IR 05000269/1976012
| ML19312C782 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/24/1976 |
| From: | Epps T, Kowalczuk A, Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19312C774 | List: |
| References | |
| 50-269-76-12, 50-270-76-12, 50-287-76-12, NUDOCS 7912190998 | |
| Download: ML19312C782 (28) | |
Text
__-
__._._______.
.
__ _ _____ _
_ _
.m.
___
,
.,
l
,-
- "49 UNITED STATES
,
s
'
-
g jo
- }
NUCLEAR REGULATORY COMMIS, N g
.
,
t
o R EGION ll
)
I
$
230 PEACHTREE STREET, N.W. SUITE 818
[
ATLANTA, GEORGI A 30303 o
- -
)
.....
i l
IE Inspection Report Nos. 50-269/76-12, 50-270/76-12 and 50-287/76-12 l
Licensee:
Duke Power Company Power Building
{
422 South Church Street Charlotte, North Carolina 28201 Facility Name:
Oconee Units 1, 2 and 3 Docket Nos.:
50-269, 50-270 and 50-287 License Nos.:
DPR-38, DPR-47 and DPR-55 Category:
C, C. and C Location:
Seneca, South Carolina Type of Inspection:
Routine, Unannounced Dates of Inspection:
September 28-October 1 and 4-8, 13-15, 1976 Dates of Previous Inspection:
September 14-17, 1976 Principal Inspector:
T. N. Epps, Reactor Inspector Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch
'
Accompanying Inspectors:
A. D. Kowalczuk, Radiation Specialist Radiation Support Section Fuel Facility and Materials Safety Branch
,
.
J. J. Blake, Reactor Inspector Engineering Support Section No. 2 Reactor Construction and Engineering Support Branch S. D. Ebneter, Reactor Inspector Engineering Support Section No. 2 Reactor Construction and Engineering Support Branch P. T. Burnett, Reactor Inspector i
Nuclear Support Section Reactor Operations and Nuclear Support Branch 0t
.
w &-.L.
.
.-
49121909 9 8
,
!
'
._
__
,, _.._
.... _. - _ _
-....,,. -
.
. -
-
-
... -.
.
.
-
..
-
-
I
-}
e)
.
,
-
o
,
IE Rpt. Fos. 50-269/76-12,
'
,
-#l 50-270/76-12 and 50-287/76-12-2-Other Accompanying Personnel':
A. F. Gibson, Chief Radiation Support Section Fuel Facility and Materials Safety Branch Principal Inspector:
.
/
a
//-//- 7d T.N.Epps,Reactoy/4Kspector Date Reactor Projects Fedion No. 2 Reactor Operations and Nuclear
-
Support Branch Reviewed by:
[. C,
///Z M 7f R. C. Lewis, Chief Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch
)
.
i l
l
-
V
- v.
- - -. -..
..,...
..
_
.,
_ _,
., - - _. _
i
-
,,
!,.
.)
'
'
IE Rpt. Nsc. 50-269/76-12,
'
F
50-270/76-12 and 50-287/76-12-3-
.
SUMMARY OF FINDINGS
.
I.
Enforcement Items Infrcctions 1.
Contrary to Technical Specification 6.4.1.g on October 13, 1976, Oconee Nuclear Station Directive 3.8.2 was not adhered to in that three persons entered radiation control zones without completing the daily exposure time card as required.
(Details II, parsgraph 2)
2.
Contrary to 10 CFR 20.203(c) access to five high radiation areas in the auxiliary building was not controlled as required, on October 13, 1976.
(Details II, paragraph 3)
3.
Contrary to 10 CFR 50 Appendix B Criterion V and the licensee's
,
Quality Assurance Program in the Duke - 1 Topical Report
,
(Section 7.2.2) as implemented by Station Directive 5.3.3 paragraph 4.2, the safety supervisor was not notified, by September 30, 1976, that several fire extinguishers were removed from specified containment areas.
(Details III, paragraph 7)
II.
Licensee Action on Previously Identified Enforcement Matters Noncompliance item I from IE Inspection Reports 50-269, 270, 287/76-4 is closed.
(Details I, paragraph 2.a)
Noncompliance items from IE Inspection Reports 50-269, 270, 287/76-6 remain open.
CDetails I, paragraph 2.bl III.
New Unresolved Items 76-12/1 Non-Destructive Examination (NDE) Acceptance Criteria NDE procedures in use at the site contain acceptance criteria for inspection which are referenced to particular
f abrication codes without listing the specific date of
-
issue of the code referenced.
The licensee ha agreed to include the required information by revision to the procedures involved prior to January 1, 1977.
(See Details IV, paragraph 41
.
%
.
_
,
. _ _ _ _ _. -... _ _ -.
.- ---W
.
_
.
.
-
'
-
,,
,
p IE Rpt. N:0. 50.c9/76-12,
-
50-270/76-12 and 50-287/76-12-4-
'.1
.
IV.
Status of Previously Reported Unresolved Items 76-9/1 Surveillance Records Verification of operational hydro records is complete.
This item is closed.
(Details V, parsgraph 2)
74-13/2 Reactor Coolant Flow Anomaly The licensee conducted examinations of fuel assemblics
'
from Unit 2 and reported no problems.
This item is closed.
(Details I, paragraph 3.a)
74-14/2 Ventilation Control Between Auxiliary and Turbine Buildings Modifications are to be completed by January 1, 1977.
This item remains open.
~
75-12, 13, 14/1 Fire Protection Program Weaknesses These specific weaknesses have been resolved.
This item is closed.
(Details I, paragraph 3.b)
76-1/1 Instrument Calibration This program is not fully implemented.
The item remains open.
CDetails I, paragraph 3.cl 76-1/2 Electrical Equipment Calibration This item remains open.
(peta 11s I, paragr.iph 3.d)
76-6/1 Definition of " Safety-Related" This item is closed.
(Details I, paragraph 3.el V.
Unusual Occurrences, Turbine Building klooding Turbine building flooding occurred on October 9, 1976, which had the potential for rendering some safety systems inoperable.
CDetails I, paragraph 5)
-
g pge e p
w
+-m--
-4-
,,)
y
-
-
.
IE Rpt. N20. 50-269/76-12,
<
50-270/76-12 cnd 50-287/76-12-5-
.i
.
,
VI.
Other Significant Findings
.
None VII. Management Interviews A meeting was held on October 15, 1976, by T. N. Epps and S. D. Ebneter with J. E. Smith and members of the Oconee staff.
Items discussed included inspection findings in Details I and III of this report.
A meeting was held on October 14, 1976, by A. Kowalczuk and A. Gibson with R. M. Koehler and members of the Oconee staff.
