 50.59(b) of 10 CFR 50
apparently was not completed. Failure to perform the safety
evaluation appears to be in violation of paragraph 50.59(b) of
10 CFR 50.
f.
CRD Motor Fault Time Delay
This change was made as shown on Steam Production Department
Notice of Design Change No. 59, dated June 19, 1973. SRC
reviewed this proposed change on June 19, 1973, and recommended
that it be made, stating thag it was an interim fix until the
auxiliaries transfer probles- is resolved. DPC design
engineering also acproved this change on Jane 18, 1973. A
procedure with a e ecklist was prepared and used to install
this change. The installation was audited by the station QC
personnel.
The change was cade contrary to Steps 1 and 3 of AP-10.
Failure
.to folicw instructions appears to be in violation of the require-
ments of Criterion V of Appendix B to 10 CFR 50.
The safety evaluation required by paragraph 50.59(b) of 10 CFR 50 '
apparently was not completed. Failure to perform the safety
evaluation appears to be in violations of paragraph 50.59(b)
of 10 CFR 50.
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1/ See RO Inspection Report No. 50-269/73-6, Details, paragraph 2.
2_/ See Design Change, " Electrical Auxiliary Transfer," above.
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The results of the RO inspector's audit indicate a general failure
to follow established procedures in accomplishing plant changes
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that affect quality. The handling of design changes is not con-
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sistent, and it appears to vary depending on who initiates the
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change and the priority of the change.
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All changes initiated within the Steam Production Department
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are not handled identically. For example, the instrument and
control section numbers design changes and uses a " Notice of
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Design Change" form to document the change. The plant maintenance
section of the Steam Production Department has no formal method for
handling design changes.
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In all cases, the inspector found that the failure to follow
instructions appears to be in viola' tion of Criterion V, " Instructions,
Procedures, and Drawings," of Appendix B to 10 CFR 50. When
this finding was discussed with the licensee's representative,
he stated that a more formal method for obtaining the necessary
reviews and approvals will be developed, and the station
instructions will be revised to cover this problem area.
In addition,10 CFR 50.59(b) requires that the licensee perform
a safety evaluation and maintain written records of the safety
evaluation which provides the bases for the determination that
the change in facility or procedures does not involve an un-
reviewed safety question.
Contrary. to this requirement, written safety evaluations were
not performed for the modifications to the safety related equip-
ment reviewed by the inspector.
4.
Review of SRC
The inspector examined the records of the SRC and compared the
committee's performance with the require =ents of Technical Specifications 6.1.2.1 and 6.2.
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The minutes of the SRC meetings held since the operating license
was issued (February 6,1973, through June 19, 1973) were examined
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by the inspector. The results are described in the following
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paragraphs:
a.
Membership of the SRC
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The SRC is composed of the assistant plant superintendent,
the operating engineer, the technical support engineer, and
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two other members of the station supervisory staff appoiated
by the plant superintendent. The maintenance supervisor and
the health physics supervisor are the two appointed members.
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The assistant plant superintendent is the SRC chairman, but
when he has been absent, an acting chairman has been appointed
by the superintendent.
This meets Technical Specification 6.1.2.1.a., which specifies
the membership requirements of the SRC.
b.
Meeting Frequency of the SRC
The SRC is required to meet at least once each month and, as
required, on call by the chairman. The committee met more
frequently than required. The inspector reviewed minutes
for 58 meetings which were held between the date the operating
license was issued, February 6,1973, and June 19, 1973.
The frequency of the meetings complies with the requirements
of Technical Specification 6.1.2.1.b.
c.
Attendance at SRC Meetings
During March, all members of the SRC attended 13 out of the 15
meetings . In April, all SRC members attended 6 out of 15
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meetings . In thy, the full committee attended 2 out of 10
meetings, and only 1 out of 11 meetings so far in June. One
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committee member attended only 36 out of 58 meetings. The
change in the record of attendance is significant but does
not violate any regulatory requirement.
