IR 05000213/1987027

Safety Insp Rept 50-213/87-27 on 871009-1207.No Violations Noted.Major Areas Inspected:Plant Operations,Radiation Protection,Fire Protection,Security Maint,Surveillance Testing & Open Items from Previous Insp

ML20149D350
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	12/28/1987
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20149D342	List: IR 05000213/1987027 ML20149D337
References
50-213-87-27, NUDOCS 8801120288
Download: ML20149D350 (17)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /87-27 Docket N License N DPR-61 Licensee: Connecticut Yankee Atomic Power Company P. O. Box'270 Hartford, CT 06101 -

Facility: Haddam Neck Plant, Haddam Neck, Connecticut Inspection at: Haddam Neck Plant Inspection dates: October 9, 1987 through December 7, 1987 Inspectors: Andra A. Asars, Resident Inspector John T. Shedlosky, Senior Resident Inspector Approved by: m

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E. C. McCabe, Chief, Reactor Projects 1B Date Summary: Inspection 50-213/87-27 (10/9/87 - 12/7/87)

Areas Inspected: This was a routine safety inspection (215 hours0.00249 days <br />0.0597 hours <br />3.554894e-4 weeks <br />8.18075e-5 months <br />) by the resident inspectors. Areas reviewed included plant operations, radiation protection, fire protection, security, maintenance, surveillance testing, events occurring during this inspection period, and open items from previous inspection Results: No violations were identified. Multiple instances of greasing of safety-related motor-operated valve motors were identified (Detail 6.3). Eight open items from previous inspections were close L l

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PDR ADOCK 05000213 l 0 DCD i

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TABLE OF CONTENTS page Summary of Facility Activities . . . . . . . . . . . . . . . . . . 1 Review of Plant Operations . . . . . . . ....... ..... 1 Plant Operations Review Committee .......... ...... 4 Observation of Maintenance and Surveill:nce Activities . . . . . . . 4 4.1 Reactor Core Support Barrel Thermal Shield Repair . ...... 4 Followup on Previous Inspection Findings . .... ........ 7 Local Leak Rate Testing By Unauthorized Test Method . . . . . . 7 5.2 Breech of Containment Integrity During Surveillance Testing . . 7 5.3 Failure to Serialize Changes to PDCRs . . .......... 7 5.4 Low Temperauvre Overpressure Protection . . . . . . . . . ... 8 5.5 Unanalyzed Post Accident Release Path . ....... ... 8 5.6 Switchgear Room Fire Door Replacement . ........... S 5.7 Change in Operating Margin Associated with Changes to Tave . .9 5.8 Engineering Fxpertise On Shif t . ............... 9 Followup on Events Occurring During the Inspection . . . . . . . . . 9 Licensee Event Reports and Safeguards Event Reports . . . ... 9 6.2 Accidental Radiation Exposure . . . . . . . . . . . . . . . . 10 6.3 Incorrect Lubrication of Motor Operated Valve Operators . . . 11 Review of Periodic and Special Reports . . . . . . . . . . . . . . 13 Steam Generator T ';e Rupture Susceptibility . ........... 13 Licensee Event Reporting to the NRC Incident Response Center . . . 16 10. Exit Interview . - . . . . . . . . . . . . . . . . . . . . 17

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DETAILS 1. Summary of Facility Activities During this inspection period, the reactor was shutdown, with all fuel removed, for an extended refueling and maintenance outage which began on July 18, 1987. Major work activities centered around repairs to the vessel thermal shield attachment device . Review of Plant Operations The inspector observed plant operation during regular tours of the following plant areas:

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Control Room --

Security Building

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Primary Auxiliary Building --

Fence Line (Protected Area)

