ML20036B326
| ML20036B326 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/02/1992 |
| From: | Wenzinger E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20036A053 | List:
|
| References | |
| FOIA-92-162 NUDOCS 9305190137 | |
| Download: ML20036B326 (23) | |
Text
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UNITED sTAl[s
,0, NUCLEAR REGULATORY COMMISSION 3 y,,
j REGios a r
47s ALLf NDALE ROAD f
sung OF PPUSSI A. PENNSYLVANIA '9406141s s,
FS 0 2 92 l
1 I am responding to the concerns that you provided to us on May 29 and June 4,1991, l
asserting that inadequate work control resulted in a vent stack monitor being off line for 10 to 15 minutes, lack of attention to detail resulted in an improper valve line-up while filling the Volume Control Tank (VCT), and work was assigned without verification of equipment 1
condition at Millstone Unit '
\\
j These concerns were referred to Northeast Utilities (NU) for their evaluation; attached for j
your information is their response. We have evaluated their response and have determined i
the following; Concern 122
\\
The maintenance effort associated with the vent stack monitor (RM 8132) was i
l performed in accordance with established site procedures and the appropriate l
Technical Specification action statement was entered as required, when the one j
sample pump was removed from service while the other one was out-of-service. No j
i work control procedure problems were noted and the actions of the Control room
\\
l watchstanders were prompt and correct.
c Concern !?9 ew Poor communications between workers led to the unplanned draining of the Volume,Q l
Control Tank (VCT) reference leg. NU's efforts to stress better communications "M
failed to eliminate the problem in this specific instance. However, because the valve j
in question remained red tagged shut, as required, this particular event was not safety : A i
'P i
)
significant.
,;; k I
i Concern 130 5m I improper spare part storage and labeling were responsible for the work being assigned; J
e without verification of equipment condition. The good work practices of the y{
}
instrument specialist, who noted the difference in gages, avoided a problem. He
. fj 0 4 j
correctly obtained and installed the proper gage. This particular incident was not
}yg safety signincant because the gage was bemg installed in a non-safety-related i
application. Further. safety-related par s are not used from stock. but rather are obtained from the 5: ores Department usmg administrative traceability controls.
l 0
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l 305190137 930216 i
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HUBBARD92-162 PDR
Based on this information, no further action is planned by the NRC in these matters, and we consider these concems to be resobed.
We appreciate you informing us of your concerns and feel that we have been responsive.
Should you have any additional questions regarding these matters, please call me collect at (215) 337-5225.
I Sin 1erel y/,
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N Edward Wenzinger, Chief Reactor Projects Branch 4' J
Attachment:
NU Response Letter A09699 of September 13, 1991.
l
m 2
Based on this information, no further action is planned by the NRC in these matters, and we consider these concerns to be resolved.
We appreciate you informing us of your concems and feel that we have been responsive.
Should you have any additional questions regarding these matters, please call me collect at (215) 337-5225.
Sincerely, Edward Wenzinger, Chief Reactor Projects Branch 4
Attachment:
NU Response letter A09699 of September 13, 1991.
l l
bec /w encl:
l Allegation File: RI-91-A-0122. RI-91-A-C ':1 PJ 91-A-0130 E. Conner's files W. Raymond/T. Shedlosky Contractor's office fines (Meeker) l l
concurrences:
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Barkjev kfen(
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UNtTED ST ATEs o,
2 NUCLEAR REGULATORY COMMISSION REGION I I
47s ALLENDAtt ROAD N
g E
aiNG OF PRUSSIA. PENNSYLVANIA 1M5141s j
m., a I am responding to the concerns that you provided to us on May 29 and June 4,1991, asserting that inadequate work control resulted in a vent stack monitor being off line for 10 to 15 minutes, lack of attention to detail resulted in an improper valve line-up while filling the Volume Control Tank (VCT), and work was assigned without verification of equipment condition at Millstone Unit 2.
These concerns were referred to Northeast Utilities (NU) for their evaluation; attached for your information is their response. We have evaluated their response and have determined the following; Concern 122 The maintenance effort associated with the vent stack monitor (RM 8132) was performed in accordance with established site procedures and the appropriate Technical Specification action statement was entered as required, when the one sample pump was removed from service while the other one was out-of-service. No work control procedure problems were noted and the actions of the Control room watchstanders were prompt and correct.
l Concern 129 Poor communications between workers led to the unplanned draining of the Volume Control Tank (VCT) reference leg. NU's efforts to stress better communications failed to eliminate the problem in this specific instance. However, because the valve in question remained red tagged shut, as required, this particular event was not safety significant.
Concern 130 Improper spare part storage and labeling were responsible for the work being assigned e
without verification of equipment condition. The good work practices of the instrument specialist, who noted the difference in gages, avoided a problem. He correctly obtained and installed the proper gage. This panicular incident was not safety significant because the gage was being installed in a non-safety-related application. Funher. safety-related parts are not.used from stock, but rather are J
obtained from the Stores Depanment using administrative traceability controis.
Irdorrnation in this reto d wm dcWd in ::.ccreane v.M th i~recdr d hdctmation Act, enmMiens
/7 roiA._ f e w
g.
i s
2 3
Based on this information, no funher action is planned by the NRC in these matters, and we consider these concerns to be resolved.
1
\\
We appreciate you informing us of your concerns and feel that we have been responsive.
Should you have any additional questions regarding these matters, please call me collect at i
l (215) 337-5225.
\\
l
~
)
f'f.3ll
<>w,n Edward Wenzinger, Chief Reactor Projects Brane,b 4' J
Attachment:
NU Response letter A09699 of September 13, 1991.
i I
I l
i 1
2 N
Based on this information, no further action is planned by the NRC in these matters, and we consider these concerns to be resolved.
We appreciate you informing us of your concerns and feel that we have been responsive.
Should you have any additional questions regarding these matters, please call me collect at (215) 337-5225.
I Sincerely, Edward Wenzinger, Chief Reactor Projects Branch 4
Attachment:
NU Response Letter A09699 of September 13, 1991.
bcc /w encl:
Allegation File: RI-91-A-0122. RI-91-A-0129, RI-91-A-0130 E. Conner's files W. Raymond/T. Shedlosky Contractor's office files (Meeker) concurrences:
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UNITED si ATEs
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I RECON I g
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47s ALLENDALE ROAD
=:NG of FRusstA. PENNSYLVANIA 19406 141s
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a C ' 92 1
1 g
l I am responding to the concerns that you provided to us on May 29 and June 4,1991, asserting that inadequate work control resulted in a vent stack monitor being off line for 10 to 15 minutes, lack of attention to detail resulted in an improper valve line-up while filling i
the Volume Control Tank (VCT), and work was assigned without verification of equipment 4
condition at Millstone Unit 2.
)
These concerns were referred to Northeast Utilities (NU) for their evaluation; attached for j
your information is their response. We have evaluated their response and have determined i
the following;
{
Concern 122 1
The maintenance effort associated with the vent stack monitor (RM 8132) was performed in accordance with established site procedures and the appropriate i
Technical Specification action statement was entered as required. when the one sample pump was removed from service while the other one was out-of-service. No work control procedure problems were noted and the actions of the Control room watchstanders were prompt and correct.
Concern 129 Poor communications between workers led to the unplanned draining of the Volume Control Tank (VCT) reference leg. NU's efforts to stress better communications failed to eliminate the problem in this specific instance. However, because the valve in question remained red tagged shut, as required, this particular event was not safety significant.
