ML20154L396
| ML20154L396 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 03/05/1986 |
| From: | Eng P, Guldemond W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20151H314 | List: |
| References | |
| 50-341-86-07, 50-341-86-7, IEB-84-03, IEB-84-3, NUDOCS 8603110464 | |
| Download: ML20154L396 (6) | |
See also: IR 05000341/1986007
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-341/86007(DRS)
Docket No. 50-341
Licensee:
Detroit Edison Company
2000 Second Avenue
Detroit, MI 48226
' Facility Name:
Fermi 2
Inspection At: Fermi Site, Newport MI
Inspection Conducted:
February 24-28, 1986
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Inspector:
atrici
L. Eng
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Approved By:
W.
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d, Chief
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Operational Programs Section
Date
Inspection Summary
Inspection on Februa_ry,i_2,4__28,_1986_(Rep' ort No. 5_0-341/860,07]p_R,S)) ion o
Areas Inspected: Rout ne, unannounced inspectTon orTicensee act
previous inspection findings; review of licensee response to IE Bulletin 84-03;
. inservice testing program implementation and inservice testing instrumentation.
The inspection involved a total of 40 inspector-hours onsita oy one NRC
inspector. During this inspection, Inspection Procedures 61700, 92701 and
92703 were used.
Resul ts: Of the areas inspected, no violations or deviations were identified.
B603110464 860305
ADOCK 05000341
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DETAILS
1.
Persons Contacted
- R. S. Lenart, ?lant Manager
- G. R. Overbeck, Superintendent, Operations
A. J. Banek, I and C Technician
- J. E. Conen, Licensing Engineer
R. C. Drouillard, Nuclear Fuel Handling Supervisor
R. J. Filipek, Acting I and C Engineer
R.
'. Mack, Plant Support Engineer
D. D. Merriman, Metrology Lab Specialist
B. J. Sheffel, ISI Programs Engineer
K. Speicher, Consultant NSS
- Denotes those who attended the exit meeting on February 28, 1986.
The inspector also interviewed others of the licensee's staff during the
course of the inspection.
2.
Licensee Action On Previous Inspection Findings
(Closed) Open Item (341/84046-02(DRS)) Determination of maximum allowed
leak rates for category "A" and "A/C" valves.
The inspector reviewed the
licensee's Inservice Testing program for valves as well as selected valve
leak test procedures and determined that valve specific maximum allowable
leak rates had been set. This item is closed.
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No violations or deviations were identified.
3.
Review of Licensee Response __to_ IE Bu_1_lp_t_in__8_4 _03
On August ?4, 1984, the NRC issued IE Bulletin 04-03, " Refueling Cavity
Water Seal", to all power reactor facilities. The IEB, which described
the events surrounding a refueling cavity water seL1 failure at the Haddam
Neck facility, required licensees to evaluate the potential for and
consequences of a seal failure and submit a sumary report supporting
their conclusions.
The inspector reviewed the licensee's response to IE Bulletin 84-03
as provided by letter dated, November 3,1984, pertinent drawings and both
normal and abnormal fuel handling procedures.
It was concluded that the
Fermi cavity seal does not contain active components, is permanently
installed and, therefore not susceptible to the type of failure described
in the bulletin.
During the inspection the inspector noted the following:
a.
The licensee does not use inflatable seals to retain water in the
reactor refueling cavity. A pern.anently installed bellows seal is
used which, on total failure, will result in a small leak rate limited
by a backup flexible plate seal. Leakage from the seal area is
directed to an alarmed flow meter which is verified operable and
calibrated periodically.
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b.
The relative elevations of the spent fuel pool, the reactor core, and
the seal are such that with a seal failure and associated draindown,
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only fuel suspended from the bridge crane and the two fuel
preparation machines could be uncovered. All remaining active fuel
would remain covered. Assuming both the loss of normal makeup
supplies and a cavity seal failure, ample time is available to place
fuel in a safe condition.
c.
