IR 05000213/1986006
| ML20210V438 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 05/30/1986 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20210V401 | List: |
| References | |
| 50-213-86-06, 50-213-86-6, IEB-81-03, IEB-81-21, IEB-81-3, IEB-83-46, NUDOCS 8606100631 | |
| Download: ML20210V438 (11) | |
Text
.
!
'
)
I U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report N /86-06 DCS NO /85-08-02 50-213/85-03-12 Docket N /85-03-17 50-213/85-03-19 License N /86-05-08 50-213/86-05-16 Licensee: Connecticut Yankee Atomic Power Company 50-213/36-05-18 P. O. Box 270 Hartford, CT 06101 Facility: Haddam Neck Plant, Haddam, Connecticut Inspection at: Haddam Neck Plant Inspection conducted: April 15 - May 27, 1986 Inspectors: Stephen Pindale, R ' dent Inspector in Training Paul D. S tlan , Senior Resident Inspector Approved by #: /- '
'/N 8 E. C. McCab @ ef, Reactor Projects Section 3B Date Summary:
Areas Inspected: This was a routine resident inspection (138 hrs) of outage acti-vities, and of plant start-up and operation following an extended refueling outag The following areas were inspected: Plant operations, radiation protection, physi-cal security, fire protection, maintenance, surveillance, open items, licensee events, and IE Bulletins /Information Notice Results: Inspector review of plant activities identified satisfactory performance in all areas. Five NRC open items were closed including licensee corrective ac-tions regarding a previously reported small break loss of coolant accident analysis error. The need for further licensee action and NRC followup was noted regarding the continued operability of certain containment isolation check valves in the component cooling water system.
l l
l l
h G
. _ . . _ . . __
.
. .
i
i j ' TABLE OF CON 1ENls i
P_ ate
- Summary of Facility Activities....................................... 1 Review of Plant Operations........................................... 1 Observation of Maintenance and Surveillance Testing.................. 3
! Followup on Previous Inspection Findings............................. 5 Followup on IE Bulletins, IE Circulars and Information Notices....... 7 Followup on Events Occurring During the Inspection................... 7 7. -Review of Periodic and Special Reports............................... 9 Unresolved Items..................................................... 9
.
., Exit Interview....................................................... 9 j
l
,
i f
I J
!
$
i
.
U f
- . ,. . . , _ , - - ~ , ,_ . - - , - - , . - , . , _ . - - - - - - _ - --
.
DETAILS Summary of Facility Activities At the beginning of the inspection period, the plant was nearing the end of an extended refueling outag The plant was in cold shutdown awaiting com-pletion of various mdifications and of equipment qualification repair After completion of the reactor coolant system (RCS) operational pressure check on April 27, 1986, plant heatup began. Containment isolation valve leakage identified on April 29 resulted in a return to cold shutdown for further testing and repairs. Plant heatup was reinitiated on May 2. A reactor startup was conducted on May 6. Low power physics testing was ac-complished from May 6 through May 8. On May 7, the licensee discovered that one steam generator (SG) U-tube in SG #2 had not been plugged as required by Technical Specifications (TS). The tube had a 55% through-wall eddy current indication characterized as a pit. The licensee isolated the #2 SG and Reac-tor Coolant System (RCS) loop and requested TS relief from NRC Licensing to allow operation for fuel cycle 14 with the defective tube unplugged. NRC Licensing granted this relief on May 8. During main turbine generator roll-off and balancing activities on May 8, the plant tripped from 10% power. This was due to operator inattention to the automatic reset of a closed main steam isolation valve trip signal which existed because of the isolated #2 S While the plant was shutdown, the #2 RCS loop and SG were unisolated and several steam leaks were repaired. The reactor was restarted on May 8 and the unit was phased to the grid on May 10. Power was held at 25% until May 16 while attempting to reduce SG secondary chemistry resultc to acceptable level During this period, a small primary to secondary leak was detected in the #2 S The leak was roughly quantified at 20 gallons per day (gpd),
well under the 150 gpd TS limit. By the end of the inspection period, this leak had been more accurately calculated to be about 80 gpd. The plant power ascension to the 80% flux mapping plateau was completed on May 18, 198 Spurious nuclear instrument power supply voltage spikes resulted in several negative flux rate rod stops and turbine cutbacks during power ascensio The plant reached full power on May 22, 1986, following the 140 day refueling outag . Review of Plant Operations Plant operation was observed during regular tours of the followin Control Room --
Security Building
--
Primary Auxiliary Building --
Fence Line (Protected Area)
--
Vital Switchgear Room --
Yard Areas
--
Diesel Generator Rooms --
Turbine Building
--
Control Point --
Intake Structure and Pump
--
Containment Building Building
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ . _ _ _ _ _ _ - _ _ _ _ _ _ - - - - _ _ - _ _ . ___- . _ - - -
.
