IR 05000213/1988015

Safety Insp Rept 50-213/88-15 on 880801-0912.No Violations Noted.Major Areas Inspected:Plant Operations,Radiation Protection,Fire Protection,Security,Maint,Surveillance Testing,Licensee Events & Open Items from Previous Insps

ML20204F333
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	10/12/1988
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20204F309	List: IR 05000213/1988015 ML20204F304
References
50-213-88-15, NUDOCS 8810210556
Download: ML20204F333 (15)

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U.S. NUCLEAR REGULATORY COMMISSION REGION I j

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Report N /88-15  ;

Docket N i License N DPR-61 l l

Licensee: Connecticut Yankee Atomic Power Company ,

P. O. Box 270 t Hartford, CT 06101  ;

Facility: Haddam Neck Plant, Haddam Neck, Connecticut  !

Inspection at: Haddam Neck Plant Ir.spection dates: August I through September 12, 1988  !

L Inspectors: Andra A. Asars, Resident Inspector [

John T. Shediosky, Senior Resident Inspector '

j Jason C. Jang, Senior Radiation Specialist, DRSS -

l Herbert J. Kaplan, Senior Rea: tor Engineer, DRS Approved by: & O. M M k E.C.McCabe, Chief,deactorProjectsSectionIB

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1. ella/03 Date l

1 Summary: Inspection 50 213/88-15 (8/1/88 - 9/12/88)

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Areas Inspected: This was a routine safety inspection of plant operations, radi- r ation protection, fire protection, security, maintenance, surveillance testing, i licensee events, open items from previous inspections, charging pump shaft fatture l and internals replacement, auxiliary feedwater pump surveillance and maintenance, l operation with elevated service water temperatures, and the effects of a contain- j ment fire on the availability of power to Containment Air Recirculation Fan motor ;

Results: No violations were identified. Five violations, unresolved items, and  :

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open items were closed. Events reviewed during this inspection included the repair i of a charging pump following pump shaft failure, operation with elevated service i water temperr.tures, and the effects of a containment fire on the availability of i power to Containment Air Recirculation Fan motor l l

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TABLE OF CONTENTS

_PAGE S umma ry o f Fa c i l i ty Ac t i v i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Review of Plant Operations........................................... 1 Plant Operations Review Committee.................................... 2 Maintenance and Surveillance......................................... 2 4.1 Charging Pump Internals Replacement............................. 2 4.2 Au xi l i a ry Feedwa te r Pump Ma i n t e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Followup on Previous Inspection Findings............................. 4 5.1 Degradation of Vital Area Barriers............-................. 4 5.2 Motor Operated Valve Torque Switch Settings. . . . . . . . . . . . . . . . . . . . . 5 5.3 Analysis of Water Samples....................................... 5 5.4 Control Room Annunciator Rev1ew................................. 6 5.5 Peactor Coolant System High Point Vents. . . . . . . . . . . . . . . . . . . . . . . . . 7 Followup on Events Occurring During the Inspection................... 9 6.1 Licensee Event Reports and Safeguards Event Reports............. 9 6.2 Charging Pump Shaft Fa11ure..................................... 9 6.3 Elevated Service Water Temperatures............................. 10 Periodic and Special Reports......................................... 11 Effects of a Hypothesized Containment Fire on the Contai,Jent Air Recirculation Fans................................................. 12 Exit Interview....................................................... 13

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D_ETAILS Summary of Facility Activities (71707)

At the beginning of the inspection period, the station was operating at full powe The "A" charging pump shaft failed on July 31. The pump rotating as-sembly was rebuilt and the pump was returned to service on August 5. A rre-chanical alignment was made of the "A" auxiliary feedwater pump and its driv-ing turbine during the week of August 7 after high vibration was detected on the turbine outboard bearing housing. On August 15, a load reduction to 90%

power was effected in response to elevated river water temperatures. Reduced power levels were maintained until river water temperature trends indicated that peak temperatures would be below the limit of 90 degrees Fahrenheit established in a Justification for Continued Operatio The plant was re-turned to full power on August 17 and continued to operate at full power through the end of the Inspection perio . Review of Plant Operations (71707)