Technical Specification 6.1.2.1.c states that the SRC meetings
shall be attended by a quorum made up of a chairman plus two
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members at each meeting. The minutes of the SRC meetings show
a quorum was in attendance in 57 out of the 58 meetings. The
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meeting on June 8,1973, was attended by two members plus the
plant superintendent, who is not an SRC member. This appears
to be in violation of Technical Specification 6.1.2.1.c.
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Records
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The master file of SRC meeting minutes did not contain the
minutes for meetings held on April 10, 1973, and May 5, 1973.
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The inspector found evidence where the committee had reviewed
die following incidents, but minutes were not available for
these meetings:
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(1) Engineered safeguards (ES) valve not fully open (Valve CF-1).l*'
(2) ES valves (BS-1 and -2) failed to open. /
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(3) Violation of TS 4.11.1 - Keowee River continuous sample. !
(4) Violation of Nonradiological Technical Specification 1.2 -
Wastewater collection basin pH.$/
e.
Responsibilities of the SRC
(1) Procedure Reviews
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Technical Specification 6.1.2.1.d (1) requires the SRC to
review all new procedures or proposed changes to existing
procedures which affect operational safety.
(According to
DPC's operational quality assurance manual, an "A" in the
procedure number indicates the procedure is safety related.)
Contrary to this requirement, the SRC meeting minutes of
March 29, April 5, 10, 16, 19, 25, 27 and 30, May 8 and
22, and June 15, 1973, state that changes to procedures
(of which approxi=ately 75 were labeled safety related by
DPC) were presented to the SRC but were not reviewed by the
SRC.
Each of the changes had been reviewed by one member of the
SRC but the results of each review had not been reported to
the committee. This apparent violation of Technical Specification 6.1.2.1.d (1) was discussed at the management
interview.
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1/ See RO Inspection Report No. 50-269/73-4, Details I, paragraph 13.
2/ See RO Inspection Report No. 50-269/73-4, Details I, paragraph 14,
3/ See RO Inspection Report No. 50-269/73-4, Details I, paragraph 4.
4/ See RO Inspection Report No. 50-2o9/73-3, Details I, paragraph 4.
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(2) Review of Station Operation and Safety Considerations
Technical Specification 6.1.2.1.d (2) requires the SRC to
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review station operation and safety considerations.
Contrary to this requirement, the minutes of the SRC
meetings do not reflect that the SRC reviewed:
(1) the
oil fire
RCP-1A1 (the second reactor coolant pump
oil fire)gy; nor (2) the lifting of the main-steam relief
valvea at 15% of full power (during whic} time part of the
turbine building siding was blown off) .2
(DPC management
was not certain whether or not the main steam relief valves
should have lifted at 15% of full power.) (See paragraph 10
below.)
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(3) Review of Abnormal Occurrences, Unusual Events, and
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Violations of Technical Specifications
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Tecanical Specification 6.1.2.1.d (3) requires the SRC to
review abnormal occurrences and violations of the Technical
Specifications and to make recommendations to prevent
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recurrence. Technical Specification 6.2.2 requires the
superintendent to cause the SRC to review abnormal occur-
rences (AO) and unusual events (UE), prepare written reports,
and make reco=mendations concerning corrective and preventive
actions.
Contrary to these requirements, the SRC meeting minutes do
not reflect that the following A0's, UE's or violations of
technical specifications were reviewed by the SRC:
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(a) Violation of Technical Specification 4.12.1 " Quarterly
Test of Control Room Filtering System Components"
This specification requires the flow to be measured
across each bank of control room filters. No in-
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strumentation was installed to facilitate measuring
the flow. The deficiency was found on April 26, 1973.
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1,/ See RO Inspection Report No. 50-269/73-3, Details I, paragraph 5.
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2/ See RO Inspection Report No. 50-269/73-6,Section VI, " Unusual
Occurrences."
