ML19340A606
| ML19340A606 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/16/1973 |
| From: | Dance H, Fishbaugher J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19340A605 | List: |
| References | |
| 50-010-72-06, 50-10-72-6, NUDOCS 8008280657 | |
| Download: ML19340A606 (15) | |
See also: IR 05000010/1972006
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U. S. ATOMIC ENERGY COMMISSION
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DIRECTORATE OF REGULATORY OPERATIONS
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REGION III
RO Inspection Report No. 050-010/72-06
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Licensce: Cocmonwealth Edison Company
P. O. Box 767
Chicago, Illinois
60690
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Dresden Nuclear Power Station, Unit 1
License No. DPR-2
Morris, Illinois
Category: C
Type of Licensee:
Type of Inspection:
Routine, Unannounced
Dates of Inspection:
November 26 through December 1, 1972
Dates of Previous Inspection: October 16 and 20, 19i2
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November 1, 1972 (Safeguards)
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Principal Inspector:
J. R'.'(F shbaughd
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Accompanying Inspc etors : None
Other Accompanying Personnel: None
( k&.w%
Reviewed Ey:
H. C. Dance, Senior Reactor Inspector
/ I lt 71
Reactor Operations Branch
(Date)
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SUMMARY OF FINDINGS
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Enforcement Action
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A.
Contrary to Technical Specifications Section B-15, the emergency
condenser water temperature has been allcwed to exceed 1000F.
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(Paragraph 5.c)
B.
Contrary 'to Technical Specifications Section J-5, failure of the
diesel generator breaker on August 25, 1972, was not reported to
the AEC as an abnormal occurrence.
(Paragraph 9)
C.
Contrary to Technical Specifications Section J-5, emergency
condenser temperature in excess of its limit was not reported
to the AEC as an abnormal occurrence.
(Paragraph 5.c)
Licensee Action on Previously Identified Enforcement Action: Not inspected
Unusual Occurrences
A.
The diesel generator breaker failed to close on August 25, 1972.
(Paragraph 9)
The diesel generator was forced out of service on October 1,1972,
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B.
due to failure of an alarm relay.
(Paragraph 9)
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O ther Significant Findings
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Current Findings
The design limitations of the shutdown cooling system were discussed.
The matter has been referred to Regalatory Operations Headquarters
for further review.
(Paragraph 5.d)
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B.
Status of Previously Unresolved Items
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For the new core spray system, completion of preoperational testing,
the new emergency procedure for loss of coolant, operator orienta-
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tion, and surveillance testing procedures were verified. These
matters are considered resolved.
(Paragraph 8)
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Management Interview
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The following topics were discussed at the conclusion of the inspection
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with Mr. W. Worden, Plant Superintendent; Mr. J. Diederich, Supervising
Engineer; and Mr. A. Roberts, Operating Engineer.
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A.
Emergency Condenser
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The emergency condenser temperature exceeding 1000F was noted to be
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in v1( ' ation of a new Technical Specifications limit.
(Paragraph. 5.c)
The licensee indicated that the problem had been recognized and that
action was in progress to obtain a relaxation of the limit.
B.
Unloading Heat Exchangers
The inspector stated that the unloading heat exchangers have a design
vulnerability to allow discharge of primary coolant activity directly
to the discharge canal in the event of leaking tubes. The licensee
concurred and identified an engineering review which is in progress.
The licensee indicated this review would be given new emphasis, but
was reluctant to make a commitment regarding its completion.
(Paragraph 5.d)
The inspector indicated that the matter would be referred to the R0
Headquarters for evaluation.
C.
Moderator Temperature Coef ficient
The inspector stated that the limit on reactivity addition by
moderator heating should be administered without discounting for
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the negative effect over the range from ambient to operating
temperature. The licensee indicated the matter will be reviewed
before the question becomes crucial to compliance with the limit.