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Vital Switchgear Room --

Yard Areas

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Diesel Generator Rooms --

Turbine Building

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Control Point --

Intake Structure and Pump 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 acknowledged. Operator awareness and response to these conditions were reviewed. Control room and shif t manning were compared to regulatory requirements. Pesting and control of radiation and high radiation areas was inspected Compliance with Radiation Work Permits and use of appropriate cersonnel monitoring devices were checked. Plant housekeeping controls were observed, including control and storage of flammable material and other potential safety hazards. The inspector also examined the condition of various fire protection systems. During plant tours, logs and records were reviewed to determine if entries were properly mada and communicated equipment status / deficiencie These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Reports. The inspecter observed selected aspects of plant security including access control, physical aarriers, and personnel monitoring. In addition to normal working hours, the review of plant operations was conducted during the 1 folicwing midnight shifts, weekends, and holidays: I October 12, 1987 6:30 AM to 1:00 PM November 1, 1987 4:00 PM to 10:00 PM December 6, 1987 5:45 PM to 7:45 PM No unacceptable conditions were identified. Operators were alert aid I displayed no signs of inattention to duty or fatigu :

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-4-3. Plant Operations Review Committee (PORC)

The inspector attended several Plant Operations Review Committee (PORC)

meeting Technical specification 6.5 requirements for required member attendance were verified. The meeting agendas included procedural changes, proposed changes to the Technical Specifications and field changes to design change packages. The meetings were characterized by frank discussions and questioning of the proposed changes. In particular, consideration was given to assure clarity and consistency among procedure Items for which adequate review time was not available were postponed to allow committee members time to review and comment. Dissenting opinions were encouraged. The inspector had no further comment . Observation of Maintenance and Surveillance Testing The inspector observed various maintenance and problem investigation activities for compliance with requirements and applicable codes and standards, QA/QC involvemert, safety tags, equipment alignment and use of jumpers, personnel qualifications. radiological controls, fire protection,

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retest, and reportability. Also, the inspector witnessed selected sur-ve111ance tests to determine whether properly approved procedures were in use, test instrumentation 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 properly dispositione *

The following activities were reviewed:

SPL 10.7-236, Main Steam - Main Feed Hydrostatic Test 4.1 Reactor Core Support Barrel Thermal Shield Repair The current refueling / maintenance outage has been extended because of necessary repairs to the Reactor Core Support Barrel, the Thermal Shield attachments and associated supports. Defects in the Thermal Shield attachments were observed when the Reactor Core Support Barrel assembly was emoved as part of the Inservice Inspection Program. Additionally, debris was found in the bottom of the Reactor Vessel. Extensive visual and ultrasonic examinations have been made of the Reactor Vessel and the l Core Support Barrel / Thermal Shield assembiv since the original findings '

were made in September. Damaged areas included the holders for the l Surveillance samples of vessel material and the bolts and pins which '

attatch the Thermal Shield and its support blocks to the Core Support ,

Barrel. The debris has been recovered from the Reactor Vessel and has J been matched with the materials identified as missing from the Core l

Support Barrel assembl )

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-5-The Thermal Shield is a stainless steel cylinder located outside of the Core Support Barrel. The shield is supported by six (6) blocks which are located in a recess on the lower outside surface of the Core Support Barrel. Each block is attached to the Core Support Barrel with five (5)

one-inch bolts and three (3) 0.75-inch dowel pins. Two (2) of those bolts and one (1) of the pins secure the Thermal Shield to the support block and the Core Barrel. The design of the block is such that it is captured in the Core Barrel recess by the Thermal Shiel Two (2) of the bolts attaching the block to the barrel are not accessible with the Thermal Shield in plac Tangential movement of the Theimal Shield was designed to be restrained by f our (4) Upper Displacement Limiter Keys. These are located near the top rim of the shield cylinder. The Reactor Vessel irradiation surveillance material specimens are located in chutes which are attached to the Core Support Barrel above the Thermal Shield. These chutes are also welded to the outside surface of the shiel The damage to the assembly includes: 1

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Three (3) accessible bolts were missing and were part of the debris found in the vesse Eight (8) of the accessible bolts were broke ,

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All twelve (12) of the inaccessible bolts (two per blocks) were identified by ultrasonic examination as broke Four (4) of the dowel pins were protruding from their fully inserted positio A 0.060-inch deep, 3 inch long indication at the lower edge of the core barrel groove, of the support block located at the 30 degree positio A measurable gap between the support block and the core barrel groove also at the 30 degree positio A measurable gap between tha bottom of the Thermal Shield and the ledge in the Support Block at the 90 degree positio l

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Wear at all contact surfaces of the Upper Displacement Limiter Key l The total gaps were found to range from between 0.090 inch to 0.180 inch. (The original nominal gaps ranged from approximately 0.020 to 0.025 inch.)