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Ob) Violation of Technical Specification Table 4.1-3,
3 1nimum Sampling Frequency"
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This spe.cification requires samples to be obtained and
analyzed for boron after each makeup to dhe core flood
tanks, the borated water storage tank, and the spent
fuel pool. A 7PC audit on June 4-6, 1973, reported
violations of this specification.
(c) Violation of Technical Specification 3.4.3, " Turbine
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Bypass System"
This specifications requires all four valves to be
operable when the reactor coolant system is above
250*F.
One valve was out of service on May 4 and
one was out of service on May 5,1973. Operation
with a turbine bypass valve out of service is a
violation of a limiting condition for operation and
is reportable as an abnormal occurrence.
(d) Abnormal Occurrence - Leak in Incore Instrument Line
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On May 18,1973, a leak was discovered on one of the
incore instrument lines.
(These lines constitute a part
of the reactor coolant system boundary and are designed
to contain the radioactive materials resulting from the
fission process.) DPC first stated that this was an AO,
then questioned whether or not it actually was; however,
in a subsequent phone conversation with the inspector
DPC management agreed to report this as an AO.
(e) Unusual Event - Oil Fire at RCP-1A1
See RO Inspection Report No. 50-269/73-3, Details I,
paragraph 5.
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The inspector noted four other items (two technical specification
violations and two UE's) that were not recorded nor discussed in
the SRC meeting minutes. Although not found in the minutes of the
SRC meetings, documentation at the site indicated these items may
have been reviewed by the SRC. The four items are:
(a) Engineered safyguards (ES) valve not fully open
TValve CF-1).1
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1/ See RO Inspection Report No. 50-269/73-4, Details I, paragraph 13.
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(b) ES valves (BS-1 and -2) failed to open.-
(c) Violation of Technical S peification 4.11.1 - Keowee
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River continuous sample
(d) Violation of Nonradiological Technical Specification 1.2 -
Wastewater collection basin pH /
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These examples of the apparent violations.of Technical Specifications 6.1.2.1.d (3) and 6.2.2 were discussed
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at the management interview.
5.
Miscellaneous Test Procedures
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a.
Genecal
DPC's Administrative Policy Manual for Operational Quality
Assurance (APM/NS), Section 4.4, provides administrative
instructions for permanent station procedures.
Paragraph 4.4.1 lists the following types of procedures
which must meet the QA requirements for permanent station
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procedures:
(1) Emergency Procedures (EP)
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(2) Instrument Procedures (IP)
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(3) Maintenance Procedure (MP)
(4) Operating Procedures (OP)
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(5) Periodic Test Procedures (PT)
At the Oconee Nuclear Station, a sixth category of procedures
has been established. These are designated as miscellaneous
test procedures and are not required to receive the level of
reviews and approvals that the permanent plant procedures
receive. Additionally, the results of the miscellaneous
tests are not required to receive the same level of reviews
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and approvals that the permanent station tests receive.
1/ See, RO Inspection Report No. 50-269/73-4,' Details I, paragraph 14.
2/ See RO Inspe~ction Report No. 50-269/73-4, Details I, paragraph 4.
1/ See RO Inspection Report No. 50-269/73-3, Details I, paragraph 4.
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The inspector reviewed the master file of miscellaneous test
procedures to deternine if any dealt with the testing of
safety related systems and, if so, whether or not the prep-
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aration of the procedures and the conduct of the tests conformed
to DPC and regulatory requirements. The extent and level of the
reviews and approvals given the procedures and test results
varied widely from procedure to procedure. The results of
the inspector's review for each specific test are summarized
below,
b.
Specific Tests
(1) 4160 v Bus Transfer Time Test
The stated purpose of the test was to determine the transfer
time of 4160 volt breakers BlTl and BlT3. These breakers
are a part of the engineered safeguards switchgear and the
automatic transfer of the breakers is required in the event
.of the loss of power to the 4160 volt bus.
The procedure had not been previously classified as
safety related and assigned an alpha-numeric designation
as required by paragraph 4.4.4.2 of the APM/NS to assure
proper reviews and approvals. The failure to follow the
approved administrative procedures appears to be in
violation of Criterion V af Appendix B to 10 CFR 50.
