ML20235J987
| ML20235J987 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Hope Creek |
| Issue date: | 03/01/1987 |
| From: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Heltemes C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| Shared Package | |
| ML20235J739 | List: |
| References | |
| FOIA-87-377, FOIA-87-A-63 NUDOCS 8710020190 | |
| Download: ML20235J987 (7) | |
Text
UNITED STATES bb*OIN h
ka tefug%
)8 NUCLEAR REGULATORY COMMISSION f 58, 6
S REGloN I
~
///// Q/
8; E
f
$31 PA.RK AVENUE
((@
g l g MING OF PRUSSI A. PENNSYLVANIA A 1940C 01 M AR 1987 MEMORANDUM.FOR:
C. J. Heltemes, Jr., Director Office for Analysis and Evaluation of Operational Data FROM:
T. E. Murley, Regional Administrator, RI
SUBJECT:
INPUT TO ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR FOURTH QUARTER CY 1986 As requested in your January 30, 1987 memorandum the Region I staff has reviewed the proposed A0's as well as the appendix items. All items appear properly classified as suggested by AE00. Attached are inputs for the Region I items, as follows:
Appendix C Items Anomalies During Loss of Offsite Power Testing at Hope Creek (PNO-I-86-78).
l A copy of the attached Hope Creek input has been faxed to the NRR program l
manager, it is requested that AE00 solicit NRR input, especially regarding the generic implications of Bailey module failures.
Conviction of Licensee (International Nutronics, Inc., Dover, New Jersey) and One Employee in Federal District Court (PNO-I-86-89A) Item Accidental Incineration of Radioactive Material at Henry Haywood Hospital (PNO-I-86-86)
Thomas E. Murley Regional Administrator
Attachment:
As Stated l
l I
\\
l GyV{('i nt N..
' ~..
871' 020190 870930 hDO A.63 PDR
l
HOPE CREEK AUGMENTED INSPECTION TEAM
)
1 The Hope Creek Nuclear Power Plant operated by Public Service Electric and Gas-Company of New Jersey, utilizes a General Electric designed boiling water reactor.
The plant is located near Salem, New Jersey.
On September 11, 1986, Public Service Electric and Gas Company performed a loss of offsite power test at the Hope Creek Plant from approximately 21.5% power.
The loss of offsite power test is an important part of the power. ascension test
)
program.
Its purpose is to demonstrate whether the plant response is satisfactory l
and in accordance with the plant design for concurrent loss of the turbine gene-
'i rator and all offsite power sources.
The Hope Creek Generating Station loss of offsite power test was initiated with the turbine generator loaded to 165 MWe.
The first indication of-an unsatisfactory i
plant response was the failure of the "C" emergency diesel generator output breaker to close automatically.
Soon after, an observed failure of the reactor auxiliary cooling system coincident with increasing drywell pressure resulted in the test being aborted by tha licensee. Normal offsite power was then manually. restored to the station. Twenty-four observations were made by Public Service Electric and Gas during this test. These observations occurred during the time from initiation of the test until the reactor vessel water level and pressure were controlled and the reactor scram was reset.
The most significant observations on September 11, 1986 were:
(1) emergency l
diesel generator "C" output breaker failed to close; (2) MSRV position indication was lost; (3) power supplies for the source and intermediate range neutron detector drives and main steam line acoustic monitors were lost; (4) 17 control rods did not provide a normal full-in position indication; (5) reactor auxiliary cooling system flow was lost; (6) emergency diesel generators "A" and "B" governors transferred isochronous (frequency control) to speed droop (load control) mode without operator action; and (7) the "B" safety auxiliary cooling system pump failed to auto-start.
l On September 19, 1986, Public Service Electric and Gas performed a cold loss of I
offsite power test (TE-SU.ZZ-313(Q)) at Hope Creek Generating Station.
The pur-pose of this test was to demonstrate that the plant response was in accordance with plant design for loss of all offsite power sources after the licensee had l
assured that the previous test observations had been investigated and resolved.