Tne inspecto. observed plant operation during regular tours of the following plant areas:

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Control Room --

Security Building

-- 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 Puep Building Control room instruments were observed for correlation between snannels and for conformance with Technical Specification requirements. The inspector ob-setved various alarm conditions which had been received and acknowledge Operator awareness and response to these conditions were reviewed. Control room and shift manning were compared to regulatory requirements. Posting and control of radiation and high radiation areas was inspected. Compliance with Radiation Work Permits and use of appropriate personnel monitoring devices were checked. Plant houseweeping contross were observed, including control and storage of flammable material and other potential safety hazards. The inspector 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 / deficiencies. 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 monitorin In addition to normal utility working hours (7:00 am to 3:20 pm), plant opera-tions was routinely reviewed during latter portions of the midnight shifts and early portions of the evening shifts. Also, extended coverage was pro-vided on the following days:

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Serember 26 at 3:00 I

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September 27 at 3:00 i

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No unacceptable conditions were identified. Operators were alert and dis- I pl4yed n: signs of inattention to duty or fatigu (

t Plant Operations Review Committee (PORC) (40700)

The inspector attended several Plant Operations Review Committee (PORC) meet-ings. Technical Specification 6.5 requirements for required member attendance were verified. The meeting agendas included procedural changes, proposed

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changes to the Technical Specifications and field changes to design chango ;

packages. The meetings were characterized by frank dfscussions and question- [

ing of the proposed changes. In particular, consideration was given w assure !

clarity and consistency among procedures. Items for which adequate review !

time was not available were postponed to allow committee members time to re- I view and comment. Dissenting opinions were encouraged. The inspector had no further comment . Maintenance and Surveillance (61725, 62703) {

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l ties for compliance with requirements and applicable codes and standards, i i QA/QC 1rvelvement, safety tags, equipment alignuut and use of jumpers, rir-l tonnel y alifications, radiological controis, fire protection, retert, and ,

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reportability. Also, the inspector witnessed selected surveillance tests to l determine whether properly approved procedures were in use, test lastrumenta- t tion was properly calibr4ted and used, technical specifications were satisfie l testing was performed by quelified personnel, procedure details were adequate, ;

and test results satisfied acceptance criteria or were properly dispositirned, i The following activities were reviewed: j

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Switchgear Built 1'ng Construction Activities (

-- Auxiliary Feedwater Pump Surveillance and Maintenance  :

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"A" Charging Pump Repairs  !'

-- SUR 5.1-4, Core Cooling Systems Hot Operational Test

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SUR 5.7-19, Inservice Inspection Pump Surveillance  !

l 4.1 Charging Pump Internals Replacement (61726, 62703)

f The inspectors observed the rebuilding of the "A" Charging Pump internals t and rotating assembly which failed on July 31. That event is discussed in report paragraph 7.2. The repair process consisted of the replacement j of the 9 ump shaft, impe11ers and wear rings and reuse of the old diffu- [

sers. A redesigned shaf t with interference (4hrink) fit impe11ers was j used during this repair. The failed shaft design had used a c apression i (locking) nut to retain the stack of thirteen impe11ers. The inspectors covered maintenance, quality assurance and radistion protection activi-ties during the repair There were no unacceptable conditions identi-fle I

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The rebuilt pump was placed in service on August 5 following successful  !'

> surveillance testin On August 23-24, a regional inspector reviewed the history of shaft f J

failures in the "A" and "B" pumps in the chemical and volume control j

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system. The pumps are of the centrifugal type with 13 stages and were  ;

i manufactured by Pacific Pump, California, currently a division of Dresser j Industries. The licensee reported that, since 1967, there have been a  !

