ML19340A603
| ML19340A603 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 05/24/1973 |
| From: | Charles Brown, Dance H, Maura F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19340A601 | List: |
| References | |
| 50-010-73-02, 50-10-73-2, 50-249-73-03, 50-249-73-3, NUDOCS 8008280655 | |
| Download: ML19340A603 (24) | |
See also: IR 05000010/1973002
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U. S. ATOMIC ENERGY COMMISSION
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DIRECTORATE OF REGULATORY OPERATIONS
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REGION III
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RO Inspection Report No. 050-010/73-02
RO Inspection Report No. 050-249/73-03
Licensee: Commonwealth Edison Company
P. O. Box 767
Chicago Illinois
60690
Dresden Nuclear Power Station
Licenses No. DPR-2
Units 1 and 3
and No. DPR-25
Morris, Illinois
Category: C
Type of Licensee:
Type of Inspection:
Routine, Unannounced
Dates of Inspection:
April 17, 19, 20, 23-25, 1973
Dates of Previous Inspection: Unit 1 - February 7, 8, 9, and 16, 1973
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Unit 3 - March 22, 1973
N C Lw/
F. A. Maura
f24/7,5
Principal Inspector:
'(Date)
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Accompanying Inspector:
C. Brown
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Other Accompanying Personnel: None
hC.k%w
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Reviewed By:
H. C. Dance, Senior Reactor Inspector
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BWR Operations Section
'(Date)
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SUMMARY OF FINDINGS
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Enforcement Action
A.
Deviations from Unit 1 Technical Specification surveillance testing
requirements were as follows:
1.
Contrary to paragraph B.16.f(1), the voltage of the pilot cell and
the temperature of the adjacent cells were not measured during the
weeks of January 1-7, 1973, and January 15-21, 1973.
(Paragraph 6.a)
2.
Contrary to paragraph B.16.f(l), the specific gravity and voltage
of the pilot cell, temperature of the adjacent cells and overall
battery voltage was not measured for the week of February 19-25,
1973.
(Paragraph 6.a)
3.
Contrary to paragraph B.16.e(l), the licensee did not have docu-
mentation to show that the Unit 1 diesel generator was started and
loaded to full load output and maintained in that condition until
the diesel engine and the generator reached equilibrium temperature
during the month of December 1972.
(Paragraph 9.a)
B.
Deviations from Unit 1 Technical Specifications reporting requirements
were as follows:
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1.
Contrary to paragraph J.3.a, the licensee failed to report the
bypassing of in-core monitoring string 113, an abnormal occurrence,
to its Production Department in a prompt manner.
(Paragraph 4.c)
2.
Contrary to paragraph J.3.a
the Station Review Board failed to
review the bypassing of in-core monitoring string 113 in a prompt
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manner.
(Paragraph 4.c)
3.
Ccatrary to paragraph J.5, the licensee failed to report to the
Cottission the fact that it operated with the emergency condenser
shell side water temperature greater than 100*F during the week
of March 5-11, 1973.
(Paragraph 3.f(l))
C.
Deviations from Unit 1 Technical Specifications operating requirements
were as follows:
Contrary to paragraph J.2.a(9), the licensee is operating without a
detailed written procedure, approved as specified in the Technical
Specifications, to assure the safe shutdown of the plant in the event
of a flood designated as a Probable Maximum Flood.
(Paragraph 2.a(16))
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Licensee Action on Previously Identified Enforcement items
A.
The corrective actions listed in the licensee's response to our letter
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of enforcement dated October 19,
972, concerning the installation of
the core spray system, were reviewed during this inspection. The items
are considered resolved.
(Paragraph 8.a)
B.
The corrective actions taken by the licensee concerning the items of
noncompliance identified in our letter of January 18, 1973, were reviewed.
The corrective action on the diesel generator breaker failure has been
completed satisfactorily.
It was noted that their commitment to report
events in accordance with Technical Specification requirements was not
carried out.
(Paragraphs 9.b and 3.f(l))
Unusual Occurrences
A.
Core spray valve CS-ll failed to open during surveillance testing.
(Paragraph 8.b)
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B.
Primary steam line isolation valve M0-169 failed to close during unit
shutdown.
(Paragraph 3.d)
C.
Emergency condenser shell side water temperature- exceeded 100*F with
reactor at pressure greater than 140 psig.
(Paragraph 3.f(l))
D.
Emergency condenser automatic initiating signals reduced during plant
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shutdown.
(Paragraph 3.f(2))
E.
In-core string 113 scram function bypassed during plant operation.
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(Paragraph 4.c)
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Otner Significant Findings
A.
Current Findings
1.
Organir cion
Effective April 2, 1973, W. Hildy replaced C. Weber, who has
terminated his employment with Commonwealth Edison Company, as
Instrument Engineer. Although Mr. Hildy will not meet all the
requirements of Section 6.1.D.5 of the Technical Specifications
until October 1973, the instrument group has two foremen who
meet all the requirements of Section 6.1.D.5 for the position
of Instrument Engineer.
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Effective June 1, 1973, J. Diederich will report to N. Kershaw at
the corporate office as a member of the Nuclear and Fossil Systems
Group.
