IR 05000245/1988002
| ML20150B605 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/01/1988 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20150B592 | List: |
| References | |
| 50-245-88-02, 50-245-88-2, IEB-87-002, IEB-87-2, NUDOCS 8803170075 | |
| Download: ML20150B605 (15) | |
Text
..
-
-
-.
. -
-.
- -
.
..
--
.
.
U.S. NUCLEAR REGVLATORY COMMISSION
REGION I
Report:
50-245/88-02 Docket No:
50-245 License No:
DPR-21 Licensee:
Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06141-0270
-
Facility:
Millstone Nuclear Power Station, Waterford, Connecticut
,
Inspection at: Millstone Unit 1 Dates:
January 1, 1988 through February 8, 1988 Inspectors:
William Raymond, Senior Resident Inspector Lynn Kolonauski, Resident Inspector Thomas Shedlosky, Senior Resident Inspector, CY
.
Leonard Prividy, Reactor Engineer, Special Test Programs Reporting Inspector:
William Raymond Approved:
[2 b
[
E.C.McCabe, Chief,[eactorProjectsSectionIB Date
,
Summary: Inspection from January 1 - February 8,1988 (Report No. 50-245/88-02)
Areas Inspected: This inspection included routine NRC resident and regional specialist inspection (107 hours0.00124 days <br />0.0297 hours <br />1.76918e-4 weeks <br />4.07135e-5 months <br />) of previously identified items, plant operations, radiation protection, physical security, the second 10 year in-service testing (IST) program, an unplanned radwaste discharge, spent fuel pool overflow, the dual role Shift Supervisor / Shift Technical Advisor issue, a potential inaccuracy of the containment high range radiation monitors, licensee event reports, committee acti-vities, and on IE Bulletin 87-02 - Fastener Testing.
Results: The inspection identified no unsafe plant conditions.
Further licensee and/or inspector followup is warranted on: (1) the 10 year IST program (Section 6.0), (ii) evaluation of OP-310, "Fuel Pool System," to determine its contribution to the Spent Fuel Pool overflow on January 29 and whether revisions are needed to prevent recurrence (Section 8.0), and (iii) resolution of the potential inaccuracy of the in-containment high range radiation monitors installed pursuant to the TMI action plan (Section 10.0).
8803170075 880304 DR ADOCK 05000245 PDR
..
.
-___ ___
.
.
TABLE OF CONTENTS Page 1.
Persons Contacted...................................................
2.
Summary of Facility Activities................................
.....
3.
Status of Previous Inspection Findings..............................
3.1 Inspection Item 87-14 Section 2.4: Policy Regarding the Use of Lead Seals.................................................
...
4.
Facility Tours and Operational Status Reviews.....................
4.1 Safety System Operability...................
..................
4.2 Plant Incident Reports.................
......................
5.
Plant Security.........
.
.......................................
5.1 Contractor Unfit for Duty..
...........................
....
5.2 Guard Inattentive at Post.......
....
..................
5.3 Safeguards Event Report 87-22-00.....
.......................
5.4 One Hour Reportable Events per 10 CFR 73.71....................
6.
Second 10 Year In Service Testing (IST) Program.
..................
.
7.
Unplanned Radwaste Discharge...........................
...........
8.
Overflow of Spent Fuel Pool.....................................
9.
High Radiation Area Doors Found Uniccked..............
............
10.
Potential Inaccuracy of Containment High Range Radiation Monitors...
11.
IE Bulletir, 87-02, Fastener Testing to Determine Conformance with Applicable Material Specificatic7s - TI 2500/26.
................
12.
Qual Role Shift Supervisor / Shift Technical Advisor (SS/STA) Issue...
13. On-site Plant Operations Review Committee (P0RC)....................
14. Management Meetings.........
j
..................................
l l
i
l
_. _ _..
_ _ _
- -
.~.
_. _
.
.
..
_
_
.
.
.
l l
DETAILS l
1.0 Persons Contacted Mr. S. Scace, Station Superintendent Mr. J. Stetz, Unit 1 Superintendent Mr. R. Palmiari, Operations Superintendent Mr. J. Summa, Assistant Engineering Supervisor Mr. M. Bigiarelli, Assistant Engineering Supervisor Mr. J. Leeson, IST Coordinator, Engineering Mr. R. Kacich, Manager, Generation Facilities Licensing Mr. J. Barnett, Licensing Mr. R. Crandall, Supervisor, Radiological Engineering Ms. P. Weekley, Security Supervisor The inspector also contacted other members of the Operations, Radiation Pro-tection, Chemistry, Instrumentation and Control, Maintenance, Engineering, and Security departments.