, total of 12 shaft failures, with the last five failures occurring in the -

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"A" charging pump. All of the failed shafts were manufactured from 17-4 i PH (precipitation hardened), 1150 d y rees F aged stainless steel, and failed within the first th eads of the threaded section of the shaf '

This portion of the shaft is retained by the locking nut. The inocard i j section of the nut is aligned with the location of the failure. The i

licensee reported that vibration testing did not identify any significant  !

l differences between pumps "A" and "B." The inspector also reviewed a :

Combustion Engineering failure analysis of one failed shaft which broke l after five months of operation. The analysis concluded that the failure was due to high cycle rotating bending fatigue. Also, no material anoma- i lies, machining marks or out of specification conditions were found to  ;

be associated with the failure. The inspector also examined the fracture j faces of the recent failure. The fracture faces appeared very similar i to the failure investigated by Combustion Engineering; generally flat (

surfaces with no discernible evidence of "necking down", except that this '

failure also exhibited some torsional feature i

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The replacement shaft far the latest failure was subjected to the foi-l lowing design changes to imorove its fatigue life:  ;

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Cold rolling of threads, f

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A more generous radius in threads and split ring I i

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Shrink fit impellers en shaft to reduce pla [

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A material i:hange from 17-4 PH to ASTM 216 Gr. 414 martensitic, $

quenched and tempered stainless steel based on Pacific Pump Com-  !

pany's laboratory data and service experienc [

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A redesigned locknut to take the load at the outboard end of the [

nut to distribute the load over all the thread !

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The inipector noted that the 414 stainless specified by Pacific Pump  ;

exhibits lower minimum tensile strength (120,000 psi) than 17-4 PH 1150 t degrees F aged (135,000 psi) and may have been used in the replacement  ;

of a failed shaft in 1977. The 414 stainless steel has better ductility,  ;

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although design features which reduce vibration (shrink fit impellers)

may be of greater importance. In this regard, Pacific Pump is working

on a new design to eliminate the threaded locking nut.

a The inspector conc 1Med that the licensee was making constructive changes

to improve the fatigue resistance of the subject shaft '

4.2 Aunt 11ary Feedwater Pump Maintenance (61726, 62703)

The inspectors observed the licensee's investigation and corrective main-

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tenance actions on the "A" Auxiliary Feedwater (AFW) pump in response to high Terry Turbine outboard bearing vibration. Earlier AN pump in-1 spection and maintenance for high vibration were previously discussed i in Inspection Report 50-213/88-11.

W l This inspection period covered two monthiy inservice inspection (101)

! surveillance intervals. The "A" A N pump was taken out of service be-

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tween August 8 and August 11 because of turbine outboard bearing vibra-

tion. At this time it was identified that the Terry Turbine discharge piping support was restricting the discharge piping and therefote pos-

{ sibly causing the vibrations on the turbine bearing. Adjustments were made to the support and the pump was ratested satisfactorily. Two bear-i ing vibration measurements were still in the ISI acceptance criteria alert range and require monthly ISI surveillance. During the second week

. of September, *he "A" AN pump was taken out of service for realignment j of the Terry Turbine-to-pump flexible drive coupling. During testing

- several parameters were monitored. Vibration was monitored at each pump i and turbine bearing housing. The pump and turbine base plates were moni-

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tored for motion relative to the equipment foundwtion. Pump shaft and j

turbine shaft vibration was monitored at each side of the drive coupling, j Following successful completion of this testing a new !$! acceptance j criteria baseline was established.

i j With the establishment of a new baseline, the "A" AN pump will resume

. a quarterly test frequency beginning in November, 1988.

) During maintenance and surveillance of the "A" AN pump, the inspectors

i verified that the requirements of Technical Specification 3.8.A.2.b. were

! met and that the "B" A N pump was verified as operable. No deficiencirs l were identified during inspector review of these activities.

, Followup on Previous Inspection Finding J E1 Degradation of Vital Area Barriers (93702)

l (Closed) Violation (88-12-01): The li ensee identified two cases of de-

graded vital area barriers. The licensee responded to this violation by letter dated August 30, 1988. The root causes of this violation have been determined to be inadequate proceduial guidelines and inadequate

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1 I instruction to construction personnel. Immediate and long term correc" ;

tive actions have been taken and included improvement of procedures, i worker briefings, and construction walkdowns. The intpectors reviewed <

l these actions and had no further concern [

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5.2 Motor-operated valve (MOV) Torque Switch Settings (61726, 62703)  !