Mr. A. Roberts, presently Operating Engineer for Unit 1,
will replace Mr. Diederich as Supervisor, Technical Staff. He meets
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the requirements of Section 6.1.D.4 of the Technical Specifications.
Mr. T. Watts, presently Operating Engineer for Unit 3 and holder of
Senior Operator's License on all three units, will replace Mr.
Roberts as Operating Engineer, Unit 1.
Mr. D. Scott, presently a
Shift Foreman, will replace Mr. Watts as Operating Engineer, Unit 3.
Mr. Scott has a Senior Operators License for Units 2/3.
Mr. G.
Abrell remains as Operating Engineer, Unit 2.
All three Operating
Engineers meet the requirements of Section 6.1.D.3 of the Technical
Specifications.
2.
Unit 1 continues operating at approximately 115 mwe (approximately
58 percent power) with a chimney gaseous release rate of approxi-
mately 44,000 uci/sec.
3.
Unit 3 has complaced core reconstitution. Return to power operation
is planned by end of May.
B.
Status of Previously Reported Unresolved Items
On March 21, 1973, the licensee submitted the analysis of stresses on
Unit 1 safety valves. The results indicate no problem exists. This
item is considered resolved.
(Paragraph 3.e)
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Management Interview
The following subjects were discussed at the conclusion of the inspection on
April 25, 1973, with Messrs. W. Worden, Station Superintendent; F. Morris,
Assistant Station Superintendent; A. Roberts, Operating Engineer, Unit 1; and
C. Sargent, Engineer.
A.
The inspector stated that the licensee's actions to correct the nine items
of nonconformance, with regard to the installation of the core spray
system, were reviewed and that he had no further questions at this time.
However, it was noted that the station records do not include the latest
"as built" drawings of the system. Also, he stated that it is our
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understanding that the QC Engineer will review the QC welding record
requirements against what is maintained by some of the piping subcontractors
for applicability into the QC program.
The licensee indicated that the latest as built drawings would be obtained
and that a review of the QC welding record requirements will be performed.
(Paragraph 8.a)
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B.
The inspector stated that the licensee's actions concerning the noncom-
pliance items covered in our letter of January 18, 1973, were reviewed.
It was noted that although both items have been resolved, the licensee
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failed to report to the Commission in a timely manner several days during
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which the energency condenser temperature exceeded 100'F which was in
violation of the Technical Specificctions, and contrary to the comaitment
made in response to our enforcement letter.
(Paragraph 3.f(l))
C.
The inspector stated that we have received and reviewed their safety
valve etress analysis report for Unit 1, and that we have no further
questions at this time concerning the results of the stress calculations,
but that we question the use of Proposed Code Cases in their analysis.
The licensee indicated our comments would be passed on to their Engineering
Department.
(Paragraph 3.e)
D.
The inspector noted that a review of Unit 1 surveillance records showed
that problems similar to those found during the last Unit 2/3 inspection
continue to occur. For example, some of the data sheets have misaing
readings, lack date and signature of data taker and reviewer, etc. As
a result of the review, the following violations of Technical Specification
requirements were noted:
1.
The weekly checks for the 125V battery bank were not done during
the week of February 19-25, 1973.
2.
During the weekly check for the 125V battery bank, the voltage of
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the pilot cell and the temperatures of the adjacent cells were not
teken during the weeks of January 1-7, 1973, and January 15-21, 1973.
3.
The surveillance test records for the December 1972 monthly diesel
generator test could not be found.
(Paragraphs 6.a and 9.a)
E.
With regard to the bypassing of the scram function of in-core string 113
on April 8-9, 1973, the inspector stated that the review indicates
inadequate procedures appear to have contributed heavily as a cause of
the event, and that we note the procedure has been corrected, but that
we question how events such as this one are incorporated into the retraining
program in an effort to reduce or eliminate the number of human errors.
The inspector stated that he would look into this during a future inspec-
tion.
(Paragraph 4.c)
F.
The inspector noted that the bypassing of in-core string 113, an abnormal
occurrence, had not been reported to the Manager of Production or his
delegated alternate in a prompt manner. Assuming that the Superintendent
of Production, Division A, is the " delegated alternate," his notification
did not occur until approximately 63 hours7.291667e-4 days <br />0.0175 hours <br />1.041667e-4 weeks <br />2.39715e-5 months <br /> after the event took place.
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The licensee asked what the inspector considered " prompt not tr tent ton"
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and was informed that this inspector considered delays of up to eight
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to ten hours a2 satisfying the " prompt notificacion" requirement.
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(Paragraph 4.c)
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G.
The inspector noted that the Technical Specifications requirement for
an emergency procedure to cover the event designated as the " Probable
Maximum Flood" has not been complied with, 10 months since such
requirement has been in effect.
(Paragraph 2.a(16))
H.
With regard to procedures, the inspector stated that the abnormal pro-
cedures for the battery bank had not been reviewed in accordance with
our understanding of August 27,1971.1/ In addition, he stated that
the time to develop adequate procedures for conducting an orderly plant
shutdown, in case all de power was lost, was now and not during the event.
The licensee could not recall any understanding regarding the battery
bank abnormal procedures.
