IR 05000313/1987029

Insp Repts 50-313/87-29 & 50-368/87-29 on 870818-20. Violations Noted.Major Areas Inspected:Containment Temps, Equipment Qualification,Process Instrumentation & Temp Effects on Containment Structure & Accident Analysis

ML20236W708
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/01/1987
From:	Ireland R, Milhoan J, Norman D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20236W693	List: IR 05000313/1987029 ML20236W692
References
50-313-87-29, 50-368-87-29, NUDOCS 8712080179
Download: ML20236W708 (19)

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APPENDIX I

U.S. NUCLEAR REGULATORY COMMISSION l

REGION IV

j NRC Inspection Report: 50-313/87-29 License: DPR-51 50-368/87-29 NPF-6 Dockets: 50-313

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50-368

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Licensee: Arkansas Power & Light Company (AP&L)

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P. O. Box 551 Little Rock, Arkansas Facility Name: Arkansas Nuclear One, Units 1 and 2 Inspection At: Little Rock, Arkansas and ANO site Inspection Conducted: August 18-20, 1987 M ers: f $.r b R. E. Ireland,' Chief, Plant Systems Section

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Division of Reactor Safety A24 %,.u~ Nbo/t7 D. E. Norman, Reactor Inspector, Plant Systems Date -

Section, Division of Reactor Safety Also participating-in the Inspection:

G. Dick, Project Manager, NRR L..Maggleby, EG&G Idaho Approved: .

. m L. Milboan, Director, Division of Reactor

/k/[87 Date Safety 8712080179 871203 3 PDR ADOCK 0500 G

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Table of Contents Page 1.0 General . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1 Bac kg rou nd . . . . . . . . . . . . . . . . . . . . . . 2 1.2 Inspection Tasks . . . . . . . . . . . . . . . . . . . 2 <

2.0 Containment Temperature . . . . . . . . . . . . . . . . . . 3 l 2.1 H i s to ry , U n i t 1. . . . . . . . . . . . . . . . . . . . 3 2.2 Unit 1 Temperatures. . . . . . . . . . . . . . . . . . 3 2.3 Unit 2 Temperatures. . . . . . . . . . . . . . . . . . 4 3.0 Equipment Quali fication . . . . . . . . . . . . . . . . . . 5 3.1 Ambient Temperatures Used for Equipment Qualification. . . . . . . . . . . . . . . . . . . . 5 3.2 Equipment Affected . . . . . . . . . . . . . . . . . . 6 3.3 Qualification Update . . . . . . . . . . . . , . . . . 6

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3.4 Maintenance Record Review. . . . . . . . . . . . . . . 10 3.5 Qualification Status at time of AIT Review . . . . . . 11 3.6 Probable Consequence of Failure to Reanalyze . . . . . 11 4.0 Process Instrumentation . . . . . . . . . . . . . . . . . . 12 5.0 Temperature Effects on Containment. Structure and Accident Analysis . . . . . . . . . . . . . . . . . . 12

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5.1 Structural Effects. . . . . . . . . . . . . . . . . . . . . 12 5.2 Accident Analysis . . . . . . . . . . . . . . . . . . . . . 13 6.0 Persons Contacted . . . . . . . . . . . . . . . . . . . . . 13 7.0 Conference Call . . . . . . . . . . . . . . . . . . . . . . 14 8.0 Summary . . . . . . . . ., . . . . . . . . . . . . . . . . . 14 .

