HomeMy WebLinkAbout851166.tiff rJ``Epn"E0Ny� UNITED STATES
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JUL 0 9 1985 ��
In Reply Refer To: � ' ar
Docket: 50-267 ��)� �t_ic“
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GREELY. Coco.
Public Service Company of Colorado
ATTN: 0. R. Lee, Vice President
Electric Production
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
We have reviewed your June 17, 1985 (P-85212) , response to our March 11, 1985,
request for resolution of our concerns over Beta radiation in liquid
effluents. We understand that you are committed to:
1. Perform releases from the Reactor Building sump by a batch method,
whereby the sump contents will be sampled and analyzed prior to
commencing a release and at 24-hour intervals during the release (we
recognized that some addition to the sump may occur during the release
period); and
2. Continue the investigation into installing in-line, Beta-sensitive,
effluent monitors.
We find these commitments to be an acceptable, interim resolution of our
concerns and will include them in the listing of commitments that will be
confirmed in connection with our authorization of plant restart.
If you have any questions on this subject, please contact the NRC Project
Manager.
Sincerely,
. H. Johnson, Chief
Reactor Projects Branch 1
cc: (cont. on next page)
851166
a. -,l , IC'.
Public Service Company of Colorado -2-
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. O. Box 840
Denver, Colorado 80201
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. 0. Box 85608
San Diego, California 92138
Kelley, Stansfield & O'Donnell
Public Service Company Building
550 15th Street, Room 900
Denver, Colorado 80202
Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Manager
Nuclear Licensing/Fuels Div.
Public Service Company of Colorado
P. O. Box 840
Denver, Colorado 80201
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 193
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
Colorado Radiation Control Program Director
°`EPa net°<qr UNITED STATES
a NUCLEAR REGULATORY COMMISSION
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9» S-I• U 611 RYAN PLAZA DRIVE, SUITE 1000
ARLINGTON. TEXAS 76011
JUL C 3 1985 !`" 1 of ,,, P ,t„,,�.,
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Docket: 50-267 i' -l
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Mr. 0. R. Lee, Vice President GREELEY. COLd,
Electric Production
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
Our September 22, 1983, letter provided the Preliminary Safety Evaluation
Related to the Post Accident Sampling System (NUREG-0737 , Item II.B.3) for the
Fort St. Vrain Station (FSV) . Your October 28, 1983 (P-83352) , and July 2 and
16, 1984 (P-84192 and P-84216) letters provided additional information to
resolve the open items in our evaluation. We have reviewed your submittals
and have determined that additional information is still required for us to
resolve this issue. The results of our review are contained in the enclosed
Supplemental Safety Evaluation (SSE) .
We have found that you now meet eight of the nine pertinent criteria contained
in Item II.B.3 of NUREG-0737 and that some revision is required in the
procedures used for estimating the extent of core damage in order for us to
find the procedure to be acceptable.
Therefore, we request that you review the attached SSE and provide your
comments on resolving our concern within 60 days of the date of this letter.
If you have any questions on this matter, contact the NRC Project Manager -
P. Wagner - at (817) 860-8127.
Since this reporting requirement relates solely to FSV, OMB clearance is not
required under P.L.96-511 .
Sincerely,
\\l
Eric Johnson , Chief
Reactor Project Section 1
Enclosure: SSE on I1.6.3
cc: (see next page)
i2 _ I _ +,. n I.4/S'C
Mr. 0. R. Lee, Vice President -2-
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. 0. Box 85608
San Diego, California 92138
Kelley, Stansfield & O' Donnell
Public Service Company Building
550 15th Street, Room 900
Denver, Colorado 80202
Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Manager
Nuclear Licensing/Fuels Div.
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 191
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
Colorado Radiation Control Program Director
Supplemental Safety Evaluation by
the Office of Nuclear Reactor Regulation
Related to Operation of
Fort St. Vrain Nuclear Generating Station
Public Service of Colorado
Docket No. 50-267
Post-Accident Sampling System (NUREG-0737,II.B.3)
I . Introduction
In our safety evaluation, we concluded that the licensee's proposed Post-
Accident Sampling System (PASS) met six of the nine criteria in Item
II .6.3 of NUREG-0737 which are relevant for a gas-cooled reactor. The three
criteria which were not fully resolved were:
Criterion (2) Provide a plant-specific core damage estimate procedure to
include radionuclide concentrations and other physical
parameters as indicators of core damage.
Criteria (9) Provide information on the procedure for taking samples of
and (10) highly radioactive coolant for gamma spectrometry in such
a manner that the activity of the sample does not exceed the
measurement capability of the spectrometer.
II. Evaluation
By letters dated October 28, 1983 and July 2, 1984, the licensee provided
additional information.
Criterion (2) :
The licensee shall establish an onsite radiological and chemical analysis
capability to provide, within the three-hour time frame established above,
quantification of the following:
-2-
a) certain radionuclides in the reactor coolant and containment
atmosphere that may be indicators of the degree of core damage
(e. g. , noble gases, iodines and cesiums, and nonvolatile isotopes) ;
b) hydrogen levels in the containment atmospheres;
c) dissolved gases (e. g. , H2) , chloride (time allotted for analysis
subject to discussion below), and boron concentration of liquids;
d) alternatively, have in-line monitoring capabilities to perform
all or part of the above analyses.
The PASS provides in-line monitoring for noble gas activity, CO and moisture
in the helium coolant, as well as for radioactivity in the reactor building
stack gas. The PASS also provides the capability to collect grab samples
of the coolant and of the reactor building atmosphere that can be transported
to the radio-chemical laboratory for CO, CO2, H2, CH4, N2 and radionuclide
analyses. These species are indicators of core damage in a gas-cooled
reactor, and their relative magnitudes indicate core temperature, fuel particle
failure, air ingress or water ingress. We find that the licensee partially
meets Criterion (2) by establishing an onsite radiological and chemical
analysis capability. However, the licensee should provide a procedure,
consistent with the clarification of NUREG-0737, Item II.6. 3, Post-Accident
Sampling System, transmitted to the licensee on July 9, 1982, to estimate the
extent of core damage based on radionuclide concentrations and taking into
consideration other physical parameters such as the concentrations of other
gases and core temperature data. Guidance for the procedure to estimate core
damage for water-cooled reactors was provided. The procedure for estimating
core damage should be consistent with those portions of these recommendations
which are applicable to a gas-cooled reactor.
-3-
The procedure for estimating core damage presented in the letter of July
2, 1984, is not acceptable because it is based solely on the Xe133 concen-
tration in the coolant. An acceptable procedure should include consideration
of (1) the concentrations of other volatile radionuclides such as xenons,
kryptons and iodines, (2) the concentration of other gaseous species, such as
H20, CO, CO2, H2, CH4 and N2, and (3) core temperature.
The procedure should indicate how these additional considerations would (1)
confirm the core damage estimate based on Xe133 (2) provide an estimate
of core damage due to water or air ingress, and (3) provide an estimate of
core temperature.
Criterion (9):
The licensee' s radiological and chemical sample analysis capability
shall include provisions to:
a) Identify and quantify the isotopes of the nuclear categories
discussed above to levels corresponding to the source terms
given in Regulatory Guide 1 . 3 or 1 .4 and 1.7. Where necessary
and practicable, the ability to dilute samples to provide capability
for measurement and reduction of personnel exposure should be
provided. Sensitivity of onsite liquid sample analysis capability
should be such as to permit measurement of nuclide concentrations
in the range from approximately 1p Ci/g to 10 Ci/g.
b) Restrict background levels of radiation in the radiological
and chemical analysis facility from sources such that the
sample analysis will provide results with an acceptably small
error (approximately a factor of 2). This can be accomplished
through the use of sufficient shielding around samples and out-
side sources, and by the use of a ventilation system design
which will control the presence of airborne radioactivity.
-4-
The radionuclides in both the helium coolant and the reactor building
atmosphere samples will be identified and quantified using the onsite
gamma spectrometer. By letter dated July 2, 1984, the licensee provided
information on the procedure to take small low pressure samples of the
highly radioactive coolant during the period of maximum activity between
approximately 5 hours and 7 days after the onset of a loss-of-cooling
accident. By controlling the sample size, the measurement capability of
the gamma spectrometer will not be exceeded. Radiation background levels
will be restricted by shielding. Ventilated radiological and chemical
analysis facilities are provided to obtain results within an acceptably
small error (approximately a factor of 2). We find that these provisions
meet Criterion (9) and are, therefore, acceptable.
Criterion (10):
Accuracy, range, and sensitivity shall be adequate to provide pertinent
data to the operator in order to describe the radiological and chemical
status of the reactor coolant systems.
The accuracy, range, and sensitivity of the PASS instruments and analytical
procedures are consistent with the recommendations and the clarifications of
NUREG-0737, Item II.6. 3, Post-Accident Sampling Capability, transmitted to the
licensee on June 30, 1982. Therefore, they are adequate for describing the
radiological and chemical status of the reactor. The analytical methods and
instrumentation are capable of operation in the post-accident sampling
environment. No additional training of chemistry personnel is required
because the same systems are used for normal and post-accident sampling and
analysis. The letter of July 2, 1984, describes provisions to limit sample
size, enabling the onsite measurement of radionuclide concentrations in the
helium coolant in the post-accident period of maximum coolant radioactivity.