Control room instruments were observed for correlation between channels and for conformance with Technical Specification requirements. The inspector observed various alarm conditions which had been received and acknowledge Operator awareness and response to these were reviewed. Control oom and shift manning were compared to regulatory requirements. Posting ind control of radiation and high radiation areas was inspected. Compliance with Radi-ation Work Permits and use of appropriate personnel monitoring devices were checked. Plant housekeeping controls were observed, including control and storage of flammable material and other potential safety hazards. The in-spector also examined the condition of various fire protection system During plant tours, logs and records were reviewed to determine if entries were properly made and communicated equipment status / deficiencie These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Reports. The inspector observed selected aspects of plant security including access control, physical barriers, and personnel monitoring. Except as noted below, no discrepancies were identifie .1 During this refueling outage, the licensee implemented several plant changes to upgrade licensee compliance with the NRC fire protection requirements of 10 CFR 50.48. Several issues for which the licensee has docketed exemption requests remain ope The licensee committed to implement a continuous fire watch to monitor those areas not in full compliance with 10 CFR 50, Appendix R. NRC Licensing determined that, with the implementation of these compensatory measures, plant startup and operation was acceptable pending final disposition of the docketed exemption requests. The inspector verified that the continuous fire watch was satisfactorily implemented prior to reactor criticality on May 6, 1986. The adequacy of implementation of fire protection require-ments will be evaluated during a subsequent NRC team inspectio .2 The licensee also was required to implement modifications to achieve compliance with the equipment environmental qualification rule, 10 CFR 50.49. During the outage, the licensee identified numerous equipment qualification (EQ) deficiencies. The licensee committed to notify NRC Region I of the scope of these deficiencies and to correct each item prior to plant startu The licensee and members of the NRC Region I staff discussed the EQ equipment discrepancies documented in a letter to NRC Licensing dated April 24, 1986. As a result of this discussion, the licensee submitted written certification on April 25, 1986 that the plant would be in compliance with 10 CFR 50.49 prior to plant startup from this outage. NRC staff acknowledged the licensee submittals and determined that NRC evaluation of the compliance and reportability aspects of these EQ component discrepancies would be performed during a special NRC team inspection of the EQ program implementatio .3 On May 7, 1986, the licenree identified that a steam generator (SG) U-tube in SG #2, with a 55% through-wall eddy current indic nion, had not been plugged. SG #2 had been returned to service, and the reactor was critical with low power physics testing in progress. Technical Speci-
_ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ .
-
l
'
fication (TS) 4.10 requires all SG tubes with flaws greater than 50%
through-wall to be plugged. The defective tube had not been plugged because of a typographical error in transposing eddy current test data to the tube plugging list. Tube 37-33 was plugged instead of tube 37-7 Upon confirmation that tube 37-73 had a 55% defect and that this tube was not plugged, the licensee isolated the #2 reactor coolant system loop and the #2 SG in order to comply with TS 4.10. The licensee determined that the defect was a pitting type anomaly and had not shown significant ;
growth over the past fuel cycles. Licensee calculations based on con- ;
servative, predicted defect growth and destructive testing results of actual SG tubes indicated acceptable U-tube strength and integrity for one fuel cycle. Therefore, the licensee requested relief from TS 4.10
.