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l 1.0 General {

1.1 Background OnAugust3,1987,theSeniorResidentInspector(SRI)attheArktasas Nuclear One site informed NRC Region IV that containment temperatures at Unit I were unusually high. This information came to the attention of the SRI when he inquired into the circumstances that would require plant cooldown before temperatures could be lowered sufficiently in certain areas to permit personnel access. Region IV personnel followed up on this information by making a request during a conference call on August 7, 1987, that the licensee document the anomalous temperature information and provide justifications for continued operation as necessary. By letter dated August 13, 1987, the licensee supplied part of the requested informatio The licensee's updated Final Safety Analysis Report states that the average containment temperature is 110 F, with expectation that personnel access to most areas of the containment building could be permitted during plant operation. In actual fact, the information supplied in the August 13 letter stated that temperatures ranged from 103 F at elevation 348 foot to 165 F at elevation 486 foot. Immediately above "A" steam generator, the temperature was reported to be about 183 F. Thes temperatures were obtained from RTDs which are normally used for integrated

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leak testing of the containment buildin The letter did not adequately address the potential deleterious effects these high temperatures might have on instrumentation and other components, nor did it adequately explain why reactor operation could continue without some remedial action. As a result it was decided by NRC l to send an inspection team to the corporate offices and the reactor site

to investigate more thoroughly the effects of these high temperatures at l AND, Unit .2 Inspection Tasks

! Region IV formed an Augmented Inspection Team (AIT) on August 17, 1987.

l The team was supported by the Office of Nuclear Reactor Regulation (NRR)

which supplied two members, one from NRR and one from EG&G Idaho. The team was assigned the following tasks: Determine how long, to what extent, and under what conditions higher than expected building temperatures have existed; Determine why the excessive temperatures exist and what corrective measures have been taken by the licensee; Determine what specific measures the licensee has taken or plans to take to assure that ambient temperatures at the vicinity of environmentally qualified electrical equipment are established;

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f i-3- Identify electrical equipment that may have.already exceeded its qualified life or will need to be replaced or repaired prematurely; and Determine whether or not the high building temperatures can affect concrete properties or accident analyse Each of these tasks was essentially completed by the team. However,'the AIT investigation was terminated on August 20, 1987, and the inspection results reported herein are for an NRC special inspectio .0. Containment Temperature 2.1 History - Unit 1 Correspondence reviewed by the NRC inspectors between Bechtel and AP&L shows a history of containment temperature problems since 197 Efforts to eliminate the problem (i.e., modifications to the containment cooling system, insulation of previously uninsulated small diameter piping, insulation of support structures around the pressurizer and filling of holes and cracks in reflective insulation) apparently have been only partly successful in lowering the temperature to the stated nominal bulk design level of 110 Although the licensee has known since 1974 that some regions of the containment building have run considerably hotter than expected, no specific analyses were conducted until recently to ascertain the effects of the higher temperatures. Correspondingly, there is no evaluation with j respect to these higher temperatures from which to determine whether the 1 conditions prevailing constituted a potential unreviewed safety questio !

Finally no evidence could be obtained through review of correspondence since 1974 and through interviews of licensee personnel,'that the licensee had informed the NRC of the temperature situation in Unit 1 prior to August 1987. This is an apparent violation of 10 CFR 50.72(b)(ii)(B) ,

which requires that a condition that is outside the design basis of the i plant be reported (313/8729-01).

The failure to evaluate the effects of high containment temperatures on 1 structures, components and accident analyses resulted from failure by the j licensee to identify and control design interfaces and to adequately i

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coordinate participating design organizations to assure that the design bases were correctly translated and that deviations were controlle Furthermore, the licensee had failed to promptly identify a significant condition adverse to quality and to take prompt corrective action. This is an apparent violation of 10 CFR 50, Appendix B, Criterion III, Design i Control, and Criterion XVI, Corrective Action (313/8729-02).  !

i 2.2 Unit 1 Temperatures Unit 1 containment temperatures measured from RTDs installed for ILRT of the containment building are summarized in Table 1. These measurements were taken during the on site inspectio Figure 1 is the licensee's

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display lof Unit:1 temperatures as used in' updating the qualified life of

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electrical; equipment. The licensee is taking'further steps to confirm as

'necessary theitemperatures reached at specific locations. through .use of 37 ,

temperature. sensitive' tape:and other temperature sensor .