Procedures are in place requiring that fuel in transit be placed
in a safe condition if leakage is indicated and makeup is
insufficient. These actions can be completed before damage occurs or
radiation levels become excessive,
d.
The spent fuel pool does not have any drains, and potential siphon
paths are defeated by installed vacuum breakers such that inadvertent
valve opening or pipe failure can not result in draining the spent
fuel pool below the level of the active fuel,
e.
Two fuel pool level alarms, the flow rate alarms mentioned in
paragraph 3.a. area radiation monitors and periodic visual inspection
are available to initiate mitigating actions on a loss of pool
inventory. Abnormal operating procedures are in place addressing
safe placement of fuel, inventory makeup and evacuation of high
radiation areas.
Based on the above, it is concluded that system design renders the
prcbability of catastrophic seal failure acceptably low.
In the event
that such a failure occurs, fuel damage is not anticipated based on
existing fuel handling procedural requirements and sufficient time to
implement such requirements.
It is, therefore, concluded that the
licensee has adequately resolved the issues identified in IEB 84-03, and
the IEB is considered closed.
A review of other potential mechanisms for loss of water was also
conducted. Short of structural failure, no credible nechanism for loss of
spent fuel pool inventory was identified.
Evaluation of other potential
leak paths fo'r the reactor cavity such as instrument installations or
access covers would result in a leak rate less than that associated with
seal failure or would be discovered prior to removal of the reactor vessel
head.
In the event of such a leak through access covers with the reactor
vessel head rerroved, personnel manning requirements assure sufficient
time to place fuel in a safe condition.
Discussions with the Nuclear Fuel Hardling Supervisor indicated that
no training or procedures are currently in place to address movement of
fuel should a loss of off site power occur. Discussions with operators
previously trained for fuel handling activities indicated that they were
aware that non-powered fuel movement could be acccmplished; however, the
details of how to conduct such operations were not clear. A training
request was promptly initiated to incorporate appropriate actions upon
loss of off site power while moving fuel into fuel handlers' training.
Verification of training conpletion and procedure revision to address
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placement of fuel in a safe condition with a loss of off-site power prior
to the first refueling outage is considered an open item
(50-341/86007-01(DRS)), and will be followed up in an inspection prior to
the first refueling outage.
No violations or deviations were identified.
Inservice Testing _P_rogr_am , Implementation
4.
o
The licensee's inservice testing (IST) program has been reviewed and
approved by the Conmission in the facility Safety Evaluation Report.
During the course of the inspection, the licensee requested copies of
various memos which provide guidance related to interpretations of the
ASME Code requirements for inservice testing. Copies of the pertinent
memos are attached to this report.
The inspector reviewed the licensee's relief requests from the ASME Code
requirements and initial program implenentation, making the following
observations
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a.
Subsection IWV-3300 of Section XI of the American Society of
Mechanical Engineers' Boiler and Pressure Vessel Code (ASME Code)
requires that those valves with remote position indicators be
observed at least cnce every 2 years to verify that valve operation
is accurately indicated. Discussions with the ISI Engineer revealed
that a program ensuring such was in place; however, valves which were
indicated on plant remote shutdown panels had not yet been checked.
The licensee agreed to initiate verification of remote position
indicators on the remote shutdown panels and to complete such
verifications within the ASME Code stipulations. Completion of
position verifications will be tracked as an open item
(50-341/86007-02(DRS)).
b.
As stated in the approved IST program, the licensee is allowed to
satisfy the vibration measurement requirements of the ASME Code by
obtaining vibration data in terms of velocity rather than amplitude.
It was noted that the licensee established a single vibration
acceptance criteria for all IST vibration measurements.
Review of
the High Pressure Core Injection (HPCI) surveillance procedure and'
initial test data revealed that past vibration test data were
unacceptably high. The licensee stated that upon completion of the
surveillance test on the HPCI pump, major maintenance was performed
on the pump, and that new reference values for HPCI pump vibration
measurements as well as associated acceptance criteria wculd be
formulated when sufficient steam was available to run the pump.