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Surveillance Specimen holder chutes were damaged at the 226.5 degree and 313.5 degree positions, a portion of the chute was missing at t the 223.5 degree position at the thermal shield upper rim. This I chute section was also recovered with the debris found in the vesse l l

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-6-From the results of a failure analysis, the licensee has determined that both repairs and modifications are necessary. The failure analysis, which was based on the inspection results, concluded that the attachment bolts were broken as the result of flow induced movement of the Thermal Shiel Impact at the Upper Displacement Limiters Keys occurs as a norni condition of the support system. The impact wear of the limiter keys har progressed over time. The increased tangential movement of the shield has resulted in increased loads at the six (6) support blocks. These loads increased to the point where fatigue cracks initiated and propogated through the long support bolts. The failure or preload reduction in the long visible bolts, due to crack propogation, led to increased loads in the short inaccessable bolts. Bolt failure results in a reduction in support block stiffness and continued wear of the displacement limiter The increased relative motion between the Thermal Shield and the Core Barrel fatigued and broke surveillance specimen holder chutes which are fixed to both the barrel and shiel During this inspection period, the repair process was completed with one exception. All eighteen (18) of the accessible bolts (three per support block) and three (3) of the dowel pins have been replaced. The removal of one dowel pin damaged a portior of the base material requiring an oversized replacement. Reactor Vessel irradiation surveillance material specimens have been moved to undamaged holder tube The replacement bolting preload has been increased with the intention of reducing the sensivity of the fasteners to fatigue. The original bolts were torqued to 180 foot pounds, the replacement bolts to over 300 foot-pounds. Since the replacement bolts have been installed and torqued under water, the licensee will evaluate the effects of a wet stainless to stainless contact on the bolt prelodd. This issue was discussed at a meeting between the licensee and NRC Materials Branch personnel at NRC Headquarters on December 1, 198 The inspectors observed significant portions of the repair proces Included in these activities was material and process control, procedures review, and radiological controls. The inspectors found that the pre-parations for work activities and the controls exercised during those activities were acceptable. Additional inspection followup of a potential radiological problem is discussed in Detail 6.2 of this repor The failure analysis indicates that additional stabilization of the Thermal Shield is necessary to arrest flow induced movement relative to the Core Support Barrel. This is to be accomplished through the addition of six (6) blocks to be attached to the upper surface of the Core Support Barrel and keyed to a second set of blocks attached to the upper rim of the Thermal Shield. These devices are being designed to limit the tangential displacement of the shiel The licensee intends to shim the gap between the support block at the 90 degree location and the bottom of

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-7-the Thermal Shield with two new dowel pins. These will be installed radially and will provide as much or more contact surface area the-original design of the support bloc This work is expected to start during the week of December 14, 1987 and will be reviewed during the next inspection perio . Followup on Previous Inspection Findings 5.1 Local Leak Rate Testino By Unauthorized Test Metho_d (Closed) Violation (86-08-04) The licensee performed containment penetration local leak rate testing (LLRT) using an unapproved test method and not in accordance with Technical Specifications (TSs).

TS 4.4.II.A required that the leak tests be performed using halon gas detection, soap bubbles, pressure decay, or other methods of equivalent sensitivit However, for numerous penetrations, the licensee performs LLRT with a water collection method. The licensee submitted exemption requests from this and other requirements of 10 CFR 50 Appendix J in March. By letter dated September 29, 1987, the NRC granted a temporary relief from Appendix J and permitted the licensee to continue perf.rming LLRTs with the water method for a period of two refueling outages. This time was provided to permit the necessary modifications to the penetration configuration Penetration modifications will be followed in future inspections; this item is close ,

5.2 Breach of Containment Integrity During Surveillance Testing (Closed) Violation (86-20-03) Violation of Containment Integrity TS by operation of four manual containment isolation valves during >1 surveillance testing. This item has been previously discussed in NRC Inspection Reports 50-213/86-20, 86-27 and 87-2 Report 87-25 !