(Paragraph 2.b)
I.
The inspector stated that an inspection of the records covering the
present refueling operations on Unit 3 indicated that the fuel and core
reconstitution was in accordance with the submittal to Licensing dated
March 5, 1973, except that the number of new elements placed in the core
periphery was increased from 28 to 52.
The licensee indicated a revised submittal to Licensing correcting the
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total number of new fuel assemblies placed in the core was recently
approved by the SRB and should be in the mail soon. The transient
analyses as described in the Safety Analysis Report remain unchanged
by the increase in new fuel acsemblies.
(Paragraph 5.a)
J.
The inspector stated that he had reviewed the repairs made to the Unit 3
torus painted surface and as far as he could tell they were similar to
the repairs performed on the Unit 2 torus in 1772; also that he under-
stands all the baffles in the Unit 3 torus have been removed and that
the Unit 2 torus baffles will be removed during its next refueling outage.
The licensee agreed.
(Paragraph 7.a)
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RO Inspection Report No. 050-010/71-07
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REPORT DETAILS
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Personnel Contacted
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W. Worden, Station Superintendent
F. Morris, Assistant Station Superintendent
J. Diederich, Supervising Engineer, Technical Staff
T. Watts, Operating Engineer, Unit 3
A. Roberts, Operating Engineer, Unit 1
G. Abrell, Operating Engineer, Unit 2
R. Bishop, Enginaer, Technical Staff
J. Bowers, Engineer, Technital Staff
T. Suchocki, Engineer, Technical Staff
W. Jack 1w, QC Engineer, Technical Staff
J. Groth, Engineer, Station Construction
S. Hlady, Engineer, Station Construction
R. Williams, Engineer, Technical Staff
R. Cozzi, Engineering Assistant, Surveillance
R. Dyer, Maintenance Foreman
C. Lawton, Office Supervisor
C. Sargent, Engineer
H. Habermeyer, Shift Engineer
D. Simpson, Refueling Foreman
J. Sullivan, Nuclear Station Operator
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2.
Procedures - Unit 1
a.
Resolution of Comments Given to Licensee During Previous Inspection
During the November 28 to December 1, 1972, inspection,2/ a number
of comments were given to the licensee for resolution. During this
inspection the items were reviewed with the licensee with the
following results:
(1) The control steps given in Technfcal Specifications section
G.2.d are being incorporated to the plant startup procedure
as indicated by Action Item 73-81.
This item is considered
resolved.
(2) All check lists will be included as part of the manual pro-
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cedures (Action item 73-72)." This~ item is co'hsidered resolved.
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RO Inspection Report No. 050-010/72-06
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(3) Administrative Procedure ADM-I will refer to present Technical
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Specifications (Action Item 73-81). This item is considered
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resolved.
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(4) ADM-II will be expanded to include specific assignments and
schedules.
(Action Item 73-81) This item is considered
resolved.
(5) ADM-VII will be rewritten to include more guidance of how log
books are to be maintained and what records are required. During
a previous inspection station management had indicated this is
one of their high priority goals for 1973. We plan to review
this during future inspections.
(6) ADM-II will refer to the specific QC procedure or method for
documenting temporary changes (Action Item 73-81). This item
is considered resolved.
(7) Table 30-C covering SRB responsibilities will be corrected so
that it will agree with Technical Specification requirements
(Action Item 73-71). This item is considered resolved.
(8) ADM-VIII is being deleted under Action Item 73-175. This item
is considered resolved.
(9) ADM-IX Appendix A is not missing. The licensee had erronecusly
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labeled it as Appendix B.
This is being corrected. This item
is considered resolved.
(10) The review of all outstanding operating orders has been
completed.
(11) Chapter 37 of the Operating Manual is being prepared under
Action Item 73-73 and will cover all our concerns about radiation
controls. This item is considered resolved. We vill review
Chapter 37 after its completion.
(12) The licensee stated their procedures will be changed to ensure
the unit is scrammed immediately whenever a ground level release
occurs. This is covered under Action Item AEC-45. This item
is considered resolved.
(13) The unloading heat exchanger system has been removed from direct
contact with the service water system. According to the licensee
the heat exchanger is now cooled by the closed cooling water
system. The system procedures must be revised to reflect this
change. We will inspect this at a future date.
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(14) The licensee is developing procedures for removal and replace-
ment of the reactor vessel head piping to rer' lect the addition
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of the core spray system. This is covered under Action Item
AEC-42 and is considered resolved.
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(15) The licensee plans to review Safety Guide No. 33 and ANS-3.2
to determine if the inservice inspection program ~ procedures
are required as part of the Station or Unit Procedures Manual.
The licensee agreed to conform to the recommendation of ANS-3.2
and Safety Guide No. 33.
It is the inspector's position that
such procedures are recommended by the above documents.
(16) The licensee has failed to provide an emergency procedure for
the event designated as the " Probable Maximum Flood" as required
by Technical Specifications paragraph J.2.a(9).
This require-
ment has been in effect for over 10 months. More recently, on
March 12, 1973, the Directorate of Licensing turned down a
request from the licensee for a Technical Specification change
to delete the requirement for such procedures.
b.