. Table 1

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. Measured Containment Temperatures August 19, 198 ANO - Unit l'

Temp.>'F TE # Location __

10 TE-2401 E1.341 Near Col. E-2'

111.4~ TE-2402- E1 341 Near Col. G-1

, L159.61 TE-2404 El.486 Near 180

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-16 TE-2405 El 486 Near 200 16 TE-2406- E1.486 Near'340*

12 TE-2407 El.381 Near' Col. H- .8 TE-2408 E1.381 Near Col. F-0 '

11 TE-2411- E1.340 Near Col. E-3

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151.0' -TE-2413 E1.405 Near Col. D- .4 TE-2414 E1.405 Near Col. H- .8 'TE-2415 El.428 Near Col. J-1 18 TE-2416 El.486 Near Col. D-2

.16 TE-2417 E1.486 Catwalk E-1 16 .TE-2418 E1. Catwalk H- .3 Unit 2 Temperatures

. Temperatures from RTDs installed in'ANO, Unit 2, taken during the NRC inspection,'are shown in Table 2. Although Unit 1 was the-object of the inspection, these data for Unit 2 were obtained so-that a determination could be made of whether a temperature problem affected Unit 2 also. It should be noted that these data show that no temperature problem is apparent for Unit ,y

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FIGURE 1 Reactor Building - Unit 1 Elevation - Temperature Profile

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TEMPERATUREREADINGS('F)

August 6, 1987

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Table 2 Measured Containment Temperatures August 19, 1987 ANO - UNIT 2 LTem F TE # LOCATION

123.6 F f2TE-8350 LE1.505 Above Polar Crane (South)

12 '2TE-8351 E1.505 Above Polar Crane (North)

112 TE-8352 E1.505 Above Polar Crane (South)

12 TE-8353 E1.482. At catwalk above north ladder-12 TE-8354 E1.482 Above Polar Crane track (NE)

12 TE-8355 E1.482 Above Polar Crane Track (SE)

123;2 2TE-8356 E1.482 At' catwalk west of south Ladder

'12 TE-8357 E1.482 At catwalk north of a/c ducts 12 TE-8362 El 434 East of Head Superstructure

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12 TE-8363 -E1.434 Top of sec. shield wall beside south .2 2TE-8364 E1.415 Inside a/c duct above equip. hatch 12 TE-8368 E1.415 Inside a/c duct above equip. hatch

'105.2- .2TE-8369 E1.395 At base of 2T2C 10 TE-8370 E1.376-65~

99.9- 2TE-8373 E1.357-270'

12 TE-8374, E1.357-27 .7- 2TE-8376 E1;365 Directly in front of equipment hatch 9 TE-8377 E1.345 Inside sec. sbiold 9 ,2TE-8378 'El.345 On sec. shield at foot of stairs 3.0.. Equipment Qualification ,

3.1 Ambient' Temperatures Used For Equipment-Qualification-The qualified life.of' safety-related, electrical equipment was based on 120 F, which was reportedly recommended by Bechtel as an average containment temperature. For purposes of determining qualified life, it was assumed by AP&L that this average temperature applied uniformly to all

, electrical equipment within the reactor buildin Temperatures recently measured by RTDs installed in the containment for integral leak rate tests (ILRT) ranged from 103 F to 183 F, depending on

' elevation and proximity'to heat generating equipment (i.e., pressurizer and steam generators). Temperatures from 21 of the 23 installed RTDs were obtained during the temperature survey. Accuracy of the RTDs was determined to be iS*at 32 F. The . calibration range was from 32 F to 200 F, and the acceptable use temperature of the RTDs was stated to be for a maximum temperature of 350 F. Calibration of the RTDs is routinely performed when the ILRT is conducted at approximately 5 year interval They are read from a' central location outside the containment on an as-needed-basis rather than continuously; therefore, there is no historical, temperature data while the plant is at powe _____._u__._._ _