Items discussed included inspection findings in Details II and Section I of the summary of this report.
A meeting was held on October 7,1976, by P. T. Burnett with J. E. Smith to discuss inspection findings found in Details V of
'
this report.
-
A meeting was held on October 8,1976, by J. Blake with J. E. Smith and members of the Oconee staff to discuss inspection findings in Details IV of this report.
l l
l
.
,
s
.
.
.
- __,
_
,** *
_
$
"4F,6 * p **
.-
-.
. -..
.
.-
.-
-
._.
_
-
'
~
']
.
r
,
IE Rpt Nos. 50-269/76-12,
.
,
50-270/76-12 and 50-287/76-12 I-1
,
,)
//-// ~ M DETAILS I Prepared byr
./
V T. N. Epps, Reac gl Inspector Date
.
Reactor Project Section No. 2 Reactor Operations and Nuclear Support Branch Dates of Inspection:
October 13-15, 1976 Reviewed byi
[ * d.
// '/Md R. C. Lewis, Chief Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch 1.
Individuals Contacted Duke Power Company (DPC)
.
Oconee Personnel J. E. Smith - Manager, Oconee Nuclear Station J. W. Hampton - Manager, Administrative Services L. E. Schmid - Superintendent of Operations i
0. S. Bradham - Superintendent of Maintenance R. M. Koehler - Superintendent of Technical Services R. T. Bond - Performance Engineer J. Cox - Senior QA Engineer J. Bracket - Assistant QA Engineer E. Hite - Engineer, Maintenance Other Operations Personnel 2.
Previous Items of Noncompliance a.
Noncompliance iten number I from IE Inspection Reports 50-269,
<
270, 287/76-4 concerned failure to shut Unit 1 down at the specified shutdown rate of 10% per hour.
The inspector verified corrective actions stated in the licensee's letter dated May 21, 1976.
There are no further questions on this item.
]
b.
Noncompliance item I.A.1 from IE inspection report 50-269, 270, 287/76-6 was discussed with the licensee.
The licensee's corrective action involved use of an operations precedure (OP/0/A/1102/06) for removal and restoration of station equipment.
Enclosure 7.1 to the procedure includes general steps requiring s
.
l
'
!
,
.
.
.
..
.
......,.. -
'
._
,
.
. _.. -
. -. _ _... _ _
_
- _ _ _
D L
,
'
~#
.IE Rpt. Nos. 50-209/76-12,
,
.
50-270/76-12 and 50-287/76-12 I-2 g
-
)
the operator to describe action taken to complete redundant checks, actions taken to remove equipment from service, reasons for equipment removal, blank valve checklists to be filled out and description of functional tests and verification of test results.
The enclosure receives approval for use from a unit superviter and a control operator.
The enclosure does not receive approval according to the requirements of Technical Specification 6.1.'2.1.
The inspector stated that licensee corrective action described above does not resolve the noncom-pliance item. This iten remains open.
3.
Previous Unresolved Items a.
Reactor Coolant Flow Anomaly (74-13/2) - This item is closed based on completion of further verification that fuel assembly motion is not occurring and a letter from Duke Power Co. to the Nuclear Regulatory Commission QUUR) dated July 21, 1976 summarizing additional surveillance conducted on Unit 2 fuel assemblies.
,
b.
Fire Protection Program Weaknesses (75-12,13,14/1) The licensee installed and tested reflash mechanisms on all units on September 1 and 10, 1976.
This completed correctiv.e action on the weaknesses identified in IE inspection report 50-269/75-12, 50-270/75-13 and 50-287/75-14.
This item is closed.
c.
Instrument calibration (76-1/1) - A program, including 83 iteme for each unit was established for calibration of non-technical specification instrumentation.
This program is not fully implemented aq this time.
This iten remains open, d.
Electrical Equipment L.libration (76-1/2) A new commitment date of November 15, 1976 was given by the licensee to resolve this administrative item. A corporate task force is evaluating the subject of administrative control over safety-related relays and breakers.
This item remains open.
e.
Definition of " Safety-Related" (.76-6/1) - The Oconee Plant Manager issued a memo to station superintendents, dated October 7, 1976, instructing personnel to broaden their thinking in determining whether an iten is safety-related or non-safety
'
related, and to perform further checks before declaring an item non-safety-related.
The checklist used in determining safety-related status was also reviewed and revised by the licensee.
l
-
.
l l
-
.
-
-
.
.-
.
... -.
-
... -. _,,
.
- - - -
_-
-
-
.
,.-
- -. _.. - - -.
.. - - -.
-. -.. -... -
. - _. -
'
-
,3
-
.
,
!
'IE Rpt. Nos. 50-269/76-12,
,'-
50-270/76-12 and 50-287/76-12 I-3
'
Thi.s unresolved item is closed based on the above actions.
..
4.
Use of Freeze Plug in 2 inch..; actor Building (RB) Sump Drain Line The inspector reviewed the licensee's plans to use a " freeze plug" in a 2 inch diameter reactor building (RB) drain line, on Unit 2, to allow repair of an RB isolation valve in the same line.
The isolation valve was closed and could not be opened for pumping the RB sump.
This review was conducted to assure that Technical Specification 3.6, requiring containment integrity, was met while the reactor was operating.
The licensee's plan involved installation of a freeze plug between the RB and the first RB drain isolation valve (failed valve) outside the containment, removal of the valve bonnet, installation of a spare valve bonnet; thence, removal of the freeze plug.
A written safety evaluation was performed that showed that an rareviewed safety question did not exist and that containment integrity could be maintained.
The freeze plug was then successfully leak tested to 60 psig to provide insurance that the plug would hold under LOCA peak pressure conditions.
Also, measures were i
taken to assure rapid installation of a valve bonnet in place of the removed bonnet, should a leak occur in the RB while the valve bonnet was removed.
The licensee used approved procedures and took the above actions and the inspector had no further questions.
5.
Turbine Building Flooding The inspector toured the turbine building to observe equipment that was affected by flooding of that building on October 9, 1976.
The auxiliary building was also toured to determine what equipment would have been affected if further flooding had occurred.
On October 9, 1975, while in a refueling outage, three Unit 3 manway covers were off the main condenser.
An electrical problem caused a 78 inch valve in the condenser outlet line to'open, allowing Keowee lake water to flow through the condenser into the turbine building. Water was approximately 1 foot deep in the turbine
-
building when the electrical problem was corrected and the 78 inch valve was closed; thus, terminating flow of water into the turbine building. Water was within 5 inches of overflowing the turbine building water retaining curb and entering the auxiliary building,
-
.
-
'
.