We find that these provisions meet Criterion (10) and are, therefore,
acceptable.
-5-
Conclusion
We conclude that the post-accident sampling system partially meets the
criteria of Item II.6.3 of NUREG-0737. Two of the eleven criteria are not
applicable to a gas-cooled reactor. The licensee' s proposed methods to meet
eight of the remaining nine criteria are acceptable. The criterion which has
not been fully resolved is:
Criterion (2): Provide a core damage estimate procedure to include
consideration of coolant concentrations of volatile radionuclides and
gaseous chemical species together with other physical parameters as
indicators or core damage.
J``tio REGp49r UNITED STATES
9
F NUCLEAR REGULATORY COMMISSION
a �" . . 3 REGION IV
m0 � S
611 RYAN PLAZA DRIVE, SUITE 1000
r�o t ARLINGTON, TEXAS 76011
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JUL 0 7985 WELD gm--vn
In Reply Refer To: EI �,1 C
Docket 50-267 `L-"-- ,-,.J'1 , 5i 1
lin, 'U
CREELEY
Public Service Company of Colorado
ATTN: 0. R. Lee, Vice President
Electric Production
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
We have completed our review of the various submittals related to the Fort St.
Vrain Prestressed Concrete Reactor Vessel (PCRV) prestressing tendon wire
corrosion problems. The results of our review are contained in the attached
Safety Evaluation (SE). We have concluded that sufficient assurance of Safe
operation of the PCRV exists to allow the resumption of plant operations,
provided the following additional items are implemented:
1. Incorporate the modified tendon surveillance program into the Technical
Specifications.
2. Provide the NRC with the results of the tendon surveillance program every
6 months, including the results of inspections related to determining
whether there is any evidence of anchorage stress washer failure.
We have reviewed your May 20, 1985 (P-85176) , and June 14, 1985 (P-85199)
letters and find that your commitments acceptably resolve the above concerns.
We will include the above in the listing of the various PSC commitments which
will be confirmed in connection with authorization of plant restart.
We have also enclosed a copy of the evaluation report, prepared by our
consultants at the Los Alamos National Laboratory, for your information and
comment.
12 r n..-1-r: -r i I cG -
Public service Company of Colorado -2-
If you have any questions on this subject, please contact the NRC Project
Manager - P. Wagner - at (817) 860-8127.
Sincerely,
A.
E. H. Johnson, Chief
Reactor Project Branch 1
Enclosures:
1. Safety Evaluation
2. LANL Evaluation
cc:
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. 0. Box 85608
San Diego, California 92138
Kelley, Stansfield & O'Donnell
Public Service Company Building
550 15th Street, Room 900
Denver, Colorado 80202
Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Manager
Nuclear Licensing/Fuels Div.
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Public Service Company of Colorado -3-
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 193
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
2J`EpF REDO UNITED STATES
Z' �1?, NUCLEAR REGULATORY COMMISSION
Cl
oyy 3 REGION IV
ll I y
611 RYAN PLAZA DRIVE, SUITE 1000
ARLINGTON, TEXAS 76011
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
PUBLIC SERVICE COMPANY OF COLORADO
FORT ST. VRAIN
DOCKET NO. 50-267
PCRV TENDON CORROSION INVESTIGATION
AND PROPOSED REMEDY
PCRV Tendon Degradation - Background:
In 1984 during the scheduled PCRV tendon surveillance, the licensee
discovered that certain PCRV tendons had broken and corroded wires. In
order to determine the extent of this problem, the licensee increased the
number of visual examinations of accessible tendons ends. He also
performed a number of lift-off tests. Metallurgical examinations of the
tendon wires and tests on the protective grease that were performed by the
licensee indicate that the corrosion is the result of microbiological attack
on the original tendon grease. The licensee has proposed to halt this
degradation by filling the tendon sheaths with an inert nitrogen "blanket".
As a mechanism for monitoring the condition of the tendons, the licensee
has proposed a surveillance program that increases the frequency of the
visual inspection and lift-off tests. The surveillance program would
compare an uncorroded tendon control group with a corroded tendons group to
establish the effectiveness of the corrosion arresting method and the trend
in the tendon wire degradation. The program would include samples of the
longitudinal , circumferential , and crosshead tendons proportional to the
population of the tendon types.
Evaluation:
1. Monitoring:
The staff evaluated the licensee' s proposed program for monitoring the
PCRV tendons and finds the proposal acceptable for assuring PCRV integrity
in the near term with certain modifications , as discussed
below:
The surveillance program, as proposed, would produce a sample of
significant size to indicate the trend of the tendon wire degradation
and the effectiveness of the corrosion arresting method. However, the
information gathered by the licensee from the past and future
surveillance activities should be integrated into a complete visual
presentation covering all tendons. The purpose of this presentation
format would be to provide better information of the extent and
significance of the tendon degradation problem. The licensee has
committed to incorporate the modified tendon surveillance requirements
into the technical specifications.
-2-
2. Corrosion Control :
The staff evaluated the integrity of the PCRV with the degraded tendons
in a safety evaluation dated May 16, 1984. The staff findings were
that the reactor vessel was capable of withstanding the operating
pressures with the degraded tendons as determined at that time. Since
May 1984, a few additional wires have broken but the reactor vessel
remains able to adequately withstand the operating pressure. The
licensee plans to use a nitrogen "blanket" in the tendon sheaths to
halt the degradation of tendons; however, our earlier evaluation
indicated concern with this approach. Accordingly, we recommend that
the licensee carefully evaluate the effectiveness of other techniques,
in terms of their ability to remove oxygen and moisture, and their long
term effects on tendon corrosion.
3. Corrosion Problems at Other Plants:
The corrosion problem at Fort St. Vrain (FSV) appears to be different
from the tendon problems recently experienced at some other nuclear
plants. In the other plants, tendons are used in the containment
structure which experiences ambient temperatures and the tendon sheaths
are filled with grease. The tendons at Fort St. Vrain are located in
the reinforced concrete reactor vessel . These FSV tendons experience
higher temperatures than other plants and are in sheaths not filled
with grease. The FSV tendon wires themselves are protected by a grease
coating and the tendon sheath annulus is coated on the inside with a
layer of grease.
A failure mechanism has been identified at the other plants related to
stress corrosion cracking of the tendon wire stress washers when
water was present, predominately in the lower end of the vertical
tendons. The stress washers are manufactured from a high strength steel
which is susceptible to stress corrosion cracking when exposed to a
source of hydrogen.
The tendon wires at FSV appear to be corroding from the attack of
formic and acetic acids generated from microbilogical sources. The
corrosion and failures seen to date at FSV seem to be limited to the
wires themselves with only one incident of corrosion occurring on
the stress washer. No evidence of failure of the stress washers has
been detected to date. The licensee has visually examined 10 of the
34 accessible bottom stress washers of the longitudinal tendons and
reported the results in a letter dated June 7, 1985. No evidence
of cracking was found. However, the possibility of stress washers
failing from corrosion cannot be ruled out. The continued presence of
moisture in the tendon tubes could lead to failure of the stress
washers as seen at other plants. The licensee has proposed an
intensified surveillance program which consists of visual inspection of
the anchorages and lift-off tests. The licensee proposed to
incorporate these inspection requirements into the plant technical
-3-
specifications under Section 3/4.6.4 "PCRV Integrity" . This
intensified surveillance program will require a visual inspection and a
report on a sample of 56 tendons at six month intervals. The
surveillance program will also require 37 tendons to be lifted-off their
shims to determine the amount of prestress available. A sample of 12
tendons are designated as a control set for visual inspection and 8
tendons are the control set for lift-off tests. The samples of 44
visual inspection and 25 lift off tendons will be rotated thru the
tendon population.
4. Restart and Re-evaluation:
The staff has reviewed the licensee's proposed surveillance and the
commitment to incorporate the surveillance requirements into the
technical specification. The staff accepts the tendon surveillance
proposal and find that the proposal would lead to maintaining the
structural integrity of the reactor vessel . The staff also finds that
the visual inspection of the tendon wire anchorage washer is currently
sufficient to determine if failure of the washers has occurred. The
increased tendon surveillance would be sufficient to show any tendency
of the stress washers to fail similar to those at another nuclear
plant. The staff finds the plant structurally ready for restart.
Principal Contributor:
H. Polk, DE
Date: July 2, 1985
Evaluation of Control Rod Drive Mechanism and
Reserve Shutdown System Failures,
and PCRV Tendon Degradation Issues
Prior to Fort St. Vrain Restart
NRC Fin No. A-7290
March 12, 1985
Los Alamos National Laboratory
Deborah R. Bennett, Q-13
Gerald W. Fly, Q-13
L. Erik Fugelso, Q-13
Robert Reiswig, MST-6
Stan W. Moore, Q-13
Responsible NRC Individual and Division
J. R. Miller/ORBS
Prepared for the
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
DISCLAIMER
Tnis report was prepared as an account of work sponsored by
an agency of the United States Government. Neither the
United States Government nor any agency thereof, or any of
their employees, makes any warranty, expressed or impliec,
or assumes any legal liability or responsibility for any
third party's use, of any information, apparatus, product
or process disclosed in this report or represents that its
use by such third party would not infringe privately owned
rights.