!
to allow operation with tube 37-73 unplugged during fuel cycle 14. This relief was granted by NRC Licensing on May 8, 1986. During this in-spection period, the licensee identified a small primary-to-secondary leak in SG #2. The leak was initially calculated to be about 20 gallons per day (gpd) with a TS limit of 150 gpd per SG. The licensee continued to monitor the leak rate during plant power escalation. At the end of the inspection period, the calculated leak rate had risen to about 80 gpd. The inspector reviewed licensee calculations and SG chemistry data to verify this conditio Primary-to-secondary leakage will be re-verified during routine resident inspection .4 On May 13, 1986, with the plant in mode 1, the inspector observed that the wide range noble gas (WRNG) stack monitor had been valved out and tagged out of service since May 10, 1986, to prevent instrument satura-tion while using the plant stack as a flow path for steam generator blowdow Technical Specification (TS) 3.238 requires the WRNG stack monitor to be operable in modes 1-4. The action associated with the Limiting Condition for Operation requires the licensee to return the channel to operable status within 7 days or submit a special report to NRC within the following 10 days. During the time that the WRNG stack monitor was out of service, the narrow range WR Stack monitor RMS-14 was in service. This monitor does not meet the extended range require-ments of TS 3.23. Following discussions with the on-shift operating staff, it was noted that the operators did not realize that they were in an Action Statement for the TS limiting condition for operation (LCO). Since this was noted only 3 days into the action statement,
,
the licensee had sufficient time (4 more days) before the WRNG Stack
! monitor was required to be placed back in service. Subsequently, a
! tag was placed on the control room stack monitor recorder which docu-mented the date when the LC0 was entered. The WRNG monitor was returned to service on May 16, 1986. The inspector highlighted this operator misunderstanding of the TS LC0 requirement to licensee management.
!
No further discrepancies were identifie . Observation of Maintenance and Surveillance Testing i
The inspector observed various maintenance and problem investigation acti-vities for compliance with requirements and applicable codes and standards, QA/QC involvement, safety tags, equipment alignment and use of jumpers, per-i __
.
'
sonnel qualifications, radiological controls, fire protection, retest, and reportability. Also, the inspector witnessed selected surveillance tests to determine whether properly approved procedures were in use, test instru-mentation was properly calibrated and used, technical specifications were satisfied, testing was performed by qualified personnel, procedure details were adequate, and test results satisfied acceptance criteria or were pro-perly dispositioned. The following activities were reviewed:
-- Rod Control System Maintenance and Testing (Work Orders 84-09403, 84-09409 and 85-03543)
-- Core Deluge Check Valve Leak Tests (Surveillance 5.7-111)
-- Component Cooling System Isolation Valve Leak Tests (Surveillance 5.7-51 and 5.7-93)
- -- Core 14 Startup Physics Testing (Procedure 1.7-33)
,
--
Control Rod Reactivity Worth Measurements (Surveillance 5.3-6)
3.1 Recurrent component cooling water (CCW) system containmer.t isolation check valve failures have been ca.used by silt build-up in the CCW syste During this refueling outage, the licensee installed a slip-stream filter in the CCW system to correct this problem. At the end of the outage, however, realignment of CCW and service water systems resulted in an accumulation of silt in the CCW system. On April 29, 1986, the licensee retested the check valve in the CCW supply to the reactor neutron shield tank heat exchanger (CC-CV-885). The measured leak rate exceeded the combined leak rate limit allowed for all con-tainment isolation valves. CC-CV-885 was disassembled, cleaned and retested satisfactorily. CC-CV-721 in the CCW supply to the reactor coolant pump seal thermal barriers also has a history of silt buildu In order to leak test this valve, the licensee cooled down the RCS to cold shutdown condit1ons on April 29, 1986. CC-CV-721 also failed the initial leak test, was cleaned, and subsequently passed several retests. The inspector discussed the continuing problems with CCW check valve leakage with licensee management on May 13, 1986. This discussion covered: (1) licensee followup ar:tions to assure that CCW check valves remain operable throughout the operating cycle; and (2)
the reportability of the leak test failures on April 29, 1986. With regard to the continued operability of CC-CV-721 and 885, the licensee stated that an action plan was being formulated for submittal with the routine containment leak rate report. The inspector stated that l corrective actions would also need to be forwarded to NRC as part of j the 30-day Licensee Event Report (LER) required by 10 CFR 50.73 (a)(2)(V).