Concern 130 Improper spare part storage and labeling were responsible for the work being assigned without verification of equipment condition. The good work practices of the instrument specialist, who noted the difference in gages, avoided a problem. He correctly obtained and installed the proper gage. This panicular incident was not j
safety significant because the gage was being installed in a no,rt-safety-related application. Funber. safety-related parts are not used from stock. but rather are obtained from the Stores Department using administrative traceability controls.
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Based on this information, no further action is planned by the NRC in these matters, and we consider these concerns to be resolved.
We appreciate you informing us of your concems and feel that we have been responsive.
Should you have any additional questions regarding these matters, please call me collect at (215) 337-5225.
pn erely, i
fY Edward Wenzinger, Ch Reactor Projects Branch 4
~
J
Attachment:
NU Response Letter A09699 of September 13, 1991.
l l
i I
2 Based on this information, no further action is planned by the NRC in these matters, and we consider these concerns to be resolved.
We appreciate you informing us of your concerns and feel that we have been responsive.
Should you have any additional questions regarding these mattars, please call me collect at (215) 337-5225.
1 Sincerely, r
1 Edward Wenzinger, Chief Reactor Projects Branch 4 i
Attachment:
NU Response Letter A09699 of September 13, 1991.
4 l
bec /w encl:
l Allegation File: RI A-0122. RI A-0129, RI-91-A-0130 E. Conner's files W. Raymond/T. Shedlosky Contractor's office files (Meeker) 1 I
~
i i
I concurrences:
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(4 Barkley he y Wenzinger
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Yes No Memo documenting why it was granted is attached?
Yes No bpscok' Position /
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September 13, 1991 Docket No. 50-336 A09699 RE: Employee Concerns Mr. Charles V. Behl, Director Division of Reactor Projects U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406
Dear Mr. Behl:
Millstone Nuclear Power Station Unit No. 2 RI-91-A-Oll3N Ve have completed our reviev of the identified issues concerning activities at Millstone Station.
As requested in your transmittal letter, our does not contain any personal privacy, proprietary, or safeguards response information.
The material contained in this response may be released to the public and placed in the NRC Public Document Room at your discretion.
The NRC letter and our response have received controlled and limited distribution on a "need to knov" basis during the preparation of this j
Additional time in which to respond to these issues was granted response.
by the Staff in telephone conversationi; of August 12 and August 30, 1991.
ISSUE ll3N:
On May 20, 1991, an operator observed an abnormal indication on the Unit 2 stack radiation monitor (RM B168).
The abnormal indication vas no variation on the meter. The operators secured and immediately reinstated power to the monitor and the meter response was noted to have returned.
On May 21, operators again observed no variation in the sonitor output. A trouble report was initiated and the technical specification action statement was entered for an inoperable monitor. The one day delay is an example of operators failing to promptly initiate a corrective action i
request and failing to enter the technical specification action statements when required.
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Mr. Charles V. Behl, Director U. S. Nuclear Regulatory Commission i
A09699/Page 2 September 13, 1991 Request:
assertions. If any deficiencies discuss the validity of the above are identified, please provide us with the corrective actions you have Please taken to prevent recurrence and assess the significance with regard to safety of the identified deficiencies.
f ISSUE 136:
failures were noted in the control June 3 to June 5,1991 repetitive room indication for the Unit 2 vent stack high range radiation monitor From RM8168A/B. On June 3 the " failure" laap was lit, and on June 5, 1991 a
' Trouble Tag" vas found to be in place.
The required technical specification action statements were not complied with during these repetitive failures.
Request:
discuss the validity of the above assertions. If any deficiencies identified, please Please in equipment availability or procedure compliance are provide us with the corrective actions you have taken to prevent recurrence and provide an assessment of the significance of the deficiencies with i
respect to safety.
Responses 113 & 136:
As issues 113 and 136 both deal with technical specification action statements relating to radiation monitor RM 8168, they vill both be answered in a single response as follows.
The chronology of observations reported in the two issues agrees with entries in the Millstone Unit No. 2 Shift Supervisor's leg, and with a chronology of Instrumentation & Controls (I&C) Department troubleshooting and repair activities.
to the specific decisions cited or implied in Issues 113 and 136, no failures to take required action occurred, ss discussed in the following Relative comments.
Taking immediate action to restore normal system output following an observed abnormal indication on RM 8168 vas an appropriate response for a Such occurrences single ' lockup' of this aferoprocessor-based instrument.
are not unusual.
Removing power to this monitor and then insediately restoring it, in effect " resets
- the device to its normal mode of the instrument is monitored routinely. It operation.
For this reason, vould not be necessary to submit a Trouble Report (TR) for such an isolated anomaly since the operator was able to immediately restore expected display the device was confirmed.
and the full operational capability of entry into an action statement vould not be appropriate since
- outputs, Furthermore, the radiation monitor operated properly once it was reset.
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Mr. Charles V. Behl, Directer i
U. S. Nuclear Regulatory Commission A09699/Fage 3 i
September 13, 1991 The RM 8168 performance anomaly observed on the morning of May 21 vas
- f not understood, and not resettable.
Evaluating the radiation re pe ti tive, monitor as 'out of service" as indicated by the Shift Supervisor's log i
entry of 0600, the operators entered the applicable Technical Specification l
action statement, and remained in that condition until May 23, 1991, when l
J l
replacement of a failed power supply was completed af ter IEC identified the 1
cause of the indication problems as a broken wire and failed 24 volt output.
Since the performance anomaly observed on the morning of May 21 was j
repetitive, not understood, and not resettable, both actions (i.e.,
submitting the Trouble Report and entering the Technical Specification i
~
action statement, Table 3.3-6. Action 17) vere clearly appropriate, During the period from June 3 to June 5,1991, Millstone Unit No. 2 was in i
Mode 5.
In Mode 5 radiation detector RM 8168 is not requ. red to be d
operable, hence under no conditions of RM 8168 performance vould the plant have entered into, or been operated in accordance with, the Technical Specification action statement for RM 8168.
The two scenarios noted above vere the result of a single probles. During the period from approximately May 24 through late July 1991, the LIC-8168 power supply anomaly caused intermittent power failure interrupts to be i
processed by the microprocessor. The intermittent lockup problem caused RM i
8168 to stop normal processing functions, recognizable in the control room by the radiation monitor display not changing and not responding to the test push button. This problem was known to the control room operators,
.i and corrective action to reset the radiation monitor was taken as needed.
Throughout this period, it was the judgment of on-shift supervisory personnel, Operations management, and I&C management (specifically discussed in a draf t Operability Evaluation approved by the IEC Manager on July 19, 1991),
that RM 8168 remained operable -i.e.
fully capable of meeting its Technical Specification functions.
In summary, af ter troubleshooting was completed, it was concluded that RM 8168 vas operated in a slightly degraded state for several veeks. This l
degradation manifested itself to control room operators as an intermittent lockup of the radiation sonitor, easily reset by on-shift operations 4
personnel.
These personnel vere alerted to the problem and checked the monitor regularly for proper operation.
on-shift supervisory personnel are tasked with initiating the appropriate 1
corrective action and compensatory seasures for equipment performance problems encountered during their shift. Judgment is frequently involved in such determinations.
Supervisors in the Operations Department are selected, trained, counseles and evaluated on their performance in such activities. The Operations Manager, other members of plant management, and specifically the Unit Duty O'ficer are available to consult with the Shif t I
1
1 Mr. Charlcs V. Behl, Director l
U. S. Nuclear Regulatory Commission A09699/Page 4 September 13, 1991 1
Supervisor concerning the level of response required for a given plant performance anomaly. Similarly, various members of the staff review plant performance and corrective actions taken on a regular basis during the workday; in this f ashion shif t operators' responses receive frequent and sultidisciplinary reviews on a continuing basis.
l At no time during the events described in Issues 113 and 136, nor at any time during the period of degraded RM 8168 operations, were the Shift Supervisors' judgments concerning operability or the need for corrective Therefore, these assertions are not valid.
action found to be in error.