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-6-In August 1987, to' correct the previous determinations of' qualified life-based on the assumed average containment temperature of 120 F the licensee'

approximated a containment temperature profile from the measured temperature data. It was found that 120 F enveloped the temperature of al1 EQ equipment below the 390 foot elevation and 150 F enveloped the equipment above the 390 foot level, with the exception of eight solenoid valves located directly above the steam generators,'which were at 180 F, and the PORV block valve which was at 160 .2 Equipment Affected

.The licensee conducted a review of equipment qualification and other documentation and identified the safety-related equipment shown in Table 3 to be above the 390 foot elevation, and thus affected by temperature levels higher than were used to qualify the equipmen .3 Qualification Update A summary of the licensee's analysis is shown in Table This Table shows the life predicted at 120 F, . life predicted at the higher level temperatt.res, and the life remaining at the higher temperatures. The analyses were conducted at 150 F for all items.except the high vent point solenoid valves (Items =c-f) which were conducted at 180 F, and the Limitorque operator.-(Item g) which was conducted at 160 The-field cables and interface connections (Items o, p, and q) between the cables and devices were also included.in the analyses. As shown by the

. years remaining column, no equipment had exceeded its qualified life; however, the pressure transmitter (Item a) and acoustic monitors (Item b)

had only-a short life remainin At the time of the inspection, the licensee had committed for contractor services to review AP&L calculations and to assist in the review and update of EQ documentatio The NRC will review these areas during a subsequent inspectic Only the methods used to justify the qualified life of the equipment based on thermal preaging tests were reviewed by the inspectors. The assumption was made, and verified verbally by the licensee, that the radiation, design bases events, and post-accident operation requirements which had been established for equipment qualification (120 F) of the containment were not affected. Much of the information provided during the inspection was ,

preliminary in nature and was handwritten or oral. The licensee stated '

that formal reports would be prepared to provide the bases and documentation for the informatio The results of the NRC inspection for individual items of equipment are summarized below. (These items correspond to those in Table 3).

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.. Rosemount Pressure Transmitter PT-2402 The. transmitter has very little' qualif_ied life remaining (0.25 years); it will be replaced as. soon as practicabl Acoustic Monitor Pre-Amplifier VBY100**

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The replacement requirement is more frequent than the normal refueling outage interval. The licensee' reported that other test data were being acquired to justify longer qualified life. Unless the demonstrated qualified life can be extended to at least one refueling outage . interval, the licensee would need to replace these preampli.fiers between refueling outage c-f. Target Rock Solenoid Valves (High Point Vents)

The previously predicted qualified life of these solenoid valves was in excess'of 40 years. The recalculated qualified life of 6 years is considered acceptabl . Limitorque Operator PORV Block Valve CV-1000 The licensee reported that the qualified ' life at the higher temperatures was established by a report by Scheider Engineer Due to the limited time of the inspection, the NRC inspectors did not review this repor .

During the licensee's investigation of the incorrect operation of the pressurizer relief valve PSV-1000, insulation damage was noticed on the cable to-the Block Valve CV-1000. The cable has been replaced

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but the qualified life, even at the higher temperature, was stated to be 40 years. Also, internal wiring of the valve operator had insulation damage even though the qualified life was reported as 35 years at the higher temperature. The licensee is well aware of the discrepancy and has established an inspection schedule for each refueling outage and cable replacement every 5 year The motor on the Limitorque valve operator was replaced. The replacement motor has Class RH insulation and is qualified. The licensee should verify that the replacement was needed because of a motor insulation failure; failure for some other reason could introduce additional questions as to the validity of the qualificatio Rosemount RTD TE10 (Hot Leg Temperature)

Material in the connector head was identified as the limiting material for qualified lif The licensee's calculations used the higher ambient temperatur The connector head, however, would be