,
2.0 Summary of Facility Activities
'
'
Millstone 1 operated at full power except for power reductions for routine maintenance and surveillance.
3.0 Status of Previous Inspection Findings 3.1 (Closed) Item 87-14, Section 2.4: policy on use of_ Lead Seals
.
As documented in NRC inspection report 50-245/87-14, an inspector noted that both a pressure switch isolation valve and a gauge drain line isolation valve had been operated during an electric fire pump surveil-lance even though lead seals were attached to the valve handles. The inspector determined that the problem did not present a significant safety concern, but identified that no station policy regarding the use of lead seals existed.
l The inspection report states that the licensee committed to developing an instruction delineating station policy regarding lead seals.
The licensee has since determined that a policy is not required because the Millstone units do not use lead seals.
The two lead seals discussed in
!
the report were isolated cases and have been removed.
The inspector agreed with the licensee's determination and has not observed any other lead seals in use at Millstone.
The inspector had no further questions.
i 4.0 Facility Tcurs and Operational Status Reviews
!
Control room instrumentation was inspected for correlation between channels, proper functioning, and conformance with Technical Specifications (TS).
Alarm conditions in effect and alarms received were reviewed and discussed with the
- operators. Operator awareness and response to off normal conditions was re-
..
_
.. -
_
__._
.
._
-. _.
..
. _
.
-2
.
viewed; operators were found to be cognizant of plant conditions and indica-tions. Operating legs and Plant Incident Reports (PIRs) were reviewed for accuracy and adherence to station procedures.
Posting, control, and the use of personnel monitoring devices for radiation, contamination, and high radt-ation areas were inspected. Plant housekeeping controls were observed, in-cluding control of flammable and other hazardous materials.
Inspections of the control room were conducted on backshif ts on January 4 at 7:30 p.m., on January 23 at 4:30 p.m., and on February 8 at 2:30 p.m.
All shift personnel were found to be alert and attentive to their duties. No unacceptable condi-
,
tions were identified. The following activities were also addressed.
'
4.1 Safety System Operability Standby emergency systems were reviewed for operability and readiness for automatic initiation.
The following systems were included in the review: low pressure coolant injection, core spray, feedwater coolant injection, and standby gas treatment.
The status of the isolation con-denser, control rod drive hydraulic control units, and the emergency diesel generator was also inspected.
The reviews considered proper positioning of major flow path valves, operable normal and emergency power supplies, proper operation of indications and controls, and visual inspection for proper lubrication, cooling, and other conditions.
References used for the review included the Updated Final Safety Analysis Report, flow diagrams, and operating procedures.
No inadequacies were identified.
4.2 Piant Incident' Reports Selected Plant Incident Reports (PIRs) were reviewed to (i) determine the significance of the events, (ii) review the licensee's evaluation of the events, (iii) review the licensee's response and corrective ac-tions, and (iv) verify whether the licensee reported the' events in ac-cordance with applicable reporting requirements.
,
No inadequacies were identified.
The following PIRs were reviewed and i
are described elsewhere in this report, as referenced: 1-88-04 (Section i
7.0), 1-88-07 (Section 8.0), and 1-87-106, 1-88-01, and 1-88-02 (Section
~
"
9.0).
l 5.0 Station Security During station tours, the inspectors verified proper implementation of selected aspects of the station security program.
These included site access controls, personnel searches, adequacy of physical barriers, reporting of security events, compensatory measures, and guard force response to alarms and degraded conditions.
Except as discussed in Section 5.4 balow, no inadequacies were identified.
The following events warranted further inspector followup.
.
. -
.
.
--
.
.
m
.
5.1 Contractor Unfit for Duty On January 27, an associate construction representative for Northeast Utilities Services Company (NUSCO) arrived at the site's south access point (SAP) while apparently under the influence of alcohol.
The guard force noted the individual's condition and denied him access to the Pro-tected Area (PA).
The licensee removed his badge, thereby removing his authorization for access to the PA.
The individual had been under ob-servation for this problem and was enrolled in the employee assistance program; the individual has since been terminated.
The inspector had no further questions.
5.2 Guard Inattentive at Post On February 3, a security supervisor discovered a security guard, who was posted as compensation for an inoperable intrusion detection device, inattentive to her post.
The supervisor described the individual as
"nodding off."
The guard had been stationed at the post for approxi-mately one hour prior to the discovery, and was removed and replaced..
A subsequent tour of the post area revealed no signs of intrusion. Also, the surveillance area of a second guard posted in the immediate area at the time did overlap somewhat with the surveillance area of the inatten-tive guard.
The guard in question is under suspension pending further licensee review.