3 Ve vere not svare that these vere issues of concern prior to receiving your letter of July 9, 1991.
4 ISSUE 114-1 (Unit 3):
l l
On May 22, 1991 during the HP-3 refuel outage a calibration error of the accumulator tank level transmitters was identified. The error was in the range of 25% due to static fluid between the transmitter and the instrument The calibration procedure did not address the error due to the level taps.
instrumentation piping configuration; therefore, the procedure was inadequate. Further, if the present instrument indication is correct, then i
it was achieved by using zero span adjustments without adhering to the calibration procedure.
Request (Unit 3):
)
Please discuss the validity of the above
,nertions. If any deficiencies in calibration procedures or procedural compliance are identified, please i
provide us with the corrective actions you have taken to prevent Please provide us with an assessment of the significance with recurrence.
l regard to safety of any identified deficiencies.
Response
l Ve have found no justification for the statements made in issue 114-1.
A calibration of the accumulator tank level transmitters was started on j
Tebruary 7,1991 and successfully completed on March 18, 1991. No vork vas performed on May 22, 1991, nor does the Shif t Supervisor's log indicate that such an error was identified on or near that date.
An error of 8.5% vas found to exist between level indications on a common accumulator after completion of the refuel outage calibration dated February 18, 1991.
This was in e: tess of the 5% desired maximum error i
between common channels and prompted a survey of "As-Built transmitter i
installations on March 16, 1991.
The Engineering Calculation and Surveillance vere revised to reflect the survey data. A second calibration vas completed on March 18, 1991 vith a noted maximum error of 0.47%.
J
. Mr. Charles V. Behl, Director U. S. Nuclear Regulatory Commission A09699/Page 5 September 13, 1991 The difference between indicated and actual level for the period of February 18, 1991 to March 18, 1991 was 13.3%.
As a result of the elevation differences discovered between channels, we know that the maximum and minimum indicated range is different for each transmitter.
An example is that one indicator vould read down to 6550 gallons while the other indicator on the same tank would stop at 6555 gallons.
There is no safety significance involved with the difference as both these indicated ranges are vell beyond the operating limits specified in Technical Specifications.
Tellow caution tags have been placed on the indicators to specify the minimum and maximum display values for each transmitter. New readout scales have been generated for the indicators to allow removal of the yellow caution tags. Ve are currently working to install these readout scales.
Ve are confident that the new method of calibration is more accurate, more repeatable and less time consuming to perform.
Indication differences between redundant channels on all accumulators are less than 44 gallons.
The present instrument indication is correct and the nev calibration method vill improve reliability. The calibration procedure was always adhered to during calibrations. No zero or span adjustments were cade unless directed by procedure, which is based on the Engineering Calculation.
This i
assertion is therefore not valid, and we vere not aware that this was a concern until notification by the Staf f's letter of July 9,1991.
ISSUE 114-2 (Unit 1):
On May 22, 1991 during the installation of the IRM cable detector assemblies under the reactor vessel, the RVP/BP controls were inadequate and resulted in the possible ingestion of radioactive material by a worker.
The cable was identified as "5K smearable" on May 22, 1991 and the RVP required workers to vear respirators.
However, on May 21, 1991, the RVP did not require respirators to do the same job.
Request (Unit 1):
Please discuss the validity of the above assertions. If any deficiencies are identified, please provide us with the corrective actions you have taken to prevent recurrence. Please provide us with an assessment of the significance with regard to safety of any identified deficiencies.
Response
l This assertion is not valid. The Health Physics controls for the under vessel IRM/SRM vork vere both adequate and conservative.
l l
l
Mr. Charles U. B:thl, Dircetcr U. S. Nuclear Regulatory Commission A09699/Page 6 s
September 13, 1991 The "5K smearable" referred to in issue 114-2 is the locee surface contamination detected during health physics surveys. This information is in terms of thousands of disintegrations per minute (dpe) over a expressed surface area of 100 square centimeters (cm2).
On May 22, 1991, the radiological data for the IRM cable work indicated a range of smearable contamination from 5K to 300K dpa/100 cm2 loose surface contasiastion. On the previous day the loose surface contaminaifon had been 20K to 50K dps/100 cm2.
The Health Physics department uses air samples in conjunction with a threshold loose surface contamination value of 100K dpa/100 cm2 for considering the required use of respiratory protection for this type of work.
On May 21, conditions vere such that the RVP required face shields and respiratory protection only if the work area contained dripping water from above.
On May 22, as a result of the work done the previous day the loose surface contamination survey results increased from the previous day's aaximum of 50K to a new value of 300K.
The air sample data obtained during and after the previous day's work did not require the use of respirators.
Bovever, based on this change in smearable contamination in the work area, Health Physics took the conservative step of requiring respirators.
The actions of Health Physics in requiring respirators on the day at issue vas a conservative step and no safety deficiencies are indicated. A review of personnel contamination events for the sonth of May 1991, reveals no personnel contamination events as a result of IRM/SRM under-vessel work.
Ve vere not aware of this concern until receipt of the Staff's letter.
ISSUE 116:
- Recently, a tagging error occurred during preparations for maintenance on the Clean Liquid Radioactive Vaste Effluent Monitor (RM 9049).
The solenoid valve isolation valves that needed to be tagged in accordance with prerequisites for the job vere not tagged.
Specifically, the valves designated to be traced by procedures IC2404AA and IC 2404AC vere not traced because the operations tag form was used to verify the tagging. The root cause of the error can be attributed to the I&C technician (who verified the tagging) not being trained and qualified as a " job supervisor".
Although there vas a qualified job supervisor associated with f
the work, this individual was allowed to leave the work area while an i
unqualified individual continued the job.
Request:
Please discuss the validity of the above assertions. If any deficiencies j
in vork control are identified, please provide us with the corrective actions you have taken to prevent recurrence and assess the significance of the deficiencies with respect to safety.
Mr. Charlos V. 5 hl Dirceter U. S. Nuclear Regulatory Commission A09699/Page 7 September 13, 1991
Response
This is a valid concern. The root cause of the tagging error at issue was a personnel error made by a plant equipment operator who placed the tags en the wrong valves.
This error was not detected by the I6C personnel assigned to the task who were expected to verify the adequacy and placement of the tagging. Verification was only made of the adequacy of the tagging documented by the completed teg log sheet in the Automated Vork Order (A90) package.
The actual placement of the tags was not verified as correct as required by procedure ACP 2.02C - Vork Orders.
Ve were aware of this issue prior to receipt of the Staff's notification.
As one action to prevent recurrence, all I&C personnel have been reminded of their responsibility to verify both the adequacy and the placement of safety tagging. There was no safety significance to the tagging error that was made. There vere no releases as a result of this event.
A task group has been formed to review tagging errors at all three Millstone units and provide an assessment of the level of performance of the station regarding the quality and implementation of the tagging program.
This group vill also provide recommendations to station management for ensuring that plant procedures and their use by our employees are adequate to minimize tagging errors in the future.
I i
This group vill present its recommendations to improve the progras along with an action plan for enhanced human performance to station management for review.
If appropriate, a meeting with Region I Staff vill be scheduled at the completion of this review to discuss the results of any actions planned.
ISSUE 122:
On or about May 29, 1991 vorkmen vere dispatched to troubleshoot a flow problem with the plant vent stack monitor (RM 8032AB) [ sic). At the time, the "A" sample pump vas running, pump "B" vas of f and flow vas as expected.