, Since the licensee also needs to update LER 86-06 submitted for leak test failures of these and other containment isolation valves earlier in 1986, the licensee planned to consolidate these reports within the 30-day reporting criteria of the last even With regard to the prompt reportability (10 CFR 50.72) of these recent check valve failures, l
l
,
l
-
l
'
the licensee had concluded that since the penetrations were isolated by other valves within the Technical Specification action statement grace period, no report was necessar However, the inspector pointed out that other 50.72 criteria similar to the 10 CFR 50.73 (a)(2)(V)
criteria were applicable to the April 29 events, because these failed containment penetration boundaries in non-seismic / missile protected systems could have compromised containment integrity during an acciden The licensee is reviewing the reportability evaluation procedures and training to assure that each event is evaluated against all the 50.72 criteria. NRC review of these valve failures remains unresolved pending licensee determination and implementation of final corrective action (UNR 213/86-06-01)
4. Followup on Previous Inspection Findings During the course of the inspection, five NRC open items were reviewe The inspector found licensee actions with regard to these areas to be suf-ficient to close the items. Details follow:
4.1 (Closed) Followup Item (213/84-28-03) The licensee was to implement improved leak test procedures for safety injection system check valve Procedure 5.7-111, Core Deluge Check Valves Leak Test, Revision 1, was approved on February 26, 198 This procedure was implemented during the plant startup on April 27, 1986. The inspector reviewed the completed test documentation to verify the adequacy of the procedure and the test results. Both check valves had zero leak rate during the test. The inspector had no further questions in this are .2 (Closed) Followup Item (213/85-11-01) The licensee was to perform follow on testing to identify the cause of potentially recurrent dropped control rod events involving rods 30 and 33. During the 1986 refueling outage, the licensee performed detailed inspection, testing and preven-tive maintenance on the entire rod control system. No abnormal condi-tions were identified. In addition, an inspection by vendor represen-tatives failed to identify a generic cause for the multiple dropped rod events. During plant startup, the rod control system was fully retested. No dropped rod failures occurred. The inspector verified the scope and results of rod cont ol system maintenance and testing completed under procedures 9.5-2, Rod Control System; 9.5-29, Checking Timing on Slave Cyclers; and 9.5-43, Hot Rod Checks. No further de-ficiencies were identified. This item is closed. Future rod control l system anomalies will be followed under the routine inspection program.
i l
4.3 (Closed) Unresolved Item (213/85-21-01) The licensee was to resolve with NRC Licensing the appropriateness of administrative controls on 3-loop operation during Cycle 13 coastdown operations. The licensee reanalyzed plant operation with 3-loops during Cycle 13 coastdown and l showed adequate shutdown margin requirements throughout the projected coastdown period. Therefore, the administrative restriction on 3-loop i
_ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
operation was remove During NRC Licensing review of the Cycle 14 reload, it was identified that the basis of Technical Specification (TS) 3.18, Power Distribution Monitoring and Control, was the reload analyses which are normally carried out only to end of-core lif Therefore, there was no basis for TS figures 3.18-la and 2a to reference applicability to coastdown operation in any fuel cycle other than Cycle 12 for which coastdown analyses had been reviewed by NRC in License Amendment 5 Amendment 74 removed TS reference to coastdown operations and documented the fact that subsequent coastdown operations will require prior NRC Licensing review. The inspector had no further questions in this are .4 (Closed) Followup Item (213/85-21-16) The licensee was to report to NRC Licensing the details of an identified failure mode for vital motor control center (MCC) No. 5. On November 8, 1985, the licensee submitted a report to NRC Licensing detailing this condition and the planned corrective actions. In addition, this event was the subject of Licensee Event Report 85-29 submitted on December 2, 198 The reliability of power to MCC 5 is being evaluated by NRC Licensin This item is close .5 (Closed) Unresolved Item (213/86-03-01) During the development of the Haddam Neck Probabilistic Safety Study, the licensee identified a scenario for certain small-break loss of coolant accidents in which the reactor core could uncover. A small break at the loop 2 cold leg or the attached charging system piping compromises the ability to recirculate emergency cooling water which normally flows through the charging system. The licensee proposed an alternate recirculation core cooling flow path through the high pressure (1500 psi) safety injection pumps to two reactor coolant loops other than loop 2. The licensee's proposal was reviewed and approved by NRC Licensing on