The procedure did not adequately define acceptance limits
contained in applicable design documents and that the test
coordinator or data taker could determine that the test
had been successful. Failure to define acceptance limits
appears to be contrary to Criterion XI of Appendix B
to 10 CFR 50.
When questioned, a member of tha station manage =ent advised
the inspector that the documentation of the test results
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had not been reviewed by the SRC but had been forwarded
to DPC design engineering for their review. Failure
of the SRC to review the tests appears to be a violation
of Section 6.1.2.1.d (1) of the Technical Specifications.
(2) Emergencv Feedvater Pump Functional. Test
The st5ted purpose of this test was:
(a) To determine the proper valve lineup of the emergency
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(b) To demonstrate the ability of the EFW pump to deliver
water to the steam generator through the auxiliary
nozzle utilizing main steam.
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(c) To demonstrate auto start and to measure the response
time of the EFW pump.
(d) To verify main turbine trip on loss of the main
feedwater pumps (MWP) .
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The EFWS is required for the addition of water to the steam
generator on a trip of the MFWP's and is a safety related
system.
The procedure had not been properly classified as safety
related and assigned an alpha-numeric designation as
required to assure proper reviews and approvals by
paragraph 4.4.4.2 of the APM/NS. The failure to follow,
approved administrative procedures appears to be a
violation of Criterion V of Appendix B to 10 CFR 50.
A revision to the procedure had been penciled into the
body of the procedure. A procedure revision form had
not been completed for this revision as required by
paragraph 4.4.6.1 of the APM/NS. Failure to follow the
approved administrative procedures appears to be a
violation of Criterion V of Appendix B to 10 CFR 50.
A revision to the procedure which changed the trip
point of the main turbine generator from 600 psig to
800 psig had been approved by M. D. McIntosh, Operating
Engineer, on June 10, 1973. The APM/NS, paragraph 4.4.2.2.3,
specifies that the plant superintendent shall approve safety
related procedure revisions. Failure to properly approve
procedure revisions appears to be a violation of Criterion VI
of Appendix B to 10 CFR 50.
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The APM/NS, paragraph 4.4.6.l(a), requires that a " Proposed
Procedure Revision" form be completed for each proposed
revision to a safety related procedure. This form re-
quires that a safety analysis be performed for the revision.
Contrary to this requirement, documentation was not avail-
able ,to indicate that safety analyses were performed for
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the revisions to this procedure. The failure to follow
the requirements of the administrative procedures appears
to be a violation of Criterion V of Appendix B to 10 CFR 50.
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Documentation was not available to verify that Section 10
and Steps 12.9 and 12.10 of the test procedure had been
completed . In addition, the test data sheets did not
indicate the dates during which the tests were conducted
nor to which power level the data applied.
Failure to
maintain adequate quality assurance records appears to be
a violation of Criterion XVII of Appendix B to 10 CFR 50.
Paragraphs 4.4.6.l(c) and (g) of the APM/NS specify that
revisions to procedyres shall be approved within seventy-two
hours and filed with the = aster file copy. Revision 4 to
the procedure had not been approved within the seventy-two
hours specified and a copy of the change had not been filed
in the master file. The failure to obtain the proper revision
approvals and to control the distribution of the revision
appears to be a violation of Criterion VI of Appendix B to
10 CFR 50.
The SRC review of the original procedure listed J. E. Smith
as one of three members present. The membership of this
committee does not include J. E. Smith. The Technical Specifications, Section 6.12.1.c. states that a quorum of
the committee is the chairman and two members. Failure to
have a quorum of the committee while conducting committee
business appears to be a violation of the technical specifi-
cations.
(3) Check of Safety and Shim Control Rod Actuators for
Frictional Binding
The purpose of this procedure was to' determine, through
weight differential measurements, that the control rod
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actuator and its coupled rod were not binding. Although
the proccdure was initialed as being reviewed by the SRC
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on February 17, 1973, the review appeared to be superficial
in that the deficiencies listed below were not identified.