found that the licensee has taken adequate corrective actions for <

all of the affected valves with the exception of containment I isolation valve SA-V-413. This valve is manipulated during l performance of SUR 5.1-6, Reactor Containment Leakage Mcnitorin I TS permits only limited and controlled manipulation of this valve I when containment integrity is required. SUR 5.1-6 has been revised. Revision 13 to that procedure now requires that an operator be continuously stationed within the Primary Auxiliary :

Building Blowdown Room and be in constant radio communications with i the control room any time valve SA-V-413 is unlocked or opene l This item is close '

5.3 Failure to Serialize Changes to Plant Design Change Records (pDCRs)

(Closed) Violation (86-01-01) Licensee failure to serialize Design Change Notices (DCNs) resulted in inadequate maintenance of files of I field changes to PDCRs. The licensee responded to this violation by

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l -8-letter dated May 12, 1986. Due to a ninety-day grace period for implementation of new Nuclear Engineering and Operations (NEO)

procedures, several groups involved with PDCRs were using different l numbering systems for DCNs and each numbering system was in accor-l dance with a revision to the governing NE0. The licensee has since l reevaluated the practice of allowing groups to take up to ninety days l to implement revisions to NE0s. To remedy this situation, NE0 1.04, Preparation, Issuance, Revision, Deletion and Control of Nuclear j Engineering and Operations Procedures, has been revised to require a specified effective date (no later than sixty days after approval)

l for universal implementation of each new procedure revision. Revision l 4 to NE0 1.04 was effective December 1, 1987. This item is closed.

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l Low Temperature Overpressure Protection (LTOP)

, (Closed) Unresolved Item (85-21-09) Resolution of the problems l encountered when placing LTOP into operation. This item was pre-l viously discussed in NRC Inspection Reports 50-213/85-21, 86-01 and 87-0 Previously, the licensee had experienced difficulties placing LTOP in service at the upper temperature and pressure limits of 330 degrees F and 340 psig, respectively. Near those values, the LTOP valves had lifted prematurely and anomalies with the temperature and pressure interlocks for the LTOP system isolation valves had p.re-vented the valves from being opened. There have been many inter-mediate actions taken to resolve this item. The final corrective actions have been to revise the system functional testing, conducted under SUR 5.2-63, LTOP System Functional Testing, and to change the Technical Specification (TS) required window for placing the system in service. TS Amendment No. 94, issued on September 10, 1987, provides a larger interval in which LTOP must be placed into operation as a result of a revision to the heatup rate specificatio With this change, LTOP must be placed in service before temperature decreases below 315 degrees The inspector reviewed the associated plant procedures and verified that the necessary changes were made to reflect this TS chang .5 Unanalyzed Post Accident Release Paths (Closed) Unresolved Item (85-13-02) Licensee implementation and NRC review of modifications to the reactor coolant system (RCS) drain l header piping. This modification is necessary to reroute the drain l header relief valve (DH-RV-1847) discharge to inside containmen This item was previously discussed in NRC Inspection Reports i 50-213/85-13 and 86-24. During the current refueling outage, the

licensee implemented Plant Design Change Re:ord (PDCR) 878, Appendix A and J Penetration Modifications. This design change modified those containment penetrations which did not qualify for exemption f rom 10 CFR 50 Appendix J. Changes to the penetrations involved installation of containment isolation velves, seismic supports, leak test vent and drain valves, leak test boundary valves, seal welds,

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. t si and relocation of relief valve To address this item, OH-RV-1847 was relocated to inside containment and its discharge rerouted to  ;

the containment sump via the containment annulus drainage system instead of to the volume control tank. The inspectors observed work-  :

in progress during the outage and reviewed the associated' work '

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orders and documentation and had no further question '

5.6 Switchgear Room Fire Door Replacement t

(Closed) Inspector Follow Item (84-14-03) Licensee to realace the

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switchgear room fire door with a National Fire Protection Association 'i Code 80 approved door. This item was initiated because the control l

. room and switchgear room did not have self closing doors and were i frequently found to be left open. The licensee replaced the control

, room door in December of 1985 and is currently replacing the ,

switchgear room door with a code approved doo (This replacement  :

p was completed after the end of this inspection period and prior co  ;

issurance of this report). This item is close ;

, 5.7 Change in Plant Operating Margin Associated with Changes in the  ;

Average Reactor Coolant System Temperature .