Abnormal Procedures - 125V Batterv Bank
During an August 25-27, 1971, inspection,3/ the licensee agreed to
review the battery bank procedures for what steps should be taken,
such as plant shutdown, in the event the battery chargers are lost.
The inspector had stated that the procedures should include the
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point in time, af ter the loss of the chargers, when a controlled .
plant shutdown would be initiated.
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A review of procedures 9800 AN I, II and III showed that the licensee
had not performed the intended ?- iew.
In addition, it was noted
that 9800 AN III called for an orderly plant shutdown if all de power
was lost (chargers and batteries). When questioned, the licensee
stated that an orderly shutdown for such an event was preferable to
a scram because without de power many of the required breaker actions
have to be performed manually and time was needed to ensure the proper
actions were taken. The inspector determined that, although the
licensee considers such an event unlikely to happen and one which if
it happens requires manual actuation of much equipment, the licensee
had not considered developing a procedure which would outline what
equipment had to be operated, in which sequence, and from where in
order to conduct an orderly. plant shutdown. The licensee's plans
were to get out the drawings and start searching after the event had
taken place. The licensee has been requested to develop such an
emergency procedure now.
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RO Inspection Report No. 050-010/71-07
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3.
Reactor Coolant System - Unit 1
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a.
Leak Monitoring
The records on airborne activity in the containment sphere and its
seven compartments were reviewed for the period from January 1, 1973,
to April 20, 1973. The frequency of sampling and response to obtained
results was in accordance with previous commitments and the proposed
addition to the Technical Specifications.
b.
A review of the reactor coolant chemistry records for the period of
January 1, 1973, to March 31, 1973, was performed. Typical range
of values was: pH - 8.7 to 6.4
Conductivity - 0.46 to 0.26 umhos
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.07 to .01 ppm
The values are within the Itaits proposed by the licensee to be
included in their Technical Specifications.
c.
Steam Generator Tuba Leak Repair - Unit 1
A review was made of the leak history and repairs associated with
the four secondary steam generators. The units are vertical U-tube
generators and have 1801 tubes each. The tubes are made of 304L
stainless steel and are supported by baffles of carbon steel (SA-7).
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Since initial startup, the following number of steam generator tube
leaks have occurred and have been plugged.
Tubes Plugged Since Startup
Latest Plugging Number /Date
S/G A - 6 tubes
5 Tubes - 2/4/73
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S/G B - 32 tubee
12 Tubes - 10/17/73
S/G C - 1 tube
1 Tube - 1/21/69
S/G D - None
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Leaks were detected by mismatch in steam flow / feed flow, activity
increase in the secondary water of the affected steam generator and
by increase in the chimney release rate of I-131.
The particular
leaking tube is located by flooding the secondary side of the
generator and observing which tube water issues from. The tube is
then plugged and a primary side hydro is conducted at 1000 psig.
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An audit of the secondary water chemistry values for January and
March indicate the following range of values:
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pH - 6.4 - 8.3
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Cl - 0.2 ppm max
Conductivity - 0.5 umho/cm max
There is no chemistry limit imposed by the facility Technical
Specifications.
According to the licensee, steam generator tube leaks have not
been a major problem and there are no plans for an engineering
study of tube failure.
d.
Failure to Close of Primary Steam Line Isolation Valve
on February 4, 1973, during a normal shutdown of Unit 1, a signal
was given to the primary steam line isolation valves to shut. The
valves, MO-169 and MD-170, isolate the primary steam lines just
upstream of the containment penetration. Valve MO-170 shut, but
MO-169 failed to close. The investigation revealed that foreign
material had prevented the opening interlock from clearing in the
valve's motor operated breaker. Simultaneous operation of motor
operator in open and close direction is prevented both electrically
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and mechanically. Following repairs MO-169 operated satisfactorily.
M0-170 operator breaker cubicle was inspected for foreign material
and appeared in satisfactory condition.
The licensee plans to inspect and clean all similar breaker cubicles
and review overhaul procedures for breakers throughout the Dresden
Station. At the time of the inspection, neither of the above items
was complete. Both items are covered by Action Items No. 73-52 and
53.
The event was reviewed by the SRB during its 433d meeting and
reported to the Directorate of Licensing on February 14, 1973.
e.
Unit 1 Safety Valves Stress Analysis
The report on Unit 1 safety valves stress analysis indicates that
while the allowable stress limit is 17,220 psi, the highest calculated
stress level is 13,750 psi. A review of the analysis showed that
the licensee utilized Proposed Code Cases to ASME,Section III,
ND-7221, to calculate the valve reaction force, account for dynamic
impact, and for uncertainties in the maximum mass flow rate. While
the analysis appears to be satisfactory and we do not question the
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results, the licensee was reminded that Proposed Code Cases should
not be used until approved by the ASME Code Committee. This item
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is considered resolved.
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f.
Emergency Condenser
(1) Operation with Shell-Side Temperatures in Excess of 100*F
The corrective measures described in the licensee's letter
(Lee to Grier) dated February 7, 1973, were not satisfactorily
carried out. A record review indicated that from March 5, 1973,
through March 11, 1973, the licensee operated with a shell side
temperature greater than 100*F, but failed to report it in
accordance with paragraph J.5 of the Technical Specifications.