..
....
-
-. _
.. -
. _ _ _ _ _
_ _ _. _ _ _. _
__
_
m.-
_.
.
~
.
/-
,
,
,
-
i
.
.
'
IE Rpt. Nos. 50-269/76-12, g1 50-270/76-12 and 50-287/76--12 I-4
'
.
It was determined from this review of the incident that additional flooding could have rendered some equipment inoperable that is required to remove decay heat from the reactor. The hot well pumps, condensate booster pumps, the main feedwater pumps and the emergency feedwater pumps (all in the turbine building) could be rendered inoperable with further flooding. The auxiliary service water pumps and LPI and HPI pumps (in the auxiliary building) would be affected first if flooding occurred in the auxiliary building.
After this incident, the licensee reinstalled the manways'on the main condenser to prevent recurrence.
The licensee assigned a task force to study this incident and a report will be submitted to the Nuclear Regulatory Comission.
The incident will then receive further review by the NRC.
.
9
<
.
', - -
,
,
.
l
'
__
. _
J
,.
.
.
..
.
.
.
J
.
.
w
i l
IE Rpt. Nos. 50-269/76-12,
.
,
,-
50-270/76-12 and 50-287/76-12 II-1 i
,
M'
DETAILS II Prepared by.
.
w x
m s,
A. D. Kowale,zuk, Radiatipn
/ Date Specialist
/
Radiation Support Section Fuel Facility and Materials Safety Branch Dates of Inspectio :
ctober 13-14, 1976
'
Reviewed by:
T II/II!7 b
_-
A. F. Gibson, Section Chief Date Radiation Support Section Fuel Facilities and Materials Safety Branch 1.
Individuals Contacted J. E. Smith - Plant Manager R. M. Koehler - Superintendent of Technical Services C. T. Yongue - Health Physics Supervisor R. T. Bond - Technical Services Engineer W. P. Deal - Assistant Health Physics Supervisor D. L. Davidson - Assistant Health Physics Supervisor H. Smith - Clerk 2.
External Exposure Control Oconee Nuclear Station Directive 3.8.2 requires that a Daily a.
Exposure Time Record be used prior to entry into radiation control zones and that the data entered reflect the applicable radiation work permit. On October 13, 1976, the inspectors observed, while accompanied by a management representative, seven persons working in the Unit 3 spent fuel pit area and asked if their Daily Exposure Time Cards were completed.
All of the individuals claimed to have initiated the card and one individual stated that his card had not been completed to reflect the applicable radiation work permit.
Upon requesting the cards for examination it was established that two additional individuals were not logged in on any applicable radiation work permit.
The remaining four individuals were logged in on Radiation Work Permit 959. Management was informed that failure to follow the Station Directive constituted noncompliance with Technical Specification 6.4.1.g.
i
.
a
p
%
,...
$..
.
e+
~
en e-w-w-
= +
'N'
-
-
_.
-
__
-__ _ _ _ _ _ _
.-
-
.. -.
-
..
. - - -.-.- -
.
. _ - -.. -,.... - -
! -
- ';
-}
.
_ j s
IE Rpt. Nos. 50-269/76-12,
'
< ^)
50-270/76-12 and 50-287/76-12 II-2
.
b.
During a tour of the Unit 3 containment, the inspectors observed two individuals sleeping in areas having about a 4 mrem per hour radiation field, two individuals taking a rest break in a posted high radiation area (the local radiation field was about 7 mrem / hour), and two individuals taking a rest break in a radiation field of about 7 mrem per hour,near reactor components reading about 1000 mrem per hour on contact. The inspectors commented that these observations were inconsistent with a policy of maintaining exposures as low as practicable. Management acknowledged this comment and stated that appropriate corrective action would be taken.
3.
Posting and Control
.
On October 13, 1976, while accompanied by a management representative, an inspector oberved that access to the following posted high
'
radiation areas was not controlled as required by 10 CFR 20.203:
1.
Bleed Evaporator Room
'A door at the entrance to this room was observed to be unlocked.
The door had been i
provided to meet the access control requirements of 10 CFR 20.203(c)(2)(iii). None of the alternative control devices described by 10 CFR 20.203(c)(2) were installed,
'
nor was the area under surveillance as described by 10 CFR 20.203(c)(4).
2.
High Pressure Injection Purp Area - A door in a corridor
-
providing access to this area was observed to be unlocked.
'The door had been provided to meet the access control requirements of 10 CFR 20.203 (c)(2)(iii). None of the alternative control devices described by 10 CFR 20.203(c)(2) were installed, nor was the area under surveillance as described by 10 CFR 20.203(c)(4).
3.
Room 160 Containing Unit 3 Low Pressure Injection Cooler - A door'at the entrance to this room was easily unlocked by the inspector without a key from outside the room.
Thus,,the intent of 10 CFR 20.203(c)(2)(iii) was not met.
The entrance to the room was not equipped with a control device as described by 10 CFR 20.203(c)(2) nor was the area under surveillance as
,
described by 10 CFR 20.203(c)(4).
4.
Janitor Storage Area, Second Level - The only door to this area was locked with a padlock.
This method of control could prevent an individual without a key from leaving a high radiation
-
area and, thus, was contrary to 10 CFR 20.203(c)(3).
.
'
./
-
.
m'=
.
,
i em ;,,
- @
-M
-
68Q
'*6"'**
-
.
.-
- -..-...-....
-.-.
.
.-.
-
' )m
'.
.
,~
s
,
c
-
[
IE Rpt. Nos. 50-269/76-12,
'
'])
50-270/76-12 and 50-287/76-12 II-3
.
5.
Corridor, Second Level, Near Valve 2HP-181 - A high radiation area was described with a rope and sign, but none of the access
'
controls marked by 10 CFR 20.203(c) were established.
4.
Internal Exposure Control a)
On October 13 and 14, 1976, an inspector reviewed occupational air monitocing results for the previous week in the Unit 3 containment and spent fuel pit areas.
The inspector noted that indications of I-131 at greater the 10 CFR 20 Appendix B concentrations were obtained from gas grab samples but were not used in exposure evaluations.
In all cases, corresponding particulate and charcoal sample results did not indicate the presence of I-131. A management representative stated that the charcoal sample results confirmed the absence of I-131.
Further investigation by the inspector resulted in sample container background data and Unit 2 containment gas grab sample data that appeared to confirm the absence of I-131 in the Unit 3 areas.
The inspector commented that closer attention to obtaining container background data and selecting uncontami-
_
nated sampling containers is needed to provide safisfactory survey data on a routine basis, b)
During a tour of the Unit 3 containment.the inspectors noted that installed connections for breathing air lines were not protected to prevent contamination and that unprotected connectors
.
on breathing air hoses were laying on the floor.