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(Closed) Unresolved Item (87-12-01) The averag9 Reactor Coolant j System (RCS) temperature (Tave) program was to be modified by .;

increasing the RCS steady state full. power average temperature from 3 557 degrees F to 562 degrees F. Plant design change PDCR 841 which j endorsed this change was not implemented during Operating Cycle 14, i a In addition, the licensee's Technical Report Supporting Cycle 15 i Operation (NUSCo 155) dated June 1987 clearly identifies the 1 proposed change in Tave within Report Sections-2-Operating History,  !

5-Nuclear Design and 6-Thermal Hydraulic Design. This report was i forwarded to the NRC as an attachment to the Cycle 15 Reload,  !

Technical Specification Change Requests and Reload Report dated June  :

1, 1987. This item is close [

5.8 Engineering Expertise On Shift  ;

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(Closed) Unresolved Item (85-13-04) The licensee's Shift Technical  !

! Advisor (STA) program does not conform to the NRC Commission Policy  !

i on Engineering Expertise On Shift. However, the existing program l l does conform to the current regulatory requirements. This issue has  :

3 been discussed with cognizant licensee personnel. The inspectors  !

! will continue to review STA qualifications du' ring routine  !

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6. Followup on Events Occurring During the Inspection l

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l 6.1 Licensee Event Reports (LERs) and Safeguards Event Reports (SERs)  !

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The following LERs and SERs were reviewed for clarity, accuracy of l l the description of cause, and adequacy of corrective action. The  !

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inspector determined whether further information was required and j l

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-10-whether there were generic implications. The inspector also verified that the reporting requirements of 10 CFR 50.73, 10 CFR 73.71, and Station Administrative, Cperating, and Security Procedures hac been met, that appropriate corrective action had been taken, and that the continued operation of the facility was conducted within Technical Specification Limit * 87-11-01 Containment Penetration Fails "As Found" Local Leak Rate Tes: (Type C)

87-16 RCS Safety Valve As-Found Lift Pressure High Oue to Setpoint Drift 87-17 Inoperable Fire Protection System Due to Personnel Error 87-S01 Detection ot' Handgun at Protected Area Access Point

• Event detailed in NRC Inspection Report 50-213/87-21 No unacceptable cor.ditions were identifie .2 Accidental Radiation Exposure A diver working in the Reactor Refueling Cavity received an unexpected whole body exposure of 2119 millirem on December 4, 198 The exposure was recorded by dosimetry worn on his left knee. At the time of the exposure he was assisting in the Reactor Core Support Barrel Thermal Shield repair activities by attempting to thread bolts through the Core Support Barrel lift rig to the top flange of the Reactor Core Support Barre NRC Quarterly radiation exposure limits were not exceede The Core Support Barrel was in the Reactor Vessel during this incident. The diver was wearing a dry suit and was working in approximately eight feet cf ve ier on the cavity floor. While working it was necessary fo iim to kneel on both knees in the area of the vessel flange and Puch to the lift rig ring, j The diver had been in the work area, just outside of the Reactor !

Vessel flange, for about ten (10) minutes when he backed off to allow '

the lift rig to be repositioned. At the time he had an accumulated dose of about 75 millirem recorded by telemetry (remote reading !

dosimetry attached at both knees). This dose agrees with the l pre-dive radiation survey results of the work area, 600 to 800 !

millirem per hou Shortly after the diver returned to % same work area, the j telemetry dosimetry attached to his ' eft knee began integrating a very high dose rate. Health Physics personnel present estimated it to be about one millirem per secon The diver was advised that he I

