ML20236A761
| ML20236A761 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/23/1987 |
| From: | Russell W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Jordan E NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| Shared Package | |
| ML20236A720 | List: |
| References | |
| FOIA-87-377, RTR-NUREG-0090, RTR-NUREG-90 NUDOCS 8710230093 | |
| Download: ML20236A761 (2) | |
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UNITED $TATES l'
' ' <g NUCLE AR REGULATORY COMMISSION
-l REGION I D
$31 PARK AVENUE
,8 KING OF PRus5IA, PENNSYLVANIA 19406 t...
MEMORANDUM FOR:
Edaard L. Jordan, Directo-Office for Analysis & Evaluation.or Crerational Data i
FROM:
William T.. Russell, Regional Administrator, RI
SUBJECT:
' CONCURRENCE.ON ABNORMAL OCCURRENCE REPORT i
TO CONGRESS FOR FOURTH QUARTER CY 1986 Region I staf f' members have reviewed the final A0 draft report enclosed in your l
memorandum dated May 12, 1987.
The information related to Region I'. licensees is complete and accurate with one exception (see attached correction). Once
.this correction is'made Region I concurs with publication of.he report.
Provided below is the information regarding A0 coordinators that you requested:
RI'A0 Coordinator:
K. G. Murphy Mail Stop - Region I FTS #488-1210 RI Backup A0 Coor:'nator:
L. H. Bettenhausen Mail Stop - Region :
FTS #488-1291 h
&)k William T. Russe'.1 Regional ad-inistrator Attachaent:
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L Libe: C41 vert Clif f diesel's are provided with a. low JWCS pressure trip. that-
.' recludes' overheating ~and engine destructive. failure, Correct Appendix C,-
.page 41,::last part;of' third paragraph as'follows:
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" Excess gas in the.JWCS can result in cavitation of the Jacket-water
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pump, low JWCS pressure,-an,d: subsequent diesel engine trip.
The problem was initially observed at a very low' level of leakage'in 1984."'
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MEMORAN0tm FOR:.C. J. Heltenesly:Jr.. Director.Offits for Ana sis and.tvaluation of PRON.
J. B. Martin:
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-Regional Administrator.
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SU5 JECT:
.ARNORMAL OCCURRENCE REPORT.70 CONGRESS FOR FOURTR QUARTER CY 19864 Your memorandum'of January 130,1987, provided a 1.isting of events proposed.for j
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the subject report.
j Region'V wesirequested to proviile a write-up regarding.the potential hand l
exposure of an individual at San Onofre Unit 3, if the. exposure.is verified to i
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saceed:the A0 threshold of 375 rees.
4 The licensee's investigation into this event has, to date, raised uncertainty' '
regarding the circumstances and extent of actual exposure to the individual, We anticipate that the licensee's investigation and Regiott V!s follow-up l
inspection activities regarding this event will extend be 1
A0 resort to Congress.- We,posure threshold be verified) yo reporting (shouldtheA0ex in the Fourth Quartar therefore, propose deleting this event from the l
Fourtl Quarter 1986 A0 report.
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Should the actual exposure be verified to exceed :the A0 threshold 6 Region V l
will provide.an appropriate writeup for reporting at that. time.
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AT Jess Crews. of my staff, has discussed the above circumstances, by telephone.-
4, with Paul Bobe, of your4 staff.
s With the deletion discussed above, we c -
'; the; proposed content,of the t
..a subject report.
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=aw g ngAA mWM FOR': Clemens J. Heltamos.- Jr. Evaluation Director Office for Anal sis and of Operations Data FROM:
Robert D.' Martin, Regional Administrator
SUBJECT:
INPUT TO TE ABNORMAL OCCURRENCE REPORT TO CONBRESS In accordance with your Jansa 30 1987 memorandum on the Abnormal e fou,rth quarter of calender year 1986 w Occurrence Report to Congress orhave reviewed AEOD proposals and activities Pursuant to the At00 abnormal occurrence critaria we have identified one issue for rewrting. That issue is 'Immediately Effective Order Modify 1 License and Jeder to Show Casse Issued to Industrial Radiogra Company, and-a draft input is enclosed. We understand that 90tSS will prov de-input to i
safluoride Cylinder and Release of Sases.', Rupture of a Uranium te or close out, as appropriate A0 84 3 Apr questions in our input may be directed to Dale A. powers at FT5: 728-4195.
i Robert D. Martin Regional AMaistrater Ensleeste:
As stated esm/ enclosure:
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( j noo 14R7 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 I
classified as suggested by AE00.
Attached are inputs for the Region I items, as follows:
a Appendix C Items Anomalies During Loss of Of f site Power Testing at Hope Creek (PNO-1-86-78).
A copy of the attached Hope Creek input has been faxed to the NRR program manager, it is requested that AE00 solicit NRR input, especially regarding the generic implications of Bailey module failures.