5.
Training Training records and security badges reviewed by an inspector indicated that seven visitors who were issued personnel dosemetry on September 20 and 21, 1976 were given health physics training or escorted as required by plant administrative procedures.
6.
Procedures
.
.
a)
The inspector observed that ites 3.1.1, 3.1.2, 3.1.3, 3.1.4 and 3.1.18 of OP/3/A/1502/05, Fuel Unloading and Reloading, were* signed off as completed.
These items are related to checks of radiation monitor operability.
b)
Based on' radiation level measurements _made by plant personnel in the presence of the inspectors, dose rated in the spent fuel pit area of Unit 3 met the commitments in FSAR paragraph 9.7.2.3.
.
.
4
.
.
.
..
_ _._.. _. _. -
. _. -.
_. -.. _ _ _ _ _ _. -
m.
.. _. _. _ _.
.-
_:
_. _ _ _ _ _...... -
__
-
-
i
.
.
-
,
,IE Rpt. No. 50-2o//76-12 III-1
,
,m DETAILS III Prepa. red by:
/
U O
'
S. D. Ebneter, Reactor Inspector
'Date Engineering Support Section No. 2 Reactor Construction and Engineering Support Branch
-
Dates of Inspection:
September 28 - October 1, 1976 and Oct 13-15, 1976 Reviewed by:
M strxr7rtn-g
//!e8
[
A. ~10. Herdtf'JBffief jf
/ Dave
'
Engineering,fSupport Section No. 2 Reactor Construction and Engineering Support Branch 1.
Persons Contacted a.
Duke Power Company (DPC)
,
J. E. Smith - Station Manager J. Hampton - Administrative Supervisor J. W. Cox - Station QA Manager R. J. Brackett - Operations QA
'
J. D. Norton - Quality Assurance Staff J. Itin - Safety Supervisor
,
b.
Babcock and Wilcox Construction Company (B&W)
G. Terning - ISI Team Leader R. Taylor - Level II Examiner
,
G. Gladney - Level II Examiner 2.
Scopa of Inspection This inspection included audits of inservice inspection (ISI)
activities related to Oconce Nuclear Station Unit 3.
ISI plans, procedures, work activities and data records were audited to determine compliance with Technical Specifications, FSAR commitments and ASME Code requirements.
An independent inspection of compliance with fire preventive directives applicable to welding and burning was I
also performed.
3.
Inservice Inspection (ISI) Quality Assurance
,
The QA program of Duke Power Company (DPC) is documented in Duke-1 Topical QA Program which was reviewed and accepted by the USNRC on April 17, 1975.
.
e
,
i
.n
.
.a.n
.,
...
.
~~**
y +-
.~;,--
____ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
_ - _ _ _ _ _ - _ _
--
. -.-
. - --
. -.
-.
..
.. -. - -
!.
,
7 3-7
-
,
- j IE Rpt. Us. 50.. /76-12 III-2
..>
>
.
,
/
The inservice inspection effort has been contracted to B&W.
B&W Construction Company has developed a QAM applicable to the ISI which parallels the requirements of 10 CFR 50 Appendix B.
It provides for procurement control, calibration of NDE instru-ments, training of personnel, certification of materials, audits and personnel certifications.
B&W procedures applicable to this ISI are:
Procedure Numbers OA Area 9QA201 Organization 9QA215 Audits 9A167 thru 9A171 NDE Personnel Qualification and Certification
!
9QA205 Document Control 9QA212 Corrective Action
'
9QA210 Calibration i
9QA214 Records The requirements of the QAM are further supplemented by detailed information contained in individual ISI procedures such as calibra-
.
i tion procedures and personnel certifications.
.
4.
ISI Program The requirements applicable to the Oconee 3 inservice inspection are specified in Section 4.2 of the Technical Specifications.
This references the ASME Code Section XI, through the Winter of 1970 Addenda as the governing code with certain exceptions as stated in 4.2.2 of the Technical Specification.
"
plan to be consistent with the ASME code and applicable Regulatory Guides.
The plan, approved by DPC, outlines examinations to be performed over the 10 year ISI interval based on nine planned outages. Section 2 of the plan contains a listing of all applicable weld areas, or par +. in each Class I category subject to examination per the ASME code. Section 3 of the plan contains a schedule of examinations which lists all examinations by specific number and the outages during which they are planned to be performed.
The ASME Code,Section XI,1970 edition lists Examination Categories in Table IS-251 and Methods of Examination in Table IS-261.
The inspector selected several items from the code and compared the code requirements with the ISI plan.
Table IS-251 in Section XI, Examination Category B specifies that circumferential welds in the vessel head shall be examined for five percent of the length of each weld during the inspection interval.
Table IS-261, Item 1.2 i
requires that these welds be examined volumetrically.
The ISI plan
/
e
=
,..
.,
...
.... -. -
.
-
. - -... -
....
.- ~
.
-
.-
_..
-
. :..
- ::
.
- - -.
-.. -...
-. _.
. -. -..... -.. - -.
!
.
['-
!l
')
'
IE Rpt. Ns. 50-2ur/76-12 III-3
-
p a
O, in Section 2 identifies this as Figure 1.2.7 and specifies an ultrasonic examination to be performed in accordance with procedures BLI-l and BLI-2.
Section 3 of the plan schedules the examination to be performed during the first outage and specifies the five percent of the weld as being from points 22 to 25.
The inspector also verified that code requirements were plarned to be met for the flange-to head weld (Figure 1.3.3) and closure nead studs (Figures
,
1.8.1, 1.8.2 and 1.8.3).
The plan also included steam generator tube eddy current examinations in accordance with Regulatory Guide 1.83 and reactor coolant pump flywheel examinations per Regulatory Guide 1.14.
Item 4.2.11 of the Technical Specification requires that two reactor coolant system piping elbows be ultrasonically inspected along their longitudinal welds for clad bonding.
The elbows to be inspected are identified in B&W Report 1364.
In resiewing this requirement, the inspector found that Report 1364 references only those elbows
'
for Oconee Unit 1 and that requirement does not apply to Units 2 or 3.
The corresponding elbow welds on Unit 3 are incorporated in the ISI plan per ASME requirements.
For example, 3Al Section Outside Long Seam Figure 4.2.4.3 and 3Al Section Inside Long Seam Figure 4.2.4.1 were planned for examination during this outage.
Section 4 of the ISI program plan contained the procedures necessary
'to conduct the examinations and record data to fulfill code and regulatory requirements.