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. i (Closed) Unresolved Item (87-22-03): This ite: encompasses clarification i j of Preventive Maintenance Procedure (PMP) 9.5-3, Crane Teledyne Valve ;

Motor Operators, and PMP 9.6-4, limitorque Valve Motor Operators, and i the inclusion of torque switch settings in the Master Setpoint List (MSL).

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The licensee has incorporated the torque switch settings for the affected i

MOVs into the MSL along with the overload heater sizes and the torque i j switch bypass setpoints. The inspector reviewed the Setpoint Change !

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Request and associated safety evaluation. The affected PMPs have been !

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revised as follows. PMP 9.5-3 no longer requires that the torque switch !

l setting be measured because, with the Teledyne design, the measurement i i would result in a changed setpoint. References to the MSL were also j i deleted from PMP 9.5-3 and a requirement that operations conduct post- !

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maintenance testing was added. A caveat stating that the job supervisor I l should be contacted if the MSL does not contain a torque switch setting l 1 and the requirement for operations to conouct post maintenance testing ;

i was added to PMP 9.5-4. The inspector noted that these PMPs do not pro- [

l vide for changing of the torque switch setting This is controlled by i J

Corrective Maintenance Procedure (CMP) 8.5-24, Crane Teledyne Valve Motor !

Operators, and CMP 8.5-25, Limitorque Valve Motor Operators. The licen- i see stated that the post-maintenance testing will include MOVATS-type *

l testing to ensure the correct torque switch settings. These actions i t satisfactorily resolved the previous concerns; this item is tiosed.

! I l 5.3 Analysis of Water _,'amples (84525)

i 1 (Closed) Inspector Follow Item (87-29-01) On completion of the analyses

of water samples (spiked samples) by the licensee and Brookhaven National

Laboratory, a statistical evaluation was to be made. The analyses were l completed and an evaluation was performed. The analytical comparirons l

l were acceptable and are itsted in the following tabl j

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Analyte Matrix Sample 10 Haddam Neck Brookhaven Fluoride Steam Generator 0.5m1 spike 62.02 1.3 ppb 76.2t 2.1 ppb 1.0ml spike 84.01 1. 4 ppt, 88.9t 0 ppb Chloride Steam Generator 0.5m1 spike 25.01 0.5 ppm (1) 932 ppb (1) i 1.0ml spike 25.01 0.5 ppm (1) 1292 8 ppb (1) '

Sulfate Steam Generator 0.5m1 spike 98.0t 0.5 ppb (2)

1.0ml spike 125.0t 0,9 ppb (2)

Wickel Steam Generator 0.5m1 spike 28.01 1.0 ppb 31.01 1.0 ppb 1.0m1 spike 50.01 0.9 ppb 57.01 2.0 ppb Iron Steam Generator 0.5m1 spike 65.0t 2.0 ppb 31.92 0.2 ppb 1.0ml spike 100.01 2.0 ppb 60.01 1.0 ppb Chromium Steam Generator 0.5m1 spike 27.0t 1.2 ppb 32.51 0.1 ppb 1.0ml spike 52.0 1.4 ppb 53.02 0.2 ppb Copper Steam Generator 0.5m1 spike 80.01 1.2 ppb 85.7t 0.1 ppb 1.0ml spike 108.0t 0.5 ppb 113.01 2.0 ppb Boron Mix Tank none 17702 2.4 ppm 1847i 6.0 ppm (1) Chloride analyses were performed by prnbe due to high phosphate in sample (Haddam Neck), and Brookhaven used ion chromatography for chloride an-alyse (2) Brookhaven was not able to analyze sulfates due to interferenc J Control Room Annunciator Review (93702, 61726)