(Paragraph 6.d) The inspector stated that future inspections
against this limit would be based on simple integration of the
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reactivity effect over the temperature region where the coefficient
is positive.
D.
Safety Circuit Trip Checks
The inspector suggested that (1) routine testing be established
for the steam isolation valve closure trips, and (2) that more
appropriate (more frequent) test frequencies be adopted for all
safety circuit trips. The inspector noted further that the new-
format Technical Specifications which are being considered will
probably bring attention to this matter. The licensee indicated
this matter uas already under discussion in connection with the
proposed new Technical Specifications.
(Paragraph 5.b)
E.
Licensee Work Load
The inspector indicated that the amount of work represented by the
list of outstanding action items seems large with respect
current
to the size of the present plant staff.
(Paragraph 3)
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The licensee stated in their view, t!.a problem was not yet acute.
F.
Incident Reporting
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The inspector stated that reporting of abnormal occurrences to
the AEC had been deficient. The diesel generator breaker failure
tvhich occurred August 25, 1972, and the emergency condenser tempera
ture exceeding its limit were cited as violations of the reporting
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requirements of Technical Specifications Section J-5.
(Paragraphs 5.c
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and 9)
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REPORT DETAILS
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1.
Persons Contacted
W. Worden, Station Superintendent
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A. Roberts, Operating Engineer, Unit 1
J. Diederich, Supervising Engineer, Technical Staff
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J. Bowers, Engineer
R. Pavlich, Supervisor, Radiation Protection
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R. Mefford, Instrument Forcaan
2.
General
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The reactor operated steadily at 180 Mwe during September 1972,
with an average off-gas release rate of approximately 37,000 uc/cc.
The reactor was shutdown from September 30, 1972, till November 2,
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1972, for installation of the core spray system. A special inspec-
tion of the new core spray system was reported previously.1/
Sub-
sequently the reactor was operated at power levels up to 160 Mwe
with power level again being limited by the off-gas release rate.
During the weekend of November 25, 1972, a reaccur scram occurred
and the unit remained shutdown for approximately one day for
correction of condenser leaks. Operation was resumed with power
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level up to 150 Mwe, and off-gas release rate at approximately
60,000 uc/sec, by December 1, 1972.
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Refueling schedule possibilities were under study. Delivery of
new fuel was expected by approximately January 15, 1973. A
refueling outage for the replacement of approximately 50 fuel
elements was expected during the winter / spring of 1973, with
schedule details somewhat dependent on shutdown planning for
Unit 3.
Additional work planned includes in-service inspection,
fuel sipping program, installation of redundant condensate valves
for emergency condenser, and miscellaneous maintenance and sur-
veillance testing.
3.
Organization / Administration
Station organization was found to comply with the functional chart
given in Figure 2 of the Technical Specifications. Shift staffing
was found to comply with Table III of the Technical Specifications.
At the time of the inspection,19 operators were licensed for
Unit 1; and approximately 16 additional staff nembers had senior
operator licenses for Unit 1.
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R0 Inspection Report No. 050-010/72-04
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The functioning of the Station Review Board (SRB) generally satisfies
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the requirements of Technical Specifications Section J.
There were
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approximately 15 SRB meetings per month during September, October,
and November. Meeting minutes indicate that while SRB attention is
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given mostly to Units 2 and 3, adequate attention is given to Unit 1
topics. Administrative procedure ADM-III describes SRB operations.
A total listing of action items from all sources is being issued
periodically to assure followup. Numerous items involved procedure
revisions.
The amount of work represented b'y the current list yields
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the consideration that with the present staffing level, many action
items may not be resolved in a timely manner.
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Investigation of abnormal occurrences, corrective action, and inter-
nal documentation are generally adequate. However, reporting of
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abnormal occurrences to the AEC has been deficient as noted in
paragraphs 5.c and 9.
Up-dated administrative procedures were reviewed as indicated in
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paragraph 4, below.
4.