- i -
Table of Contents
1.0 Background
1 .1 Assessment Report Restart Issues
1.2 PCRV Tendon Restart Issues
1.3 Purpose of the Technical Evaluation
2.0 Control Rod Drive and Orifice Assemblies
2.1 Failure Mechanisms
2.1.1 Motor Brake Malfunctions
2.1.2 Reduction Gear Mechanism Malfunctions
2.1.3 Motor and Motor Bearing Malfunctions
2.2 Refurbishment Program
2.2.1 CRDOA Refurbishment
2.2.2 Control Rod Cable Replacement
2.2.3 Reserve Shutdown System Material-Related Failure
2.2.4 Purge Flow and Seal Replacement
2.3 CRDM Temperature Recording and CRDM Requalification
2.4 CRDM Preventive/Predictive Maintenance and Surveillance
2.4.1 CRDM Preventive/Predictive Maintenance
2.4.2 CRDM Interim Operational Surveillance
3.0 Moisture Ingress Issues
4.0 PCRV Post-Tensioning Tendon System
4.1 Tendon Accessibility, Extent of Known Degradation and Failure
Mechanism
4.2 Tendon Corrosion Corrective Measures
4.3 PCRV Tendon Interim Surveillance
4.4 PCRV Structural Calculations by Los Alamos National Laboratory
5.0 Conclusions
6.0 References
- ii -
Evaluation of Control Rod Drive Mechanism and
Reserve Shutdown System Failures,
and PCRV Tendon Degradation Issues
Prior to Fort St. Vrain Restart
1.0 Background
On June 23, 1984, following a moisture ingress event resulting in a
. loss of purge flow to the Control Rod Drive Mechanism (CRDM) cavities, 6
of 37 control rod pairs in the Fort St. Vrain (FSV) High Temperature Gas-
- Cooled reactor failed to insert on a scram signal. Subsequently, all six
control rod pairs were successfully driven into the core.
In July, 1984, an assessment team consisting of Nuclear Regulatory
Commission (NRC) personnel from Headquarters, Region III and Region IV,
and their technical consultant, Los Alamos National Laboratory, conducted
an on-site review of the Control Rod Drive Mechanism failures, overall
conduct of plant operations, adequacy of technical specifications and a
review of the continued moisture ingress problem. An additional plant
visit in August, 1984, reviewed CRDM instrumentation anomalies.
1 .1 Assessment Report Restart Issues
Tne results of both assessments were reported in the "Preliminary
Report Related to the Restart and Continued Operation of Fort St. Vrain
Nuclear Generating Station"1, in October, 1984. Tne report concluded
that Fort St. Vrain should not oe restarted until modifications and/or
other corrective actions had been taken, or until all control rod drive
mechanisms had been inspected and refurbished to provide reasonable as-
surance that the control rods would insert automatically on receipt of a
scram signal . More specifically, and as included in this technical eval-
uation, Reference 1 required Public Service Co. of Colorado (PSC) to com-
plete the following, prior to restart:
a. The licensee must identify the CRDM failure mechanism(s) and
take corrective actions, or, if the mechansm(s) cannot be posi-
tively identified, take other compensatory measures to provide
assurance of control rod reliability, which could reasonably
include refurbishment of all CRDMs.
- 1 -
b. The licensee must outline and commit to periodic inspection ,
preventive maintenance and surveillance programs for control
rod drive mechanisms and associated position instrumentation.
A change in the Technical Specifications shall be proposed to
implement a weekly control rod exercise surveillance program
for all partially or fully withdrawn control rods. A Limiting
Condition for Operation should define control rod operability,
and the minimum requirements for rod position indication.
c. The licensee must functionally test one-20 weight % boron and
one-40 weight % boron hopper from the Reserve Shutdown System
(RSS) , to assure the full availability of the RSS, prior to
restart. The licensee must outline and commit to periodic in-
spection, preventive maintenance and surveillance programs for
Reserve Shutdown System material. A change in the Technical
Specifications shall be proposed to implement the RSS surveil-
lance program. A Limiting Condition for Operation should define
and confirm the operability of the Reserve Snutdown System.
d. The licensee should develop a procedure requiring reactor shut-
down when high levels of moisture exist in the primary coolant,
or when CRDM purge flow is lost.
e. The licensee should implement a procedure for recording repre-
sentative samples of CRDM temperatures at all operating condi-
tions, until continuous recording capability is available.
f. The licensee should implement procedure to prevent overdriving
the control rods past the "Rod-In" limit.
g. The licensee must develop a plan to implement any modifications
recommended by the PSC Moisture Ingress Committee that are
determined, by PSC, to have a high potential for significantly
reducing the severity and frequency of moisture ingress events .
1 .2 PCRV Tendon Restart Issues
As a result of previously identified tendon degradation in the Pre-
stressed Concrete Reactor Vessel (PCRV) post-tensioning system, PSC must
complete the following, as comfirmed by Reference 2, prior to restart:
- 2 -
a. The licensee should submit documentation evaluating the mechan-
ism(s) causing corrosion on and failure of tne PCRV tendon
wires, and corrective measures to eliminate further tendon
degradation, thereby assuring the continued structural integ-
rity of the PCRV and its post-tensioning system.
b . The licensee should propose and implement a tendon surveillance
program that determines the extent of current tendon degrada-
tion in the PCRV, and that systematically monitors the rate of
tendon corrosion.
1 .3 Purpose of tne Technical Evaluation
This document provides a technical review of tne restart issues
identified above, and the corrective measures and/or actions proposed by
licensee, based on the licensee.'s January 31, 1985 submittals (References
given as used in this document), and the meeting between the licensee and
NRC at the FSV plant site on February 20-22, 1965, as transcribed in
References 3, 4 and 5.
2.0 Control Rod Drive and Orifice Assemblies
This section includes a review of CRDM failure mechanisms, Control
Rod Drive and Orifice Assemblies (CRDOA) refurbishment, CRDM temperature
recording and requalification testing, CRDM preventive/predictive main-
tenance and surveillance.
2.1 Failure Mechanisms
The failures of control rod pairs to scram, under various operating
oonditions , has been documented since 1982,6'7 and are as noted in
Table 1 by region, CRDOA number and CRDM purge flow subheader (total of 8
purge flow subheaders) .
- 3 -
Table 1. Control Rod Failures
Date 2/22/82 6/23/84 1/14/85
Region 7 28 6 7 10 14 25 28 28 31 32
CRDOA # 18 44 29 18 14 25 7 44 36 17 15
CRDM Purge 1 1 6 1 7 2 5 1 1 2 3
Subheader #
High moisture content in the primary coolant and loss of purge flow
were common modes during the 2/22/82 and 6/23/84 events. Substantial
descriptions and operating characteristics of the drive motor, friction
braKe and dynamic braking, the reduction gear mechanism, the cable drum
and cable, and the bearing lubricant are provided in Reference 6. Tne
licensee reviewed those CRDM components that could have caused the fail-
ures to scram, and postulated various failure mechanisms that could have
interacted on each component, as described below.
2.1.1 Motor Brake Malfunctions
During a scram, the motor brake is de-energized and released, thereby
freeing the motor rotor shaft and gear train assembly to rotate under the
torque applied by the weight of the control rods. In the motor brake
assembly, failure of the scram contactor to de-energize dc power to the
electromagnet was discounted because the operator had removed the brake
fuses following the CRDM failures to insert the control rod pairs.
According to the licensee, electromagnetic remanence and reduced
spring constant in the brake spring plungers (due to elevated tempera-
tures) were eliminated as possible failure mechanisms. Some corrosion
and rust was identified on the brake disks of CRDOAs 25, 18 and 29. How-
ever, the disks of a CRDM motor brake assembly with "discoloration and
whatever surface variations„3, p.149
could not be made to stick in an
elevated temperature helium environment with high moisture content (test
T-228) . The licensee concluded that the motor brake was not instrumental
in the failures to scram.
Los Alamos agrees with the licensee that the motor brake assembly
was probably not related to the CRDM failures.
- 4 -
2.1.2 Reduction Gear Mechanism Malfunctions
The reduction gear train is driven by the motor rotor shaft, and
rotates the cable drum with a gear ratio of 1150 between the motor and
drum. The condition of the reduction gear mechanism was postulated by
the licensee to potentially contribute to a failure to scram through gear
tooth or bearing damage, by the presence of large particulate matter pre-
venting gear rotation, and/or the presence of particulate matter in the
gears or gear bearings reducing the gear train efficiency--i.e. , the
torque transmitted from the gear train to the motor rotor shaft might
have been insufficient to overcome the friction of the motor bearings.
The licensee stated that no major damage has been identified on sev-
eral inspected reduction gear mechanisms, even though some wear and debris
were observed. The licensee's analyses indicated that particulates with
a size of 0.030 inches in diameter or greater, and with a comparable
material composition as the reduction gear mechanism (implying comparable
hardness) , would be required to inhibit gear or gear bearing rotation.
Analyses of CRDOA debris8 showed the presence of rust, molybdenum di-
sulfide and traces of silicon particles, which are relatively soft mate-
rials. The average particle of 0.020 inches was uniform in size, and
tended to be smaller than that thought to innibit rotation, even though
rust particles on the order of 0.0625 to 0.125 inches were scraped off
the ring gear pinion housing of CRDOA 18. However, the presence of debris
in the gears and gear bearings tended to support the licensee's case of
reduced gear train efficiency when sensitivity studies indicated that the
motor bearings were only three times more sensitive to debris than the
first pinion gear mesh of the reduction gear assembly, and 500 times more
sensitive to debris than the cable drum bearings.