(Closed) Unresolved Item (87-22-04): The licensee committed to review all of the control room annunciators to assure annunciator operabilit This item was opened in response to the identification that the tempera +

ture switches for the residual heat removal pump seal water coolers were not wired correctly and would not provide a control room alarm if an over-temperature condition occurred. The licensee elected not to perform special functional tests for each annunciator. The verification of an-nunciator operability was by operator documentation when an alarm was received either as a result of station operations or surveillance Alaras that an operator specifically remembered having annunciated earlier in the refueling outage were also noted. The inspector reviewed the verification sheets and noted that the First-Out and the Con rol Room Auxiliary Annunciator Panels were not included in this verificatio Of the annunciators surveyed, one was identified as ineffectiv The

"High Level Transformer Oil Suep" alarm was determined to not be wired and not represented in design documents. After review, the licensee removed this annunciator from service. The First-Out and Auxiliary Panels were handled separatel The Control Room Auxiliary Annunciator Panels include the Safety System Lockout, Post Accident, Emergency Diesel Generator Auxiliary, Fire Pro-tection, Environmental, and Low Pressure Steam Dump Panel Annunciator These annunciators were separated into three categories; those that are received on a routine basis through normal operations or surveillances,

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those that were recently installed by Plant Design Change Records and vhich were verified through post-mod'.fication testing; and th>se that require an individual action plan to assure operability. Action plans are required for two alarms on the Safety System Lockout Panel, six alarms on the Fire Protection Panel, and two on the Environmental Pane The licensee is currently performing these vertftcation The Haddam Neck First-Out annunciators nre divided into two panels, a reactor side (B2) and a turbine side (H2). The First-Out annunciator function is designed to allow the first trip signal to lock in and, through relay action, lock out any other alarms on these two panel Testing of these annunciators was performed during the refueling outage

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under Preventive Maintenance Procedure (PMP) 9.5-152, Functional Testing of Annunciators B2 and H2. The test included a simulation of a trip signal at the annunciator and verification of the lock-out and lock-in I features. During the initial performance of this PHP, under Automated Work Order (AWO) 87-6256, several annunciators did not perform as ex-pected. AWO 87-9787 was issued for inspection and repair of the faulty annunciator The inspector reviewed the completed AWO and the Produc-l tion Maintenance Management System (PMMS) record of this work and dis-

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cussed the job with maintenance personnel. Several relays and alarm j modules required replacement. The failures were determined to have been l caused by age. Some of these components were original plant equipmen >

Subsequently, the PMP was successfully completed. The inspector noted j that the licensee has elected not to enter this FMP into the PMMS work

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list for performance during refueling outage During discussions with maintenance and operations personnel, the in-

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spector was informed that the First-Out annunciator panels are not ef-fective if simultaneous or near-simultaneous trip signals are receive This is due to the time delay inherent in the relay action for the lock-1 out feature. Operations personnel stated that they are comfortable with the ability of the plant computer to record post-trip informatio This is used with other data, such as the First-Out annunciators, to complete the post-trip review For the past several years, the licensee has conducted reviews of this system performance. At this time, plans for modification of the First-Out annunciator system have not been mad Licensee actions in this area were determined to be acceptable. The inspector had no further concerns at this tim .5 Reactor Coolant System High Point Vents (71707)

(Closed) Inspector Follow Item (84-03-05): This item was opsned to track the completion of several actions relating to TMI Action Plan Item II.B.1, Reactor Coolant System (RCS) Vents. At the tiee of the initial inspec-tion, the following actions requi "d completion: verification and docu-mentation of the environmental qualification of Litten connectors, change

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l of Normal Operating Procedure (N05) 2.1-1 Cold $hutdown to Hot Standby, I and incorporation of procedures for RCS ventino into the symptom-oriented Emergency Operating Procedures (EOPS). The issue of RCS high point vents was also addressed in the Haddam Neck Safety Evaluation Report (September 5, 1983) and the Integrated Safety Assessment Program, Topic 1.4 l The qualificetion of these valves had previously been resolved. Howeve l at the close of this inspection period, the licensee made the determina- ,

tion that corrosion recently found in the r.onnectors on the pressurizer -

vents constitutes a substantial safety haraM ($$H). The licensee sent ,

a 10 CFR Part 21 Report to NRC dated September 27, 1988. The licensees actions and impact on the equipment qualification will be discussed in

the NRC Inspection Report 50-213/88-1 '