'
April 28, 1986. This approval included a one fuel cycle exemption from the single failure criterion requirements for emergency core cooling systems, because a failure of either of two motor-operated valves in the proposed flow path could prevent fulfillment of core cooling. Manual operation of these two valves is available as a back-up means of operation, and the licensee committed to a monthly surveil-lance test of the valves to verify continued component operabilit In addition, the licensee committed to make two changes to the Loss l of Coolant Accident emergency procedure and to conduct operator training
'
with regard to the procedural and emergency response changes necessitated by this issue. The inspector verified that revision 30 to procedure 3.1-4, Loss of Coolant Accident, incorporated both the direction to manually operate either of the MOVs if it fails to function properly, and the caution to continue injecting the remaining water from the reactor water storage tank should the core cooling recirculation flow path be unavailable. The inspector also verified that operator training on revision 29 to procedure 3.1-4 had been conducted in April 1986, on each operating shift. Upon approval of revision 30 to the procedure, each operator was required to review the procedure (with the changes I
.
'
highlighted) prior to assuming the watch. MOVs 24 and 874 were satis-factorily tested on April 28, 1986 in accordance with surveillance procedure 5.1-4, Hot Operational Tests. The inspector identified no discrepancies in the implementation of the above licensee commitment The follow-on M0V testing during this cycle and the implementation of final core cooling system corrective actions for subsequent fuel cycles will be reviewed during routine NRC inspection . Followup on IE Bulletins (IE8s), IE Circulars (IECs), and Information Notices (ins)
5.1 IEB 81-03 and IEN 81-21 and 83-46 discuss instances of safety-related heat exchanacr fouling due to clams or mussels. In order to evaluate the effectiveness of licensee response to these and other industry notifications of problems related to bio-fouling of open-cooling water systems, the inspector reviewed the licensee actions responding to these notifications and evaluated the effectiveness of these action Licensee response to IEB 81-03 concluded that bio-fouling at Haddam Neck had not been a problem to date and that existing system monitoring, chlorination and inspection would detect any change in the marine environment which may cause future problems. This response was previously reviewed in NRC Region I Inspection Report 50-213/84-03. Since that report, two refueling interval inspections of service water supplied heat exchangers have identified no significant fouling problems, and there remains no evidence of clam / mussel migration to this are Between refueling outages, service water performance is monitored continuously by the containment air recirculation cooler flow indicators and closed cooling water system performance; and monthly during emergency diesel generator testin The inspector had no further questions in this area.
,
6. Followup on Events Occurring During the Inspection 6.1 Licensee Event Reports (LERs)
The following LERs were reviewed for clarity, accuracy of the description of cause, and adequacy of corrective action. The inspector determined l whether further information was required and whether there were generic implications. The inspector also verified that the reporting require-ments of 10 CFR 50.73 and Station Administrative and Operating Procedures i had been met, that appropriate corrective action had been taken, and l that the continued operation of the facility was conducted within l Technical Specification Limit Post Loss of Coolant Accident Release Paths - Revision 2 (see details in Paragraph 6.2 of this report)
86-15 -- Control Rod Cladding Wear and Cracking
__
.