On March 12, 1973, a change in the Technical Specifications
allowing shell-water temperatures as high as 212*F was approved
by the Directorate of Licensing. The present limit cannot be
exceeded since the system is open to the atmosphere. The
highest recorded temperature so far has been 210*F on March 18,
1973.
(2) Reduction of Emergency Condenser Effectiveness in Terminating
The licensee reported that on March 17, 1973, during a controlled
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shutdown of Unit 1, the emergency condenser key switch was placed
in bypass position with 47 control rods withdrawn and pressure at
650 psig. With the switch in bypass, high reactor pressure and
closure of primary steam isolation valves do not initiate
emergency condenser operation; however, the bypass did not affect
the auto oparation of the emergency condenser if needed for
emergency core cooling which is initiated by signals from high
sphere pressure and low primary drum level. The normal shutdown
procedure, which does not cover shutdown to hot standby, was
used, as modified by instructions from the Operating Engineer.
The licensee stated that procedure inadequacies were the major
cause of this incident. As corrective action, the licensee is
in the process of generating additional, more detailed procedures
to cover shutdown to hot standby and cooldown from hot standby.
This is covered under Action Item 73-105. The SRB reviewed the
event during its 452d meeting and it was reported to the
Directorate of Licensing by letter dated March 27, 1973.
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4.
Reactivity and Power Control
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a.
Control Rod Drives - Unit 1
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CRD F-1 remains out of service in the fully inserted position. No
further attempts to wichdraw it have been made and the plans are to
investigate and correct the problem during the refueling outage
scheduled for next fall. All other rods continue to operate satis-
factorily.
The performance of surveillance tests as required by the Technical
Specifications was reviewed as noted below:
Item
T/S Requirement
Results
Time required
At intervals dC
Performed 4/11-14/73. Min,
to fully
six months
withdraw time was 15.5 sec.
withdraw and
between refueling
After adjustment min. time
insert each CRD
outages Due 5/1/73
to withdraw was 17.0 sec.
Exercise up
Same as above
Performed 4/11/73. No
and down.
Stop
abnormalities recorded.
at each latch
po sition. Check
for proper
latching,
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unlatching,
position switch
and indicator
operation.
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Friction test
Same as above
Performed 4/10-11/73. Ten
CRD's recorded A P in excess
of 100 psi required to move
drive. Max was 174 psi
(E-5).
Lowest recorded was
49.5 psi (C-7 and B-6).
Scram times
Same as above.
Performed 4/11/73. Longest
Max. time
scram time recorded 1.68 see
allowable 2.5
(H-2).
Shortest 1.08 sec
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sec.
(F-8).
Time in buffer ranged
from 0.63 to 0.15 sec.
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b.
Scram Accumulators - Units 1, 2, and 3
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A review was made of the maintenance history of the accumulators for
Units 1, 2 and 3.
There is no regularly scheduled inspection of the
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'te accumulators are replaced with a spare if a
problem occuta with the installed unit. The replaced accumulator is
then inspected, repaired (if the problem was in the seals) and then
placed in stores as a spare. Unit 1 has 53 accumulators, 26 pairs
operating three control rods and a single accumulator operating two
control rods, and 73 change outs have occurred since the unit was
started up in 1960. Of these 73 accumulators, 25 of the cylinders
have been junked. According to the licensee, the major cause for
rejection was minute pitting of the chromium plating on the inside
of the cylinder. A few had the chromium plating pealing off wherc
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the piston had sat for long periods of time.
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Units 2 and 3 each have 177 accumulators and the records show that
Unit 2 has had 16 accumulator change outs and Unit 3 has had 12
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change outs. A total of 14 cylinders have been scrapped from the
two units because the chromium plating had become pitted. No peeling
of the ' plating has been observed.
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Units 2-3 also experienced another problem whereby "0" rings made
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of Buna-N appeared to be dissolving and " smearing" on the cylinder
surface when the piston moved. The cylinder had to be cleaned
with CCl . The licensee representative stated all the Buna-N
"0"
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rings were replaced with new "0" rings made of Viton. The Buna-N
"0" rings appeared to have been mistakenly installed on a few
accumulators by the manufacturer.
c.
Bypassing of In-Core String 113 - Unit 1
On April 8, 1973, the scram function of in-core string 113 was
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bypassed, in violation of Section B.9.a of the Technical Specifications,
following an upscale failure of all four chamber amplifiers. This
action left a four foot diameter, vertical cylindrical core volume
without scram protection for approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The event was
reviewed by the SRB during its 468th reeting on April 17, 1973,
approximately nine days after its occurrence, and the details were
reported to the Directorate of Licensing on April 17, 1973. The
SRB review approximately nine days after the event cannot be con-
sidered to have been conducted in a prompt manner as required by
the Technical Specifications.
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The inspector verified that the procedures in effect at the time
of the event were utilized by the shift personnel, but that the
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procedures did not identify the strings which constituted " critical
('
pairs" and therefore were a contributing factor in the erroneous
.
decision to bypass the scram function of all four chambers. The
,
procedure has been revised and should no longer be a cause of con-
fusion to the operating personnel.