! 1he inspector reviewed the current revision of NOP 2.1-1, Revision 2 >

Step 6.35 provides for cycling of the downstream vent valves in each pair i

each 100 degrees F or 500 psig during station hentup to prevent over-pressuritation of the piping between the valve RCS venting procedures were previously reviewed in NRC Inspection Report 50-213/87-02, detail 7.2. During that insper. tion the !ncorporation of the venting procedures into E0P FR-l.3 was verified. It was also stated t that a similar procedure, Abnormal Operating Procedure (AOP) 3.2-22, Pc5 j Venting of Non-Condensible Gases, would be cancelled. However, AOP ,

3.2-22 has not been cancelled and has recently bean revised and improved l

] as part of the procedure upgrade program. The Itcensee is evaluating ['

the incorporation of AOP 3.2-22 into the corresponding E0P, FR-I.3. The

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resident inspectors du not consider the existence of an AOP in addition l

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The inspector also reviewed the testing of these valves by the Inservice ,

Testing (!$T) Program. The program requires that, when the plant is in '

cold shutdown, these valves are tested in accordance with SUR 5.7-105,  !

Inservice Inspection and Operability Test of Reactor Head and Pressurizwr [

Vent Line Isolation Valves. This test includes a local of valve opera- r j tion and valve remote position indication verifications. The inspector l

reviewed the refueling outage maintenance and testing records for these l

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valves. No deficiencies were identifie *

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The previously mentioned $$H had resulted from the identification of .

ground on pressurizer head vent PR-50V-552A during post-maintenance 1 testing. This valve had failed to close during the refueling outage IST [t

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test and was removed from the system for shop maintenanc During the 1 post-maintenance test, the licensee identified the ground which resulted I in dual indication and discovered corrosion on soldered wires at the l Litton connector, The area was cleaned with an electronic contact ,

! cleaner spray and the ground cleared. The valve was retested satisfac- t torily and returned to servic The SSH will be further discussed during  !

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. , Followup on Events Occurring During (ne Inspection 6.1 Licensee Event Reports (LERs) and Safeguards Event Reports (SERs)

[H76T~Td?l2,~93701)

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

determined whether further information was required and whether there !

were generic implications. The inspector also verified that the report-ing requirements of 10 CFR 50.73, 10 CFR 73.71, and Station Administra-tive and Operating, and Security Procedures had been met, that appro-priate corrective action had been taken, and that the continued operation of the facility was conducted within fechnical Specification Limits, r

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88-16, Diesel Fire Pump Declared Inoperable Due to Water Pump Leakage

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• 88-17, River Water Temperature Exceeded Deign Basis Limit

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88-18, Stack Wide Range Gas Monitor Determined Inoperable D ring

, Testing

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88-19, Switchgear Room F'.e Watch Improper!y Secured Due to Persnn-nel Error

-- 88-505, Sateguards Event Report

• Event detailed in NRC Inspection Reports 50-213/88-11 and 88-15 No unacceptable conditions were identifie .2 Ch_arging Pump Shaft Failure (40700, 61726, 62703, 71707, 93700)

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On July 31 at 4:30 pm, the plant was operating at full power when opera-i tors received a low pressure alarm for the reactor coolant pump (RCP)

seals and a RCP high seal water temperature alarm coincident with indi-cation of higher than normal operating amps for the "A" charging pump.