-
86-16 -- Inoperable C02 Fire Protection System 86-17 -- Degraded Fire Barrier Seals 6.2 In August 1985, the licensee reported the identification of unanalyzed ,
post-LOCA release paths from the reactor coolant pump seal return !
line. Subsequently, the licensee amended LER 85-17 to add another unanalyzed release path through the reactor coolant system drain lines if the alternate letdown flow path using the drain lines is in servic During the 1986 refueling outage, the licensee re-evaluated the capabil-ity to isolate these rele u e paths using motor-operated valves (M0Vs)
inside containment. Revision 2 to LER 85-17 reported that only 4 of the 5 MOVs were fully qualified for the post-LOCA containment environ-ment. The drain header isolation valve (DH-MOV-310) used to isolate the alternate letdown path could not be qualified. Therefore, the licensee committed to disable this valve in the closed position during cycle 14, foregoing the alternate letdown capabilities. Environmental qualification of DH-M0V-310 will be completed during a subsequent outage. The inspector verified licensee implementation of administrative controls in procedure 2.1-1, Plant Startup, to verify DH-M0V-310 is disabled prior to reactor criticality, and procedure 2.6-2, Volume Control: Normal and Alternate, to prohibit the alternate letdown mode when the reactor is operating. No discrepancies were identifie The inspector had no further questions in this are .3 On May 8, 1986, the plant tripped during main turbine testing following the refueling outage. One of four reactor coolant loops (#2) was isolated during the plant startup because of steam generator (SG)
operability considerations detailed in paragraph 2.2 of this repor As a result, the #2 Main Steam Trip Valve (MSTV) was closed, causing a trip signal to the reactor protection system (RPS) for main steam line break protection. This trip signal is automatically bypassed when plant load is less than 10% power. Operator inattention to the automatic reset of the 10% permissive (P-7) caused the trip when turbine power (measured by turbine first stage pressure) was raised above 10%
during turbine balancing and P-7 was cleared. All systems functioned properly following the trip, and the reactor was subsequently restarted on May 9, 198 During the shutdown, reactor coolant loop #2 was recovered after receiving NRC relief from SG tube plugging limits (see Detail 2.2 of this report). Licensee corrective actions for the event will be reviewed when the licensee event report (LER) is submitte .4 During the plant power escalation from 25% to 80% power on May 16-18 1986, several spurious rod motion blocks / turbine runbacks occurred due to power supply spikes on two of the power range nuclear instru-l ment drawer These spikes triggered the negative rate actuation of the dropped control protection (rod block / runback) of the reactor protection system. After verification of the spurious cause of the trip, the operators reset the actuation signal and stabilized plant powe Power supply anomalies in the Nuclear Instrumentation drawers
!
._
.
- 1
is a continuing problem. The licensee is closely monitoring nuclear instrument performance by weekly verification of system operabilit No further problems have been identified. In addition, the licensee is evaluating upgrading the equipment. NRC evaluation of this matter will be conducted upon receipt of the LER for this even .5 On April 30, 1986, the licensee discovered that the d.c. power supplies (battery chargers A & B) to two d.c. distribution panels (DCP-1C &
DCP-1D) were cross-wired. The cross-wiring took place during a 1977 modification of the d.c. distribution system. Power was available to both panel Post accident sampling system (PASS) and reactor and pressurizer head vent solenoid valves are affected by this circui There was no affect on operability because the two d.c. distribution panels are redundant to one another. The panels were promptly wired correctly via plant design change request (PDCR) 836 and work order no. SS-00028 on May 7, 198 Modification control changes instituted in 1984 and 1985 provide a safeguard against recurrence of problems such as this one. PDCR paperwork closecut remains to be completed, at which time the drawing revisions and jumper /tagout controls will also be complete The inspector verified correct wiring and testing of the panel No further discrepancies were identifie . Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review verified that the reported information was valid and included the NRC required data; that test results and supporting information were consistent with design predictions and performance specifications; and that planned corrective actions were adequate for resolution of the problem. The inspector also ascertained whether any reported information should be classified as an abnormal occurrence. The following periodic reports were reviewed:
-- Monthly Operating Report 86-04 - April 1 - April 30, 1986
-- Annual Report - Haddam Neck Plant - January 1 - December 31, 1985 8. Unresolved Items
.
l Unresolved items are matters about which more information is required in order to determine whether they are acceptable items or violations. An unresolved item identified during this inspection in discussed in Paragraph . Exit Interview l
l During this inspection, meetings were held with plant management to discuss the finding No proprietary information related to this inspection was l identified.
_ _ _ . _ _ _ _