The guard had an excellent work record previous to the incident.
The inspector had no further questions.
5.3 Safeguards Event Report _87-022 Safeguards Event Report (SER) 87-022-00 describes the unauthorized and inadvertent entries by two contractors into a VA.
The event itself has already been discussed in inspection report 50-245/87-33. Another con-cern is the completeness of the SER.
It lacked sufficient detail to allow a reader to fully understand the circumstances of the event or the
,
analysis conducted by the licensee to determine the cause.
Specifically, j
the SER did not state that the individuals inadvertently entered the VA
'
and therefore failed to substantiate the statement "Safety Systems Af-fected or Threatened: none." This SER has been discussed with the Se-curity Supervisor, who agreed to include more detail in future SERs.
The inspector will continue to review the licensee's reporting of safe-guards events.
5.4 One Hour Reportable Events per 10 CFR 73.71c A number of recent security events, incleding the event addressed by Section 5.3, involve violations of VA access procedures.
requires reporting, within one hour, of sabotage attempts.
Regulatory Guide 5.62, "Reporting Safeguards Events," states that safeguards events
e
.
reportable within one hour under 10 CFR 73.71(b) include any entry of an unauthorized person into a vital area.
This Regulatory Guide provi-sion is more conservative than the regulatory requirement.
Discussions between the resident and NRC regional inspectors and head-quarters security personnel revealed specific guidance based on recently developed NUREG 1304, "Reporting ~of Safeguards Events," which will be distributed to all power reactor licensees in April 1988.
If an indi-vidual unintentionally enters a VA, and prompt corrective action is taken, the event is to be logged, but it is not reportable to the NRC within one hour.
The NUREG 1304 emphasis is on the intentions of the individual involved.
If an individual who enters a VA represents a threat to se-curity, or if the licensee is unable to determine the intentions of the individual, then the event is reportable within one hour. Also, an un-authorized entry into a VA would not be reportable within one hour if the individual would have been authorized VA access had it been requested, or if access authorization were delayed merely because associated paper-work had not been completed.
The above information was discussed with the Security Supervisor.
The inspector will continue to monitor the licensee's reporting of safeguards events, including applications of the guidance provided by NUREG 1304.
i Pending the dist-ibution of NUREG 1304, the licensee may reference this inspection report in justifying reportability decisions for future ap-plicable safeguards events.
6.0 In Service Testirrg (IST) Program
)
6.1 Purpose and Background On January 20 and 21, the inspector reviewed the licensee's second 10-
{
year program for inservice testing (IST) of pumps and valves.
In para-
'
llel with this review, the inspector assessed the status of the licen-see's response to four unresolved items (UNR 87-09-03, 04, 05 and 07)
l that originated in a special check valve inspection, j
The second 10 year IST program interval began on December 28, 1980.
Since that time, several licensing actions have occurred. The most re-cent licensing actions were the issuance of a draft safety evaluation report (SER) to the licensee for comment on May 22, 1985 and receipt of comment; and a revised IST program from the licensee on September 26, 1985.
i 6.2 Inspection and Review The inspector discussed his individual comments and concerns with the licensee's IST coordinator. Many of these comments and concerns were developed by reviewing the licensee's IST program positions (as stated in their submittal of September 22, 1985) and comparing these to current NRC staff positions.
It was apparent that the IST program needed modi-
-
-
,
..
~
.
.
.
- _ _ -
_ _ - _ _ _ _ _ _
.
.
fication of various items which ranged from changes in the IST program valve tables to changes of valve and pump relief requests.
Two of the more significant items were (1) testing of containment isolation valves (CIVs) and (2) flow instrumentation for pump testing.
6.2.1 CIV Item The inspector and the licensee discussed the basis for not.in-cluding or not leak rate testing certain power-operated and check valves in the IST program.
It appeared that certain valves in "closed loop" systems (e.g., Core Spray valves 1-CS-6A and 68) serve as CIVs but were not listed to be leak rate tested in the IST program valve table.
The licensee responded that prior NRC approvals authorized the use of only one CIV for "closed loop" systems and Valves 1-CS-6A and 6B do not require leak rate testing because other valves in the'same piping are designated as CIVs to be leak rate tested. The inspector noted that unresolved items 87-09-04 and 87-09-05 are involved.
The licensee indicated that a future submittal would identify specific NRC approvals substantiating their position. Also, the licensee noted that a draft submittal concerning various CIVs had been prepared and would soon be sent to the NRC.