The inspector examined the following procedures to verify applicable code requirements had been incorporated in then:
Procedure Procedure Number Title BLI-l Ultrasonic Examination of Similar Metal Weld Seams and Attachment Welds BLI-22 Dye Penetrant Inspection of Welds ISI-252 Magnetic Particle Examination of Reactor Coolant Pump Motor Flywheels ISI-401 Eddy Current Examination of OTSG
.
Tubes Several other procedures including BLI-2, ISI-64 and ISI-403 were reviewed during the course of the inspection due to interfacing requirements.
.
w..
'
.
b
_f -.
. -.
,
...
...
...
.
.
.
,_
..
~
,_
.
.
..
_ _. _ _
-- --
.
.
._
.
. - - - _.. -.
_-
-
.
.
.j}
r}
'
'
.
.
!
IE Rpt. No. 50-287/76-12 III-4
~
!
'
-
.
P
,-
The procedures appeared to be adequate and addressed code require-ments such as calibration, calibration blocks, equipment character-istics, data records and personnel qualifications. Acceptance criteria for examination were specified and reportable conditions were defined. Procedure ISI-64 delineated a means of reporting potentially unacceptable conditions.
The interface responsibilities between B&W and DPC was not adequately defined with regard to dispositioning of reportable indications.
The licensee reviewed this and revised Procedure PT/3/A/200/22 to incorporate QCK-1 nonconformance control procedure, to formalize the disposition of reportable indications.
5.
Observation of Work i
The inspector observed B&W personnel perform ultrasonic examination on two reactor head welds and several pressurizer nozzle-to-vessel welds.
ISI plan Figure 1.2.7 is the circle seam closure head ring to center disc weld.
It required a voir atric examination (UT) per BLI-1 and BLI-2 between points 22 and 25.
The inspector observed the UT scans, verified such items as scan overlap, technique accepta-bility, data recording, and calibration verification.' The inspector
,
verified that the weld area scanned was identified as points 22 to i
25 (previously marked by low stress stamps during PSIl and correlated with drawing B&W 149916E.
The examinations (;0, 45 and 60 ) were
conducted with instruments 12014, 12017 and 12015 using Hamikleer couplant.
The reactor head weld examinations were restricted or limited due to the head geometry.
For example, a large lifting lug and part of
,
the service structure which were integral to the head prevented performing all the scans required by the procedures.
The NDE
.
instrument and transducer identification and calibration were l
verified and cross checked against calibration records and data j
sheets.
Personnel performing the examinations were certified as Level I and Level II examiners in accordance with SNT-TC-1A and B&W-procedure 9A 169 with the certification documented on form P05 32102-2.
-
'
l l
DPC personnel from the station QA staff and corporate QA staff l
audited / monitored the above examinations.
In addition, several I
DPC auditors from the QA audit groups conducted part of a Level II audit during these examinations.
The inspector also witnessed UT examinations of the closure head-to-flange circle seam weld and pressurizer nozzles.. No deviation from procedural steps or code requirements were noted.
-
.
'w)
%*
.
-
i i
.
.!
-
-
J
.
-
-
a
_ _ _ _ _ _ _ _
.
,
-
1:
.
y ~}
r7
,
!
s'
.J l
IE Rpt. No. 50-287/76-12 III-5 AV The inspector observed B&W personnel conducting eddy current examina-tion of the steam generator tubes. Approximately 480 tubes, which exceeds the three percent requirement of RG 1.83, were scheduled for examination as specified in procedure PT/3/A/200/22.
The examinations witnessed were conducted on steam generator 3B tubes36-100, 36-103,36-104, 36-105,37-100, 37-101, and 37-102 in accordance with ISI-401 which conforms with RG 1.83.
Previous examination of the OTSG tubing had been performed with the steam generator drained.
B&W conducted several examinations on the same tubes with the steam generator drained and with it partially full of water. No differences in data were observed.
The procedure was revised to permit examination of the tubes with the generator partially filled with water and the probe drive system was modified to minimize removal of the contaminated water.
The examinations were conducted by Level II examiners whose certifications were in conformance with SNT-TC-1A requirements. Equipment met the technical requirements of RG 1.83 and objective evidence of calibration 'as available. No departure from procedural requirements was noted.
6.
Records The B&W procedures which define record requirements are 9QA208, 9QA214 and ISI-64. DPC record requirements are primarily delineated in procedures QA504 and QA502.
The inspector audited records, data sheets and calibration sheets applicable to inservice activities which included verification of personnel certifications, material certifications, equipment calibration and data records.
Personnel certifications were available for all personnel performing examinations.
Certifications for ultrasonic examiners were based on SNT-TC-1A recommendations and conformed with B&W procedure 9A169.
Form PDS 32102-2 included documented evidence of training, experience, and current eye examination.
No discrepancies were noted during the review of these personnel certifications.
Couplant materials used during the ultrasonic examinations were procured from Hamikleer and identified as Batch No.1711.
Certifi-cations were available attesting to less than 50 ppm of halogens and sulfur constitutents in the couplant.
,
The NDE instruments used in performing ultrasonic examinations were identified as serial numbers 12014, 12017 and 12015.
Records were available which provided c5jective evidence of calibration of the instruments.
Records conta!ned quantitative information related to horizontal linearity, verticci linearity, attenuation, and resolution.
v
.
e
_
......
_
, _.. _.
...
...,,
.... _ _ _.
. _. _
_
,. _
__
_
.
.-.
.
.
._
_._.-__m
..
_ _ _ _. _ _ - _ _ __
_.. _
_ _.
'
f
'
.
f IE Rpt. No. 50-287/76-12 III-6
-
-,
'
l l
Calibration and data sheet records, see table below, were evaluated for the ISI exam and compared with the PSI data.
Calibration Sheet Data Sheet Weld / Component 090019, 090020, 090021 Figure 1.27 Closure Head 090019, 090020, 090021 Figure 1.33 Closure Head 090001, 090002, 090003 Figure 4.2.4.3 3Al Suction (OLS)
N/A Figure 4.2.4.4 3Al Suction (OLS)
090001, 090002, 090003 Figure 4.2.4.1 3Al Suction (ILS)
N/A Figure 4.2.4.2 3Al Suction (ILS)
N/A Figure 1.8.1 STUDS 090040, 090036, 090044 Figure 1.8.2 STUDS N/A Figure 1.8.3 STUDS N/A Figure 1.8.3.1 STUDS The times, dates and examiner data on the calibration sheets were consistent with the entries on the data sheets with the exception of an isolated data sheet serial number which was corrected immedi-ately.
The data entries appeared to be consistent with the conditions observed during the examination.
For example, limited scan areas identified on the data sheet were confirmed by visual observation of physical ILuitations.