In
addition, the licensee stated that the decision whether to
permanently install the vibration transducers or to use hand held
instruments had not yet been made. Establishnent of the method
of vibration testing and appropriate acceptance criteria, and marking
of the data points on the pump will be tracked as an open item
(50-341/86007-03(DRS)).
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c.
The inspector noted that, in most cases, maximum valve stroke times
had been set based on an evaluation of valve data provided by the
manufacturer, the design engineer and plant test data. Review of the
IST program revealed that a relatively small number of valves have
the system response time defined as their maximum allowed stroke
time. The licensee agreed with the inspector's observations and
stated that valve stroke times would be reviewed and revised to
reflect individual valve characteristics and test results. The
definition of specific maximum valve stroke times which are more
indicative of component degradation, is considered to be an open item
(50-341/86007-04(DRS)). The licensee agreed to complete this effort
by the first refueling outage,
d.
A number of administrative procedures addressing the IST requirements
delineated in the ASME Code were in place. The inspector reviewed
the procedures for adequacy and consistency.
The inspector
determined that Code requirements were appropriately and clearly
addressed.
No violations or deviations were identified.
5.
Inservice Testing Instruments
A review of the adequacy of instruments used to obtain inservice testing
data against established requirements was performed by the licensee prior
to program implementation. The inspector evaluated a number of
instruments and discovered that the equipment history file for the HPCI
permanently installed tachometer defined the instrument as non-seismic and
Quality Assurance (QA) level 3; however, the Master Instrument List (MIL)
identified the same instrument as seismic category 1 and QA level 1.
The
licensee stated that this discrepancy was probably due to the fact that at
the time of purchase, it was not clear as to how the tachemeter would be
used.
The loop calibration procedure for the tachometer was not located
during the course of the inspection; however, the inspector reviewed the
calibration procedure for the tachometer sensor and noted that a one point
calibration was performed. The licensee stated that a multi-point
calibration for the loop was probably performed. The inspector also noted
that the tachcmeter was overdue for scheduled calibration. Discussions
with the licensee revealed that the tachometer had been used to obtain
initial reference data for the HPCI pump; however, due to extensive punp
mcdifications and the need to establish vibration acceptance criteria as
discussed in paragraph 3 above, new reference values would have to be
taken.
In effect, data taken with the tachometer had not been used to
verify pump operability under the auspices of the IST prcgram.
The
licensee was unable to identify any other data taken with the tachometer.
Resolution of the calibration status and requirements, and classification,
both seismic and QA level, for the tachometer, as well as evaluation of
the validity of any data taken using the tachometer, is considered an
unresolved item (50-341/86007-05(0RS)).
No violations or deviations were identified.
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6.
Oyen Items
Open items are matters which have been discussed with the licensee, which
will be reviewed further by the inspector, and which involve some action
on the part of the NRC or licensee or both.
Open items disclosed during
the inspection are discussed in Paragraph 3 and 4.
7.
Unresolved It_ ems,
Unresolved itemt. are matters abcut which more information is required in
order to ascertain whether they are acceptable items, open items,
deviations, or violations. Unresolved items disclosed during the
inspection are discussed in Paragraph 5.
8.
Exit Interview
The inspectors met with licensee representatives (denoted in Paragraph 1)
on February 28, 1986, to discuss the scope and findings of the
inspection.
The licensee acknowledged the statements made by the
inspectors with respect to items discussed in the report. The inspector
also discussed the likely informational content of the inspection report
with regard to documents or processes reviewed by the inspectors during
the inspection. The licensee did not identify any such documents /
processes as proprietary.
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NAR 171980
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HEM)RANDUM FOR:
R. C. Lewis, Acting Chief. RO&tS tranch, Region II
FROM:
Smuel E. Bryan, A/D for Field Coordination DR0!, IE
SUSJECT:
OPERASILITY REQUIRE'TNTS FOR PUMPS (AITS NO. F02-700028-N07)
As we understand then, the questions in your February 1 memo are:
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1.