The pumps were svitched to permit the workers to investigate the flow probles. Pump "A" vas stopped, but "B" did not start due to a preventive i
l maintenance action that was still in progress.
As a result, the stack monitor was out of service for 10-15 minutes.
Request:
Please discuss the validity of the above assertions. If any deficiencies in vork control are identified, please provide us with the corrective actions you have taken to prevent recurrence and assess the significance of the deficiencies with respect to safety.
l 1
Mr. Charles V. Exhl, Director U. S. Nuclear Regulatory Commission A09699/Page 8 September 13, 1991 Response-This assertion is not valid.
On May 27, 1991 a trouble report was submitted to the Maintenance department to determine why the RM 8132 sample fan would not develop proper flow. Later that same evening the sample fan was tagged out of service. On May 29, 1991 II,C personnel worked en RM 81325, using AVO M2-91-05446, to check the low flow problem identif d on May 27.
The " Tagging Required" section of tbe AVO indicated t!st a Technical Specification action statement was involved. This entry was made by the control room operator at the time the AVO was released.
May 29, at 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br />, the plant entered Technical Specification action On s tatement 3.3.3.10.a. Table 3.3-13, Action 2 for RM 8132 being out of t
service.
The plant was logged out of the action statement at 1740 hours0.0201 days <br />0.483 hours <br />0.00288 weeks <br />6.6207e-4 months <br /> that same day. Nothing in the Shif t Supervisor's log indicates this was i
anything other than a planned event. Realizing that one sample pump was out of service for preventive maintenance and that the other might have flow
- problems, it was proper to enter the action statement and trouble-shoot the remaining pump.
We find no work control deficiency associated with this mainte-nance / trouble-shooting activity. We were not aware that this was an issue i
of concern prior to receipt of the Staf f's letter.
'l ISSUE 128:
On June 1, 1991 a worker learned that he had been assigned duty as the on-call IEC technician (Unit 2 Emergency plan) for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period from the morning of May 30 through the morning of May 31, 1991. The worker was unaware of this assignment on May 29 when he informed his supervisor that he vould not be at work on May 30 for personal reasons. The worker did not pick up the department radio paging device and no one else was assigned as l
his replacement. Lapses in on-call coverage such as this example occur on a routine frequency.
~
Request:
Please discuss the validity of the above assertions. If any deficiencies in the on-call coverage for emergency planning are identified, please provide us with the corrective actions you have taken to prevent In addition, please assess the frequency and significance with recurrence.
respect to safety of lapses in on-call coverage by the Instrument and Controls and Maintenance technical staffs.
l l
Response
This is a valid concern, of which Northeast Nuclear Energy Company (NNECO) is well aware.
A lapse in on-call coverage for this particular I&C Technician position did occur on May 30, 1991.
However, three I&C Technicians and three Maintenance Technicians, one per unit, are on call at any time.
I s-..
-,..-.~
n.
i Mr. Charles V. Bahl, Directer U. S. Nuclear Regulatory Commission 1
A09699/Page 9 j
Septenbar 13, 1991 i
j On-call schedules are published sonthly and cover a period of one sonth and They are distributed at the end of each month so that the five days.
personnel know their assignments for the upcoming month. A person on-call assigned to be on-call May 30/31 vould have been made avare of that by receiving a copy of the on-call list in late April.
It is assignment responsibility of the individual to review the list on a regular basis the to ensure that they pick up a radiopager on their assigned days.
Being excused from vork for personal reasons does not automatically release an individual f rom on-call responsibilities. Energency Plan Implementation (EPIP) 4211 directs an individual on-call but unable to fulfill Procedure their on-call obligations to arrange for a qualified substitute themselves.
An exception to this is if a person calls in sick on the day they are to the on-call responsibilities. Then supervision vill assign another individual. If an individual becomes incapacitated or otherwise 2nable to assume fulfill their on-call responsibilities outside of normal vorb ng hours, EPIP 4211 directs that individual to notify the Millstone Unit No. 1 Shift Supervisor (SS) who vill assign the Millstone Unit No. 1 Shift Supervisor Staf f Assistant (S$$A) to find a qualified relief.
The purpose of the on-call Station Emergency Organization (SEO) is to provide augmentation of shift personnel to provide adequate and timely to abnormal and emergency conditions. Any one system has failure response probabilities, e.g., individual pager failure, auto. accident or breakdown during response, etc. In view of this, Millstone Station has developed a response in-depth program which provides reasonable assurance that adequate SEO staffing is available in a timely aanner. The I&C and Maintenance Supervisors also supplement the SEO thereby exceeding Emergency Plan requirements.
Lapses in on-call coverage for certain technician positions occur more frequently than ve consider acceptable from a management perspective but from a safety perspective. Ve have not had a total lapse in coverage not for any of the Maintenance or I&C technician positions this year because of our response in-depth approach.
If an individual from Millstone Unit No. 2 not respond to a radiopager message during an emergency, the Millstone did Unit No.1 SSSA, upon notification by the Millstone Unit No.
2 SS, vould j
call that individual at home using the telephone. If the individual could be reached or was not able to respond, the Millstone Unit No. 1 SSSA not vill contact the next person on the on-call schedule for the same position determine availability to assume the on-call assignment. If necessary the SSSA vill continue to call until a qualified relief is found. This to process limits the significance of any lapses in coverage.
NNECO has recently upgraded the Emergency Notitication System to automatically verify the on-call SEO positions that have been notified of the event (called into the station system).
This enables the on-shift communicators to make back-up calls to alternate SE0 members.
emergency Each SEO position has a minimum of five trained staff and most non-aanager positions have between ten and twenty. Ve have taken further steps to strengthen the on-call assignment to the SEO, dissemination of on-call schedules to individuals, and have a traceable means of verification:
Mr. Charics V. B;hl, Dircetst U. S. Nuclcar R:gulstery Commissica
'A09699/Page 10 l
september 13, 1991 J
1.
A major revision is planned to EPIP 4211, "On Call Procedure",
clarifying and strengthening the responsibilities of the Lead Managers and on-call individuals.
2.
The station's Emergency Plan Coordinator has been assigned i
responsibility for maintaining and monitoring of the on-call schedule.
j 3.
A new procedure, EPIP 4617, ' Station Emergency Organization Response 3
Verification Drill", to require a quarterly unannounced activation of y
the SE0 is under final review.
i I
ISSUE 129:
On June 3, 1991, the periodic evolution of refilling the volume control i
tank (VCT) level instrument reference leg was performed in accordance with procedure IC-2428F.
During the reference leg fill, a worker noted an unexpected increase in VCT level. Because of this unexpected increase, it suspected that the evolution actually drained the VCT reference leg.
vas This observation was reported to supervision. Pressure in the primary 4
makeup water supply was checked, and it was discovered that valve 2CH-195 in the supply path was red tagged closed instead of being in the open position as specified by step 6.2 of procedure IC-2428P.
The valve alignment check had been performed by a Plant Equipment Operator. At that time the PE0 did not perform a hands-on position check of valve 2CH-195 and failed to notice the red tag indicating the valve was closed. There was a l
4 conflict between the work procedure IC-2428F, which required valve 2CH-195 to be open, and the requirement to prevent boron dilution during reactor s
shutdown, which required the valve to be closed.
a 1
Request:
I Please discuss the validity of the above assertions. If any deficiencies J
in vork control, attention to detail, or vork procedures are identified, please provide us with the corrective actions you have taken.to prevent and provide an assessment of the significance of the deficiency 1
l recurrence t
with respect to safety.
Response
In stating that valve 2CH-195 vas tagged closed, as required to prevent boron dilution during reactor shutdovn, the assertion is accurate.