Conviction of Licensee (International Nutronics, Irc., Dover, New Jersey) and One Empicyee in Federal District Court (PNO-I-56-89A)
' Item I
l Accidental Incineration of Radioactive Material at Henry Haywood Hospital (PN0-I-86-86) i Thomas E. Murley Regional Administrator
Attachment:
As Stated
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i HOPE CREEK AUGMENTED INSPECTION TEAM 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.
1 The plant is located.near Salem, New Jersey.
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On September ll,,1986 Public Service' Electric and Gas Company performed a loss
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of offsite1)ower 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.
i program.
Its. purpose is to demonstrate whether the plant response is satisfactory and in accordance with the plant design for concurrent loss of the turbine gene-
- rator and all offsite power sources.
- The Hope Creek Generating Station loss of offsite power test was initiated with
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the turbine generator loaded.to 165 MWe.
The first indication of an unsatisfactory plant response was the failure-of the "C" emergency diesel generator output breaker I
to close automatically.
Soon af ter, an observed failure of the reactor auxiliary.
cooling system coincident with increasing drywell pressure resulted in the testi i
being aborted by the licensee. Normal of fsite 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.
1 The most significant observations on September 11, 1936 were:
(1) emergency 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 i
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.
On September 19, 1986, Public Service Electric and Gas performed a cold loss of of f site 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
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' with plant design for loss of all offsite power sources af ter the licensee had 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-
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down.
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The significant observations during this test were:
(1) the "B" safety auxiliary t
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-4 well fan also failed to start. Hope Creek station personnel observing the test identified a total of 17 observations.
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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 ef festiveness 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-1 tion was a repeat of a previous observation and involved a Bailey logic module.
i The inspection been on September 25 and ended October 3,1986.
Of the 41 observations reported from the two loss of offsite power tests, the overall safety significance was concluded to be relatively min'or except for the 3
Bailey solid state losic module failures.
These modules, manufactured by the Bailey Meter Co., are multipurpose electronic devices used extensively throughout the plant fer contro' and safety functions.
Of eight hardware. failures identified i
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 eqcipment; (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 t
actuation functions.
t A number of minor plant design, construction, and manufacturing problems were also
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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
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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 observations 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 of fsite 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 surveillance testing and a failure rate which was higher than expected.
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' 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 component which' failed and, to tt e degree possible, the cause of the failure, An assessment, by the manufacturer, of module failures at installations of other users.
An accelerated aging and cycling test ?rogram, with a final reliability analysis report by the end of the second yiarter of 1987.
A monthly trending program that will provide a report bi-monthly indicating:
the number of module failures having an adverse affect r 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-lE 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.
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 1
without removing staple jumpers or the FPLA.
The determination of the feasibility and implications of modifying the exist-i ing Bailey system to permit in-situ testing.
i This program is underway and the results are under continued NRC review.
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i 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 millicuries of molybdenum-99 (as of' noon October 19,1986) had been inadvertently incinerated in' the hos-
-pttal'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 astumed that the molybdenum had vaporized and was released through-the stack.
The licensee's surveys of the grounds sur-i rounding the hospital likewise revealed only background radiation levels.
The incinerator was cleaned out and the debris held just in case it was contam-inated.
The licensee subsequently used the incinerator two more times.
Radiation surveys' conducted by NRC inspectors upon arrival at the site i
revealed radiation levels exceeding 200 mR/hr in the incinerator and the con-
- tainer holding the debris' from the incinerator.
Immediate corrective action taken by the licensee. included roping off the area surrounding the incinerator, j
removing the container containing the debris to a restricted access area, and i
shielding it with lead sheets.
Use of the incinerator was suspended until radiation levels decrease'to background.
About 1492 millicuries of techne-tium-99m was estimated to have been released.
The resulting concentration of l
technetium-99m released, 0.22 E-7 uti/ml when averaged over a one year period, was -less than the limits specified in 10 CFR 20, Appendix 8. Table II, Column.I for technetium-99m (5 E-7 uti/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 resulting concentration was'about 16 times the value of 10 CFR 20, Appendix B, 1
Table II.
1 A number of factors contributed to the incident, and include inadequate train-ing of the nuclear medicine technician who performed the initial surveys and
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the personnel who were expected to handle and control radioactive materials.
Also, inadequate management involvement in the ' program contribut:d to the licensee's in, effectiveness in correctly evaluating the affect of the event, j
No personnel exposures were attributable to this event, and no substantial i
hazard resulted to personnel in unrestricted areas.
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Conviction of International Nutronics, Inc., and one Employee in Federal District Court International Nutronics, Inc., (INI) a California corporation, and Eugene i
O'Sullivan, a Ccrporate 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 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.
l The charges resulted from a December 1982 incident involving a spill of radio-actively contaminated water at the INI' irradiation facility in Dover, New Jersey.
The spill resulted in widespread contamination of the facility, in-cluding 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.
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