Los Alamos agrees with the licensee that the presence of debris ,
especially in the first pinion gear mesh and the gear bearings, could
reduce the efficiency of the reduction gear train, and thereby contribute
to CRDM failures.
2.1.3 Motor and Motor Hearing Malfunctions
During a scram, the motor is de-energized and does not directly con-
tribute to the scram process, even though it operates as an induction
generator. However, because 1b-20 inch-ounces of resisting torque on the
- 5 -
motor rotor shaft can forestall scram,9 the friction from the motor
bearings can be a significant contributor to the failure to scram. Pos-
sible contributions to increase the friction include debris in the bear-
ing race, wear on the bearing ball or race, and changes in the lubricant
properties during adverse conditions.
The licensee reported that debris was observed in the bearing races
of CRDOAs 7, 18 and 44, "roughness in rolling the bearing balls was noted
in virtually all of the unrefurbished bearings examined",6 and minor
race wear was identified. Reference 8 verified that the major debris
constituents could be attributed to the motor bearing materials (which
includes bearing balls, races, and other bearing components), whereas
minor constituents were indicative of the motor itself. The analysis
provided little evidence to support the theory that debris had been
"washed" into the bearing races. The licensee also determined, because
of the relatively close bearing tolerances and because rod weight alone
might not produce sufficient "crushing force" to deform bearing particu-
late, that bearing operation could be reduced with the presence of par-
ticulate matter. Tne licensee therefore concluded that internally gener-
ated wear byproducts in the CRDM motor bearings contributed significantly
to tne failures to scram.
Los Alamos agrees with the licensee that increased friction in the
motor bearings, caused by the presence of internally generated debris,
could have been a likely contributor to the failures to scram. Los Alamos
also agrees with the licensee that the "wash in" theory of debris into
the motor bearing races is not supported.
Los Alamos contends that the loss of CRDM purge flow allowed primary
coolant with high moisture content to enter the CRDM cavity. An indepen-
dent literature search indicates that the dry film lubricant, molybdenum
disulfide, MoS2 , experiences an increase in its coefficient of fric-
tion in the presence of moisture38. Therefore, the increased frictional
coefficient of the lubricant on the motor bearings, MoS2, may have also
contributed to the CRDM failures by resisting motor rotor shaft rotation.
2.2 Refurbishment Program
Tne cause of tne failures to scram could be attributed to several
mechanisms such as reduced reduction gear train efficiency, internally
- 6 -
generated debris in the motor bearings causing increased friction on the
motor rotor shaft, and possibly an increased frictional coefficient in
the dry film lubricant in the presence of moisture. Because the CRDM
failure mechanism cannot be specifically delineated, and because of CRDM
cable failures, the licensee has undertaken a refurbishment program, in-
volving the CRDM motors and reduction gear mechanisms, on all 37 CRDMs.
The licensee reported that the CRDM refurbishment process and a testing
program will ensure the ability of the control rods to scram under oper-
ating conditions.
In addition, the licensee has elected to replace the control rod
cabling and other connecting hardware in light of recently identified
stress corrosion problems, to replace the Reserve Shutdown System material
due to the discovery of material "bridging" during hopper discharge, anc
to install seals around certain penetrations into the CRDM cavity to
mitigate the effects of primary coolant ingress by natural circulation.
2.2.1 CRDOA Refurbishment
The licensee has proposed complete refurbishment of all Control Rod
Drive and Orificing Assemblies to ensure that the CRDOAs will perform
their intended safety functions, and to avoid potential operability prob-
lems that could limit plant availability. As specified in Reference 10,
the following major components are to be inspected, tested, refurbished
or replaced, as necessary:
1. Control Rod Drive (200) Assembly--shim motor and brake assembly,
bearings, reduction gears, limit switches/potentiometers.
2. Orifice Control Mechanism--orifice control motor, bearings,
potentiometer, gears, drive shaft and nut, drive shaft housing.
3. Control rod clevis bolts.
4. Reserve Shutdown System--boron balls, rupture disks, DP switch.
Design modifications include the replacement of control rod cables ,
cable end fittings, and cable clevis bolts, the installation of new purge
seals into the CRDM cavity, and the installation of RTDs (Resistance Tem-
perature Detectors) in all CRDOAs--the impact of tnese design changes
will be evaluated later in this report.
Each CRDOA will undergo the following series of scram tests in the
refurbishment process6: a pre-refurbishment, in-core full scram test; a
- 7 -
pre-refurbishment full scram test in the Hot Service Facility (HSF) ; a
scram test with refurbished reduction gear mechanism and unrefurbished
shim motor, using dummy weights; a full scram test using a "standardized"
motor, using dummy weights; a scram test with completely refurbished 200
assembly, using dummy weights ; a post-refurbishment, full scram test in
the HSF; and finally, a post-refurbishment, full in-core scram test.
As designated by the licensee in Reference 6, back-EMF voltage meas-
urements from the shim motor will be taken for the series of scram tests
conducted before, during and after refurbishment, and should define the
CRDM operating characteristics. From the back-EMF voltage measurements,
the licensee states that they can generate the following information--
voltage versus time, frequency versus time, voltage versus frequency,
acceleration versus time, torque versus time, peak angular velocity, time
to peak back-EMF and angular velocity, average torque on motor rotor dur-
ing acceleration to peak velocity, maximum torque on motor rotor each 10
second interval, maximum deviation of torque values each 10 second inter-
val , and gear train efficiency.
The licensee has proposed a CRDOA refurbishment acceptance criterion,
taking into account the results of the back-EMF voltage measurements and
the resulting calculations of acceleration and torque such that6:
1. Tne minimum calculated average torque during acceleration to
peak velocity will be 17.0 inch-ounces ; this value corresponds
to an average acceleration to peak velocity of 98.83 radians/
second
2. The maximum torque calculated during "steady-state" will be 7.0
inch-ounces.
According to the licensee, final acceptance of a refurbished CRDOA will
be based upon the results of its in-core full scram test.
Los Alamos agrees with the mechanical refurbishment of all CRDOAs,
as the program is currently being implemented by the licensee. In par-
ticular, the replacement of shim motor bearings3' pp' 174-75 is con-
sidered essential to the refurbishment process. However, the current
program of mechanical refurbishment alone cannot ensure CRDOA
operability.
- 8 -
From the documentation presented by the licensee and reviewed earlier
in this section, Los Alamos believes that the proposed back-EMF testing
and acceptance criteria have potential in providing a data base from which
control rod operability might be determined. But, an element of uncer-
tainty, as to CRDOA operability based on back-EMF testing, is introduced
because the test method and interpretation of its results are still in
the developmental stages, and because in-core full scram testing of re-
furbished CRDOAs has not yet taken place.
Los Alamos recommends that the back-EMF testing method continue to
be developed, that the further collection of back-EMF information be used
in preparing a statistical data base for possibly defining CRDOA opera-
bility, and that more attention be paid to the initial, start-up scram
characteristics of the CRDOA, in developing a better understanding of
break-away torque effects. In line with Region IV's increased inspection
of the refurbishment process, we suggest a review, by Region IV, of all
testing results pertaining to CRDOA refurbishment acceptability, after
in-core testing is complete, but prior to startup. As an additional
method to ensure CRDOA operability during scram, a procedure requiring
control rod run-in is recommended.
As a post-startup item, Los Alamos recommends that a final determina-
tion be made as to the suitability and acceptability of back-EMF testing
in defining CRDOA operability.
2.2.2 Control Rod Cable Replacement
In September, 1984, the control rod cable on CRDOA 25 was severed in
several places during an investigation of a slack cable indication.11
A subsequent metallurgical examination12 of the austenitic 347 stainless
steel cable indicated that the cable surface was pitted and cracked, that
the delta-like material cracks were typical of stress corrosion cracks ,
and that the fracture surfaces were brittle in nature. Further investi-
gation revealed that the 347 SS cable material was susceptible to stress
corrosion when under the existing stressed conditions, and in the presence
of chlorides and moisture.
Tne potential sources of the chlorides in the primary coolant'con-
tributing to the chloride stress corrosion are reviewed in Reference 13.
The licensee states that the chlorine occurs as two different species--HC1
- 9 -
gas and a salt; the sources of the gas species include the fuel rods ,
H-327/H-451 graphite, PGX/HLM graphite and the Ti sponge, whereas the
sources of the salt species include the ceramic insulation, concrete and
water, all to varying degrees.
As part of tne overall CRDOA refurbishment program, the licensee
elected to replace the control rod cable with Inconel 625, which is con-
sidered resistant to chloride stress corrosion, and has increased strength
and fatigue properties over the former 347 SS. Cable components and con-
necting hardware that were made from materials susceptible to stress cor-
rosion, and are being replaced with materials more resistant to stress
corrosion include:
Component Material
1. Cable and rod portion Inconel 625--high strength
of the ball end and resistance to oxidation
2. Anchor, set screw Martensitic steel-high
strength, ability to be
nitrided, resistance to
oxidation
3. Spring, connecting bolt Inconel X-750--high yield
strength, resistance to
oxidation.
Drawing numbers and material information are available in Reference 12.