This submittal addresses other issues related to 10CFR50. Appendix J such as exemption requests with justi-fication for not leak rate testing certain CIVs and designation of different valves as CIVs for certain containment penetra-tions. While.the review of this submittal is un Appendix J review which is separate from the IST program review, the de-terminations from the Appendix J review will be directly ap-plicable to the IST program.
The inspector emphasized to the licensee the need to closely coordinate this future Appendix J submittal with the IST program.
6.2.2 Flow Instrumentation For the second 10 year IST program the licensee has committed to implement (as nearly as possible)Section XI of the 1980 ASME Code including the 1980 Winter Addenda requirement for the measurement of individual pump flow rates.
However, there is no individual pump flow instrumentation for the condensate, condensate booster, reactor feedwater seal injection, service water, reactor building closed cooling water, and secondary closed cooiing water pumps.
In 1985, the licensee submitted relief requests with technical information to justify the ab-
'
sence of individual flow instrumentation for these pumps.
Discussion of this item with the licensee indicated that indi-vidual pump flow measurement may be obtainable by procedural changes while using existing instrumentation or by installation of new flow instrumentation.
Some preliminary work in this regard has been performed since 1985.
,.,. _ _ _.
.
._
_
-
- -.
.
..
... - -
!
--
- - -
. _=.
.
=
~
..
.
.
. -
.
.
l I
l
.
)
6.3 Summary and Conclusions On January 21, the inspector and the licensee's IST coordinator agreed to a detailed list of items that need resolution in order to issue a SER'
for the second 10 year IST program.
These items would also facilitate resolution of unresolved items 87-09-03, 04, 05 and 07, which remain open.
The inspector met with the IST coordinator's immediate supervisor and discussed the major items.
In the exit interview with the licensee's station superintendent, the licensee committed to providing the NRC with a submittal addressing the items discussed.
The licensee expected this submittal to be issued by May 1988.
7.0 Unplanned Radwaste Discharge The inspector reviewed PIR 1-88-04 which describes the unplanned discharge of 230 gallons from an unsampled Floor Drain Sample Tank (FDST) to Long Island Sound. The inspector obtained the following information from discussions with
,
licensee personnel and review of the radwaste log and other documentation.
On January 20, a Plant Equipment Operator (PE0) prepared the "A" FOST for discharge.
The tank had been properly recirculated and sampled in accordance with Operations Procedure (0P) 3128, "Liquid Radwaste Discharge - Floor Drain System." OP-3128 directs the operator to shut the discharge vai of the s
sample tank not to be discharged and then to open the discharge valve of the tank to be discharged and enter the valve lineup in the radwaste log.
The procedure then requires a second operator to independently verify and log the lineup.
Contrary to the procedure, both discharge valves were initially closed and independently verified and logged as such prior to the discharge.
At the commencement of the discharge, the radwaste PE0 opened the discharge valve for the opposite FDST ("B") ana then started the opposite FOST pump ("B").
Af ter a few minutes, the discharge was automatically terminated when the effluent radiation monitor tripped on high-high radiation,-causing the isolation valve for the discharge flow control valve to automatically close.
The Shif t Supervisor (SS) directed that the "A" FOST be resampled and the isotopic data be compared'to the initial sample in an effort to determine the cause of the effluent monitor trip.
The PE0 then noticed a decrease in the
"B" FOST level while the "A" FOST level did not change. After reviewing the log and the valve lineup, the PE0 realized his mistake and contacted the con-trol room.
He then changed the log entry for the "A" discharge valve from
"CL" to "0 PEN."
The mistake should have been single-lined, initialed and dated; this practice was overlooked by the PEO.
In follow-up discussions,
'
the PE0 stated that it was not an attempt to falsify a record; he was at-tempting to reconcile the log with the actual valve position at the time of the discharge.
The PE0 admitted that he had not followed the procedure but instead followed a lineup entered in the log for a previous discharge. As disciplinary action, the PE0 was suspended without pay for several days.
f J
g.
-
,.~----,-n---
--
n--
-.. -,,
. - -,
,, - - -
-,
,.-,v,.
,w,n,.n,.....--
-,,,,
,.,,,....+m,
&w e,
-,,.,,. - -,
.
_
_.
._
.
l
.
Subsequent sampling of the "B" FDST for radioactive isotope concentrations identified 1.717 E-04 uCi/ml of Cobalt-60.
The tank level had decreased from 1267 to 1037 gallons during the discharge, at a flow rate of 150 gpm, as specified on the liquid discharge permit (LOP). When converted, this yields a total of 1.49 E-04 Ci released to Long Island Sound.
The inspector verified that the amcunt was within the release limits of the LDP.