The inspector compared the ISI data with the PSI data, the latter being recorded in June of 1972.
The data correlated quite well probably due to the fact that the areas examined were relatively free of indications.
Limited scan areas were identified in the same areas during both examinations and no significant changes from PSI to ISI were detected.
However, there apparently was an error in the data with regard to the head thickness which was recorded as 7.5 inches on the baseline and 6.2 inches during ISI.
This discrepancy was also detected by the B&W computerized data bank when entered and checked automatically.
A re-examination of the head was conducted and found to be 7.5 inches.
-
Several other instances of data entry inconsistencies were noted.
For example, on calibration sheet 090003 for 3Al Suction Outside Long Seam, 60 u1* * onic examination, the instrument serial number does not correspond to the instrument used to conduct the examina-tion which is recorded on data sheet 4.2.4.1.
This was discussed with the licensee and audits were made available which showed that the licensee had identified the discrepancy.
For the situation discussed above, page 15 of the audit specifically identified the inconsistencies between the calibration and data sheets.
The inspector reviewed additional pages of the licensee data audits and
_
$
gri -
e
- .,**
= ' * * * *
- - - *
,
_ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _
._ _ _ _ _ _ _ _. _. _
_ _ _ _ _ _
___ _ __ _
_ _ _ ___
__
..
. _, _
. - -. ~.
. -.-.
- -
--
- --
IERpt.No.50-Q.,//76-12 r}-
,
,
.
i III-7
-
.
,
,
f
!
{
,
)
i i
I noted that they were comprehensiv.e and effective in identifying
'
.
'
inconsistencies and/or problems areas.
The resolution of identified discrepancies between the licensee and B&W was in process and being monitored by DPC QA personnel.
The B&W data for ISI is entered into a computer data bank for storage and checking.
The data entered by field personnel includes the leader information, calibration data, examination scan informa-tion and examination results.
The computer compares the inputted
!
data /information with the information in storage and then prints l
out any discrepancy between the two thus providing a check on the i
ISI crews and/or procedures. As an example, the data entered for Figure 1.2.7 immediately detected and flagged the invalid head thickness.
In addition, the printout also indicated invalid beam direction and invalid surface both of which were readily explained by limited scans.
7.
Fire Prevention The inspector observed that many scheduled modifications to plant.
systems were in process during the plant outage.
For example, a modification to the steam generator recirculation system to permit recirculation, sampling, and chemical additions to the steam generators
on an individual basis was being accomplished under NSM-00-31-S.
I The inspector observed part of this work being performed in the
-
basement of the Unit 3 containment on September 23, 1976. The work crew was welding materials and cutting with o' pen flames with apparently i
no fire extinguishers in the immediate area.
The inspector questioned several of the crew members and the foremen about the lack of fire
.
extinguishers and the location of the Flame Safety Permit. The crew members stated that there was no fire extinguisher in the area
,
and the Flame Safety Permit was posted on the personnel hatch.
The
!
crew immediately obtained a fire extinguisher from elsewhere in the
,
containment. ONS Station Directive 5.1.4, Welding and Burning
!
Procedure, paragraph 3.1 requires a completed Flame Safety Permit in containment areas and paragraph 4.2.8 requires portable fire extinguisher to be concentrated at the work area.
Upon exiting the containment at approximately 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />, tihe inspector observed that the Flame Safety Permit had been posted on the personnel hatch as allowed by procedure, but the fire watch had signed the form in block 9.
Block 9 is a sign-off by the fire watch inspector to verify that a fire inspection was conducted 30 minutes af ter completion of each work period.. Subsequent investiga-tion by DPC QA and Safety personnel revealed that the individual had signed it in the wrong place.
a
-
.
S
,e
,
..
g
ed.
san t-p eqeeq.
e N
_ _ _ _ _
_ _.._
__
_ _.-
._
_ _ _
_ _.. _
_._ _ _ _
i a
a
- -
-
.
IE Rpt. Ns. 50 _d7/76-12 III-8 i
.
,
,
/*g
'
\\)
,
.
ONS Station Directive 5.3.3, Reporting of Fire Protection Impairment, paragraph 4.2 specifies that the Safety Supervisor or his assistant will be notified before any piece of Firefighting Equipment is removed from its designated place for any reason, except for the purpose that it was intended (fighting fires).
In isolated areas l
where a fire watch requires a fire extinguisher for welding or burning permits, extinguishers will be checked out from Supply Department for this purpose.
,,
The inspector observed on September 23, 1976, that extinguishers had been removed from designated areas on the refueling floor and in the basement of the containment.
In addition, on one extinguisher the safety pin had been pulled and it appeared that the extinguisher had been operated. DPC Safety did not have any record of impairment or removal of extinguishers as required by paragraph 4.2 of Station
,
Directive 5.3.3.
DPC Safety investigated further and found that the pins on all extinguishers in the containment basement had been removed but that none had been discharged (verified by weighing).
.
Audit 76-5 dated February 10, 1976, was conducted by DPC to determine compliance with Administrative Procedure 38 (~ superseded by Station Directive 5.1.4).
It identified a problem wherein construction supervisors failed to sign Flame Safety Permits.
The cause was determined to be inadequate training of temporary construction personnel in station procedures. DPC conducted training exercises to rectify the situation and a re-audit confirmed the effectiveness of the corrective action.
The list of construction personnel receiving training in ONS procedures did not contain the names of those on the crew in question.
The above items are summarized as:
(a)
Failure to locate extinguishers in an open flame area per paragraph 4.2.8 of Station Directive 5.1.4.
(b)
Signing of Flame Safety Permit in a manner contrary to Station Directive 5.1.4
(c)
Impairment / removal of fire, extinguishers and failure to report such events per Station Directive 5.3.3, paragraph 4.2.
All represent an apparent failure to follow Station procedures and thus appear to be a noncompliance to Criteria II and IV of 10 CFR 50, Appendix B.
-
O
n _
..
...
..n.
.. -..,. -..
._,
.
,. - -
-
...
-.
..
- -
-
,,_
.,
,
,)
'
(
}
-
.
l IE Rpt. No. 50-287/76-12 III-9
,
fs t
I
.
Criterion II states in part that activities affecting quality shall be accomplished under suitably controlled conditions and Criterion V states that activities affecting quality shall be prescribed and accomplished in accordance with documented instructions, procedures, or drawings.
Section 17.2.2 of Duke-1 Topical QA Program states that these procedures and instructions may be contained in manuals, station procedures and directives, administrative instructions and/or other documents.
This is implemented by station directives among which are the previously identified Station Directives 5.1.4 and 5.3.3.