A safety analysis of tne material changes in tne reactor control rod drive
and orificing assembly, which are classified as Class I, Safety Related
and Safe Shutdown components, is included in Reference 14.
Los Alamos metallurgical analyses on a sample of the corroded control
rod cable15 also indicate pitting on the cable surface, ductile and
brittle fracture surfaces, and to a lesser degree than the licensee,
cracking indicative of stress corrosion cracking. Qualitative measure-
ments confirm the presence of chlorine on fracture surfaces. Therefore,
Los Alamos agrees tnat chloride stress corrosion contributed to the de-
graded condition of the control rod cable. The Los Alamos analysis also
observed that a certain particle removed from between the individual cable
strands of the Los Alamos sample had a "shaved" appearance, and was
- 10 -
identified as a 7000 series aluminum alloy--the licensee noted that the
control rod cable drum is constructed of 7075 aluminum alloy4' p.20,
and that no excessive drum wear had been noted.
Los Alamos agrees that the licensee's recommended material changes
tend to improve the overall resistance of the CHDOA cable components and
connecting hardware to chloride stress corrosion. However, Los Alamos
also recommends a continued analysis into the sources of tne chlorine and
its effects on other reactor components, especially components potentially
subjected to high chlorine concentrations such as the bottom plenum or
other areas where water could accumulate.
2.2.3 Reserve Shutdown System Material-Related Failure
In November, 1964, during the required testing of a 20 weight %
boron and a 40 weight % boron hopper in the Reserve Shutdown System, only
half of tne RSS material in CRDOA 21 (40 weight % boron) was discharged.
The licensee's examination of the undischarged material revealed that the
B4C boronated graphite balls had "bridged" together through a crystal-
line structure on the ball surfaces. Analyses on the crystalline material
indicated that it was boric acid.16 The formation of tne boric acid
crystals was caused by moisture reacting with residual boric oxide in the
RSS material. It was concluded that the moisture had entered the RSS
hopper through the CRDOA vent/purge line by "breathing", and/or by water
contamination in the helium purge line.
In Reference 16, the licensee proposed a threefold corrective action
to the RSS material problems. First, new RSS material, manufactured by
Advanced Refractory Technologies (ART) in late 1984 and early 1985, has
an order of magnitude less residual boric oxide in the B4C material,
and will be installed in all RSS hoppers as part of the overall CRDOA
refurbishment program. No effort will be mace to use ART blended RSS
material currently in stores4' p.32 unless NRC is notified. Second, an
expanded RSS material surveillance program, which will be incorporated
into the Technical Specification, will test one 20 weight % boron hopper
and one 40 weight % boron hopper during each refueling outage, and will
include visual examinations for boric acid crystal formations, chemical
analyses of RSS material for boron carbide and leachable boron oxide con-
tent. Tnird, efforts will be made to mitigate or eliminate the ingress
- 11 -
of moisture into the RSS hoppers by installing a knock-out pot, moisture
elements, and a back-up helium source for the main CRDOA purge and Reserve
Shutdown System purge lines 17 Each knock-out pot will be equipped
with a sight glass and a high level alarm in the Control Room.
Los Alamos concurs that the crystalline structures on the surface of
the B4C RSS balls is meta-boric acid,18 most probably formed by mois-
ture reacting with leachable boric oxide in the B4C material. In light
of the new RSS material to be used, the increased surveillance efforts ,
and measures to mitigate the ingress of moisture in the RSS hoppers, Los
Alamos agrees that the refurbished RSS should be able to reliably perform
its function.
2.2.4 Purge Flow and Seal Replacement
Just prior to the June 23, 1984 event when 6 of 37 control rod pairs
failed to insert on a scram signal , a high moisture content in the primary
coolant resulted in the loss of purge flow into the CRDM cavities. The
loss of purge flow may have allowed the additional ingress of moist pri-
mary coolant into the CRDM cavities, resulting in mechanisms that may
have contributed to the CRDM failures. Because the exact CRDM failure
mechanism has not been determined, and to alleviate the possibility of
purge flow loss and/.or hign moisture content in the primary coolant con-
tributing to future CRDM failures, the licensee has proposed several cor-
rective measures19 as part of the overall CRDOA refurbishment program.
To provide an accurate measure of the purge flow into the CRDM cavi-
ties, tne licensee has proposed that new flow indicators with a range of
0-20 scfm be installed on each helium purge line, providing local indica-
tion, remote indication in the Control Room, ana an alarm in the Control
Room to indicate low flow conditions.20 A minimum of 8 bypass lines
(one line serviced by each of the 8 purge flow subheaders) will be in-
stalled prior to restart. Tne licensee intends to install the flow
instrumentation4' pg 3-4 on these subheaders when the devices are
available.
As mentioned in section 2.2.3, to reeuce tne possibility of moisture
ingress into the CRDM cavities via the helium purge lines, the licensee
will install a knock-out pot, moisture elements and a back-up helium
source for the main CRDOA purge and RSS purge lines, prior to criticality
- 12 -
following the fourth refueling outage4, pg 5. The knock-out pots will
be equipped with a sight glass and a high level alarm in the Control Room.
The helium trailer, which will act as the back-up source of dry helium
for purge, can provide helium at a rate of 7.4 acfm (4.5 lbms/hr per pen-
etration at 700 psig) for approximately 2 hours .17
To mitigate the ingress of primary coolant, which could contain
moisture, into the CRDM cavity, seals will be installed on four large
flow passages into the CRDM cavity--the two passages in the reserve shut-
down tube holes , and the two passages over the eye bolts that penetrate
the floor of the CRAM cavity.21 Cover plates with integral gaskets
will also be installed on the four access openings on the lower CRDM
housing. Thermal and mechanical analyses22 have determined that the
seal additions will not interfere with the RSS performance under the in-
fluence of mechanical , thermal or seismic loadings. Tne flow calculations
in Reference 22 conclude that addition of tne mechanical seals to tne RSS
pressure tubes and the lifting eyebolts will reduce naturally convective
ingress of primary coolant into the CRDM cavity from a flow rate of 0.68
acfm to less than 0.006 acfm. Additional calculations have confirmed
that the seals are able to withstand both a design basis slow depressuri-
zation transient and a design basis rapid depressurization transient.
The licensee has proposed a procedure in Reference 23 tnat basically
requires reactor shutdown in the event CRDM purge flow is lost, or if
high moisture content is present in tne primary coolant.
Los Alamos agrees with tne efforts of tne licensee in monitoring the
flow and moisture content of tne helium purge into the CRDM cavities, in
restricting the ingress of moisture into the CRD cavities via the purge
lines, and in providing a back-up source of helium in case of purge flow
loss. From the review of the provided documentation, Los Alamos agrees
that the addition of seals and coverplates with integral gaskets will
indeed mitigate the ingress of primary coolant and moisture into tne CRD
cavities through penetrations.
In addition, Los Alamos believes tnat the procedure requiring reactor
shutdown with loss of purge flow or high moisture levels in the primary
coolant fulfills the requirements of the assessment report1. The
licensee defines "high moisture levels" in Reference 23.
- 13 -
2.3 CRDM Temperature Recording and CRDM Requalification
The lack of direct measurements of CR0M temperatures curing the June
23 event, and during steady state and other transient operating condi-
tions, has prompted the installation of RTDs to monitor the CRDM cavity
closure plate (ambient) , orifice valve motor plate and control rod erive
motor temperatures. Strip chart recorders will continuously record tne
three temperatures for each CRDM,24 and will provide a CRUM operating
temperature data base. The old data collection surveillance procedure25
will be modified to collect data on a continuous basis.4' p.60 The
licensee intends to install the permanent recorders prior to
restart.4' p'57
The licensee postulates in Reference 26 that "the maximum temperature
rating of the drive mechanism which might inhibit the scram function is
272°F", and in monitoring CRDM_temperatures "the maximum temperature
rating of 272°F should not be exceeded during power operation".
The licensee has also proposed a CRDOA requalification testing pro-
gram that is designee to establisn a temperature at which the CRDOA is
qualified for operation.27 The helium test environment will be operated
at 250°F, 260°F, 270°F, 2b0°F, 290°F and 300°F with a goal of qualifying
all CRDOA components for 300°F operation. Results of tne requalification
testing are anticipated by the end of 1985.
Los Alamos agrees tnat the placement of CRDOA thermocouples, and the
continuous data monitoring at all operating conditions is sufficient to
provide a CRDOA temperature data base curing steady state and transient
operating conditions.
In addition, Los Alamos believes that tne CRDOA is currently only
qualified to operate up to 215°F based on the original mechanical CRDOA
qualification tests, an NRC recommendation,28 and previous Los Alamos
calculations.29 The licensee's argument tnat the CRDOA is qualified
for 272°F operation oased on analytical calculations4' p.49 is not sub-
stantiated. Tnerefore, Los Alamos recommends tnat CRDOA operation be
limited to 215°F until mechanical requalification supports a higher oper-
ating temperature.
- 14 -
2.4 CRDM Surveillance and Preventive/Predictive Maintenance
The licensee has proposed a set of preventive/predictive mainte-
nance tests and surveillance inspection procedures that are intended to
monitor the performance of the CRDOAs and to determine the overall opera-
bility of the CRDOAs during reactor operation. Initial development of
these operating tests are considered part of the CRDOA refurbishment pro-
gram, and will utilize the data base and resultant trends formulated dur-
ing refurbishment.