TS 3.8.C.1, "Liquid Effluent Concentration," requires that the concentration of radioactive mate-rial released from the site does not exceed the concentrations specified in
'
10 CFR 20, Appendix B, Table II, Column 2.
For Cobalt-60, this concentration is S E-05 uCi/ml. When corrected for the dilution provided by the four cir-culating water pumps and two service water pumps, which have a combined total flow of 420,000 gpm, the requirements of TS 3.8.C.1 are satisfied.
The in-spector verified the calculations and found no discrepancies.
Members of the Millstone Human Performance Evaluation System (HPES) reviewed the event and concluded that the existing procedure was adequate to prevent inadvertent discharges.
The inspector concurred with the HPES determination.
Millstone 1 Technical Specification (TS) 6.8.1.h requires that written proce-dures fcr radiological effluent monitoring be established, implemented, and maintained. This requirement was not satisfied because OP-3128 was not fully implemented.
In consideration of 10 CFR 2, Appendix C, no enforcement action will be taken at this time because the licensee identified the event, reported it in accordance with the regulations, and took appropriate and timely cor-rective action to prevent recurrence.
The inspector observed the Plant Operations Review Committee (PORC) meeting when the revisions to OP-3128 were reviewed.
The PORC'monbers approved the deletion of the stamp originally used for documenting the valve lineup in the radwaste log and approved the imple-mentation of a new form which more clearly directs the operators on performing the valve lineup.
In addition, a design change (suggested by the HPES report)
is to be implemented to provide a common selector switch for the discharge valves.
This will prevent both valves from being open or closed at the same time.
The inspector had no further questions and will verify the implementa-tion of the revised procedure and the control switch modification in a future inspection.
8.0 Overflew of Spent Fuel Pool The inspector reviewed PIR 1-88-07, which discusses the overflow of the Spent Fuel Pool onto the refueling floor and into reactor M! ding ventilation ducts.
The inspector obtained the followir.g information from personnel involved in and investigating the event.
On January 29, a PE0 placed the demineralizer for the Fuel Pool Cooling System (FPCS) into service.
OP-310, "Fuel Pool System," requires the operator to first vent the demineralizer. When the PE0 called the control room.for guid-ance on the venting ope ation, both the Shift Supervisor (SS) and the Super-vising Control Operator (SCO) were involved with other issues, so the Reactor Operator (RO) taking the call advised the PEO.
The R0, unaware that the de-mineralizer contained new resin, advised the PE0 not to vent the demineralizer
..
-
-.
.
.
.-
.
._
=
.
.
.
because he thought it had previously been 'in service.
His objective was to reduce radiation exposure to the PE0 because the associated valves are located in a high radiation area.
When the demineralizer was placed in service, the entrapped air in the de-mineralizer was forced out of the FPC spargers at the bottam of the fuel storage pool at such a rate that a wave was generated. The wave caused pool water to splash onto the northwest corner of the refueling floor and into the reactor building ventilation ducts located near the surface of the fuel pool.
Health Physics was contacted to conduct surveys and restrict access to the affected areas.
Radiation surveys showed general area contamination ranging from 450K dpm/100 cm2 to less than 1K dpm/100 cm2.
Dripping water-c.ontamin-ated the northwest corner stairs from the refueling floor to the ground floor
'
in the reactor building.
In spite of the contamination, the area radiation monitors on the refueling floor did not reach their alarm setpoints. Also, no personnel were contaminated.
Decontamination efforts began on January 29.
All general areas were cleaned by February 4.
j An interim change to OP-310 requires the operator to slowly open 1-FPC-19, the dsmineralizer outlet manual stop valve, in order to safeguard against fuel
_
pool overflow.
In follow-up discussions, the PE0 indicated that he was con-fused by the interim change, and that he never opened 1-FPC-19 as required.
No enforcement action is being taken at this time in spite of the failure to follow established procedures because the licensee has not yet determinad what corrective actions will be taken.
The inspector will discuss the actions with the licensee and will evaluate their effectiveness.
In addition to inadequate venting, further investigation revealed a second problem causing air entrapment. A leaking air-operated valve allowed con-tinuous introduction of air to the demineralizer. A trouble report has since been submitted and the service air supply has been isolated, Concern about OP-310 remains. The effective cete of the current issue is October 7, 1985. The interim change noted above (Change 2, dated November.
11,1985) is one of three separate changes att>ched to the procedure (Change 1 is dated October 29, 1985 and Change 3 is dated July 1, 1987); these changes increase the difficulty in using OP-310.
Procedure OP-260, "Biennial Review-of Operations Procedures," states that the review should be scheduled prior to March 15 of each even-numbered year.
It is possible that the procedure was reviewed but not reissued in the two years since the original issue date.