The failure to review safety related procedures appears to
be contrary to the requirements of Technical Specification 6.1.2.1.d (1) .
Operation of '*te control rods and actuators are required to
effect orderly control of the reactor and are classified as
safety feature syste=s.
Contrary to the requirements of
paragraph 4.4.4.2 of the APM/NS, this procedure had not
been properly classified as safety related and assigned
an alpha-numeric designator to assure proper reviews and
approvals. The failure to follow the approved administrative
procedures appears to be a violation of Criterion V of
Appendix B to 10 CFR 50.
The procedure was inadequate in that the following were not
included:
(a) The minimum boron concentration of the reactor coolant
was not specified.
(b) 'The verification of proper operation of the nuclear
instrumentation was not specified.
(c) The qualifications of the hoist operator were not
specified.
(d) There was no limitation on the number of control rods
that could be in the "Out" position at any one time.
(e) There was no requirement that the reinsertion of the
control rod be verified.
(f) There was no requirement that the position indication
be verified at any time during the operatica.
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(g) The procedure did not provide any method, such as check
sheets, to assure that the procedure steps were followed
or to document that the procedure was followed.
(h) The procedure did not specify the steps to be followed
in the event of an emergency.
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(i) The procedure did not require that a licensed operator
manipulate the controls or to be present when the con-
trols were manipulated.
(Controls in this case being
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the air hoist.)
Failure to provide adequate procedures appears to be contrary
to the requirements of Criterion V of Appendix B to 10 CFR 50.
The method prescribed in the procedure for the conduct of the
test specified that the hoist operator lif t each control rod
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with an air hoist and, by measuring the pounds of force
required, determine the friction drag on the drives and rod.
The procedure required that voice co=munication be main-
tained between the hoist operator and a licensed reactor
operator in the control building.
Part 50.54(i) of 10 CFR 50 provides that the licensee
not permit the manipulation of the controls of any facility
by any one who is not a licensed operator or senior operator
unless the manipulation is done under the direction of and
in the presence of a licensed operator or senior operator.
An unlicensed maintenance technician operated the hoist in
conducting this test. The manipulation of the control rods
by an unlicensed person appears to be a violation of
Part 50.54(1) .
(4) Auto Transfer From IT to CT1 Transfor er Without Generator
Lockout
The purpose of this test was to verify that the time required
for the transfer of station power from the unit station service
transformer IT to the reserve station service transfer CT1
would not cause a lockout of the main generator.
This procedure was approved by Smith on February 3,1973,
and was perfor=ed on February 4,1973. The procedure
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required the disabling of the emergency start relay for
the Keowee Hydro Station. The operating license for
Oconee 1 was issued on February 6,1973, and the Keowee
.
Hydro Station provides emergency power for the Oconee
Nuclear Station.
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The procedure did not require that the emergency start
relays be restored to service nor whether their correct
operation need be verified. Failure to identify the status
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of the individual items of equipment appears to be contrary
to the requirements of Criterion XIV of Appendix B to
10 CFR 50.
(5) Inspection of Retainer Nuts on Engineered Safeguards (ES)
Valves
.
On April 9, 1973, the yoke locking nuts on two ES valves
in the core flood system were found to be loose preventing
proper operation of the valve position indication switches
and proper operation of the valves. The purpose of this
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procedure was to determine if other ES valves also had
loose and/or improperly staked lock nuts. The following
valves were checked:
BS-1, BS-2, CS-5, HP-24, HP-25, HP-26, LP-3, LP-17,
LP-19, LP-20, LP-21, LP-22, PR-1, PR-6, PR-7, PR-9,
CC-7, CF-1, CF-2, FDW-103, EDW-104, GWD-12, LPSW-5,
.LPSW-6, and LPSW-18.
The procedure was approved on April 14,1973, by the
maintenance supervisor. Documentation was not available
to verify that the procedure had been reviewed by the SRC.