All corrective actions listed in the licensee's report' have been
completed. The inspector questioned how the licensee's retraining
program incorporated hunan error in order to reduce the number of
such errors. The licensee was not able to answer this question,
and the inspector stated that he would follow the matter at a
future inspection with the Training Supervisor.
While reviewing the event, the inspector noted that the licensee
did not report the event to the Superintendent of Production,
Division A, until 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> on April 11, 1973, or approximately
63 hours7.291667e-4 days <br />0.0175 hours <br />1.041667e-4 weeks <br />2.39715e-5 months <br /> after the event had taken place. This is an extremely
loose interpretation of the requirements of paragraph J.3.a of
the Technical Specifications, which requires prompt notification
of the event to the Manager of Production or his delegated alternate.
5.
Core and Internals - Unit 3
a.
Refueling Operations
The inspectors observed Unit 3 refueling operations on April 17,
-
1973. The observation, which included refueling procedure steps
253 through 270, covered:
(1) Transfer of fuel from the storage well to the reactor core
(2) Transfer of fuel from one core position to another
(3) Testing of the fuel grapple interlocks (at the request of the
inspector)
,
(4) Control room operation, including rod withdrawal following the
loading of each fuel cell
(5) Fuel element reconstitution, including the handling of fuel
rods, and eddy current and ultrasonic testing of fuel rods
The inspectors verified that the refueling operations were being
conducted in accordance with the requirement established by
Operating Orders 17-73 and 21-73, the unit Technical Specification
requirements, and the liceneee's Proposed Modification 73-2, entitled
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"Dresden Unit 3 First Reload Fuel," dated March 5,1973; the only
('
devistion being that 52 new fuel elements were loaded in the core
-
1
internals of 28 elements as stated in Proposed Modification 73-2.
.
!
The licensee stated the change does not affect the safety analysis
prea'. ted in the document and that a supplementary .1.etter informing
of the change would be submitted to Licensing before Unit 3 restart.
Communications between the refueling area and the control room were
satisfactory. The inspector noted that during the rod withdrawal
following the loading of each fuel cell the operator verified
coupling integrity by attempting to withdraw the rod past position
48.
The operator stated this is normal practice following withdrawal
of any control rod.
1
b.
Fuel Inspection Results
4
A review of the results of fuel sipping and reconstitution showed
that 103 fuel assemblies contained failed fuel rods. Of these,
seven were discovered during partial in-core sipping, and 90 during
100 percent out-of-core sipping. Six additional assemblies were
rescheduled for sipping due to suspected initial results. Of these,
five were rejected as a result of ultrasonic (UT) and eddy current
(EC) tests and were not resipped. The sixth was resipped with satis-
factory results, but UT and EC testing gave positive indication on
two of its fuel rods. Of the 103 rejected assemblies, 52 were replaced
with new fuel assemblies and the remaining were replaced with recon-
!
stituted assemblies in which failed fuel rods were replaced with sound
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rods scavenged from within the 103 fuel assemblies. In addition to
passing a satisfactory UT and eddy current test, a fuel rod could
only replace a failed rod if the criteria outlined by a document
titled " Bundle Repair Rules," dated February 16, 1973, and its
addendum dated Feburary 28, 1973, were met.
The Station Review
Board approved the criteria at its 440th meeting on March 7,1973.
In addition, all reconstituted fuel assemblies were resipped with
satisfactory results prior to being returned to the reactor. The
inspectors randomly selected fuel assemblies 84, 134 and 440 and
determined that all the criteria for bundle repair had been complied
with.
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All fuel rods, except for the spacer capture rods and 18 stuck rods
which were rejected, comprising the 103 failed fuel assemblies were
ultrasonically (UT) and eddy current (EC) tested. The criteria for
rod rejection consisted of a positive UT and/or greater than 35
percent EC indication. Of the 5,047 rods, 603 were rejected, an
average of approximately 5.9 rods per assembly. A summary of the
rejected rods follows:
1
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(1) By Enrichment
-
Enrichment _
Failure
-
,
1.2%
80
1.69%
168
1.69% (Tie Rods)
90
2.44%
226
2.44% (Tie Rods)
39
603
(2) By Testing Criteria
EC
Failure
35%
Positive
187
'
35%
Positive
86
35%
Negative
269
35%
Negative
61
at spacer
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603
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The locations of the 197 fuel roda with greater than 35 percent
EC and positive UT withir. the fud assembly is shown below.