Facility Procedures
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A complete review, up-dating, and reorganization of the operating
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procedure manual was nearing completion at the time of.the inspec-
tion.
All of the new procedures had received SRB approval.
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final task of duplicating and assembling new manuals was in progress.
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The new procedure manual is organized according to systems nomencla-
ture which has been standardized company-wide. For each of the
various plant systems, all proceudres (startup, normal operation,
shutdown, abnormal operation, and surveillance testing) are included
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in a single 'section of the canual. Additional general sections of
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the manual cover administrative, general plant operation, and certain
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plant emergency procedures.
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Scope of the new procedures manual was reviewed. Tables of contents
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for the various -sections' indicate adequate coverage for those systems
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and components involving nuclear safety. The " Generating Station
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Emergency Plan" and maintenance procedures are outside the scope of
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the new procedures manual and were not inspected.
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The following new procedures were selected as a sample for detailed
review:
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a.
Plant Startup
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b.
Administrative procedures (entire section)
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c.
Radwaste Disposal to discharge canal
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d.
Reactor coolant alarms (eleven procedures)
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Emergencies (loss of coolant, loss of instrument air,
containment isolation, fire)
f.
ECCS surveillance (six procedures)
In addition, approximately 20% of the entire manual was scanned.
In general, the procedures were found to be adequate. A number
of comments were given to the licensee for resolution during a
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future inspection. These were as follows:
The control step given in Technical Specifications Section
a.
G-2-d, should be added to the plant startup procedure.
b.
All check lists should be included. For example, pre-startup
nuclear instrument checks, and daily / weekly surveillance
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sheets should be included.
c.
Administrative procedure, ADM-I, refers to new form Technical
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Specifications which are not yet issued. Reference should be
to existing Technical Specifications.
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d.
ADM-II states, " procedures will be periodically reviewed by
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operators and supervisor." To be meaningful, this needs to
be supported by specific assignments and schedules.
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For ADM-ll, the records that are required should be specified,
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such as daily / weekly surveillance records, completed startup
check lists, etc.
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For ADM-II, elaboration on acceptable aethod(s) for documenting
temporary changes is suggested.
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g.
Review Table 30-C (SRB responsibilities) with respect to Technical
Specifications Section J (see page 11 of Change 21) .
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ADM-VIII - To be meaningful, specifics are needed on what
reports, when, by whom, etc.
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ADM-IX - Note that Appendix A is missing.
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Review all outstanding operating orders to be sure they are
included in the up-dated procedures.
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ADM-VI - Include these additional topics under radiation controls:
- Administrative limits for radiation exposure
- Designation of controlled access zones
- Provisions for radiation surveys
- Provisions for calibration of radiation protection
instruments
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The reactor should be scrammed in event of a ground level release
of off-gas (such as the one which occurred in August 1971).
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The procedure given in 1000-0-I for placing an unloading heat
exchanger in service does not assure against release of radio-
activity due to tube leaks.
(Additional discussion in
paragraph 5.d, below.)
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Procedure for recoval and replacement of reactor vessel head,
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including disconnect and replacement of core spray piping, to
be included.
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Definition of the in-service inspection program should be
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included.
An emergency procedure is needed for the event designated
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" Probable Maximum Flood" (see page 13 of Technical Specfications
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Change 21).
These comments include some f rom an earlier preliminary review 2/
which were not resolved by the new procedures. Asterisk items were
noted to be included in the licensee's current listing of action
items.
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5.
Reactor Coolant Sys tem
a.
Safety Valves
A review of the design of the safety valves to ensure that reaction
forces are recognized had been requested previously.3/ The licensee
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RO Inspection Report No. 050-010/72-01
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Letter, RO:III to CE, dated April 28, 1972
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indicated that the review has been completed, and that preliminary
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examination of the report at licensee headquarters indicates the
design to be satisfactory. A copy of the report was expected at
the site within a few days. The inspector indicated the report
would be examined during a future inspection.
b.