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-8-expected to be at a temperature higher than ambient due to conduction from the primary coolant. Also, the calculations included a determination of-the equivalent normal operating time to account for the accident condition.- This calculation raises the question of whether the RTD has been tested under accident conditions. Lack of an accide a test would not be in agreement with the general oral report'from the license Verification is needed that conduction was '

considered in establishing the qualified life and that the detector-has tested under accident conditions, Reactor Vessel Level Detector UE11 These detectors could have the same concern of conduction as described for the Rosemount RTDs. Verification is needed that the 170 F temperature given in the qualification report is not for some lower' ambient temperature (lower than 170 F) with a margin added for conductio Acoustic Monitor Sensor VBE 100 The licensee reported that these sensors were not susceptible to thermal agin k&l. Conax RTD TEll- (Hot Leg Temperature)

These detectors could have the same concern of conduction as described for the Rosemount RTDs. Verification is needed that the 250 F given in the qualification report is not for some lower ambient temperature (lower than 250 F) with a margin added for conductio Hydrogen Recombiner RE 8060 A concern was identified about the qualification of the cables and what was meant by " extrapolation." The licensee presented the Okonite report that showed " extrapolation" of the curve for 40 percent elongation to the 40 year life. The licensee also verified that the test specimen used for the accident simulation was preaged to at least the equivalent of the 40 percent elongation curve. This explanation resolved the concer Radiation Monitor RE 8060 The cable to this component is a Rockbestos coaxial cable. The Inspection Report for the July'14-18, 1986, Equipment Qualification inspection identified this type of cable as not having a similarity statement justifying that the tests on similar cable applied. The licensee reported that a similarity statement had been received from l

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b -9-o q. Cable, Tape, and Splices i

The evaluation of many of the cables relied on the condition that

'there is little or no internal heating from conducting curren The justification given by the licensee was that all the components located in the region with above design temperatures are instruments that carry very low currents or that carry a power current for only a short time. Cables other than those for the identified components may pass through the high temperature areas. The licensee should determine if any other cables are routed through these areas. If other cables are routed through these areas, the validity of the assumption of little or no internal heating should.be verifie Degradation of cables in the his. temperature areas has been observed by the licensee, and provisions for periodic change out have bee made. A concern similar to the one discussed under the Limitorque Operator, Item f above, exists for the other cables. The Arrhenius calculations indicate that all cables should be qualified for 40 years-even at the high temperatures; yet insulation damage has been observe The investigations to resolve this discrepancy should continu A concern was also identified with regard to instrument. loop error analyses. One of the parameters in these analyses is the error from leakage current in the cables, .especially during a postulated accident. Confirmation ~is needed that efforts to verify cable life will also include maintenance of the required instrument loop accurac _ _

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TABLE 3 Qualified Life of Limiting Interface / Component in Years

. Tag Number Function Connection Previous Revised Remaining PT-2402 RPS RB Pressure Conax Connectors, 10 .25 Trip Raychem Splice Conax Sealsd Cable N/A 40- 5. 8 3.7S'

Connectors Interface

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Conax Quick Disconnects N/ '40 40 Disconnects VBY 1000A Acoustic Monitor Raychem Splice 4 1.04 VBY 1000B PORV Position VBY 1001A (Pre-amp)

VBY 1001B VBY 1002A VBY 1002B SV 1071 ZS 1071 Reactor High Point Direct Connection >40 6 SV 1072 ZS 1072 Vent-SV 1073 ZS 1073

'SV 1074 ZS 1074 SV 1091 ZS 1091 Hot Leg High Direct Connection >40 6 SV 1092 ZS 1092 Point Vent SV 1093 ZS 1093 SV 1094 ZS 1094 SV 1077 ZS 1077 Pressurizer SV 1079 ZS 1079 High Point Vent Direct Connection >40 6 )

' SV 1081 25 1081 Hot Leg High Direct Connection >40 6 2. 2 j SV 1082 ZS 1082 Point Vent 3 SV 1083 ZS 1083 1 SV 1084 ZS 1084 )