However, it would be prudent to reissue the procedure to incorporate the in-terim changes to avoid further confusion in implementing OP-310.
The inspec-tor has discussed this topic with the Operations Supervisor, and will follow the remaining licensee actions in resolving the concerns with 0F-310.
This item is unresolved (50-245/88-02-01) pending' ovaluation of the licensee corrective actions and further examination of the procedure review process.
.
l
.
.
9.0 High Radiation Area Doors Found Unloched The inspector reviewed the following PIRs involving three separate discoveries of unlocked High Radiation /.rea (HRA) doors.
~
.PIR No.
Date/ Time Door 1-87-106 Cecember 28, 4:30 p.m.
Solid to Liquid Radwaste Access Door 1-88-01 January 6, 8:04 a.m.
NE Turbine Deck Access Door 1-88-02 January 6, 11:15 a.m.
N Turbine Deck Access Door Inspection report 50-245/87-27 covered an additional unlocked HRA door (northwest corner ioom) which was discovered on September 23 (PIR 1-87-86).
At thst time, the !icensee added chains and padlocks to HRA doors to require conscious involvement from workers in accessing these areas. However, this method can only be used for cage-type doors; the three cases listed above -
involve solid doors where the use of padlocks is not practicable.
(No HRA cage-type doors have been discovered unlocked since the addition of padlocks.)
Also, as a result of safety concerns over the potential need to evacuate an HRA, the licensee now uses plastic breakaway tiewraps with the metal chains to maintain the need for conscious effort without delaying area egress in an emergency.
Another contrit>uting factor is that all HRA dcors have local alarms that sound when the door is open, but the limit switch sensitivity is such that it is possible to extinguish the alarm and have the door appear closed andflocked when it is unlocked.
Licensee efforts to locate more sensitive limit switches
-
have been unsuccessful.
The licensee also considered the replacement of cur-rent automatic closure devices on the hRAs with automatic opening devices.
This would result in a local HRA alarm until the individual took positive action to close the door.
The 6)or to the Unit 1 radwaste control room had the automatic closure hinges modified such that it automatically opened, but this proved to be ineffective.
TS 6.12.2 describes HRA requirements.
The whole body radiation doses used in determining the designation of HRAs at Millstone 1 is based on contact readings. More recent reaulatory guidance allows the use of readings at 18 inches. Millstone 3 TS specify that "areas accessible to personnel with~
radiation levels greater than 1000 mR/h at-45 cm (18 ir.ches)...shall be-pro-vided with locked doors.~"
The licensee plans to convert to the 18 inch cri-terion through TS amendments for Unit 1 and Unit 2 and to implement changes to Unit 1 HRA doors during the next refueling outage, scheduled for April 1989.
The licensee expects that reduciag the number of locked HRA doors will also reduce the number of HRA doors lef t unlocked.
No inadequacies were identified with the licensee's. response and corrective actions.
The inspector had no additional questions. During reactor building tours,-the inspector has identified no other instances of unlocked HRA doors.
Therefore, these three instances will be treated as licensee identified per 10 CFR, Appendix C, similar to the item described in Section 7.0 Further i
inspection will be routinely ccenducted.
!
_..
_...
-
__ _
_
-_
_
_-
. _..
__
.
.
..
.
10.0 Potential Inaccuracy of Containment High Range Radiation Monitors-On February 23, 19P7, General Atomic (now Sorrento) issued a 10 CFR Part 21 report concerning the coaxial signal cable for the in-containment high range-radiation monitors (HRRMs) installed pursuant to the TMI Action Plan.
The HRRMs in use'at Units 1, 2, and 3 and Haddam Neck were manufactured.by General Atomic.
High temperature testing identified that the HRRMs could indicate lower than actual radiation levels under accident conditions due to cable and penetration current leakage as a result of high temperatures.
The error for-Unit 1 is 41 R/hr, which appears to be consistent over the range.of the moni-tor.
Regulatory Guide 1.97 specifies that the HHRMs have a range of 1 R/hr to 1-E7 R/hr and be accurate within a factor of two.
The 41 R/hr error for Unit 1 appears to provide less than the specified accuracy up to about 100 R/hr.
While other containment and reactor parameters are available to detect fuel failure, and the drywell high range monitor readings provide no automatic control function, these readings are used.in the Emergency Plan Implementing Procedures in determining the Emergency Classification and in assigning Emer-
-
gency Action Levels.
The Unit 1 Barrier Failure Table uses the HRRMs in assessing the loss of the Reactor Coolant System barrier at readings as low as 5 R/hr without fuel clad barrier loss, and between 15 and 370 R/hr with fuel clad barrier loss. An error of 41 R/hr would be unacceptable for use in these cases.