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-11-had entered a higher field and instructed to move. At that time the left knee telemetry dosimetry then began integrating at a very high dose rate and the diver was directed to immediately leave the cavit The high dose rate dropped off as he moved from the work area to the ladder. In an approximate thirty second time period the telemetry dosimetry integrated approximately 1700 millire The diver was wearing seven (7) dosimetry packages, both knees, both ankles, both hands and upper trunk. These were in addition to the two (2) telemetry dosemeters, one at each knee. Each package contained a thermoluminescent dosimeter (TLD) and self reading pocket ton chamber (PIC). His assigned whole body dose of 2119 millirem is based on the TLD reading at his left knee. T'ne TLD '

attached at his lef t ankle recorded an exposure of 1811 millire The 1500 millirem PICS in each of these packages were reading off scale high. All other dosimetry recorded approximately 200 millirem for the dive and correlates with the pre-dive and post-dive survey The licensee suspended all diving activities to conduct a thorough investigatio In an effort to find the source of exposure, extensive surveys were performed and the cavity floor was cleaned by underwater vacuu Nothing which departed from pre-dive surveys was immediately identified . Detailed surveys of the cavity floor con-tinued in an effort to the locate the material. A hot particle was eventially found on the floor of the refuel cavity with a Cad-Tel PR-2 High Range instrument. It was found in the area of the path which the diver took when moving from the work area to the ladde That particle was retrieved with double backed tape at the end of a reach rod. It had not been removed from the Reactor Refuel Cavity at the end of this inspection period. The particle, which is too small in size to be seen, reads approximately 1400 Rem per hour on contact and 50 Rem per hour at six inches. Both readings are approximate values and were taken under wate The licensee has upgraded their radiation protection practices based on their findings from this even The plant Residual Heat Removal (RHR) system will not be in use during diving evolutions. (The Refuel Cavity level is adjusted with the RHR system). This is to prevent circulating material from the Reactor Coolant System or Reactor Vessel. Additionally the cavity filtration system will be shut down and the cavity floor will be cleaned with the underwater vacuu This event will also be addressed during future NRC inspections of the station radiation protection program. No un- .

acceptable conditions were identified at this tim l

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6.3 Incorrect Lubrication of Motor-Operated Valve Operators On November 17, 1987, the licensee discovered that lubricating grease had been inproperly applied to safety-related valve '

operator These were environmentally qualified electrical equipment (EEQ) associated with safety-related valve . ..

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-12-Work orders were processed for preventive maintenance of valve motor operators during the current refueling / maintenance outage. While examining the operators, the electricians concluded that the valves required additional grease to be added to the valve operator pinion gear housing This was reported to Maintenance Department supervision and the personnel were instructed to add the lubricant in accordance with the station procedur During the post maintenance inspection of EEQ equipment licensee personnel discovered that the motor "T-drains" were improperly installed at the top of some valve operator motor This led to further examinations and the identification that grease had been injected into the electrical motor rather than the pinion gear housin There is an open space near the plugs; however it is expected that some grease was injected into the motor winding (This type of motor is supplied with sealed bearings).

The style of qualified valve operator which has recently been installed at the station has a series of threaded plugs which allow the operator to be installed in several positions relative to vertical. The "T-drain" plugs are repositioned to allow a moisture drain path. The fact that the electricians improperly replaced those plugs, as noted during the post maintenance inspections, was the method of discovery. There were approximately twenty (20) safety-related EEQ valve motor-operators involved with this inciden The Unit Superintendent has been directing the licensee's investigation, which was not completed at the end of this inspection period. Actions and findings to date have included the following:

The valve motor-operators which we'.'e improperly lubricated have been removed from service; their power supply breakers are tagged ope The Unit Superintendent has directed that no work will be '

performed on any EEQ equipment without his specific authoriza-tion. This will continue until he has verified procedures and training are adequat Valve motor-operator procedures are to be upgraded to include specific instructions and graphics for each type of valve operato There are several different types of valve operators installed in the plant. The current station procedure provides general directions applicable to all operators manufactured by Limitorque.