2
1
5
1
5
0
6
'
2
2
1
1
0
5
1
1
1
0
1
0
1
4
1
0
1
2
1 10
2
1
1
0
3
2
24
1
1
1
5
1
3
2
O
3
2
9
19
7 45
- Not tested
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The locations of the 603 rejected fuel rods within the fuel
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assembly are as shown below.
s
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9
7
12
8
10
9
11
11
8
7
10
6
11
17
7
9
9
8
5
8
9
10
4
9
12
9
22
13
9
8
8
15
13
37
8
7
7
19
9
18
7
13
12
11
26
33
17
56
- Not tested
During the EC and UT scanning of the fuel rods, defect signals may -
have been observed in more than one axial location. The licensee
recorded the area where the highest signal was observed for each
rejected rod. The axial distribution of these " worst" areas is
given below:
-
Axial Location
Above Lower End
Rods With 35% EC
Plug (inches)
and Positive UT
All Rejected Rods
0-10
36
255
11-20
12
58
21-30
4
8
31-40
11
22
41-50
5
10
51-60
5
10
,
61-70
6
8
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Axial Location
Above Lower End
Rods With 35% EC
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Plus (inches)
and Positive UT
All Rejected Rods
,
,
71-80
3
8
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81-90
14
23
91-100
20
34
101-110
19
30
111-120
15
22
121-130
8
13
131-140
11
25
Location not identified
18
77
187
603
The licensee's records indicated that at least 1.0 fuel rods separated
circumferentially as follows:
(1) Six fuel rods separated at lower end plug.
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(2) One fuel rod separated at approximately one-third from lower
end plug.
(3) One fuel rod separated approximately eight to 12 inches from
lower and plug.
1
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(4) Two fuel rods separated approximately 12 inches from upper
and plug.
6.
Electrical Systems - Unit 1
a.
Unit 1, 125V Battery
The licensee's performance of surveillance tests required by the
Technical Specifications was reviewed as noted below:
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T.S.
Period
Itam
Requirement
Inspected
Results
(
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Cell voltage to
Every three
1/73-3/73
Satisfactorily
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nearest .01 volt,
months
performed
1/12/73
of each cell cnd
temperature of
every fifth cell.
We' ekly
1/73-3/73
Voltage of pilot
and voltage of
cell and tempera-
pilot cell.
ture of adjacent
Temperature of
cells not taken
adjacent cells
during the week
and overall
of 1/1-1/7/73 and
battery voltage.
1/15-1/21/73.
,
Entire record
missing for week
of 2/19 to 2/25
.
The above violations are probably the result of the lack of adequate
data sheets as previously noted in RO Inspection Report No.
050-237/73-01. During this inspection it was noted that the licensee
has prr. pared new data sheets which provide space for all the required
_
readings and is in the process of inn.orporating them in the Unit 1
Operating Manu21. The new data sheets were approved by the SRB
during its 461st meeting.
b.
Essential Service Power Supply
The inspector reviewed the normal and emergency supplies of power
to the essential service buses 15 and 16.
Bus 16 is supplied through
its transformers from bus 11, and bus 15 from bus 12 via its trans-
former. Bus 11 is supplied through transformer 11 from the main
generator, or section 1 of the 138 KV bus. Bus 12 is supplied
through transformer 12 from section 2 of the 138 KV bus. As an
alternate supply, transformers 11 and 12 can supply either bus 11
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or 12 through an automatic throwever. Bus 11 can also be supplied
from transformer 13 which is supplied from the 34.5 KV bus. The
low voltage signal on buses 11 and 12 that closes the breaker from
transformer 13 also starts the diesel-generator.
If the buses 15
and 16 are not energized by the time the diesel-generator is up to
normal speed and voltage, buses 15 and 16 are electrically isolated
from the rest of the system and the diesel-generator is automatically
closed into the buses.
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Control of the diesel-generator and the above breakers can also be
accomplished manually from the control room.
(w
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7.
Containment
.
Unit 3 Torus
The licensee stated that the following work was completed in the Unit 3
torus during the present outage.
a.
All baffles were removed and disposed of.
4
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b.
An inspection of the surface below the water line indicated a large
i
amount of small blisters (1/4 to 1/2" in diameter) covering
approximately 25 percent of the surface area. A few large blisters
were also noted.
c.
The surface below normal water level and a band up to one foot above
the water level were sandblasted to bare metal. This area was then
painted with Carbozine-11. A bellyband of Carboline's Phenoline 368
winter grade primer was applied approximately four inches above and
below the transition line to overlap the existing paint and the new
Carbozine-ll. This repair procedure is similar to that used on
Unit 2 during 1972.
d.
An inspection of the above water line surface indicated a few areas
of pin-hole rust. No flaking or blisters were noted. As before,
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the repair procedure consisted of touch-up.
1
Improved bracing of the relief valve lines similar to that performed
e.
,
on Unit 2 in 1972 was completed.
f.
Paint test coupons were installed.
These repairs, with the exception of the baffle removal, update Unit 3
torus to the condition of the Unit 2 torus. Unit 2 baffles will be
removed during its next refueling outage scheduled for 1974
8.
ECCS - Unit 1
a.
Response to Items of Noncompliance
The corrective measures, to items of nonconformance, cutlined in the
licensee's (Lee to Grier) letter of November 21, 1972, were inspected
as noted.
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(1) The cable installation for spare core spray pump 1C was enclosed
in conduit prior to initial operation of the core spray system,
and now satisfies IEEE-279 criteria. This item is considered
(
resolved.
.
(2) As-built drawings for pipe hangers and seismic restraints were
reviewed with Construction personnel.
The new drawings are
dated April 19, 1972, and November 21, 1972. As.a result of
the problems identified during the February 1972 inspection,
a review of the installation, by the Architect-Engineer,
determined that the piping routing was not identical to what
was originally analyzed. A new analysis was performed and a
copy of the results transmitted to CE Company on April 19, 1972.