Surveillance Testing
Surveillance testing frequency and recent results were exar led
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for the following safety circuit trips:
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- High Reactor Pressure
- Closure of Primary Steam Sphere Isolation Valves
- Closure of Turbine Stop and Bypass Valves
- Low Condenser Vacuum
- Low Water Level in. Primary Steam Drum
- Low Water Level in Reactor Vessel
In general, none of these trips are designed to be tested during
operation.
No requirement for testing these trips has been
established in the Technical Specifications. Typically, these
trips are being tested during refueling outages, or approximately
once per year, except that no routine testing has been established
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for the steam isolation valve closure trips.
Satisfactory results were recorded for the testing that was
examined. The inspector stated that the licensee should review
surveillance testing practices to be sure that the testing pro-
gram is comprehensive, and that the test frequencies are appropriate.
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The licensee indicated that this matter is being considered.
c.
Emergency Condenser
A new maximum temperature limit of 1000F has been established
for the emergency condenser heat sink.
(See ,1ge 4 of Technical
Specifications Change No. 17.) This limit is being exceeded.
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On November 30, 1972, this temperature was observed to be 1100F.
The licensee stated that the condenser shell side temperature
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has never been controlled below 1000F; that in fact the ambient
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temperature near the condenser is typically greater than 1000F.
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The licensee recognized the problem and started action on
November 10, 1972, to obtain relief from the limit. However,
operation outside the limit was not reported to AEC as an
abnormal condition. Operation outside the limit was continued.
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Subsequently, on December 15, 1972, the licensee reported that
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emergency condenser temperature had decreased below the limit
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due to the effect of winter weather conditions on ambient
temperatures in the sphere.
d.
Unloading Heat Exchanger
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The B Unloading Heat Exchanger was leak checked at the start
of the October 1972 outage and found to be leaking. An activity
release by way of the plant service wate'r effluent occurred.
The licensee stated that the amount of the release was measured,
that release limits were not exceeded, and that the entire
release was conducted according to plan. The A Unloading Heat
Exchanger was used for shutdown cooling while the B unit was
repaired.
The B unit experienced a similar leak in September 1971.1/5/
During the refueling outage which ended in February 1972,
several leaks in both A and B units were repaired.
The inspector stated that the existing design may have limita-
t ons with respect to shutdown cooling in that:
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- Unloading HX's constitu'te a single boundary between
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reactor coolant and the discharge canal.
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- Integrity of the Unloading RX tubing has been shown
to be poor.
- In event of RX leaks, there are no holdup provisions
for the secondary coolant; and no alternate cooling
systems are available during shutdown.
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Discussion indicated that the licensee has an engineering review
in progress to identify options and determine a course of action.
The licensee indicated this review would be reemphasized, but
was reluctant to make a commitment regarding its completion.
6.
Reactivity and Power Control
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a.
Rod Worth - Reactivity Insertion Rate
Individual control rod worth is determined by a computer code.
A cold calibration of one rod has been performed at the start
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Letter, CE to DRL, dated September 22, 1971
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RO Inspection Report No. 050-010/71-08
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of each operating cycle to verify the computer predictions.
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The current prediction shows the strongest rod to be worth
approximately J O mk; assuming a correct rod sequence is used.
For an erroneous rod pattern, a worst case worth higher by a
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factor of five has been postulated.
The fastest time for control rod withdrawal is maintained at
18 seconds by individual speed control valves. Technical
Specifications Section F-1, requires that rod withdrawal
times be measured during each refueling outage and at inter-
vals of not more than 6 months between refueling ritages.
The most recent measurements were made on November 1, 1972,
and on May 13, 1972. They inCicate withdrawal times in the
range of 18 to 28 seconds.