Terminal Block Cable Connections N/A >40 60 40 j u CV 1000 ZS 1000 PORV Block Valve Direct Connection >40 35 27

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Qualified Life of Limiting Interface / Component in Years Tag Number Function Connection Previous- Revised Remaining

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. TE 1013 RCS hotleg Temp- Direct Connection >40 >40 27 TE 1012 Monitor.(RTD)

TE.1040 TE 1041 UE'1188 Reactor Vessel ICC Vendor Supplied 30 30 29-UE 1187 Device (Temp, Level) Connector VBE 1000A Acoustic Monitor Raychem Splice >40 >40 33 VBE 10008 PORV Position VBE 1001A (ACCELMTR)

VBE 1001B VBE 1002A VBE 1002B k. -TE 1111 RCS Hotleg Temp Raychem Splice >40 >40 38 TE 1112 Monitor (RTD)

TE 1139 TE 1140 TE 1189 Ref Leg Hotleg ICC Conax Disct, >40 >40. 39 Raychem Splice TE 1190 Level XMTR (RTD) Sealant TE'1191 TE 1192 TE 1197 TE 1198 :j M55A Hydrogen Recombiner Okonite Taped >40 >40 40 M55B Splice RE 8060 Hi Range Contnmt Rad Raychem Tubing on >40 >40 40 Monitor Connector (Vendor) ,

J Raychem Splice Cable Splice with N/A >40 >40 40 j Heat Shrink Tubing I l Okonite Tape Tape Over Cable N/A >40 >40 40 1 Splice ' Cable Transmit Power, N/A >40 >40 40 Control and Signals ,

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-10-3.4 Maintenance Record Review The NRC inspectors sampled maintenance records for the following equipment in order to determine if trends of early failures, due to high temperatures, could be established:

• Target Rock Solenoid Valves

• - Limitorque Valve Operator (PORV Block Valve)

• Acoustic Monitors

• Rosemount Pressure Transmitter The only adverse trend was associated with the Limitorque Valve Operator which had the following identified problems: February 1980 Contacts on both torque switch and limit switch were severely corroded and prevented valve operation. The switches were replace March 1984 Valve would not operat The motor and torque switch were replace December 1984 Field cable to valve was found to be badly deteriorate The wiring was replace (Month not recorded by the inspector)

Motor lead wire was grounde Condition was repaired, September 1986 (During MOVATS Testing) '

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• Internal wiring was badly cracke Wiring was replaced with Rockbestos SIS wirin * Cracks and erosion were observed on limit switch rotors and finger boards. The limit switch was replace * Conduit insulation was damage Conduit was replace * Grease was hardened in the limit switch gear box. The 1 limit switch was replaced and the gear box repacked with greas * Grease was hardened in the main gear bo The gear box was flushed and repacked with new greas :

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Table 3 shows this Limitorque operator to be qualified for 35 years'

at 160 F. However, the operator is presently scheduled to be inspected each refueling outage, and the field! cable is now scheduled to be changed.each 5 years. The additional hee. ting effect of the limit switch compartment space heater was not considered in the-analysis, and is considered an unresolved item for Limitorque operators, as identified in previous NRC Inspection Reports 50-313/86-23 and 50-368/86-2 .

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3.5 Qualification Status at Time of NRC Inspection Although the licensees reanalyses, reflecting the effects of higher temperatures than used in the licensees original equipment qualification program, were preliminary in nature, they were performed using the conservative assumption that the equipment in question was continuously exposed to the temperatures discussed in Section ?.1. It was the judgement of the inspectors that the revised qualified life for each component, shown in Table 3, is sufficiently accurate and/or conservative to support the conclusion that no currently installed component had exceeded its '