Interviews with the I&C and Operations personnel involved have determined that there was a miscommunication regarding whether or not the valve lineup i
had been completed. The Plant Equipment Operator (PEO) had not previously told the 16C technician that the valve lineup had been completed when he was informed that the valve had been found closed.
The importance of complete and precise communications is stressed regularly to Millstone Unit No. 2 operators, and examples of intra-and ir.t e r-departmental communication shortcomings are used in training and counseling sessions.
4
w-Mr. Charlos V. Behl, Director U. S. Nuclcar R:rulatory Commissien A09699/Page 11 September 13, 1991 As this was the required valve position for the reactor conditions, and procedure IC 2428F is designed to ensure that the reference leg filling evolution does not adversely impact the VCT level indication process, there was no safety significance involved. We vere not aware that this was an issue of concern prior to receipt of the Staff's letter.
ISSUg 130:
On May 31, 1991, during the replacement of a local pressure indication gage PI8167 in the condensate recovery system a worker was issued the wrong part (diaphraga isolated liquid filled gage [ sic]) to replace a conventional gage that was already in service.
Instrument and Controls supervision is responsible to verify plant and equipment conditions, such as replacement part suitability before authorizing vork on a system.
Reques t:
Please discuss the validity of the above assertions.
If any deficiencies are identified, please provide us with the corrective actions you have taken to prevent recurrence and provide an assessment with respect to safety of the deficiency.
Response
The issue of the vrong gauge being issued to be installed is accurate. The difference in gauge type was noted by the instrument specialist and he obtained and installed the correct model gauge.
Issuing replacement parts is not a normal activity for the first-line supervisor.
Typically, replacement parts are identified and drawn from those maintained in stock. In this case the parts vere kept in the I&C shop and the box in which the parts vere stored was mislabeled. The supervisor mistook the diaphragm isolated gauge as one appropriate to be installed in this application.
There is no safety significance to this event. The pressure gauge monitors the discharge pressure of the auxiliary steam system condensate recovery tank.
This system has no safety function and the proper gauge was identified and installed. For safety-related systems, the parts required for maintenance are obtained from the Stores Department via a Material Issue Form which documents traceability of the parts issued. No additional action to prevent recurrence, other than review of the issue with the supervisor, is planned.
After our reviev and evaluation, ve find that these issues did not present any indication of a compromise of nuclear safety. Ve recognize the need to strive for a higher level of performance in these areas and we are aggressively vorking toward that objective. Ve appreciate the opportunity
1
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Mr. Charles V. B;hl, Director U. S. Nuclear Regulatory Consission A09699/Page 12 September 13, 1991 to respond and explain the basis of our actions and ve appreciate your granting additional time beyond the original 30 days for us to complete our l
vork.
Please :ontact my staff if there are further questions on any of l
these matters.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANT E. J.~ Kroczka 6/
Senior Vice President l
cc:
V.
J. Raymond, Senior Resident Inspector, Millstone Unit Nos.
1, 2, and 3 E.
C. Venzinger, Chief, Projects Branch No.
4, Division of Reactor j
Projects E. M. Kelly, Chief, Reactor Projects Section 4A J. T. Shedlosky, NRC, Millstone Nuclear Power Station i
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MRTFORD. CONNECTICUT 06141-o270 1
t J C 0".".7..C 003N5N0 Septeaber 13, 1991 Docket No. 50-336 A09699 FI: Employee Concerns l
Mr. Charles V. Behl, Director l
Division of Reactor Projects U. S. Nuclear Regulatory Coraission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406
Dear Mr. Behl:
I i
l 1
Mil' stone Nuclear Fover Station, Unit No. 2 RI-91-A-Oll3N Ve have co:pleted our reviev of the identified issues concerning activities at Millstone Station.
As requested an your transmittal letter, our response does not contain any personal privacy, proprietary, or safeguards information.
The material contained in this response may be released to the public and placed in the NRC Fublic Document Room at your discretion.
The NRC letter and our response have received controlled and limited distribution on a "need to knov" basis during the preparation of this response.
Additienal time in which to respond to these issues vas granted by the Staff in telephone conversations of August 12 and August 30, 1991.
ISSUE 113N:
On May 20, 1991, an operator observed an abnormal indication on the Unit 2 stack radiation monitor (RM 8168).
The abnormal indication vas no variation on the meter. The operators secured and immediately reinstated power to the monitor and the meter response was noted to have returned. On May 21, operators again observed no variation in the monitor output. A trouble report was initiated and the technical specification action statement was entered for an inoperable monitor. The one day delay is an example of operators failing to promptly initiate a corrective action request and failing to enter the technical specification action statements when required.
Mr. Char 13s V. 81hl, Dirceter U. S. Nuclear R:gulotory Commission i
A09699/Fage 2 September 13, 1991 Request assertions.
If any deficiencies Please discuss the validity of the above corrective actions you have are identified, please provide us with the taken to prevent recurrence and assess the significance with regard to safety of the identified deficiencies.
i ISSUE 136:
failures were noted in the control From June 3 to June 5, 1991 repetitive room indication for the Unit 2 vent stack high range radiation monitor RM8168A/B. On June 3 the " failure" lamp vas lit, and on June 5, 1991 a
" Trouble Tag" vas found to be in place.
The required technical specification action statements were not complied with during these repetitive failures.
Request:
assertions. If any deficiencies Please discuss the validity of the above identified, please in equipment availability or procedure compliance are provide us with the corrective actions you have taken to prevent recurrence and provide an assessment of the significance of the deficiencies with respect to safety.
Responses 113 & 136:
As issues 113 and 136 both deal vith technical specification action statements relating to radiation monitor RM 8168.
they vill both be ansvered in a single response as follovs.
The chronology of observations reported in the two issues agrees with entries in the Millstone Unit No. 2 Shift Supervisor's log, and with a chronology of Instrumentation & Controls (I&C) Department troubleshooting and repair activities.
to the specific decisions cited or implied in Issues 113 and 136, no f ailures to take required action occurred, as discussed in the following Relative j
comments.
Taking immediate action to restore normal system output following an abnormal indication on EM 8168 vas an appropriate response for a i
l observed Such occurrences
' lockup' of this microprocessor-based instrument.
are not unusual.
Removing power to this monitor and then insediately l
single restoring it, in effect " resets" the device to its normal mode of the instrument is monitored routinely. It operation.
For this reason, would not be necessary to submit a Trouble Report (TR) for such an isolated anomaly since the operator was able to immediately restore expected display the device vas confirmed.
and the full operational capability of entry into an action statement vould not be appropriate since
- outputs, Furthermore, the radiation monitor operated properly once it was reset.
1
i Mr. Charics V. B;hl, Dirceter U. S. Nuclear R:gulatory Cctaissien j
- A09699/Page 3 September 13, 1991 The RM 8168 performance anomaly observed on the morning of May 21 was repetitive, not understood, and not resettable.
Evaluating the radiation monitor as "out of service" as indicated by the Shift Supervisor's log entry of 0800, the operators entered the applicable Technical Specification action statement, and remained in that condition until May 23, 1991, when replacement of a failed power supply was completed af ter IEC identified the cause of the indication probless as a broken wire and failed 24 volt i
output.
)
Since the performance anomaly observed on the morning of May 21 was repetitive, not understood, and not resettable, both actions (i.e.,
submitting the Trouble Report and entering the Technical Specification i
action statement, Table 3.3-6, Action 17) vere clearly appropriate.
During the period from June 3 to June 5,1991, Millstone Unit No. 2 was in Mode 5.
In Mode 5 radf9 tion detector RM 8168 is not required to be operable, hence under no conditions of RM 8168 perfo ance vould the plant have entered into, or been operated in accordance with, the Technical Specification action statement for RM 8168.