] The "A" charging pump was shutdown and the "Ed charging pump starte l Locally, operators observed water spraying f em the "A" pump inboard sea Water supply to the pump was isolated and the pump declared inoperabl Technical Specification (TS) 3.5 specifies that the reactor not be cri-tical without one charging pump and one metering pump operable. TS also specifies that the reactor not be critical without one charging pump 1 operable. Vnile the TSs require prompt reporting of these occurrences,

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neither of these limiting conditions has an action statement limiting

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the time plant operation may continue. However, several years ago, the Itcensee implemented Administrative TS (ATS) which are similar to the

Standard TSs (STSs). ATS 3.6 allows power operation with one train of

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Emergency Core Cooling System (ECCS) inoperable for up to 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Otherwise, the plant must be in hot standby within six hours, and hot shutdown within the following six hour Pump repair activities commenced immediately; these are discussed in detail 4.1 of this report, j

Repair efforts continued through the week. On Wednesday, August 3, it ,

became evident that the pump would not be returned to service before the l 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock expired. The licensee elected to administratively handle '

the ATS as if it were an NRC approved TS. A one-time, temporary waiver of ATS 3.6.A - I.1 was processed permitting an extension of the 72-hour clock by 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Safety Evaluation performed for this waiver de- l termined it not to be an unreviewed safety question. A portion of the evaluation was based on the eff1ct on core melt frequency as determined i by the Probabilistic Safety Study (PSS). This review concluded that,  !

with the charging pump out of service for an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the core melt frequency would be increased by less than one percent. To off- -

set this increase, the PSS group recommended that operatient perform sur- ,

veillances of ECCS equipment to verify operability. SUR 5.1-4, Core  !

Cooling Systems Hot Operational Test, and SUR 5.7-19. Inservice Inspec-  !

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tion Pump Surveillance, were successfully performed on August Although the Itcensee's actions did not require NRC approval, the resi-dent inspectors and representatives from the NRC Rev on ! Office and the t Offir.e of Nuclear Reactor Regulation were kept informed of the licensee's planned actions.

! On Friday, August 6, maintenance personnel were installing the pump in l the charging pump cubicle but it was questionable whether installation, r alignments, and testing could be completed by the 4:30 pm Friday deadlin !

The site requested the PSS group to perform M additional analysis to re-extend the clock. This request was denteu. Operations personnel  !

commenced a load reduction at 2:00 pm and the appropriate notifications [

were made to the State of Connecticut and the NRC. During the Ined re-  !

duction, raintenance personnel continued their efforts to restore the  !

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pump. At about 5:00 pm the power reduction was suspended (at about 60%

] pwer) while operations personnel supported post maintenance testing of  :

the charging pump. The pump tested satisfactorily and was declared I operable. The licensee elected to perfom high risk surveillances before resuming full power operations. A power increase to full power was then I started at about 7:00 p l 6.3 Elevated Service Water Temperatures (40700, 71707, 93702)

During the inspection period, the Service Water Syssen (SW) temperature t was elevated as a direct result of the increased temperatures of the l Connecticut River (the ultimate heat sink). SW is required under acci- l dent conditions to cool the Emergency Diesel Generators, Containment Air l

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Recirculation Fan Motors and Heat Exchangers, s Residual Heat Removal Heat Exchangers and pump seals and lubricating oil coolers. This matter was previously discussed in NRC Inspection Report 50-213/88-1 On August 5, with SW approaching the design basis and accident analysis assumption of 85 degrees F4hrenheit (85F) the licensee prepared a Justi-fication for Continued Operation (JCO) with SW temperatures up to 90 The equipment affected by SW temperatures and considered under this JC0 were containment pressure and temperature Residual Heat Removal and Containment Air Rectreulation System operation, and Emergencu Diesel Generator performance. The licensee concluded that temporary operation with SW up to 90F was acceptable. The inspectors attended Plant Opera-tions Review Committee Meeting No.88-188 on August 11 concerning this 4 issue and identified no additional concerns. Also, the JC0 was provided to NRC Region I and NRR; no inadequacies were identifie The Connecticut River water temperature at the SW inlet reached 90F on August 15. A power reduction to 90% power was made in accordance with the JCO. The licensee notified the NRC and the State of Connecticu SW temperature was above 90F for approximately 14 minutes. The licensee elected to maintain 90'4 power until August 17. This decision was based upon temperature profiles and trends, and consideration of the effects of power swings on fuel clad integrit Based on information provided to the NRC Region ! Project. Sectior. Chief by the licensee's Vice President, Nuclear and Environmental Engineering, the JC0 for the elevated water temperatures was only valid for the equip-ment conditions existing this summer. A specific JC0 fc each future high SW temperature period or appropriate long-term action will be needed to address future high terrperature conditions. The licensee is evalu-ating actions to resolve this issu . Periodic ar.d Special Reports (71707)