As a result of the analysis four additional seismic restraints,
CSR 15A, 17, 12 and 13, the latter two being located in the
reactor cavity, were installed. This item is considered resolved,
however, it was noted that the licensee's technical staff and
QA records do not include copies of the latest drawings.
Station
management was informed and an Action Item has been generated to
ensure the copies are obtained.
(3) The inspector verified that the licensee has documentation from
Atwned and Morrill Company, manufacturers of the two cast check
valves in the system, which gives the dimensional measurements
of the valve body castings made by C. A. Wills on October 28, 1971.
The dimensions meet the minimum wall thickness specified in the
drawings.
In addition, the licensee has documentation from
Sargent and Lundy. Engineers concerning a trip to Velan Engineering
~
for the purpose of auditing Velan's records for the four high
pressure forged valves. The licensee is satisfied that the
forged valves meet the engineering specifications. This item
is considered resolved.
(4) The liquid penetrant inspection procedure was approved by the
Architect-Engineer firm by letter dated February 29, 1972. In
addition, the qualifications of the two Level II inspectors
involved were reviewed and approved by the Architect-Engineer.
'
The documentation is now in the site files. This item is
considered resolved.
(5) Three welders were requalified to F. Conry's procedure WSSTM-3,
Revision 4.
They were reviewed with no comments by Sargent and
Lundy on August 17, 1972. They had also been approved by
Pittsburgh Testing Laboratory. This item is considered resolved.
(6) The licensee's QC Procedure 9-51.1 provides a Quality Control
Weld Check Sheet in which a visual inspection of joint fit-up
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is documented. The check sheet does not list all the specific
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fit-up elements that were checked. The licensee plans to
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compara his check sheets against the check sheets used by
'
other piping contractors, and will modify his to include at
least as much information as is being documented by the piping
contractors' check sheets. This items is considered resolved.
(7) The licensee's QCP 9-51.1 check sheet documents prehcat and
interpass temperature. The above comment on a check sheet
comparison study also applies to chia item. This item is
considered resolved.
(8) A procedure was developed for welding electrode control and,
according to the licensee, used in the completion of the job.
l
The licensee's QCP 9-51.1 controls the issuance and handling
of welding rod and wire during future safety related jobs.
!
This item is considered resolved.
(9) A weld repair procedure was developed. The licensee stated
that all future safety related jobs, and whenever the ASME
Code requires it, will be supplied with a weld repair proce-
dure. This item is considered resolved.
b.
Core Spray Valve CS-ll Failure to Open
During surveillance testing on December 29, 1972, and January 12,
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1973, core cpray valve CS-11 experienced temporary failures to open.
Given a closed signal followed by a second signal to open, the
valve would operate satisfactorily.
Satisfactory operation of the
two valves on the redundant injection line assured the operability
of the system. After consulting with the valve manufacturer, the
licensee has reset the open torque switch on all four valves from
a setting of "1" to "2.5" and the closing torque switch on all four
7
valves-to "2.0".
This was completed on February 26, 1973, at which
time the valves were cycled satisfactorily,
the event was reviewed
by the SRB on its 425th meeting and reported to the Directorate of
Licensing on January 25, 1973.
I
c.
Barton Differential Pressure Transmitters
The inspector questioned the licensee regarding the use of Barton
Model 368, 384 and 386 for which a manufacturing deficiency had
been identified. The licensee identified the use of Barton Model
386 for the transmittal of sphere level information (readout plus
,
alarm in control room) in connection with the core spray system.
The licensee indicated no other Barton d P equipment is in use for
Unit 1, and that the mcnufacturer would be contacted for instructions
on the required corrective action for Model 386.
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9.
Emergency Power - Unit 1
a.
Diesel Generator
,
The licensee's performance of surveillance tests required by the
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Technical Specifications was reviewed as noted below:
Item
T/S Requirement
Period Inspected
Results
2100 gal of
Record Monthly
1/73-4/73
Performed.
fuel for diesel
Quantity
generator
> 3500 gal
Quality of
Check Monthly
1/73-3/73
Water con-
diesel
tent and
generator
microorganism
fuel
growth
c
results
negative
Manual start
Monthly
11/72-4/73
12/72
and load of
Records
diesel
missing
generators
,
Operability
Monthly
11/72-4/73
12/72
2
of diesel
Records
_
fuel oil
missing
,
j
transfer pump
b.
Diesel Generator to Bus 15 Breaker Failure
The corrective measures described in the licensee's letter (Lee to
<
Grier) dated February 7, 1973, have been completed. The breaker was
replaced with an identical unit on April 13, 1973. The replacement
was satisfactorily tested. The breaker to bus 16 was not replaced.
The licensee stated that in the past failures have been experienced
with the breaker to bus 15 while no failures have ever been experienced
with the breaker feeding bus 16. The item was reported in the semi-
'
annual report as indicated in the letter. This item is considered
resolved.
10.
Radioactive Waste Systems - Units 2 and 3
Reactor Building CAM Alarms
The inspector verified that the reactor building continuous air monitor
alarm has been wired to the annunciators of Units 2 and 3 in the control
room. This item is considered resolved.
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