The APM/NS, paragraph 4.3.2.2.8, requires that safety
related procedures receive the approval of the plant
superintendent and be reviewed by the SRC. The failure
to follow approved administrative procedures appears to
be a violation of Criterion V of Appendix B to 10 CFR 50.
The data for approximately ten valves had been marked
over such that the record of the "as. found" condition
of the locking nut could not be determined. The revisions
to the data were neither initialed nor dated. Failure to
provide records to furnish evidence of activities affecting ,
quality appears to be in violation of Criterion XVII of -
Appendix B to 10 CFR 50.
(6) Hydro of RC-48
The. purpose of this procedure was to hydro a replacement
valve located in the 1B loop drain and pressure transducer
line. This valve is one of two valves that would form the
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second boundary of the reactor coolant system (RC System)
if the line were removed from service. The valve normally
sees the reactor coolant system pressure.
This procedure is for the test of a safety related component
and the development of the procedure and the hydro of the
valve and it appe es that this test should have conformed
to the requirements specified in the APM/NS. An alpha-
numeric designator indicating that the test was safe +.y
related and would obtain the proper reviews and approvals
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was not assigned this test. A centrolling procedure for
conducting hydro tests on the RC and SC systems is available.
The procedure actually used for conducting this test is
entitled " Controlling Procedure for Hydrostatic Test
(Excluding RC System and SC System) ." The apparent failure
to follow the approved administrative procedures is a violation
of Criterion V of Appendix B to 10 CFR 50.
(7) Shuffling Control Components in Spent Fuel Pool
This procedure was designed to control the removal and
reinstallation of control rods in the fuel assemblies.
The procedure was approved by McIntosh and initialed by
Smith. No documentation was available to verify that the
procedure had been reviewed by the SRC. The apparent
failure to review the procedure is a violation of Technical Specification 6.1.2.1.d (1) . The procedure had not received
an alpha-numeric designator as required by the APM/NS. The
apparent failure to follow approved administrative procedures
is a violation of Criterion V of Appendix B to 10 CFR 50.
The above deficiencies were discussed in detail in the management
interview. Smith and Powell were advised that time had not been
available to review all of the miscellaneous tests, but that
numerous others appeared to have safety significance. Smith
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assured the inspectors that a review of all of the tests would
be made to formalize those required by the APM/NS.
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6.
NSRC Minutes
The inspector reviewed the minutes of the NSRC which were in the
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Oconee master file. The minutes of the most recent meeting con-
tained in these files were for a meeting conducted on January 30, 1973.
Attached to these minutes was a list of recommendations and requests
for information that the NSRC indicated had not yet been resolved.
This list contained eighteen items which dated from November 1971
to January 30, 1973, and which, according to the technical specifi-
cations, are provided to appropriate members of management. The
failure to obtain resolution of these items is indicative that the
management reviews as required by Criterion II of Appendix B to
10 CFR 50 appear to be inade,quate.
The inspector questioned the station management about other NSRC
meetings . He was advised initially that a second meeting had been
held in May 1973, but in the management interview, two additional
meetings were indicated. Minutes of these meetings were not available.
Section 6.5.o. of the technical srecifications requires that copies
of the NSRC minutes be retained in the plant files.
Failure to
retain these files appears to be a violation of this requirement.
The inspector advised members of DPC management during the management
interview that the records did not indicate that the NSRC had reviewed
any of the deficiencies revealed by the AEC inspections.
Although
this is not a specific requirement of the committee, the inspection
reports and letters could be a source of information that would aid
the com=ittee in its operation. The inspector was advised that his
comment would be considered.
The inspector, in reviewing the NSRC minutes, also observed that
the committee had only. requested to review two safety related
procedures in recent months. He urged that the committee give
more attention to this area since the review of such procedures
is a responsibility of the group. He was assured that this would
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be done.
7.
Failure to Obtain Clearances for Maintenance
Time did not permit a complete review of the station logs during
this inspection. The inspector quickly read the shif t supervisor's
log for the period from January to April 1973.