The two scenarios noted above vere the result of a single problem. During the period from approximately May 24 through late July A991, the LIC-8168 power supply anomaly caused intermittent power failure interrupts to be processed by the microprocessor. The intermittent lockup problem caused RM 8168 to stop normal processing functions, recognizable in the control room by the radiation monitor display not changing and not responding to the test push button. This problem was known to the control room operators, l
and corrective action to reset the radiation monitor was taken as needed.
Throughout this period, it was the judgment of on-shift supervisory personnel, Operations management, and I&C management (specifically discussed in a draft Operability Evaluation approved by the I&C Manager on July 19, 1991),
that RM 8168 remained operable, i.e.
fully capable of i
meeting its Technical Specification functions.
In summary, af ter troubleshooting was completed, it was concluded that RM 8168 vas operated in a slightly degraded state for several veeks. This l
degradation manifested itself to control room operators as an intermittent lockup of the radiation monitor, easily reset by on-shift operations personnel.
These personnel vere alerted to the problem and checked the monitor regularly for proper operation.
On-shift supervisory personnel are tasked with initiating the appropriate corrective action and compensatory measures for equipment performance problems encountered during their shift. Judgment is frequently involved in such determinations.
Supervisors in the Operations Department are i
- selected, trained, counseled and evaluated on their performance in such activities. The Operationr Manager, other members of plant management, and specifically the Unit Duty Officer are available to consult vith the Shift i
j.
Mr. Charics V. I:bl, Dircetor U. S. Nuclear Regulatory Commission A09699/Page 4 September 13, 1991 Supervisor concerning the level of response required for a given plant performance anomaly.
Similarly, various members of the staff review plant
]
performance and corrective actions taken on a regular basis during the workday; in this fashion shift operators' responses receive frequent and
];
sultidisciplinary reviews on a continuing basis.
At no time during the events described in Issues 113 and 136, nor at any time during the period of degraded RM 8168 operations, vere the Shift Supervisors' judgments concerning operability or the need for corrective action found to be in error. Therefore, these assertions are not valid.
Ve vere not aware that these vere issues of concern prior to receiving your letter of July 9, 1991.
ISSUE 114-1 (Unit 3):
2 On May 22, 1991 during the MP-3 refuel outage a calibration error of the accumulator tank level transmitters vas identified. The error was in the range of 251 due to static fluid between the transmitter and the instrument taps. The calibration procedure did not address the error due to the level instrumentation piping configuration; therefore, the procedure was inadequate. Further, if the present instrument indication is correct, then it was achieved by using zero span adjustaents without adhering to the calibration procedure.
i 1
Request (Unit 3):
i j
Please discuss the validity of the above assertions.
If any deficiencies in calibration procedures or procedural compliance are identified, please provide us with the corrective actions you have taken to prevent Please provide us with an assessment of the significance with recurrence.
regard to safety of any identified deficiencies.
Response
Ve have found no justification for the statements made in issue 111.-l.
A calibration of the accumulator tank level transmitters was started on l
February 7, 1991 and successfully completed on March 18, 1991. No work was performed on May 22, 1991, nor does the Shift Supervisor's log indicate that such an error was identified on or near that date.
l An error of 8.5% vas found to exist between level indications on a common accumulator after completion of the refuel outage calibration dated j
February 18, 1991.
This was in excess of the 5% desired aaximum error j
between common channels and prompted a survey of "As-Built" transmitter installations on March 16, 1991.
The Engineering Calculation and Surveillance vere revised to reflect the survey data. A second calibration vas completed on March 18, 1991 vith a noted saximum error of 0.47%.
4 4
l
Mr. Charles V. Behl, Director
, U. S. Nuclear Regulatory Commission A09699/Page 5 September 13, 1991
~-
The difference between indicated and actual level for the period of February 18, 1991 to March 18, 1991 was 13.3%.
i As a result of the elevation differences discovered between channels, we know that the maximum and minimum indicated range is different for each transmitter.
An example is that one indicator would read down to 6550 gallons while the other indicator on the same tank would stop at 6555 There is no safety significance involved with the difference as gallons.these indicated ranges are vell beyond the operating limits specified both in Technical Specifications.
Tellow caution tags have been placed on the indicators to specify the j
minimum and saximum display values for each transmitter. New readout l
scales have been generated for the indicators to aller removal of the yellow caution tags. Ve are currently vorking to install these readout i
scales.
Ve are confident that the new method of calibration is more accurate, more l
repeatable and less tire consuming to perform.
Indication differences between redundant channels on all accumulators are less than 44 gallons.
The present instrument indication is correct and the new calibration method The calibration procedure was alvays adhered to a
improve reliability.
vill calibrations. No zero or span adjustments vere made unless directed during by procedure, which is based on the Engineering Calculation.
This l
assertion is theref ore not valid, and ve vere not avare that this was a concern until notification by the Staf f's letter of July 9,1991.
i ISSUE 114-2 (Unit 1}:
On May 22, 1991 during the installation of the IRM cable detector assemblies under the reactor vessel, the RVP/EP controls vere inadequate l
and resulted in the possible ingestion of radioactive material by a vorker.
The cable was identified as "$K smearable" on May 22, 1991 and the RVP l
required vorkers to vear respirators.
Eovever, on May 21, 1991, the RVP did not require respirators to do the same job.
Request (Unit 1):
Please discuss the validity of the above assertions. If any deficiencies are identified, please provide us with the corrective actions you have taken to prevent recurrence. Please provide us with an assessment of the significance with regard to safety of any identified deficiencies.
j
Response
This assertion is not valid. The Eealth Physics controls for the under vessel IRM/SRM vork vere both adequate and conservative.
Mr. Charles V. B:hl, Director U. S. Nuclear Regulatory Commission
~ A09699/Page 6 September 13, 1991 The "5K smearable" referred to in issue 114-2 is the loose surface contaminetion detected during health physics surveys. This information is expressed in terms of thousands of disintegrations per minute (dpe) over a surface arms of 100 square centimeters (cm2).
On May 22, 1991, the radiological data for the IRM cable work indicated a range of smearable contamination from 5K to 300K dpa/100 es2 loose surf ace contamination. On the previous day the loose surface contamination had been 20K to 50K dps/100 ca2.
The Health Physics department uses air samples in conjunction with a threshold loose surface contamination value of 100K dpa/100 ce2 for considering the required use of respiratory protection for this type of vork.
On May 21, conditions vere such that the RVP required face shields and respiratory protection only if the work area contained dripping vater from above. On May 22, as a result of the work done the previous day the loos e surface contamination survey results increased from the previous day's maximum of 50K to a new value of 300K.
The air sample data obtained during and after the previous day's work did not require the use of respiratore.
Bovever, based on this change in smearable contamination in I
the work area, Bealth Physics took the conservative step of requiring respirators.
l The actions of Health Physics in requiring respirators on the day at issue vas a conservative step and no safety deficiencies are indicated. A reviev of personnel contamination events for the sonth of May 1991, reveals no personnel contamination events as a result of IRM/SRM under-vessel work.
Ve vere not aware of this concern until receipt of the Staff's letter.
1 ISSUE 116:
Recently, a tagging error occurred during preparations for maintenance on the Clean Liquid Radioactive Vaste Effluent Monitor (RM 9049).
The solenoid valve isolation valves that needed to be tagged in accordance with prerequisites for the job vere not tagged.
Specifically, the valves designated to be traced by procedures IC2404AA and IC 2404AC vere not traced because the operations tag form vas used to verify the tagging. The root cause of the error can be attributed to the I&C technician (who verified the tagging) not being trained and qualified as a " job l
supervisor". Although there vas a qualified job supervisor associated with l
the work, this individual vas alloved to leave the work area while an I
unqualified individual continued the job.