This loss of offsite power test was initiated with the reactor at cold (T <200 F) shutdown temperature and pressure conditions with the reactor mode switch in shut-down.
The significant observations during this test were:
(1) the "B
safety auxiliary cooling system loop head tank level indicator failed; (2) one control room emer-gency ventilation (air recirculation) system fan failed to start; and (3) one dry-well fan also failed to start. Hope Creek station personnel observing the test identified a total of 17 observations.
I
e 2
N As a result of'the unsatisfactory test results an Augmented Inspection team was formed and sent to.the site to (1) independently assess.the root cause of each observation; (2) review the effectiveness of the corrective action planned or taken; and (3) assess the overall. implications of the test results.
A.second cold loss of power test was conducted on October 2, 1986.' The Augmented Inspection Team witnessed this test and assessed the~results.
One test observa-tion was a repeat of a' previous observation and involved a Bailey logic module.
The inspection began on September 25 and ended October 3, 1986.
Of the 41 observations report'ed from the two loss of offsite ' power tests,-the overall safety significance was concluded to be relatively minor except for the Bailey solid state logic module failures.
These niodules, manufactured by the Bailey Meter Co., are multipurpose electronic devices used extensively throughout:
the plant for, control and safety functions. Of eight hardware' failures identified-during this review, six were attributable to various malfunctions with Bailey, logic modules.
Three weaknesses with the Bailey logic modules were found:
(1) the dependency on common equipment for accomplishment of automatic and ' manual safety actions. for the actuated safety system equipment; (2) limited test provisions to assure the online operability of the Bailey logic modules after their installation into the equipment cabinets; and (3) the usefulness of the bench test equipment in assuring that the Bailey logic modules are operable.
The team was also concerned that the failure rate of the Bailey logic modules appeared high. These weaknesses are especially significant since all of the balance ~of plant safety-related. systems (and a part of one NSSS system) use Bailey modules to develop the' safety system logic and actuation functions.
A number of minor plant design, construction, and manufacturing problems were also identified. Several specific weaknesses in the scope.of various system preopera-tional tests were revealed since the loss of.offsite power tests were the first integrated demonstration of the plant response to this event.
Several subtle interactions involving the dependency of various systems on cooling'and instrument air supporting systems were revealed.
A number of 6servations resulted because instruments or other equipment-lost power during the test. A number of these instances involved the apparent failure to meet FSAR commitments to provide reliable power to specific instruments or equipment.
In summary, the team determined that the results of the loss of offsite power tests.
indicated certain weaknesses in the design, construction, and testing programs for Hope Creek. With the exception of the Bailey modules the team found the weak-nesses to be minor in nature.
For the Bailey modules, however, the team identified ~
concerns with the adequacy of bench and su'rveillance testing and a failure rate which was higher than expected.
L
i
~
3-I N
1 s
Following the team inspection, the. licensee presented a program to resolve concerns
]
related to the Bailey modules. The program will include:
An in-house data assessment program which will include a review of each in-plant module failure and a determination by the manufacturer of the individual i
component which failed and, to the degree possible, the cause of the failure.
1 An assessment, by the manufacturer, of module failures at installations of other users.
An accelerated aging and cycling test program, with a final reliability analysis report by the end of the second quarter of 1987.
i l
A monthly trending program that will provide a report bi-monthly indicating:
the-number of module-failures having an adverse affect on system function; resulting time in an LCO' and the number of failures determined by surveil-lance. This program will apply to both IE and non-IE systems.
A report of Bailey's recommendations _to improve module reliability based upon their observations at Hope Creek of site environment, handling, and testing techniques.
l The modification of existing module test equipment and procedures to permit module testing without staple jumper removal.
The development and procurement of a test rig capable of bench testing modules for all utilized functions prior to November 1987.
Testing would be conducted without removing staple jumpers or the FPLA.
The determination of the feasibility and implications of modifying the exist-ing Bailey system to permit in-situ testing.
This program is underway and the results are under continued NRC review.