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This log contained almost daily entries pertaining to Unit 2
events that were not related to Unit 1 operations. Licensee
manatement assured the inspector that this discrepancy had been
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recently corrected and should not occur again.
An entry on March 23, 1973, stated that an equipment manufacturer's
representative and an engineer from the DPC Design Department had
removed the strainer on the oil lift system for the 1B1 reactor
coolant pump. This operation was performed without obtaining
clearance from the control room operator and without the pump
being deenergized. Oil was spilled from the system as a result
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of these operations. This event was discussed in the management
interview. Licensee management stated that the individuals who
had removed the strainer were violating plant administrative pro-
cedures, the individuals were severely reprimanded and that
all personnel were cautioned against unauthorized operation of
equipment .
8.
Steam Relief Valve Operation
During plant testing the week of June 10, 1973, the main steam
relief valves had relieved subsequent to reactor scrams and the
released steam had blown the siding off the auxiliary building in
the area of the valves. The valves had operated when the turbine
had tripped from power levels as low as 15%. As a part of the
inspection followup, Murphy questioned members of the plant staff
about these occurrences. None of the staff members could answer
relatively basic questions about the operation of the valves such
as:
a.
Which valves relieved during scrams from 15% power level and
from 40% power levels?
b.
Did the valves relieve in the expected se,quence?
c.
Was relief valve operation expected on a trip at 15% power,
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i.e., did analyses indicate that the valves should release at
trips from this power level?
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If valves were expected to relieve, did the proper number
relieve?
e.
In every case, did the valves relieve at the set pressure?
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Was the relieving of the valves at 15% pcwer analyzed to
determine if the event was considered to be safety related
as an unusual event or abnormal occurrence as defined in the
technical specifications?
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The inspector was given the following information relating to
the safety valves.
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a.
The plant staff had previously determined that the set points
of the valves had drifted downward from the points at which
they had been originally set to relieve, i.e. , they lifted
at a lower system preesure. The staff was not concerned since
this drift was in the " conservative" direction. It had not
been determined the extent of the drift or if the drif t was
continuing, nor had the undesired effects of the unanticipated
operation been considered.
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b.
The SRC had not reviewed the occurrence.
c.
The cause of the valve setpoint drift had not been determined.
It had not been ascertained how much might be "one time" drif t,
such as would occur by the permanent relaxation of the spring
after being heated, nor how much might be repeatable and cyclic
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such as changes caused by the temperature fluctuations experienced
by the valves.
The inspectors expressed concern during the management interview
about the lack of action as related to the valve operations.
Licensee's management agreed that a review would be made of this
area and any corrective actions required would be taken. RO:II
will be advised of the resuits of the study, and if the study
indicates that the operation of the valves at the lower power
levels is a reportable event, then the report will be issued.
This item will be included as an unresolved item until DPC
completes this study.
9.
Plant Reporting Reouirements
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The inspector was advised by members of the plant staff that each
event which was potentially reportable to the AEC was investigated
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by a me=ber of the plant staff to determine if the event was indeed
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reportable. The plant superintendent assigned a staff member to
perform the 12vestigation. Upon further questioning, the inspector
determined that there was no formal method for assuring that the
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superintendent was made aware of the various events. Based upon
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the number of reportable occurrences that this inspection had
revealed which had not been reviewed, it does not appear that
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the licensee has an adequate procedure to assure that reportable
occurrences are recognized at the time of the event and that
management is apprised of the occurrences (paragraph 4). The
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apparent failure to provide adequate methods to identify
deficiencies, deviations and malfunctions appears to be contrary
to the requirements of Criterion XVI of Appendix B to 10 CFR 50.
In addition, recorder charts were not indexed with actual time
marks at a frequency that would permit correlation of the charts
accurately enough to reconstruct the details.of an event. One
method of indexing the charts would be to have the operators mark
the actual time at eac;i shift change. Management agreed to consider
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this method. These items were discussed in the management interview.
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