Request:
Please discuss the validity of the above assertions. If any deficiencies in vork control are identified, please provide us with the corrective actions you have taken to prevent recurrence and assess the significance of the deficiencies with respect to safety.
l
l Mr. Charles V. H:hl, Dirceter U. S. Nuclear Regulatory Ccr.zission l
A09699/Page 7 September 13, 1991 ls I
Response
This is a valid concern. The root cause of the tagging error at issue was l
a personnel error made by a plant equipment operator vbo placed the tags on the wrong valves.
This error was not detected by the IEC personnel assigned to the task who vere expected to verify the adequacy and placement of the tagging. Verification was only made of the adequacy of the tagging i
documented by the completed tag log sheet in the Automated Vork Order (A90) package.
The actual placement of the tags was not verified as correct as required by procedure ACP 2.02C - Vork Orders, j
Ve vere avare of this issue prior to receipt of the Staff's notification.
As one action to prevent recurrence, all I&C personnel have been reminded
[
of their responsibility to verify both the adequacy and the placestnt of i
safety tagging. There was no safety significance to the tagging error that was made. There vere no releases as a result of this event.
A task group has been formed to review tagging errors at all three Millstene units and provide an assessment of the level of performance of the station regarding the quality and implementation of the tagging program.
This group vill also provide recommendations to station management for ensuring that plant procedures and their use by our employees are adequate to minimize tagging errors in the future.
I This group vill present its recommendations to improve the program along l
with an action plar' for enhanced human performance to station management l
for reviev.
If appropriate, a meeting with Region I Staff vill be scheduled at the completion of this review to discuss the results of any actions planned.
ISSUE 122:
On or about May 29, 1991 vorkmen vere dispatched to troubleshoot a flow problem with the plant vent stack monitor (RM 8032AB) [ sic]. At the time, the "A" sample pump was running, pump "B" vas of f and flov vas as expected.
The pumps were svitched to permit the workers to investigate the flow problem. Pump "A" vas stopped, but "B" did not start due to a preventive maintenance action that was still in progress.
As a result, the stack monitor was out of service for 10-15 minutes.
Request:
Please discuss the validity of the above assertions. If any deficiencies in vork control are identified, please provide us with the corrective actions you have taken to prevent recurrence and assess the significance of the deficiencies with respect to safety.
Mr. Charles V. 5:bl, Dirceter
~
l U. S. Nuclear Regulatory Commission A09699/Page 8 September 13, 1991
Response
This assertion is not valid.
On May 27, 1991 a trouble report was submitted to the Maintenance department to determine why the RN 8132 sample fan vould not develop proper flow. 14ter that same evening the sample fan l
vas tagged out of service. On May 29, 1991 IEC personnel worked on RM 8132B, using AVO M2-91-05446, to check the lov flow problem identified on May 27.
The " Tagging Required" section of the AVO indicated that a Technical Specification action statement was involved. This entry was made i
by the control room operator at the time the AVO was released.
l On May 29, at 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br />, the plant entered Technical Specification action statement 3.3.3.10.a. Table 3.3-13 Action 2 for RM 8132 being out of service.
The plant was logged out of the action statement at 1740 hours0.0201 days <br />0.483 hours <br />0.00288 weeks <br />6.6207e-4 months <br /> that same day. Nothing in the Shif t Supervisor's log indicates this was anything other than a planned event. Realizing that one sample pump vas out of service for preventive maintenance and that the other might have flow
- problems, it was proper to enter the action statement and trouble-shoot the remaining purp.
Ve find no work control deficiency associated with this mainte-nance / trouble-shooting activity. Ve vere not aware that this was an issue of concern prior to receipt of the Staff's letter.
ISSUE 128:
l l
l l
On June 1, 1991 a worker learned that he had been assigned duty as the I
on-call I&C technician (Uni: 2 Emergency plan) for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period from the morning of May 30 through the torning of May 31, 1991. The worker was unavare of this assignment on May 29 vhen he informed his supervisor that he vould not be at vork on May 30 for personal reasons. The worker did not pick up the department radio paging device and no one else was assigned as his replacement. Lapses in on-call coverage such as this example occur on i
a routine frequency.
l Request:
Please discuss the validity of the above assertions. If any deficiencies in the on-call coverage for emergency planning are identified, please provide us with the corrective actions you have taken to prevent In addition, please assess the frequency and significance with j
recurrence.
respect to safety of lapses in on-call coverage by the Instrument and i
Controls and Maintenance technical staffs.
Response
This is a valid concern, of which Northeast Nuclear Energy Company (NNECO) is well avare.
A lapse in on-call coverage for this particular I&C Technician position did occur on May 30, 1991.
- Bovever, three I&C Technicians and three Maintenance Technicians, one per unit, are on call at
{
any tice.
-Mr. Charics V. Behl, Directcr U. S. Nuclear Regulatcry Commissicn A09699/Page 9 September 13, 1991 on-call schedules are published monthly and cover a period of one month and five days. They are distributed at the end of each month so that the on-call personnel know their assignments for the upcoming month. A person assigned to be on-call May 30/31 would have been made aware of that by receiving a copy of the on-call list in late April. It is assignment responsibility of the individual to review the list on a regular basis the to ensure that they pick up a radiopager on their assigned days.
Being excused from work for personal reasons does not automatically release an individual from on-call responsibilities. Energency Plan Implementation (EPIP) 4211 directs an individual on-call but unable to fulfill Procedure their on-call obligations to arrange for a qualified substitute themselves.
An exception to this is if a person calls in sick on the day they are to the on-call responsibilities. Then supervision vill assign another assume individual. If an individual becomes incapacitated or otherwise unable to their on-call responsibilities outside of normal working hours, fulfill4211 directs that individual to notify the Millstone Unit No.1 Shif t EPIP Supervisor (55) who vill assign the Millstone Unit No. 1 Shift Supervisor Staf f Assistant (SSSA) to find a qualified relief.
The purpose of the on-call Station Emergency Organization (SEO) is to provide augmentation of shift personnel to provide adequate and timely to abnormal and emergency conditions. Any one systen has failure response probabilities, e.g., individual pager failure, auto accident or breakdown during response, etc.
In viev of this, Millstone Station has developed a response in-depth program which provides reasonable assurance that adequate SEO staffing is available in a timely manner. The I&C and Maintenance Supervisors also supp'ement the SEO thereby exceeding Emergency Plan requirements.
Lapses in on-call coverage for certain technician positions occur more frequently than ve consider acceptable from a management perspective but f rom a safety perspective. Ve have not had a total lapse in coverage not for any of the Maintenance or I&C technician positions this year because of our response in-depth approach. If an individual from Millstone Unit No. 2 not respond to a radiopager zessage during an emergency, the Millstone did Unit No.1 SSSA, upon notification by the Millstone Unit No.
2 55, vould call that individual at home using the telephone.
If the individual could be reached or was not able to respond, the Millstone Unit No.1 SSSA not vill contact the next person on the on-call schedule for the same position determine availability to assume the on-call assignment. If necessary tothe SSSA vill continue to call until a qualified relief is found. This process limits the significance of any lapses in coverage.
NNECO has recently upgraded the Emergency Notification System to automatically verify the on-call SEO positions that have been notified of the event (called into the station systes).
This enables the on-shift communicators to make back-up calls to alternate SE0 sembers.
emergency SEO position has a sinimum of five trained staff and most non-sanager Each positions have betveen ten and twenty. Ve have taken further steps to the on-call assignment to the SEO, dissemination of on-call strengthen t
schedules to individuals, and have a traceable ceans of verification:
Mr. Charles V. 5:hl, Directer U. 5. Nuclcer Rcgulatsry Consissica A09699/Fage 10 l
September 13, 1991 l
1.