qualified life as of August 20, 1987. Nevertheless, it was also concluded that becauce of the short remaining qualified life for some components (e.g., items a and b of table 3), preventive mair,tenance and/or replacement would have to occur during the mid-cycle outage scheduled for October 1987 in order to assure continued qualificatio .6 Probable Consequences Of Failure To Reanalyze As stated previously, the SRI brought the Unit I containment temperature disparity to the attention of NRC in early August 1987. Had this not occurred, the licensee would not have updated the thermal aging analyses, as reported herein; and no one would have been aware that the qualified life of a number of electrical components had been severely shortened by exposure to higher than expected temperatures. The probable consequence would have been that plant operations would have continued with the expectation that the replacement schedule for electrical components derived from the original analyses was still adequate. Under these circumstances, there is little doubt that the plant would have operated in the near future with a number of components in an unqualified stat From installation and replacement schedules, the inspectors did conclude that the reactor did operate with unqualified acoustic monitor  ;

preamplifiers (Table 3, Item b) in place. This resulted from the fact that the qualified life of these components had been reduced from 4 years to 1.04 years. These components, installed in March 1984, were rep' aced in January 198 Thus they were in use for approximately 21 months beyond their qualified life. This is an apparent violation of 10 CFR 50.49(g)

which requires that electrical equipment important to safety used in a plant after November 30, 1985, be environmentally qualified (313/8729-03). ,

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The inspectors did not have enough time on site'to review-installation records of other components which'had significantly shortened life. Thus,

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beyond their; qualified. life-(e.g., PT-2402 and various solenoid ~ valves)..

14.0 Process Instrumentation Early in its onsite investigation,: the : inspection . team' requested that' the licensee ~ provide.information on the effects.of the'high containment temperatures on the operability and accuracy,of instrumentation relied upon by the operators to control the reactor and its associated system ; This- request covered :all key process instrumentation whether or not..it was required to be environmentally qualifie . At the.'close of the inspection on August 20, 1987, the licensee reported that the' requested.information had been assembled-from plant records, and that the' results indicated that the operability of process instrumentation was not affected and that accuracy (e.g., steam generator level) was not affected_by more than one percen The written summary was not made available to the team while it was onsit ~

However,;it was. discussed at a meeting between AP&L and NRR at NRC .

Headquarters.on August.21, 198 l 5.0 Temperature Effects On Containment and Accident Analysis-

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5.1' Structural Effects From the; standpoint of potential effects of higher than expected

- temperatures on theEcontainment building,- it was apparent from review of-RTD temperature records'and' discussion with the licensee that the principal areas of concern were the steam generator cavities and the containment walls'and dome above an elevation of about 400 foot. All other' locations appeared to be within normal operating design temperatures for concrete and structures. The suspect areas could see temperatures in the range of 150 to 165 F. At such temperatures, concrete strength .;

should not be significantly affected by dehydration or bound water los l Heating of the containment liner in the upper side wall and dome areas would increase compressive stresses and buckling tendenc When questioned

. about this, the licensee stated that routine inspections during refueling ,

outages confirmed that buckling had never occurred, thus demonstrating that liner anchorage was adequat While the information obtained while the inspectors were onsite was incomplete, it did lead the team to conclude that the effects of elevated

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temperatures on containment and internal structures has been slight and L that structural integrity is unlikely to have been adversely affecte ]

The licensee stated that more detailed analyses would be performed to

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5.2 ' Accident Analysis

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', At the. time' of the-NRC inspection visit. AP&L had performed no .large break LOCA -

analyses reflecting the-higher containment temperatures beyond those. older.-

l analyses cited.in its August 13, 1987, letter. Therein.it was stated tha parametric ^ analyses with.Bechtel's COPATTA Code (Revision 3) as set forth ' , .