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-13-Personnel training is to be upgraded to include the qualified valve motor operato The preservation of the qualification is to be a specific issu The training provided in the past used older style valve operator A search is being made of the computerized mcintenance management system to identify all work performed on EEQ equipment. That work activity is to be reviewed to verify that the activity performed did not compromise the equipmen All valve motor-operators will be inspected to ensure that they were not degraded by an inappropriate activit Inadequate supervision contributed to the incident in that there was inadequate instruction or oversite and the personnel had not properly adhered to the procedur The vendor has determined that the type of grease used is compatible with the motor windings and has not recommended cleaning the electrical motor stators with a solvent. The motors will be opened and wiped clean by a representative of the motor manufacturer. In  ;

addition, a root cause analysis of this incident is to be conducted '

by personnel from the licensee's Independent Safety Review Grou ;

The licensee's actions will be followed by the NRC resident inspectors and their findings will be included in subsequent inspection report . 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

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l 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 87-09, Covering the Period September 1, 1987  ;

through September 31, 1987  ;

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Monthly Operating Report 87-10, Covering the Period October 1, 1987 through October 31, 1987 '

No unacceptable conditions were identifie . Steam Generator Tube Rupture Susceptibility The inspector attended a November 19, 1987 meeting at NRC Headquarters  !

at which the licensee met with members of the Materials Branch of NRR to i discuss the susceptibility of Haddam Neck's steam generators to a tube '

rupture similar to that of North Anna Unit l l

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-14-On July 15, 1987, North Anna Unit 1 experienced a steam generator (Model 51) tube rupture. The break was a complete, double-ended, circumferential severance of the tube located at Row 9, Column 15 at the top support plate on the cold leg sid Approximately 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> before the break, primary to secondary leakage was detected and determined to be increasing until the break occurre In response to this event, Westinghouse conducted an evaluation of the failure mechanism and concluded that:

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The failure was caused by flow-induced vibration,

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The ruptured tube was subjected to high local flow velocities due to the high general velocity at North Anna Unit 1, and a funnel effect caused by variations in the Antivibration Bar (AVB) penetrations,

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Tube denting resulted in high mean stress and minimum damping, and

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The ruptured tube had a combination of unusually low fatigue resistance and dampin Because of potential generic implications, Westinghouse reviewed the flow and vibration conditions experienced by other utilities in their steam generators of the same and different model The evaluation was based on a one dimensional model for flow cotiditions, the assumption that the AVBs penetrated one row deeper than the required design depth of Row 14, and the dented condition of the tubes at the top support plate. A susceptibility ratio-to-break was calculated for each tube based on these characteristics and an equation for flow induced vibraticn in single phase fluid From this study, Westinghouse concluded that the Haddam Neck Steam Generators (Model 27) could be susceptible to this type of failur By letter dated September 22, 1987, Westinghouse informed Haddam Neck of this conclusion. It was recommendeJ that the licensee conduct an indepth review of the conditions of rows ten through fifteen in all steam generators with the following emphasis:

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Identify those tubes with any denting at the top support plate in either the hot leg or cold leg,

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Quantify the AVB insertion depths for each column from the eddy current data,

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Identify any tube wear at any AVB or top support plate intersection, and

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Upon receiving this notification, the licensee established a special committee consisting of both station and corporate personnel to conduct the analysis of the Haddam Neck Steam Generators. A review was performed using the eddy current data from the 100% inspection conducted this refueling outage. Also, data was reviewed from the previous two outages to determine any trends in tube conditions, especially denting. After the full review of the data, Combustion Engineering was contracted to conduct en independent party review of the data and conclusion Based on the licensee's evaluation of the data and CE's confirmation, the following conditions were identified:

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Steam generators #1, #2, and #3 have several locations where the tubes are unsupported in the fourteenth row. In steam generator