A major revision is planned to EPIP 4211 "On Call Procedure",
e clarifying and strengthening the responsibilities of the land Managers and on-call individuals.
Emergency Plan Coordinator has been assigned 2.
The station's responsibility for maintaining and monitoring of the on-call schedule.
3.
A new procedure, EPIP 4617, ' Station Emergency Organisation Response Verification Drill", to require a quarterly unannounced activation of the SE0 is under final review.
i ISSUE 129:
i the periodic evolution of refilling the volume control June 3, 1991, On (VCT) level instrument reference leg was performed in accordance with tar.k procedure IC-2428F.
During the reference leg fill, a worker noted an unexpected increase in VCT level.
Because of this unexpected increase, it l
suspected that the evolution actually drained the VCT reference leg.
This observation was reported to supervision. Pressure in the primary l
vas makeup vater supply was checked, and it was discovered that valve 2CE-195 in the supply path was red tagged closed instead of being in the open position as specified by step 6.2 of procedure IC-2428F.
The valve At. tha t check had been performed by a Plant Equipment Operator.
alignment time the PE0 did not perform a hands-on position check of valve 2CH-195 and to notice the red tag indicating the valve was closed. There was a failed between the work procedure IC-2428F, which required valve 2CE-195 conflict to be open, and the requirement to prevent boron dilution during reactor l
I shutdovn, which required the valve to be closed.
Request:
Please discuss the validity of the above assertions. If any deficiencies in vork control, attention to detail, or vork procedures are identified, provide us with the corrective actions you have taken to prevent please and provide an assessment of the significance of the deficiency recurrence with respect to safety.
Responser In stating that valve 2CH-195 was tagged closed, as requi. red to prevent boron dilution during reactor shutdovn, the assertion is accurate.
with the I&C and Operations personnel involved have determined Interviews that there vas a miscommunication regarding whether or not the valve lineup had been completed. The Plant Equipment Operator (FEO) had not previously told the IEC technician that the valve lineup had been completed when he was informed that the valve had been found closed.
The importance of complete and precise communications is stressed regularly to Millstone Unit No. 2 operators, and examples of intra-and inter-departmental communication shortcomings are used in training anc counseling sessions.
(
I I
i l
Mr. Charles V. 5:bl. Director U. S. Nuclear Regulatory Commissicn A09699/Page 11 September 13, 1991 i
As this was the required valve position for the reactor conditions, and procedure IC 2428F is designed to ensure that the reference leg filling evolution does not adversely ispect the VCT level indication process, there was no safety significance involved. We vere not svare that this was an issue of concern prior to receipt of the Staff's letter.
ISSUg 130:
On May 31, 1991, during the replacement of a local pressure indication gage PI8167 in the condensate recovery system a worker was issued the wrong part (diaphraga isolated liquid filled gage [ sic]) to replace a conventional gage that was already in service. Instrument and Controls supervision is responsible to verify plant and equipment conditions, such as replacement part suitability before authorizing vork on a system.
Request:
Please discuss the validity of the above assertions. If any deficiencies are identified, please provide us with the corrective actions you have taken to prevent recurrence and provide an assessment with respect to safety of the deficiency.
Response
The issue of the vrong gauge being issued to be installed is accurate. The difference in gauge type vas noted by the instrument specialist and he obtained and installed the correct model gauge.
Issuing replacement parts is not a normal activity for the first-line supervisor.
Typically, replacement parts are identified and dravn from those maintained in stock. In this case the parts vere kept in the I&C shop and the box in which the parts vere stored was mislabeled. The supervisor mistook the diaphragm isolated gauge as one appropriate to be installed in this application.
There is no safety significance to this event. The pressure gauge monitors i
the discharge pressure of the auxiliary steam systes condensate recovery tank.
This system has no safety function and the proper gauge vas J
identified and installed. For safety-related systems, the parts required i
for maintenance are obtained from the Stores Department via a Material Issue Form which documents traceability of the parts issued. No additional i
action to prevent recurrence, other than review of the issue with the supervisor, is planned.
After our reviev and evaluation, ve find that these issues did not present any indication of a compromise of nuclear safety. Ve recognize the need to strive for a higher level of performance in these areas and we are aggressively vorking toward that objective. Ve appreciate the opportunity i
t Mr. Ch;rles V. Behl, Director U. S. Nucl:ar Regulatory Coc. ission m
A09699/Page 12 September 13, 1991 to respond and explain the basis of our actions and we appreciate your granting additional time beyond the original 30 days for us to complete our l
vork.
Please contact my staff if there are further questions on any of these matters.
Very truly yours, NORTHEAST NUCLEAR ENERGT COMPANT E. J.~ M'roczka 6/
Senior Vice President cc:
V.
J. Raymond, Senior Resident Inspector, Millstone Unit Nos.
1, 2,
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ud3 E.
C. Venzinger, Chief, Projects Branch No.
4, Division of Reactor Projects E. M. Kelly, Chief, Reactor Projects Section 4A J. T. Shedlosky, NRC, Millstone Nuclear Power Station i
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- f JUL 3 01991 9..
MEMORANDUM FOR:
Cbester W. White, Field Office Director Office of Investigations Charles W. Hehl, Director FROM:
Division of Reactor Projects 3,
N l' NOTIFICATION REGARDING ALLEGATIONS RI-91-A-0091, f
SUBJECT:
RI-91-A-0134, AND RI 91-A-0029 l
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Charles W. Hehl, Director Division of Reactor Projects l
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E. Wenzinger j
E. Kelly J. Stewart l
i Files RI-91-A-0091 l
RI-91-A-0134 RI-91-A-0029 i
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WAY 101991 i
i Jose Calvo, Assistant Director for Region I l
MEMORANDUM FOR:
Reactors l
Division of Reector Projects - I/II i
Office of Nuclear Reactor Regulation I.
i Charles W. Mehl, Director FRON:
Division of Reactor Projects i
Region I l
PROPOSED TASK INTERFACE RBGARDING
SUBJECT:
COMPLIANCE OF MILLSTONE UNIT 2 NITN GENERAL DESIGN CRITERION 17 i
f Youi assistance is requested in resolving a concern originally identified by an SSFI conducted by the licensee approximately two l
years ago and subsequently classified as an allegation by the 4
NRC.
j The technical concern involves the compliance of Millstone Unit 2 a
5 with General Design Criterion Number 17; specifically, a scenario l
t wherein a single fault on an onsite bus and eventual loss of the i
J Reserve Station Service Transformer (RSST) could cause the simultaneous disruption of power to both Unit 2 redundant 4ky I
i class 1E divisions.
Our preliminary assessment has identified two issues:
one involving a protective relaying question of j
breaker coordination; and, the other a licensing question regarding GDC17 acceptability.
The implications -- if true --
l are of such significance that an independent assessment by the i
Enclosed is an internal Northeast Utilities l
NRC is warranted.
document dated May 2, 1991, containing an indapendent, third-J l
party analysis of GDc-17 compliance for Millstone Unit 2.
l i
We would appreciate your prompt attention to this matter and a f
l preliminary assessment by July 1991 so that, if a significant weakness is identified, action can be taken to ensure that the j
station can mitigate the ef fects of a blackout.
The Region I l
point of contact is Edward Wenzinger, Chief, Reactor Projects 4
j Branch 4 (FTS 346-5225).
This TIA proposal has been discussed I
with John Stola of NRR.
I 1
Char Ds W. Heh Directak.
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Division of Reactor Projects 1
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Enclosure:
As stated i
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