Contrary to the above, welding activities on a safety related modification were not accomplished in accordance with the requirements of the directives.
This is an apparent noncompliance of the infraction category and is identified as 76-12-Al (II).
,
p
.
.
m.
,en e
mm * ~ +
qm m m mqq.
--
.
. _. _
_
_
_
._..
- _ _..
,
,
{
h b
s
IE Rpt; No. 50 *w//76-12
.IV-1
'
I
'
o
.
y l
DETAILS IV Prepared by:
'u
- L Ll1
'/ Y N'
,.4J. J. Blake, Metallurgical Engineer Date
Engineering Support Section No. 2
-
'
j Reactor Construction and Engineering
,
l Support Branch Dates of Inspection:
cto 8, 1976 Reviewed by:.;
) AhWp/
-
//
(
A. R. Herd @ ection Chief M
[Dap
Engineering Support Section No. 2
'
Reactor Construction and Engineering Support Branch 1.
Persons Contacted
.
Duke Power Company (DPC)
'
J. E. Smith - Site Manager O. Bradham - Maintenance Supervisor
'
-
j L. Wilkie - Maintenance Engineering Support J. W. Cox - Site QA Manager i
j J. Dunlop - Site QC Engineer j
C. Freeman - Senior QC Inspector, NDE
.
2.
Scope
,
The purpose of this inspection was to review the licensee's program
'
relative to welding and nondestructive examination (NDE) activities during maintenance and/or modification operations on safety related systems. 'The piping systems at this plant were fabricated to either USAS BIl.7 1968 Edition with June 1968 Addenda or USAS
'
B31.1.0 as described in Paragraph 1C.3.1 of the FSAR.
3.
Welding
This part of the inspection involved a eview of the station directives and procedures involved with welding activities; observation of l
.
welding activities in process; review and inspection of welding filler material controls; review of welding documentation; and
,
review of welding personnel qualifications.-
J The directives and procedures reviewed during this inspection in
'
cluded the following:
-
'
.
.
~~~
.,,,..
.....
.
~~-
,.
.
~. -,
_
.
q.
..-
-
-
,
-
.. :
-
u- - -,,,,
.
. -, -
.
-
.
,
}
(
.
s IE Rpt. No. 50-287/76-12 IV-2 (I)
-
,
I a.
Oconee Nuclear Station Directive 3.6.l(M) Welding Program b.
MP/0/A/1810/1 - W' elder Qualification Records c.
MP/0/A/1810/2 - Identification and Control of Welding Materials
~
d.
MP/0/A/1810/3 - Identification and Control of Field - Fabricated Pipes and Welds e.
MP/0/A/1810/4 - Heat Treatment of Welds f.
QC D1 - Designating and Maintaining Cleanliness Level in Areas Containing Nuclear System Components g.
QC D2 - Cleanliness Control of Piping Systems (Primarily Field Fabricated) at Oconee Nuclear Station i
h.
QC F2 - Piping System Layout and Hanger Installation Inspection 1.
QC G1 - Receipt, Inspection, and Control of Materials, Parts,
,
and components Important to Nuclear Safety j.
Welding Procedure L-200 - Gas Tungsten Arc delding Process Specifications
k.
Welding Procedure L-300 - Shielded Metal Arc Welding Process Specifications
~
The welding activities observed during this inspection included the following:
-
Shop welding of 1 1/2-inch diameter carbon steel pipe tail a.
pieces to various valves for the modification of the steam generator drain valve piping.
This work involved the Gas Tungsten Arc welding of socket welds.
b.
Field welding of a 6-inch diameter carbon steel backing ring butt weld joint in the Emergency Feedwater System.
The welding observed was a repair weld attempt on Joint No. lE of Isometric No. 97.
This welding was the Shielded Metal Arc welding process.
During the observation of the weld repair activities on the weld joint in the Emergency Feedwater System, the inspector performed a visual examination of two adjacent welds, Nos. 1C and 1D.
These welds had been shop welded prior to the outage to make up a spool piece for installation, thereby minimizing the inplace fabrication effort.
.]
.
.
,,
.%
s..
....
..
~..
_.. -
__
..
--
,. -
_. _,
. _.
_
- -
_ _
__
'
-
-
~
j.
,
,]
)
,
IE Rpt. No. 50..//76-12 IV-3
/
'
m).
.
<
,
I An inspection was made of the welding material issue station to observe storage conditions, temperature controls on electrode holding ovens, and the operations of issuing new materials and dis-position of returned materials.
The following documentation was reviewed during the course of this
.
inspection:
a.
In-process documentation of welding operations observed during this inspection.
b.
Weld history records for completed welds inspected.
,c.
Welding personnel certification for all welds inspected.
,
d.
Filler material documentation for all welds inspected.
e.
Material certifications for piping. materials bef.ng installed.
There were no items of noncompliance noted within this area of inspection.
4.
Nondestructive Examination (NDE)
<
This part of the inspection involved a review of NDE procedures, documentation of inspection results, and personnel qualification certifications.
The procedures reviewed included the following:
a.
QC H1 Visual Worknanship Standards for Welds b.
QC Il Documentation of Miscellaneous NDE c.
QC 12 Liquid Penetrant Examination Technique d.
QC 15 Alternating Current Yoke Method Magnetic Particle Inspection Techniques e.
QC 16 General Radiography Procedure f.
QC I7 Direct Current Prod Method Magnetic Particle Inspection Technique g.
QC 19 Liquid Penetrant Examination Technique (Water Washable)
e I
\\
-
l
.
- 9
ge v=
e
,m
- - em m
~**
,
,
.- --
,
,
v
i
)
- )
-
.
.
l IE Rpt. N3. 50-z87/76-12 IV-4 i
l
/-')
-
.
l i
.
h.
QC K1 Control of Nonconforming Items 1.
QC K2 Work Stoppage
'
J.
QC K3 Corrective Action k.
QC L1 Procedure for Inspection of Nuclear Safety Related Field Fabricated Pipe and Welds (Oconee Nuclear Station Only)
1.
QC L2 Densitometer Adjustment and Calibration Check m.
QC L3 Densitometer Calibration MacBeth Model The NDE documentation reviewed consisted of the final inspection packages including radiographs for the shop welds No. 1C and 1D and in-process inspection records for Field Weld No. lE in the Emergency Feedwater System.
During the review of the NDE procedures, the inspector noted that acceptance standards listed in the procedures were referenced to fabrication codes without listing the date of issue of the referenced codes.
The inspector informed the licensee that the question of
,
what issue of the code was applicable to the work at the station would be listed as an unresolved item.