Do the Technical Specification ACTION stat m ent time perios run
consecutive or concurrently with the data evaluation time (96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />)
given in IW-3220 cf Section XI of the ASK So11er and Pressure Vessel
l
Code, l'374 Edition with Addenda thru the Sweer 1975, and
2.
When should the test results be reviewed and, if out-of-specification,
the associated pay declared inoperable?
The answer to the first goestion is the Technical Specification ACTION state-
.~ent tim period starts after the detamination is made that the pay is
inoperable as defined inWlon XI. WP-3230fc). If the data is within the
l'equired Action Rance of Table IW-3100-2 and it is de;;1ded to recalibrate
the ir.strtr.ents and rerun the test, as provided for in IWP-3230(b), the
Technical Specification ACTION statement time starts when the determination
is cade that the data is within the Required Action Range. The reasoning
behind the preceeding statement is that once the detennination is r.ade that
the data is within the Required Action Range the pure must be declered
inoperabic. The provisions in IW-3230 to recalibrate and rerun the test to
show the ptr9 is still capable of fulfilling its function are interpreted by
us as an alternative to replacement or repair, not an additional action that
can be taken before declaring the pump inoperable.
The answer to the second question is that'as soon as the data is recognized
as being within the Required Action Range the pump must be declared inoperable.
Section XI. WP-6230, ' Inservice Test Plans", stateNst the test plan shall
include 'The reference values ' Table IW-3100-1)Is Subsection." limits of Pg and Tb (Tab
IwP-3100-2), and any other values required by th
This statement
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C0'iTACT:
J. C. Stone, IE
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MAR 171980
then requires the acceptance criteria to be included in the test plan.
With that indomation available, the shift supervisor should be able to
make the detemination as to whether or not the data meets the requirements.
The important point is that once the data becomes available that shows the
pop cannot meet the inservice inspection requirements and by dedinition
cannot fulfill its function then the pomp must be declared inoperable.
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We have discussed the abon interpretations with 00R personnel and the
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Standard Technical Specification Group and they agree. If you have any
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further questions, please call.
Samuel E. Bryan
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Assistant Director
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for Field Coordination
Division of Reactor
Operations Inspection. IE
cc:
81. C. Moseley. IE
J. 5. Wetmore. STS
G. Johnson. E8
J. C. Stone. IE
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F. J. Nolan IE
J. I. Riesland. IE
B. R. Messitt, RI!
E. J. Brunner. RI
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G. L. Madsen RIV
J. L. Crews, RV
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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WASWNGTON, D. C. 20555
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Docket No.: STN 50-483
1
MEMORANDUM 'OR: Richard L. Spessard, Dir(ctor
F
Division of Reactor Safety
Region III
FROM:
Hugh L. Thompson, Jr., Director
Division of Licensing
Office of Nuclear Reactor Regulation
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SUBJECT:
CLOSURE VERIFICATION OF NORMALLY CLOSED CHECK VALVES
DURING PREOPERATION TESTING AT CALLAWAY (TIA 83-117)
REFERENCE:
Letter from R. L. Spessard to D. G. Eisenhut on the
above subject, dated November 8, 1983.
The referenced letter requested the staff position regarding testing of
normally closed check vpives for closure capability during preoperational
testing and during plant life.
The staff position is that normally closed
check valves, that have a safety function in the closed position, should
have the closure function verified both during preoperational testing and
periodically throughout the plant life.
In addition, the staff verifies
that closure verification testing is specified in their normal review of
the IST program, and if not, measures are taken to see that the program
is revised.
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In an attempt to have the ASME clarify ambiguities within Section XI of the
ASME Code regarding valve categorization and leak testing, the staff submitted
an inquiry to the society. The response time from the society (approximately
^
one year) was a major factor in the delay of this response to you.
Enclosed
is a more detailed staff evaluation of the subject.
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]\\/Hugh L.
hompso
Jr., Director
Division of Licensing
Office of Nuclear Reactor Regulation
Enclosure: As stated
cc:
T. Martin
P. Bemis
R. Denise
T. Bishop
P. Wohld
, P. Pelke
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