NUSCO produced a Justification for Continued Operation (JCO) while the oper-ability of the HRRMs was still under evaluation.
The JC0 relies on the es-
-
tablishment of alternate monitoring methods and the completion of associated procedures and operator training.
Presently, no alternate method for moni-toring in-containment radiation levels is available at Unit 1.
The licensee has initiated a design change which would provide alternate radiation monitor-ing capability by relocating an area radiation monitor to the containment penetration for Control Rod Drive (CRD) removal.
The present Unit 1 TSs do not address the in-containment high range radiation monitors: Unit I has in process but has not yet submitted a proposed TS
amendment to address the monitors.
The resident will continue to follow the licensee's analysis and~ action on the problem. Outstanding concerns include the deviation from Reg Guide 1.97, the establishment and accuracy of the alternate monitoring method, and the absence of TSs addressing the operability of the HRRMs.
11.0 Followup of IE Bulletin 87-02, Fastener Testing To Determine Conformance with Applicable Material Specifications - TI 2500/26 The NRC issued IE Bulletin 87-02, Fastener Testing to Determine Conformance with Applicable Material Specification, dated November:6,1987, to rcquest-licensees to review their receipt inspection requirements for fasteners, and to determine through independent testing whether fasteners in stcak meet re-i j
.
__
-
-
- - - - - -
_ _ _ _. _
.
-
-
J
D
,
quired mechanical and chemical specifications.
Item 2 of the bulletin re-
~
quired that the licensee draw a sample of fasteners from safety-related and non-safety-related stores, and to obtain the samples with the participation of the resident inspector.
NRC Inspection Reports 50-245/87-33 arc 50-336/
87-29 document resident inspector review of licensee actions to withdraw fastener samples from station stores and to submit them to an offsite labora-tory for analysis.
Inspector review of licensee actions to withdraw duplicate sets of safety grade and non-safety fasteners and nuts for all three Millstone units, for a total Millstone station sample size of 80 items, verified con-formance with Bulletin. items 2 and 3.
The licensee submitted his response to IE Bulletin 87-02, which was due on January 10, 1988, by letter dated January 12, 1988.
The inspector r,eviewed the response and found it to be responsive to the information requested by Bulletin items 1, 4, 5 and 6 concerning receipt inspection practices, controls for storage and use of fasteners in safety-related and non-safety-related applications, the results of chemical and physical testing performed in ac-cordance with the specifications grade and class, and the need for additional actions based on the test results.
The licensee.'s response addressed the test results for 160 fastener specimens, which included samples drawn from stores
at tFe Connecticut Yankee facility.
The inspector reviewed Purchase Order 864178 dated December 9, 1987 and sub-mitted to an independent laboratory, J. Dirats and Co., to request testing in accordance with the bulletin.
The inspector verified that the purchase order instructed the laboratory to complete testing per Bulletin Item 4 in accordance with the specification grade and class applicable to each fastener,
,
including the verification of ultimate strength, hardness and chemical pro-perties as required by the appropriate specifications.
The inspector also
'
reviewed the sample data sheets included in the licensee's January 12 response, along with the test results provided by the laboratory.
The review of the purchase order and test results verified that the description of the tested material matched the sample descriptions recorded by the NRC inspector as each item was withdrawn from stores and tagged on December 7, 1987.
The licensee's test program identified a total of 7 discrepancies out of the 160 samples tested.
Of the 7, only two were judged to be nonconforcing and nonconformance reports were issued to disposition the items.
The licensee concluded that both of the nonconforming items were minor specification de-viations and were judged to not be safety significant.
The nonconformances involved Millstone samples drawn from QA and non-QA stores: MP-21A/B, ASTM A307 Gr B 5/8 inch bolt; and MP-3A/B, ASTM A19J-78A Gr B 3/8 inch bolt.
In both samples, the test results showed that the material deviated from the ASTM specification.
The licensee determined that all samples met their specified functional re-quirements and could be used "as is".
Specifically, the chemical composition for one bolt showed chromium outside the specified range of 18.00 to 20.00 at 17.29, and nickel outside the specified range of 8.00 to 10.5 at 10.8.
The out-of-specification chemistry was acceptable because the slight vari-i
._
_
.
-
.-
.
- _
-.
.__
__-
.
.
ations were not enough to significantly degrade the alloy's strength, duct-ility or corrosion resistance.
The measured Rockwell B hardness for a bolt was in excess of the specified maximum of 95 at 95.5.