Thus for the strongest rod being withdrawn at the fastest rate,
an average reactivity addition rate would be 10 ek/18 sec =
0.55 mk/sec. A factor of five for the most adverse rod pattern
would yield a worst case insertion rate of 2.8 mk/sec, which
compares with a limit of 2.9 mk/sec given in Technical Specifi-
cations Section D-5.
b.
Procedural Control of Reactor Startups
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The licensee displayed the standard rod withdrawal sequences
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and the critical predictions which have governed all reactor
startups. Actual critical was compared with predicted critical
for a typical hot startup (August 7, 1972) and a typical cold
startup (November 2, 1972). The difference in both cases was
less than 10 mk.
Rod verification records (verificatica that the control blade
moves together with its drive) are filed with the critical
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prediction sheets for each startup.
Records for the above two
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startups indicate that all rods were verified except F-1.
The
licensee stated there were no unverified rods, except F-1, at
the time of the inspection.
The licensee stated that one of the nuclear engineers is on
duty in the control room for every reactor startup,
c.
Stuck Control Rod
Rod F-1 remains out of service in the fully inserted position
because it could not be uithdrawn during June 1972.5/ The
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RO Inspection Report No. 055-010/72-03
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licensee plans to investigate and correct the problem during
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the next refueling outage. The licensee indicated there have
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been no similar symptoms with other rods during recent months.
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d.
Moderator Temperature Coefficient of Reactivity
This temperature coefficient is positive at ambient temperature.
C'.th increasing temperature, the coefficient decreases and turns
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negative. With increasing fuel burnups, the coefficient becomes
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more positive at ambient temperature and turns negative at a
higher tc< 3rature, such that the potential reactivity addition
from nuc ear heating becomes greater during each period between
refueling outages.
Technical Specifications Section D-5, limits the potential
reactivity addition to less than one dollar, and requires that
the coefficient not be positive above 5500F. Also, measurements
are required to verify that these conditions are satisfied.
Moderator temperature coefficient measurements have been per-
formed during the present fueling period at the start of the
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cycle on February 1, 1972, again on June 12, 1972, and on
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November 4, 1972. The temperature at which the coefficient
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turned from positive to negative was approximately 1200F on
February 1,1972, and approximately 2500F on November 4,1972.
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The reactivity addition for November 4,1972, (maximum case)
determined by integrating from ambient
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temperature (1000F) to
250cF was approximately 30c. For this determination, a delayed
fraction of 0.0054, cot. responding to 12,000 MWD /T average fuel
exposure was used.
While these measurements were judged to be satisfactory, the
inspector noted that additional measurements at lower tempera-
tures might be necessary in the future in order to confidently
integrate the positive reactivity effect when the coefficient
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becomes more strongly positive.
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The licensee indicated that the limit on net reactivity addi-
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tion has been administered on the basis of integracing the
moderator temperature coefficient from ambient
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temperature to
5500F, thus taking the net result of whatever positive and
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negative components occur between those temperatures.
The
licensee stated further that this interpretation had received
concurrence by the Nuclear Review Board. The inspector
reitterated his interpretation that the limit should be applied
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to only the positive component. The licensee indicated that a
review of the matter would be conducted. The licensee under-
stood that the matter could become crucial toward the end of
the fuel cycle.
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e.
Peak Heat Flux
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Power distribution within the reactor is determined on a monthly
frequency using wire monitors. Most recently this was done on
November 6, 1972. The monthly data is used to check the cali-
brations of the in-core flux monitors. Each shift, the in-core
monitors are used to determine flux peaking factors, and the
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peaking factors are applied to total heat rate to yield the peak
heat flux within the reactor. Records indicate that typical
peak heat flux is in the range of 75 - 85% of the Technical
Specifications limit of 360,000 BTU /hr/ft2, The licensee stated
that peak heat flux up to approximately 90% of the limit has been
experienced.
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f.