I 1
i
\\
w i
l 1
N Incineration of Molybdenum-99/ technetium-99m Generator at Hospital On October 21, 1986, Henry Heywood Memorial Hospital reported that a moly-bdenum-99/ technetium-99m generator containing 880 mil 11 curies of molybdenum-99 (as of noon October 19,1986) had been inadvertently incinerated in the hos-pital's incinerator on the evening of October 19, 1986.
Initial surveys of the
)
incinerator performed by the licensee revealed only background radiation levels.
Therefore, the licensee assumed that the molybdenum had vaporized and was released through the stack.
The licensee's surveys of the grounds sur-rounding the hospital likewise revealed only background radiation levels.
The t
incinerator was cleaned out and the debris held just in case it was contam-l inated. The licensee subsequently used the incinerator two more times.
Radiation surveys conducted by NRC inspectors upon arrival at the site revealed radiation levels exceeding 200 mR/hr in the incinerator and the con-j i
tainer holding the debris from the incinerator.
Immediate corrective action taken by the licensee included roping off the area surrounding the incinerator, removing the container containing the debris to a restricted access area, and shielding it with lead sheets.
Use of the incinerator was suspended until radiation levels decrease to background. About 1492 millicuries of techne-i tium-99m was estimated to have been released.
The resulting concentration of i
technetium-99m released, 0.22 E-7 uCi/ml when averaged over a one year period, was less than the limits specified in 10 CFR 20, Appendix B, Table II, Column I j
for technetium-99m (5 E-7 uC1/ml). When average over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period the i
resulting concentration was about 16 times the value of 10 CFR 20, Appendix B, I
Table II.
A number of factors contributed to the ir.cident, and include inadequate train-ing of the nuclear medicine technician who performed the initial surveys and the personnel who were expected to handle and control radioactive materials.
Also, inadequate management involvement in the program contributed to the I
licensee's ineffectiveness in correctly evaluating the affect of the event.
No personnel exposures were attributable to this event, and no substantial hazard resulted to personnel in unrestricted areas.
l l
t
~
~
r 1.
N Conviction of International Nutronics, Inc., and one Employee in Federal District Court International Nutronics, Inc., (INI) a California corporation, and Eugene O'Sullivan, a Corporate Vice President and Corporate Radiation Safety Officer of INI, were convicted on October 29, 1986, in Federal District Court in Newark, New Jersey. They had been charged with two counts of willful violation of the incident notification requirements of the Atomic Energy Act, one count of willfully furnishing false information to a government agency, one count of j
conspiracy to conceal the incident from the NRC, and five counts of mail and wire fraud. Bruce Thomas, the Plant Manager and Radiation Safety Officer of INI's Dover, New Jersey, facility was acquitted on all nine counts.
INI was fined $35,000, the maximum fine.
Mr. O'Sullivan was given a suspended sentence and two years probation.
Both convictions have been appealed.
The charges resulted from a December 1982 incident involving a spill of radio-actively contaminated water at the INI irradiation facility in Dover, New l
Jersey. The spill resulted in widespread contamination of the facility, in-l ciuding the ground immediately under and adjacent to it.
Decontamination was begun in 1983 and completed in early 1986. The facility has been released for unrestricted use, and the INI license has been terminated at their request.
l l
l l
L
1 APR 02 '87 15:05 USNRC 8 AVE DOC P02
,, j-
[c.'e l L e
N
]
ABNORMAL occtfRRENCE WPBATE On July 15, 1986, Region I, with the consent of the licensee, isssed an amendment to the Mercy Hospital license which did not include the individual who directed the f ailure to make required reports to the NRC. While this individual will continue to work at the hospital, his activities will be under the supervision of an authorized user and he will not sanage the Nuclear medicine program and is no longer Radiation Safety Officer.
Region I has agreed to consider again adding this individual as a authcrized user, if the hospital formally requests it, af ter a period of one year.
On December 15, i
l 1986 the license of Valley Radiology Associates was amended to delete this l
same individual, agata with the consent of the licensee.
The technologist i
involved in this matter no longer herks in the field of Nuctaar Medicina.
1 1
i i
l i
,i
,i i
6 i
k i
t