2.4.1 CRDM Preventive/Predictive Maintenance
Tne licensee's CRDOA preventive/predictive maintenance program is
proposed in Reference 30. According to the licensee, the normal preven-
tive maintenance (PM) program will be implemented on a refueling basis
rotational cycle for CRDOAs that would normally be removed for refueling,
unless the predictive maintenance (PDM) program indicates tne need for
more frequent maintenance. The PM program woulo emphasize the mechanical
examination and refurbishment of the shim motor/brake assemoly, the drive
train, control rod cable, reserve shutdown system, position potentiom-
eters, limit switches, orifice drive motor assembly, orifice drive lead
screw, assorted seals, valves, electrical components , bolts and the ao-
sorber string.
On the other hand, the predictive maintenance techniques would be
used to monitor the most important aspect of CRDOA performance--the
"scram capability"--by determining the shim motor/brake and gear train
performance. The tests proposed in the PDM program include wattage
requirements, back-EMF voltages, delivered torque at the motors, scram
times, rod drop rates and torques to rotate motor/brake assemblies. Cer-
tain aspects of the PDM program would be implemented on a weekly basis to
determine scram capability and temperature performance during power oper-
ation. The licensee has also proposed that testing information be
acquired during reactor shutdown for trending purposes.
Los Alamos concurs with the proposed preventive maintenance pro-
gram as outlined by the licensee, on the assumption that data acquired
during reactor operation will show that predictive maintenance tecnniques
can be used to detect a reduction in CRDOA performance. Tne PDM testinb
tecnniques are closely linked to tne techniques tnat are being used for
- 15 -
the acceptance criteria in the refurbishment program, and will therefore
be dependent on the suitability and acceptability of back-EMF testing for
determining CRDOA operability, as discussed in section 2.2.1.
2.4.2 CRDM Interim Operational Surveillance
Tne licensee's CRDOA interim surveillance program is proposed in
Reference 31. The surveillance tests are scheduled on a weekly basis ,
using a 10" rod drop method on all withdrawn and partially inserted con-
trol rods, except the regulating rod.5' pg 82 The surveillance tests
will obtain data for analysis and long term trending, exercise the rod,
test selected circuitry, verify FSAR (Final Safety Analysis Report)9
assumed scram times, and confirm control rod operability. In addition,
CRDOA temperature and purge flow information will be collected.
For a fully withdrawn rod, analog and digital position
information will be obtained, "Rod-Out" lights will be verified on,
"Rod-In" and "Slack Cable" lights will be verified off, and the rod will
be dropped approximately 10" by de-energizing the brake, while back-EMF
data are obtained for future trending. The "Rod-Out" light indication
will be verified off, and analog and digital information will be compared,
with an acceptable deviation of 10 inches between position indications.
The rod will then be withdrawn to the full out position, so that analog
and digital positions can again be obtained. Control rods that are par-
tially or fully inserted will undergo variations of this method.
Quarterly surveillance tests are intended to supplement weekly sur-
veillance information, and to verify redundancy of selected control rod
position limit switches. Refueling shutdown surveillance will acquire
the same information as the weekly and quarterly tests, except full stroke
insertion tests will be performed.
The operability acceptance criteria, according to the licensee, will
be based on distance and time rod drop data used to calculate a conserva-
tive average full length scram time. A CRDOA will be considered inoper-
able if it does not meet the maximum scram time of 160 seconds as defined
in the FSAR9. Such an indication would warrant back-EMF testing in
confirming scram operability.
Los Alamos agrees that the basic surveillance methodology is suffi-
cient to exercise the control rod, verify FSAR scram times, and to test
- lb -
selected circuitry. However, references to 272°F as the maximum CRDOA
operating temperature are still considered inappropriate as discussed in
section 2.3, and a 10 inch deviation is not considered acceptable between
digital and analog position indications--such a deviation could inadver-
tently lead to control rod overdrive through a misinterpretation of rod
position. Also, the bacx-EMF testing methods and interpretation of the
results are still in the developmental stages , and an engineering deter-
mination of the suitability and acceptability of tnis testing method in
determining continued CRDOA operability will need to be mace before the
licensee can finalize tnis portion of tne surveillance program.
3.0 Moisture Ingress Issues
The licensee has submitted32 a listing of tne issues considered,
and actions taken, by the FSV Improvement Committee (formerly the FSV
Moisture Ingress Committee) in significantly reducing the frequency and
severity of moisture ingress events. The issues were divided into four
categories:
1 . Issues currently under consideration by the Fort St. Vrain Im-
provement Committee.
2. Circulator Auxiliary System modifications yet to be completed
prior to startup.
3. Circulator Auxiliary System modifications to be completed prior
to startup, provided material availability and schedule permits .
4. Items identified by tne Moisture Ingress Committee wnicn are
installed and operational .
Los Alamos believes that a listing of intended and installed modifi-
cations does not provide any indication as to what any given modification
really is, wny they contribute to tne reduction in potential for moisture
ingress events , nor which improvements will substantially reduce the
severity and frequency of moisture ingress events. Tne licensee nas com-
mitted to submit a more explanatory version of the actions to mitigate
moisture ingress, prior to restart4, pg 8d.
4.0 PCRV Post-Tensioning Tendon System
In the spring of 1984, during scheduled PCRV tendon surveillance,
tendons with corroded and broken wires were found. Since that time, the
- 17 -
licensee has evaluated the corrosion mechanism, has performed lift-off
tests on selected tendons to determine their load-carrying capability,
and proposed corrective actions and an increased surveillance procedures.
4.1 Tendon Accessibility, Extent of Known Degradation and Failure
Mechanism
The licensee, in determining the extent of tendon corrosion in the
PCRV, determined what fraction of the tendons were available for visual
examination and lift-off tests. The tendon system is subdivided into
four major groups: the 90 longitudinal (vertical) tendons have 169 wires
per tendon; the 210 circumferential tendons in the PCRV sidewall have 152
wires per tendon, and the 50 circumferential tendons in both the top and
bottom heads have 169 wires per tendon; tne 24 bottom cross-head tendons,
and 24 top cross-head tendons nave 169 wires per tendon. Of the four
groups, the licensee states the following accessibility3d:
Tendon Group Both Ends Acces. One End Acces. Neither
End Acces.
Longitudinal
•
Visual 20 69 1
Lift-off 0 74 16
Circumferential
Visual 261 27 2
Lift-off 236 62 12
Bottom cross-head
Visual 20 4 0
Lift-off lb 4 4
Top cross-nead
Visual 17 7 0
Lift-off 16 6 2
The number of tendons witn known broken wires as identified in the
licensee's 1904 surveillance,34 include° 10 longitudinal tendons with 1
to 22 broken wires, 2 circumferential tendons witn 2 and 15 broken wires,
8 bottom cross-head tendons with 1 to 19 broken wires, and no top crDss-
head tendons with broken wires. In some cases , tne total number of cor-
roded, broken wires include wires broken during lift-off tests , or during
retensioning.
- 18 -
The results of 74 longitudinal lift-off tests35 indicated that
tendons with identified broken wires generally had a slightly smaller
lift-off value than intact tendons. Thirty lift-off tests on circumfer-
ential tendons snowed little change in lift-off value. Some of the fif-
teen bottom cross-head tendon lift-off tests showed a definite reduction
in lift-off value for tendons with multiple wire breaks. The value of
the lift-off test on one top cross-head tendon was nominal. All lift-off
test values exceeded tne minimum limits.
The licensee conducted metallurgical investigations into the cause
of the corrosion, and determined that microbiological attack on the tendon
NO-OX-ID CM organic grease caused the formation of formic and acetic
acids.34'36 Tne acids, in conjunction with moisture in the tendon tube,
vaporized and recondensed on the cooler portions of the tendons--in this
case, usually toward the tendon ends. Tne acidic attack resulted in re-
duced cross-sectional wire area, stress corrosion cracking, localized
tensile overload and wire breakage.
Los Alamos believes, based on the documentation presented by the
licensee, that microbiological attack of the tendon grease and the resul-
tant formation of acetic and formic acids , in the presence of moisture,
is a probable cause for the currently observed tendon corrosion, and has
led to the subsequent wire breakage through tensile overload. However,
Los Alamos believes that the extent of known tendon corrosion, breakage
and previous surveillance have not been clearly defined by the licensee.
Los Alamos therefore recommends that a complete map oe mane tnat lists
each tendon, its visual examinations and lift-off values, and the number
and location of corroded and broken wires. An indication of tne degree
of wire corrosion would also be desiraole.
4 .2 Tendon Corrosion Corrective Measures
The licensee evaluated several methods for arresting the corrosion
process,34,36 including the use of ozone as a biocide to kill the micro-
organisms, the use of an alkaline grease wnich should not be conducive to
microbiological growth, and tne use of an inert blanket consisting of
nitrogen gas. The licensee's consultants found that the nitrogen atmo-
sphere arrested the growth of the microbes in the NO-OX-ID CM organic
- 19 -
grease34,35, and eliminated the oxygen which is necessary for the cor-
rosion process to continue. eased on these results, and as a snort term
action, the licensee has proposed that nitrogen blankets be establisheu
on the longitudinal and bottom cross-head tendons. Long term actions
would include further investigations into the corrosion process and ar-
resting techniques, and the possible installation of additional load cells
in monitoring the PCRV behavior.