The out-of-specifica-tion hardness was acceptable because the reading indicated above average strength with no significant effect on ductility or corrosion resistance, and because the reading was within the ASTM E10 measurement accuracy for the pro-perty of +/- 2%.
Based on the above results, the licensee concluded that no additional actions relative to the fasteners in stock were warranted.
No further actions are planned. The inspector identified no inadequacies in the licensee's conclu-sions or plans.
The inspector aad no further question on this item at the present time.
This
'
item will be reviewed further on a subsequent routine inspection following further NRC Staff review of the licensee's response to Bulletin 87-02 and evaluation of the reported test results. The NRC has an action item to com-pile and evaluate the reported results.
Further review per TI item 05.01 and 05.02 regarding receipt inspection program proce.dures and implementation will be conducted on a subsequent routine inspection.
12.0 Oual Role Shif t Supervisor - Shif t Technical Advisor (SS/STA) Issue On January 21, the Office of Nuclear Reactor Regulation (NRR) informed the inspectors of concerns involving the use of dual-role Shift Supervisor / Shift Technical Advisor (SS/STA) personnel.
The concern included the apparent non-compliance with administrative Technical Specifications (TS).
Unit 1 TS Table 6.2-1, "Minimum Shift Crew Composition" lists a total of seven individuals for Modes 1, 2, and 3, but because the SS serves in a dual capacity, these seven positions are filled by six individuals.
The current Unit 1 TS (ori-ginal issue, license number DPR-21, dated October 31,1986) do not explicitly identify the dual role or otherwise contain a reference acknowledging that the STA role may be filled by the on-shift SS.
The inspectors discussed this issue in meetings with station management on January 21.
The issue was subsequently expanded to include Unit 2 and the Haddam Neck Plant.
In response to a licensee request to amend t!.e TS for both Units 1 and 2 by letter dated October 4,1983, NRR processed License Amendment 95/92, by two letters dated February 16, 1981. One would have issued a version of TS Table 6.2-1 which recognized the SS/STA.
The text of the associated transmittal letter stated that, since the Commission policy statement on engineering ex-
pertise on shift was still in draft form, the proposed TS for the SS/STA was deemed unacceptable at the time.
This inconsistency was recognized and a second version of TS Table 6.2-1 was issued without a reference to the SS/STA.
The transmittal letter was also revised to state that the issue of an on-shift SR0 filling the STA position would be addressed by a separate licensing action.
. _. -
-. _,
__
_ __ -
__ - ____
. _
_ _
....
- _ _
.
,
On January 1,1984, Unit I was the first Northeast Utilities (NU) plant to switch from the dedicated STA to the SS/STA and proceeded with this program without a change to the TS.
In an unrelated licensing action, the NRC staff issued Amendment 104/102 for MP-1/2 on August 6,1985 to approve an organizational change as found on page 6-3 of the TS.
By mistake, a copy of page 6-4 which contained the version of TS Table 6.2-1 which addressed the dual-role SS/STA, was also included in the transmittal.
The licensee recognized the mistake and did not distribute this version of Table 6.2-1.
On January 22, the Manager of General Facility Licensing (NNECO) met with the inspectors to discuss the issue's history since 1979 and to explain.the lic-ensee's basis for maintaining that no TS noncompliance existed.
On January 27, 1988, the NRC and the licensee met in Bethesda, Maryland to discuss the resolution of this issue.
The NRC subsequently asked the licensee to develop a plan to achieve compliance with the applicable TS.
The staff stated that there was no compelling safety reaso.n to take immediate action to address compliance with the present TS prior to Commission review and approval of a license amendment.
NU has prepared a proposed TS amendment which would allow the current SS/ STAS to serve in the dual role without ed-ditions to current shift staffing.
The proposed TS amendment also states that future STAS will meet the Commission policy statement.
(This proposed amend-l ment dated February 17, 1988 was sent to the HRC subsequent to the inspection
,
period.)
The inspector will continue to follow the progress of the newly proposed l
amendment.
13. On-site Plant Operations Review Committee (PORC)
The inspector attended Plant Operations Review Committee (PORC) meetings 1-88-01 through 1-88-11.
Technical Specification 6.5 requirements for committee quorum were met. The meeting agenda included reviews of Plant Design Change Records (PDCRs), procedure revisions, interim changes to procedures, and Licensee Event Reports (LERs).
The inspector noted active participation by each member and thorough attention to the importance of safety for the matters under review. No inadequacies were identified.
14. M_anagement Meetings At periodic intervals during this inspection, meetings were held with senior plant management to discuss the findings.
No proprietary information was identified as being in the inspection coverage.
No written material was pre-vided to the licensee by the inspector.
1