1. 2, unsupported tubes were identified in rows fifteen and sixtee No AVB wear indications on the tubes were identified in the area of concer Tube denting at the top support plate was identified in both the hot and cold legs of all four generators. Generators #1 and #2 displayed more prominent denting than #3 an( # Compilation of this data identified that steam generator #2 is the only generator which displays the precursors to a tube rupture similar to North Anna's; namely, tubes which are both dented at the top support plate and unsupported by the AVBs to produce a funnel effect in the steam / water flo There are four tubes which fit into this category, R14,C86, R15,C85, R15,C84, and R14,C79. While these tubes are of particular concern, the licensee is also focusing attention on six more tubes in this generator which currently are not dented but are unsupported and in a configuration which could experience the funnel effec Currently, the licensee is evaluating possible methods of three dimensional thermal hydraulic analysis to more accurately determine the flow conditions in the area and the susceptibility ratios of these i unsupported tubes. Until such a time as the analyses hre completed, i Haddam Neck is relying on leak-before-break and the leak detection l systems and procedures which are already in place to preclude an event similar to North Anna' This reliance is based upon reviews of the sequence of events for the North Anna event in relation to procedures and leak detection systems at Haddam Neck. There are three leak detection systems currently in use, air ejector noble gas monitor, steam generetor blowdown radiation monitors, and grab sampling. Under normal operating conditions, the radiation j monitors are observed for trends as part of routine control room '

operations and steam generator grab samples are counted on a weekly basi If primary to secondary leakage is indicated by the air ejector monitor,

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-16-the samples are counted daily to determine the leak rate in accordance with CH0P.4.21, Calculating Primary to Secondary leak Rate Based on Tritium. If the leak rate should exceed ten gallons per day the sampling frequency is increased by this procedure. As the leak rate increases, the sampling and counting frequencies also increase. Station management is continuously updated on the status of any leakage. This was demonstrated in June and July of 1987 when a 1.5 gallon per day leak was identified and tracked in steam generator #2. Technical Specification 3.14.A.5 limits primary to secondary leakage for all steim generators not isolated from the RCS to 0.4 gpm, and to 150 gallons per day through any one steam generator not isolated from the RCS. Leakage must be reduced in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or the plant must be placed in mode 4 in the following 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> The licensee has superimposed their actions, by procedure and TS require-ments, onto the sequence of events for the North Anna event and determined that, had this leakage occurred at Haddam Neck, the plant would have been shut down six hours before the tube ruptured. This supposition did not include any management actions to shutdown the plant prior to reaching the TS leakage limit of 150 gallons per da The licensee has been requested to provide further information on the one dimensional model and assumptions used to calculate the tube rupture susceptibility ratios, which tubes have a susceptibility ratio greater than 0.9, and the mean stress levels and fatigue curves for these tube This information is to be submitted before the end of 1987, before plant startup occur . Licensee Event Reporting to the NRC Incident Resporse Center NRC Inspection Reports 50-423/86-21 and 87-12 and 50-245/86-13 and 87-05 addressed a problem with the Millstone Plants' event reports to the NRC Incident Response Center. The problem involved the use of the term

"General Interest Event". This phrase is used by the State of Connecticut but not by the NRC, and not familiar to NRC officials who have received such non-emergency reports on the Emergency Notification System (ENS).

Corrective actions proposed by Millstone management include revision of the Emergency Plan Implementing Procedures (EPIP) Form 411201 and training of shift personne The issue remained open in NRC Inspection Report 50-245/87-12. This problem also has existed for Haddam Neck event report l By letter dated October 23, 1987, Northeast Utilities outlined their corrective actions as follow All Millstone Shift Supervisors (SSs) and SS Staf f Assistants (SSSAs) have received a memorandum from the Station Superintendent explaining the problem and giving explicit instructions that the '

term "General Interest Event" is not to be used when making 10 CFR l 50.72 (b)(2)(v) non-emergency reports to the NRC, '

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The above instructions are being included in the formal training of emergency response personnel at Millston ;

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NNECO is working to alter their EPIP and reporting forms to derive a reporting system which has a more reliable human performance interface. The expected completion date was November 15, 198 CYAPC0 initiated similar actions for Haddam Nec These actions were completed on December 8, 198 The inspectors reviewed the revisions to EPIP 1.5-33, Shift Supervisor's Staff Assistant, and verified that it clearly stated that State of Connecticut posture codes are not to be relayed to the NRC Response Center Duty Officers. This procedure revision has been reviewed by the necessary station personnel and incorporated into Emergency Response Training. The inspectors had no further-concern . . Exit Interview During this inspection, meetings were held with plant management to discuss the findings. No proprietary information related to this inspection was identifie !

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