Los Alamos believes that the use of a nitrogen blanket to halt tne
corrosion process may be suitable, but difficult to implement as proposed.
The tendon tubes are not likely to be leaktight, and maintaining an inert
gas atmosphere at a set over-pressure may prove difficult. Consideration
might be given to maintaining an intermittent or continuous purge flow
through the tendon tubes, as needed, rather than to maintaining a speci-
fied overpressure. However, Los Alamos recommends that initially the
nitrogen be purged through the individual tendon tubes to remove as much
moisture as possible, and that gas samples be used to monitor moisture
and oxygen reduction. Further investigation into the long term effects
of a nitrogen blanket on tendons, tne corrosion process and currently
available corrosion acids are also recommended.
4 .3 PCRV Tencon Interim Surveillance
Because the total extent of tendon corrosion in tne PCRV is unknown,
because the rate of existing corrosion is unknown, and because tne use of
a nitrogen blanket as an arrest to the corrosion process is an unknown,
the licensee has proposed an interim surveillance program designed to
address eacn of these issues. ' pp•16k-7' The interim tendon surveil-
lance program would include increased visual and lift-off surveillance
for tnree years, or until effective corrosion control has been estab-
lished. Two populations of tendons would be inspected--a population of
tendons that have not been previously identified as being corroded, and a
control population with known corrosion. On a six-month frequency, visual
surveillance of both tendon ends, when accessible, would include:
- 20 -
Tendon Group No. of New Tendons No. of Control Tendons
Longitudinal 24 6
Circumferential 13 3
Bottom cross-head 6 2
Top cross-head 1 1
Lift-off tests would be performed on two frequencies--an to month
frequency for the population of new tendons, and a b month frequency for
the control population. The number of tendons for lift-off will include:
Tendon Group No. of New Tendons No. of Control Tendons
Longitudinal 12 3
Circumferential 13 3
Bottom cross-head • 3 1
Top cross-head 1 1
As an acceptance criteria, the licensee proposed that, based on vis-
ual examinations, a mandatory engineering evaluation be conducted on any
tendon that has 20% of its wires broken. For any tendon that has only
one accessible end, the mandatory engineering evaluation will be con-
ducted when any tendon has lob; of its wires broken. The control tendon
population will include those tendons with the worst known corrosion with
ready accessibility.
Los Alamos agrees that the increased tendon surveillance program of
the nature proposed by the licensee will provide more information on the
extent of corrosion in the PCRV by inspecting new tendons each surveil-
lance, and at the same time, monitor the rate of corrosion with the con-
trol tendon population. Tne increased surveillance should also determine
the effectiveness of the nitrogen blanket in arresting corrosion, or any
other corrective measure the licensee may propose. Los Alamos recommends
that the licensee submit an outline of the intended mandatory engineering
evaluation, which should include all lift-off, load cell and relaxation
data incorporated into a safety evaluation. The licensee should define
the extent of the visual and lift-off testing procedures, and could use
US/NRC Regulatory Guide 1.3537 for guidance.
- 21 -
4.4 PCRV Structural Calculations by Los Alamos National Laboratory
The PCRV tendons are intended to apply sufficient compression in the
concrete to balance or exceed the circumferential and vertical tension in
the concrete that results from the internal pressure. A combined analyt-
ical and numerical study39 was undertaken by Los Alamos National
Laboratory to evaluate the evolution of these stresses, both to the ini-
tial prestressing and to subsequent partial and total rupture of tnese
tendons. At the stress levels anticipated in the concrete, and for the
anticipated operating life span of tne PCRV, the concrete benavior was
modeled as a linear viscoelastic solid with the creep strain varying pro-
portionally with the logarithm of time at constant stress tnrougnout the
projected reactor lifetime.
A one-dimensional model of a long concrete column of rectangular
cross-section, with an embedded prestressing tendon along the length, was
used to evaluate the concrete and steel stresses as well as tne hold-down
and lift-off forces. Tnese were evaluated for the intact tendons and the
degraded tendons. The degree of tendon degradation is described through
the ratio of the number of unbroken strands to the original number of
strands . Initial time of rupture was varied from the time of initial
prestressing to 400 days after emplacement. Tne formulation led to an
integral equation, which was solved numerically. The hold-down forces
decayed approximately with the logarithm of time and for both the extreme
observed degradation (21 broken strands) and for a more extreme case (40
broken strands) , the hold-down force still exceeded the minimum safety
design requirements.
In addition, several finite element calculations, using the finite
element code N0NSAP-C, were made to evaluate complete tendon failure in a
60° sector of the Fort St. Vrain PCRV. This code has an extensive
material library of constitutive relations to model the various properties
of concrete, together with a specialized element model to simulate pre-
stressing tendons. Two rows of vertical and an arc row of circumferential
tendons were incorporated in the model as a baseline calculation. Tne
tendons were prestressed to 70% of the ultimate and an internal pressure
of 775 psi was applied (tnis pressure is the internal pressure of tne
helium coolant in the HTGR) and the creep of the concrete and slow decay
of tne tendon stresses were evaluated out to 30,000 days. Then, three
- 22 -
cases wherein one tendon was removed at one day were evaluated. First
the middle vertical tendon in the outer row and in line with the outer
buttress was removed. Second, an inner vertical tendon opposite the
thinnest portion of the PCRV wall was removed. Finally, an inner layer
circumferential tendon at midheight was removed. Stress redistributions
at 300 days after ruptures were calculated and shifts of tne remaining
tendon loads to accommodate the broken tendon were calculated. Regions
of local tensile and snear stress in the concrete portion of tne PCRV
were identified and related to overall structural integrity.
With all tendons present, the mean vertical stress was about -7b0
psi, the radial stress decreased from the applied internal pressure of
-705 to about -1200 psi at the ring of circumferential tendons and tne
tangential stress ranged from -2400 psi at tne inner wall to about -2200
psi at the same place. Removal of a vertical tendon reduced the mean
axial stress by about +40 psi, the local tangential stress by -10 psi and
did not materially affect the radial stress. Removal of a circumferential
tendon reduced the mean tangential stress by +30 psi and the local axial
stress by -80 psi. The vertical hold-down force from zero days tnrough
30,000 days decreased linearly and remained above the prescribed safety
limit, as did the circumferential hold-down force.
Comparison of the analytical solution and a small finite element
proolem simulating the analytical problem was made to verify the visco-
elastic creep models and the tendon element in the NUNSAP-C code. Excel-
lent agreement for stresses, strains and nold-down forces was ootained.
5.0 Conclusions
Los Alamos concludes that the licensee, Public Service Co. of
Colorado, has made a conscientious effort to address all of the restart
issues listed in the assessment report.1 The refurbishment program on
all CRLOAs provides confidence in CRDOA operability during reactor opera-
tion and the ability to scram, even if the exact "failure to scram" mech-
anism has not been defined. Questions concerning the reliability of the
back-EMF testing procedure on the shim motor/brake assembly in determining
control rod operational acceptability still exist, but further method
development, more experience with result interpretation, and in-core
- 23 -
testing may alleviate the questions. Until CRDOA operability can defin-
itely be ascertained with these methods, we recommend that the licensee
have backup measures such as rod run-in following scram.
Control rod cable and connecting hardware material replacement, along
with replacement of the Reserve Shutdown System material, serve to rectify
the material problems brought on by corrosive mechanisms.
In light of chloride stress corrosion problems, Los Alamos also
recommends that all reactor components exposed to the primary coolant be
reviewed for susceptibility to chloride attack, especially the PCRV liner.
Review should continue into the source of cnlorine and methods to elimi-
nate its generation and presence.
The effects of purge flow loss have not been determined to be in-
strumental in CRDOA failures to scram, yet the licensee has committed to
maintaining purge flow by external means, and to reducing the effects of
primary coolant naturally convecting into the CRDOA cavity with extra
seal installation.
Even though current qualified CRDOA operating temperatures are very
much in question, the licensee is in the process of requalifing the mech-
anism for temperatures more in line with those anticipated during reactor
operation.
From a mechanical standpoint, CHDOA preventive/predictive mainte-
nance procedures are certainly reasonable, but like the proposed surveil-
lance program, they are dependent on back-EMF testing methods wnicn are
still in the developmental stages.
Evaluation of moisture ingress corrective measures was difficult due
to the lack of information with which to understand the measures taken.
The licensee has committed to submit a more explanatory version of the
actions to mitigate moisture ingress prior to restart.
The extent of PCRV tendon degradation is not well known, even if the
licensee may nave determined the cause of the corrosion. Further investi-
gation into arresting measures is definitely required, especially because
the nitrogen blanket technique may be so difficult to employ. However,
the interim surveillance program should provide information on the degree
and rate of corrosion, in addition to establishing a tendon wire loss
acceptance criteria. The tendon acceptance criteria should ensure PCRV
margins to safety.
- 24 -
6.0 References
1 . "Preliminary Report Related to the Restart and Continued Operation
of Fort St. Vrain Nuclear Generating Station, " Docket No. 50-267,
Public Service Co. of Colorado, October, 19d4.
2. "Review of Dallas Meeting (1/15/85) and Restart Committments" ,
letter from Martin, NRC/Reg IV, to Lee, PSC, 1/17/85.