Rod Scram Time
Technical Specifications Section F-2, requires that rod scram
times be measured during each refueling outage and at intervals
of not more than 6 months between refueling outages. The last
two sets of measurements were made on May 13, 1972, and
November 1, 1972. Scram times from start of motion to buffer
ranged mostly from 1.4 to 1.7 seconds. Rod E-4 was the slowest
at 1.82 seconds.
For reference, the Technical Specifications
limit (Section B-4) is 2.5 seconds.
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7.
Auxiliary Systems
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Potential failures and general experience with (1) the instrument
air system and (2) the various demineralizers were discussed.
Loss of the instrument air system has not been experienced. Several
backup provisions are in place and would be activated if needed by
abnormal procedures in response to low pressure alarms. Consequences
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of a complete loss of instrument air are believed to be generally
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fail-safe, but a comprehensive analysis has not been performed. The
licensee indicated that system reliability has been judged to be ade-
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quate basad on (1) the numerous backups and (2) the extended period
of favorable experience.
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For the various desineralizers, no serious difficulties have been
experienced with (1) resin escape into coolant systems, (2) gross
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breakthrough of poor quality water, or (3) adequate control of
regeneration chemicals.
8.
A special preoperational inspection of the new core spray system,
conducted during the October shutdown, identified certain items
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for attention prior to startup.2/ Resolution of these items was
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as follows.
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Prooperational tests, including automatic system actuation, times
for valve operation, and flow testing, were completed and documented.
SRB revieu and approval of test results were recorded on November 2,
1972. The temporary cap on the discharge of valve CS-17 was removed
prior to November 2, 1972. Deletion of a position indicator for
valve CS-22 was formalized by SRB review prior to November 2, 1972.
Emergency procedure PEOP-I covering use of the core spray system for
the loss of coolant event was completed and approved by the SRB; and
supplementary video tape orientation on the new procedure was con-
ducted prior to November 2,1972.
Incorporation of the core spray system into the startup check list
and the daily / weekly surveillance sheets, and the test procedures
for weekly / monthly surveillance tests were completed and approved
by the SRB prior to November 2,1972.
9.
Diesel Generator
On August 25, 1972, with the reactor operating at 180 Mwe, the diesel
generator was started as a precautionary step due to tornado warnings.
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The diesel generator breaker failed to close.
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Immediately the breaker was racked down and back up, and a second
attempt to close the breaker was successful within 14 minutes of
the initial failure. Cause of the malfunction could not be deter-
mined by immediate laspection. Subsequently, the failure could not
be reproduced and a more complete inspection did not reveal any
faults.
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The breaker has been determined to be operable based an extensive
inspection and testing. Further diagnostic /correctivs steps are
being considered by the licensee. A similar event occurred on
April 22, 1969, and could not be explained.
On October 1, 1972, with the reactor in a cold shutdown, the diesel
generator had to be taken out of service due to failure of an alarm
relay.
(Control power was disconnected because smoke was flowing
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out of the control cabinet.) The faulty relay was replaced and sent
to the vendor for analysis.
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RO Inspection Report No. 050-010/72-04
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Neither of these events were reported to the AEC as abnormal
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occu rrence.s . Final corrective action by the licensee remains
pending for both events.
Failure to report the August 25, 1972,
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failure is in noncompliance with Technical Specifications Section
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J-5.
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10.
Radioactive Waste Controls
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The licensee has been requasted previously to demonstrate that the
off-gas isolation valve can perform its intended function.$/ Basis
for the request was a gross leakage problem with the equivalent
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valve at another facility. During the October shutdown, the licensee
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observed a systems pressure response to valve closure, and thereby
established that a gross leakage condition does not exist. Further
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inves.igation including inspection of the valve is planned for the
next refueling outage.
Measurements to determine the loss of particulates from the stack
sample stream were previously reported to be in progress with com-
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pletion expected by October 1,1972.9/ These ceasurements were
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not completed. The licensee indicated a new target date of
January 15, 1973.
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R0 Inspection Report No. 050-010/72-03
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Ibid.
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