-in~ BN-TOP-3 indicate .that raising' containment bulk temperature from 110*F, to 150 F would increase ~ peak pressure by about 1 psig and peak temperature by about 3 F. These increased values leave peak pressure and. temperature

- within the containment' design values of 59 psig and 286 F. respectivel These' older analyses, together with more recent analyses'done by Bechtel

..for other plants _and discussed with AP&L suggested that the increased temperatures.for ANO, Unit 1 would have only a small and unimportant effect on the capability of the building'to retain its integrity in the'

event of'a bounding LOC !Nevertheless,.it was noted that all of those analyses:were indirect.. The'

licensee is performing a plant specific analysis for Unit 1, using the Bechtel COPATTA code (Revision 4). An analysis is also being done to confirm

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that'an excessive negative containment pressure would not'be developed after cooldown following a LOCA. -The results of these analyses were not available prior to the end of the team's effort .0 Persons Contacted Licensee W. Cottingham, Supervisor', I&C Engineering.

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D. Williams, Electrical' Engineer G.'Dobbs, Supervisor, Electrical Engineering .

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C.' Turk', Supervisor, Nuclear Engineerin R. Barnes, Manager, Electrical Engineering j R. Oakley, Supervisor; I&C Engineering D. Howard,'Special~ Projects Manager D. Baxter, Plant Licensing Engineer J. Levine, ANO Site Director l D. Lomax, Supervisor, Plant Licensing

.E. Ewing,. General Manager, Plant Support P. Michalk, Plant Licensing Engineer P.7 Jones, Superintendent, Site Maintenance A'. Wrape III, Manager, Nuclear Services L

x" NRC C. Harbuck, Resident Inspector 1 W. Johnson, Senior Resident Inspector j i

In addition, a number of other AP&L employees were interviewe ;

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'7.0 Conference Call On August'20, 1987, a conference call tock place, involving AP&L management, NRR and RIV management, AIT team members, and the NRC resident

inspector During this conference, NRC staff informed the licensee of its principal concerns regarding the licensee's evaluation of the Unit I containment temperature problem and it was agreed that AP&L management would meet with NRR management on August 21, 1987, at NRC headquarter Since it was concluded by the AIT (in conference with Region IV managemen immediately after the conference call described above) that the containment temperature situation posed a potential unreviewed safety question, the Regional Administrator passed lead responsibility to NRR and terminated the AIT investigation. The team suspended its activities but remained onsite through August 21, 1987, to assemble documents provided by the licensee. In light of the fact that the AIT effort had been terminated, it was decided that the information developed would be reported as an NRC special inspection effor .0 ' Conclusions The licensee has been aware that a temperature problem has existed in Unit 1 since 1974. Higher than expected temperatures exist.in the steam generator cavities, including the vicinity of the pressurizer and in the upper part of the reactor building including the dome area. The high temperatures exist at all times during plan operations and under hot shutdown condition The high temperatures exist because reflective insulation on the .

steam generators and pressurizer' proved inadequate to assure design heat load. Efforts to reduce the heat load through improvements to the insulation have been largely unsuccessful. The high containment temperatures continue to exis The licensee did not report the temperature problem to the NRC until August 1987, after the SRI became aware of its existenc There is no evidence that the licensee conducted an evaluation of the effects of the high temperatures on equipment operability or made an attempt to update the FSAR prior to August 198 Because of inadequate internal communications, AP&L based qualification of electrical equipment (aging) on 120 F without taking into account the actual temperatures, which in some cases were considerably higher than 120 .

i None of the currently installed equipment had gone beyond its qualified life at the time of the inspection, nor would it before the planned mid-cycle outage in October 1987, when remedial actions were to be take .

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-15-g. Prior to replacement in January 1987, preamplifiers for the PORV acoustic monitors had exceeded their qualified lif Some other i equipment may have gone beyond its qualified life.

l h. Internal wiring and cables for the Limitorque operator on the PORV block valve has needed replacement, apparently because of thermal damag The licensee is taking steps to develop temperature

information at or near EQ components, as necessar i. Subject to confirmation, the effects of high temperatures on containment integrity and accident analyses appear to be negligibl j. Unit 2 containment temperatures do not represent a proble .

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