3. "Fort St. Vrain Meeting, NRC-PSC, February 20, 1985," Volumes I, II
and III, recorded and transcribed by Koenig & Patterson, Inc.
4. "Fort St. Vrain Meeting, NRC-PSC, February 21, 1985," Volumes I and
II, recorded and transcribed by Koenig & Patterson, Inc.
5. "Fort St. Vrain Meeting, NRC-PSC, February 22, 1985," Volumes I and
II, recorded and transcribed by Koenig & Patterson, Inc.
6. "Engineering Report on CRDOA Failures to Scram-Control Rod Drive and
Orifice Assemblies, " PSC suomittal P-85037, 1/31/85.
7. "Failure of Three CRDOAs to SCRAM, " PSC submittal P-85029, 1/28/85 .
8. "bearing Debris Analysis, " PSC submittal P-85017, 1/18/85.
9. "Fort St. Vrain Nuclear Generating Station, Updated Final Safety
Analysis Report, " Public Service Co. of Colorado.
10. "CRDOA Refurbishment Program Report, " PSC submittal P-85040-2,
1/31/85.
11. "Control Rod Drive Cable Replacement, " PSC submittal P-85032-2,
1/20/85.
12. "Control Rod Drive Cable Replacement Report, " GA Technologies
Document 907822, Attachment 1 to PSC submittal P-85032-2, 1/31/85 .
13. "Investigations into Sources of Chloride in FSV Primary Circuit,"
PSC submittal P-86036, 1/31/85.
14. "Safety Analysis Report--Change in Material of the FSC Control Rod
and Orifice Assemblies, " Attachment 2 to PSC submittal P-85032-2,
1/31/85.
15. "FSV Control Rod Cable Metallurgical Examinations," draft report
from Los Alamos National Laboratory, 3/85.
16. "Report on Reserve Shutdown Absorber Material ," PSC submittal
P-85027, 1/28/85.
17. "Moisture Control in CRDOA Purge Lines, " PSC submittal P-85032-9,
1/20/85.
- 25 -
18. "FSV Reserve Snutaown System Material Metallurgical Examinations, "
draft report from Los Alamos National Laboratory, 3/85.
19. "CRDOA Moisture/Purge Flow, " PSC submittal P-85032-6, 1/20/85.
20. "Modifications to CRDOA Helium Purge Supply, " PSC submittal
P-85032-8, 1/20/65.
21. "Control Rod Drive Cavity Seals, " PSC submittal P-85032-7, 1/20/85.
22. "FSV CRD Cavity Seals Design Report, " GA Tecnnologies Document
907604, Attachment 1 to PSC submittal P-65032-7, 1/20/85.
23. "Operations Order No. 84-17 Describing Operator Actions Upon a Loss
of Purge Flow and or Detection of Hign Moisture Levels in Primary
Coolant, " PSC submittal P-85040-8, 1/31/85.
24. "CED Temperature and Helium Purge Flow Recorders, " PSC submittal
P-85032-3, 1/20/65.
25. "Current CRD Temperature Data Collection Procedure Which Requires
Station Manager Notification Upon Discovery of a Measured CRD
Temperature in Excess of 250°F, " PSC submittal P-85040-9, 1/31/85.
2b. "Control Rod System Operability Evaluation Report, " PSC submittal
P-85040-1 , 1/31/85.
27. "CRDOA Mechanism Temperatures Environmental Requalification, " P.C
submittal P-85032-1, 1/20/85.
26. Letter from Robert A. Clark, Chief, ORd3, to 0. R. Lee, PSCo. ,
December 2, 1982.
29. Meier, K. , "Fort St. Vrain Reactor Control Roo Drive Mechanism Over-
Temperature Problem, " Los Alamos National Laboratory, 1962.
30. "CRDOA Proposed Preventive/Predictive Maintenance Program Report, "
PSC submittal P-85040-3, 1/31/05.
31. "CRDOA Interim Surveillance Program Report, " PSC submittal P-85040-5,
1/31/85.
32. "FSV Improvement Committee Actions," PSC submittal P-85022, 1/24/85.
33. "Tendon Accessibility Report," PSCo. letter from Warembourg, PSC, to
Jonnson, NEC/Reg IV, PSC submittal P-84523, 12/14/64.
34. "Lab Report No. 52--Examination of Failed Wires from Fort St. 'drain
Unit No. 1," PSC submittal P-o4543-4, 1/24/65.
35. "Liftoff Tests, " Attacnment 1 to "Engineering Report on Fort St .
Vrain Tenaons, " PSC submittal P-84543, 12/31/84.
- 26 -
36. Thurgood, Roberts and Epstein, "Evaluation of the Causes of Corrosion
in the Fort St. Vrain Post-Tensioning Tendon Wires, " GA Tecnnologies ,
PSC submittal P-84543-5, 1/24.85.
37. US/NRC Regulatory Guide, Rev. 2, January 1976 .
38. Clauss, F. J. , "Solid Lubricants ano Self Lubricating Solids",
Academic Press, 1972.
39. Fugelso, E. and Anderson, C. , "Evaluation of Concrete Crrep and
Stress Redistribution in the Fort st. Vrain PCRV Following Rupture
of Prestressing Tendons", Los Alamos National Laboratory, October
31 , 1984.
- 27 -
SSINS NO. : 6835
IN 85-49
UNITED STATES WELD C0""iv rc"1MlSS!RNERS
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT 5, n
WASHINGTON, D.C. 20555 DJ5. SC��?S�? 'i� JUL 1 01985
July 1, 1985 i
IE INFORMATION NOTICE NO: 85-49: RELAY CALIBRATION PROBLEM
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice provides information on a potentially significant
problem pertaining to relay orientation during calibration and operation. It
is expected that recipients will review the information for applicability to
their facilities and consider actions, if appropriate, to preclude a similar
problem occurring at their facilities. However, suggestions contained in this
information notice do not constitute NRC requirements; therefore, no specific
action or written response is required.
Discussion:
South Carolina Electric and Gas Company, at the Summer nuclear power plant,
recently discovered that there was a significant error in the calibration of
several E-7000 series Agastat time-delay relays. The licensee' s procedure for
replacement of Agastat timing relays did not require calibration of these
relays in the installed position. During relay replacement, these time-delay
relays were "bench" calibrated in a horizontal position, then field mounted in
a vertical orientation. Subsequent time-delay measurements determined that the
delay times of the installed devices were as much as 30% greater than that
established during "bench" calibration.
Licensee investigation of this anomaly determined that the manufacturer' s data
sheet for installation and operation of these relays identifies a potential for
the observed calibration error. This data sheet states that a dial calibration
error may result if the device is mounted horizontally without the manufacturer' s
supplied horizontal operation options. Because these relays were mounted
vertically at this facility, the potential for error as a result of horizontal
"bench" calibration was overlooked. Because of this oversight, the licensee' s
relay replacement procedure did not specify device orientation during calibration.
Several of these relays were functionally arrayed (sequentially or in parallel)
in safety-related applications, such as in the emergency diesel generator under-
voltage start circuitry. The above identified time-delay error went undetected
for a period of approximately 4 months for some of these relays. This occurred
because the licensee' s replacement procedure specified only an operational or
functional postmaintenance test of the system containing the relays. This method
of testing did not detect the individual relay time-delay calibration errors.
8506260654
1,a ,r-+4 rhs18
IN 85-49
July 1, 1985
Page 2 of 2
The licensee has subsequently revised their relay calibration procedures to
calibrate or check the calibration of relays after mounting in place, where
practical . The procedures were also revised to ensure "bench" calibrations
are performed in the same orientation as mounted, where applicable. Personnel
involved in relay calibration have received training on the revised procedures.
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional
Administrator of the appropriate NRC Regional Office or this office.
4gEdwar Je Director
Divisio of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: H. Bailey, IE
(301) 492-9006
Attachment: List of Recently Issued IE Information Notices
Attachment 1
IN 85-49
July 1, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-48 Respirator Users Notice: 6/19/85 All power reactor
Defective Self-Contained facilities holding
Breathing Apparatus Air an OL or CP, research,
Cylinders and test reactor,
fuel cycle and
Priority 1 material
licensees
85-47 Potential Effect Of Line- 6/18/85 All power reactor
Induced Vibration On Certain facilities holding
Target Rock Solenoid-Operated an OL or CP
Valves
85-46 Clarification Of Several 6/10/85 All power reactor
Aspects Of Removable Radio- facilities holding
active Surface Contamination an OL
Limits For Transport Packages
85-45 Potential Seismic Interaction 6/6/85 All power reactor
Involving The Movable In-Core facilities holding
Flux Mapping System Used In an OL or CP
Westinghouse Designed Plants
85-44 Emergency Communication 5/30/85 All power reactor
System Monthly Test facilities holding
an OL
85-43 Radiography Events At Power 5/30/85 All power reactor
Reactors facilities holding
an OL or CP
85-42 Loose Phosphor In Panasonic 5/29/85 All power reactor
800 Series Badge Thermo- facilities holding
luminescent Dosimeter (TLD) an OL or CP
Elements
85-41 Scheduling Of Pre-Licensing 5/24/85 All power reactor
Emergency Preparedness facilities holding
Exercises a CP
OL = Operating License
CP = Construction Permit
Hello