The licensee agreed to resolve the matter and revise the procedures.by January 1, 1977.
There were no items of noncompliance in this area of inspection.
,
.
l l
'
,
_
)
.
'
-
_.. _ - -. - -. ~. - - -
w
,
-.
..
.
..
.
_.
_
)
' '
'
.
i j,
IE Rpt. No. 50-287/76-12 V-1
'
'
v
".
DETAILS II Prepared by: h
/ooq M
//hr-M <
P. T. BuFne~tt, Reactor Inspector Date Nuclear Support Section s
Reactor Operations and Nuclear Support Branch Dates of Inspection: October 4-7, 1976
!
.
Reviewed by:
6ww
"/
l-
'
H. C. Dance, Acting Chief Date
'
Nuclear Support Section Reactor Operations and Nuclear Support i
Branch 1.
Personnel Contacted J. E. Smith - Manager, Oconee Nuclear Station O. S. Bradham - Superintendent of Maintenance
L. E. Schmid - Superintendent of Operations
J. W. Hampton - Manager, Administrative Services
R. T. Bond - Performance Engineer i
T. S. Barr - Technical Services Engineer j
G. Findley - Maintenance Engineer i
J. N. Pope - Operating Engineer
2.
Previously Unresolved Item
1
As reported in Inspection Report No. 50-287/76-09, the procedure records for Units 1 and 2 did not confirm that the systems were
'
leak tested at 2285 psig following opening of the primary systems.
,
During this inspection the shift supervisor and reactor operator i
logs for the periods of concern were reviewed by the inspector.
These records did confirm that the tests were performed as required.
The quality of the log records varied quite extensively from one
!
individual to another, and in some cases entries in both logs were
'
required to ascertain that the test had been properly conducted.
The station directives on recordkeeping and keeping of logs were reviewed and discussued with management.
The inspector expressed the view based on review of directives and of the logs that the former were not adequate to assure that sufficient records of station activities were maintained.
He pointed out that records are required of activities required by technical specifications and that the system in use had nearly lost the records of the pressure
-
test.
.
uf
,
.
.
,
, ge,
et e
e-We -*
- f
"
- ' *,
'
_
.. -. _
. -. - - -
--,
~
--
--
-
-
--.
--2-m-
2 yap,
-
gr v.
--w--
e
.-
..
.
-. -
-
-.
-
-
-
-
D
'
.
)
e, I
IE Rpt. Ns. 50-287/76-12 V-2 j
\\ _3 " *
/v
,
f
/
i
.
,
3.
Preparation for Refueling By review of OP/3/A/1502/03, " Fuel Unloading and Reloading," and
.
OP/A/0/ 1503/1, " Preparation for Refueling," the inspector verified that all of the requirements of Technical Specification 3.8,
" Refueling," were addressed in procedures with the exception of part 3.8.8.
That specification addresses the requirements for separation of fuel bundles being moved simultaneously in the refueling canal.
The planned sequence of operations appeared to preclude violating that technical specification. The latter procedure also addressed i
the checkout of refueling equipment and testing of the refueling interlocks.
Crane testing in accordance with ANSI-30.2 was performed by comple-tion of MP/0/A/3000/10.
The inspector confirmed that OP/3/A/1104/06,
" Spent Fuel Cooling System," had been performed to satisfy the i
requirements of Technical Specifications Table 4.1.
'
l 4.
Fuel Handling h
The inspector observed the activities in the spent fuel pool area.
None of the planned activities of removing burnable poison rods and
'
replacing them with orifice assemblies were in progress due to the
,
failure of a handling tool.
That situation continued throughout the inspection. However, the inspector did observe part of the activity of moving spent fuel from the Unit l' and 2 pool to the Unit 3 pool. Since the entire evolution takes about 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> only a i
small fraction of it was observed.
The inspector noted that the
-
fuel handling crew consisted of a senior reactor operator, a reactor operator and two auxiliary operators.
The inspector observed
that egress from the spent fuel pool was a slow process under normal conditions because of security and safeguards requirements.
,
'
Provision exists for rapid exit under emergency conditions.
However, there were no posted instruction on how and when to use that provision, nor was emergency egress addressed in the appropriate emergency procedure EP/0/A/1800/13 " Spent Fuel Damage." This was discussed with the licensee, who committed to remedying the situation consistent with safeguards and security requirements.
,
In the control room che inspector confirmed that the master copy of OP/3/A/1502/05 was complete and up-to-date.
However, the master
,
copy of the Preparation for Refueling Procedure was not up-to-date in the control room; varicus enclosures to that procedure were still at the locations for which the enclosures applied such as fuel handling bridges.
The inspector did confirm by discussions with the appropriate shift supervisor that he had in fact confirmed thac all enclosures to the procedure were completed
,
s-
e I
,...,
.. _. _
.._
'l i
r e.,-
=
n..,r,.
,..
.
...
, -
,-
-
--e,
,._.
n
, ---
-. _
_
_ -_ _ _ - _ _ _ _.
.-. _-
-
_ -.. -
-.-.- - - ---
-.~
---
~--
.
.,
,
...
.
IE Rpt. No. 50-287/76-12 V-3 before signing off the appropriate step in the Fuel Unloading and i
Reloading Procedure. The inspector confirmed from records in the control room that core monitoring and containment integrity were maintained as required during fuel handling, and that the boron concentration of the primary system and spent fuel pool was being monitored with the required frequency.
5.
Maintenance During Refueling
.
l The inspector reviewed the modification package, SMR-0-282-S for installing flanges on lines to the relief valves in the LPI System.
!
,
An approved procedure was provided as part of the package of informa-tion. It included QA inspection hold points and provisions for nondestructive examination.
The inspector also reviewed maintenance procedure MP/0/A/3000/24 i
for the removal and installation of the, hydraulic snubbers. All 106 of the snubbers in Unit 3 were to be removed, repaired and replaced during this outage.
The inspector witnessed several stages of the snubber repair including decontamination of the snubbers being removed from the containment building. He also witnessed repair of the snubber valves and the replacement of "O" ring seals at one work station; the flushing of the hydraulic fluid
,
from the snubber and reservoir, where it was an integral part of
'
the snubber, and replacement with new fluid at a second station; and, finally, the testing and setting of the stroke of the snubbers at a third station. Work performed at the third station was under continuous QA surveillance, and the personnel performing the work had received 3 days instruction in the use of the testing machine.
Both the machine and instruction were provided by the vendor of the
,
snubbers. Snubber repair activity was carried out under MP/0/A/3000/30.
>
!
l
'
-
.
'wW e
.
..
. -.
-..
_
.... -
a
._
_.
-,
.-
,.
-,
.
,,