Periodic and special reports submitted pursuant to Tech',ical Specification 6.9 were reviewed. This review verified that the repor;ed information was valid ano included the NRC required data; that test results and supporting information were consistent with design predictions and performance specift-cationt; 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 re-ports were reviewed:

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New Switchgear Building Construction Bi-Monthly Progress Report No. 11, dated July 29, 1988

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Haddam Neck Monthly Operating Report No. 88-07, for the period July 1, 1938 through July 31, 1988

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Haddam Neck Monthly Operating Report No. 88-08, for the period August

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! No unacceptable conditions were identified.

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' Effects of a Hypothestred containment Fire on the Containment Air Rectreula-tion Fans (4D700DDD7, 61726)

On September 13, the licensee informed the NRC that, during design reviews for the nr.w Appen,itx R Switchgear Building, a new locetion for the most limit-ing 10 GR Part 50, Appendix R fire was identified, lhts postulated fire i location could prevent the plant from reaching reactor cold shutdown condi-tiens within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> It was determined that a hypothetical fire inside the containment and affect-ing one safeguards penetrations area will result in the loss of three of the four containment air recirculation (CAR) fans. Loss of the third fan is due to inadequate cable separation discovered during the above mentioned review; the No. 2 CAR fan cables pass through both penetration areas. Analysis has shown that at least two CAR fans are necessary to maintain containment tem-peratures low enough for personnel entry. In this scenario, placing RHR into service requires manual alignment of va'ives inside the containment. Due to heat from the fire and loss of three CAR fans, containment entries are assumed to be impossible. The licensee would be unable to put the plant into cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, but the hypothetical fire would not interfere with placing the reactor into hot standby. A Justification for Continued Operation (JCO) was prepared and the Plant Operation Review Committee (PORC) met on September 13. The PORC discussed the fire scenario, reviewed the JCO, deter-mined that there is no unreviewed safety question, and developed a list of actions to be take The JC0 describes in detail the conditions resulting in the inoperability of three of four CAR fans. Continued operation until the next refueling outage was justified by the following:

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A fire in containment large enough to affect three CAR fans is unlikel The area of concern is monitored by four ionization smoke detectors and a cable insulation fire is characterized by significant smoke generatio Operations personnel have been made aware of the significance of a fire in containment and the necessity for a prompt respons The simultaneous loss of offsite power and significant fire inside con-tainment in not likel The plant can be maintained in a stable hot thutdown condition until containment temperatures decrease enough to permit personnel entry for RHR alignmen T

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The PORC recommended that the following immediate actions be taken:

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Verification of operability of the four smoke detectors in the are Inform operations personnel of this scenari Remind operations personnel of the Abnormal Operating Procedure for using the diesel fire pump for makeup to the Demineralized Water Storage Tank (to supply Auxiliary Feedwater).

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Inspect the area of concern inside containmer t and assure that there are no combustibles in the are All of these actions were completed by September 16. Additionally, the lic-ensee is evaluating the feasibility of using the containment spray system as a source of containment cooling in this fire scenari The licensee plans to correct this cable separation problem during the next refueling outage (September 1989).

The resident inspectors attended the PCRC meetings and reviewed licensee ac-tions. The JC0 was also made available to representatives of the NRC Office of Nuclear Reactor Regulation. No unacceptable conditions or NRC concerns were identifie . Exit Interview (30703)

During this inspection, meetings were held with plant management to discuss the findings. No proprietary information related to this inspection was identifie ,