ML20236A758
| ML20236A758 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/15/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236A738 | List: |
| References | |
| NUDOCS 8710230089 | |
| Download: ML20236A758 (9) | |
Text
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4 s na,u UNITED STATES if cf()
NUCLEAR REGULATORY COMMISSION j
W ASHINGTON, D. C. 20555 s
-.p THE OFFICE OF NUCLEAR REACTOR REGULATION l
EVALUATION OF THE JUSTIFICATION FOR CONTINUED OPERATION l
RELATED TO HIGH REACTOR BUILDING TEMPERATURES FACILITY OPERATING LICENSE NO. DPR-51 ARKANSAS POWER AND LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT NO. 1 DOCKET NO. 50-313 I
~_. 0 INTRODUCTION l
On August 3,1987, the Arkansas Nuclear One resident inspector raised questions regarding the reactor building temperatures.
Temperatures i
ranged from 103 F to 183 F with an average of 140 F.
This was compared i
to an average normal operating temperature of 110 F assumed in the FSAR and a temperature of 120 F used for equipment qualification.
A confer-j ence call was held on August 7 with the licensee and the staff to discuss the issue.
The conference call was followed by a submittal dated August l
13, 1987.
Staff review of the submittal determined it to be inadequate and an Augmented Inspection Team was sent to the site to review the licensee's supporting documentation and determine if any specific technical j
safety issues existed.
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On August 28, 1987, Arkansas Power and Light (AP&L, the licensee) delivered a Justification for Continued Operation (JCO, dated August 27, 1987).
The staff performed an initial review of the document at that time and discussed it with the licensee staff.
Following the discussions the staff concluded that there were no immediate safety concerns that would preclude the plant from continuing to operate.
This Safety Evaluation documents the s'.aff review.
It also includes commitments made by the licensee during tle discussions.
2.0 REACTOR BUILDING AIR TEMPERATURE BACKGROUND
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During startup testing (hot functional) in early 1974, the licensee first i
noted that the insulation surface temperatures on the Reactor Coolant I
System (RCS) components and the air temperatures in the reactor building were higher than expected.
Since 1974, the licensee and licensee contrac-tors have investigated the causes of the elevated temperatures.
An evaluation of the reactor building heat removal by the main chillers determined that the heat load during normal operation was about 5 million J
BTU /hr versus the original calculation of 2.9 million BTU /hr.
Further investigation to determine heat losses from the steam generators disclosed i
insulation design and installation deficiencies.
Improvements of RCS insula-
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tion and acid cleaning of the main chillers reduced the reactor building j
temperatures somewhat but they remained higher than expected.
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Based on measurements taken in August 1987, tne licensee presented maximum j
expected reactor building temperature profiles expected during normal j
operation.
Outside of the steam generator cavity, the temperature varied h0 871015' P
0500 3
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from 110 F at elevation 330 ft.' to.162 F at elevation 410 ft. and'above.
j In the. steam generator cavity, the temperature range was 103 F to 183 F.
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The measurements-were compared with other temperature data obtained-in lj 1975,1978,-and'1987 and it was determined _that it represented the most' 4
severe conditions.
- 3. 0 EVALUATION 3.1 Accident Analyses
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i The licensee performed a LOCA containment analysis using.the computer j
code COPATTA and an initial temperature of 150 F.
The effect on.the q
peak containment pressure was negligible.
There is about a'3 F in-i crease (from 280 F'to 283 F) on the peak containment temperature.
The.
j containment design pressure (59'psig) and temperature (286 F) were not
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exceeded.
The licensee further evaluated the impact of the slight in-J crease of reactor-building temperature following a LOCA to~be within the margin of environmental qualification of all the safety-related equipment.
The licensee also evaluated the effect of operating with elevated reactor building temperatures on accidents where core flooding tank
. injection was calculated to occur.
As a' result of the increased reactor building temperature, the temperature in these tanks could.
be as high as 150 F, as opposed to the.110 F'used in the original analyses.
The licensee concluded that although the core. flooding j
tanks' inject for. a main steam line break, all regulatory criteria are satisfied prior to this injection and the analysis conclusions are not affected by the increased temperature, j
For a LOCA, the licensee concluded that-since the fluid in the downcomer, during the refill and reflood phases of a LOCA, would be saturated, the increased temperature would not impact.the core. refill and reflooding rates.
This was discussed-in more detail with the staff on August 28, 1987.
The staff concludes.that the'effect of.
the elevated building temperature on a LOCA would be minimal and does not pose an undue risk on the public health and safety.
The licensee also performed an analysis'of reactor building negative pressure caused by inadvertent actuation of.the reactor building spray system.
The original analysis referenced in the FSAR addressed a sealed reactor building with an internal temperature of 110 F and subsequent cooldown to an internal temperature of 50 F.
The resultant dif ferential pressure was determined to be. 2.5 psi.
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To evaluate the potential impact of the elevated _ temperature, addi-tional analyses were performed.
In the base case analysis an initial temperature.of 150'F and an initial pressure of 0 psig were used.
Relative humidity was assumed to be 100%. -The minimum spray tempera-ture allowed by Technical Specifications,.40 F, was conservatively
-i used as the post-event reactor building temperature, f
The base-case analysis results -showed a. maximum differential pressure j
across the reactor building wall of -5.60 psid. Although this value j
is in excess of the criginal analysis result of -2.5 psid.the licensee performed evaluations which shewed the reactor building and i
plate liner could withstand the -5.60 psi differential.-
l An additional analysis was performed with the same initial conditions 1
as the base case, except that initial pressure was. assumed to be -2.7 psig. This is the minimum value of reactor building pressure allowed
'3 by the Technical Specifications.
The results for this case showed a final pressure differential of -7.80 psid.
Further evaluation showed that this pressure would not cause failure of the reactor building or plate liner.
The steam generator sub-cupartment differential pressure was reanaly-
-l zed using the computer code C0PDA and assuming the initial temperature of 180*F compared to 110 F, there is a small increase from the FSAR-value of 16 psid to 17.1 psid.
Based upon the above analyses, it is concluded that the reactor building can withstant an inadvertent actuation of the Reactor Building Spray System :nd associated cooldown transiert.
3.2 Reactor Building Structural Intevity According to Section CC-3440 " Concrete Temperatures" of the Code for Concrete Reactor Vessels and Containments,Section III, Division 2, ASME Boiler and Pressure Vessel Code, the existence of temperatures higher than 150 F for normal operation or any other long term period is prohibited without justification.
High temperature could cause buckling of the steel liner, deteriorate concrete material strength, accelerate concrete creep and tendon relaxation.
With respect to the subject of concrete materials deterioration, the licensee had performed an inspection on the exterior surface of the containment on August 18, 1987. and found no sign of deterioration and proposed, in the JC0, to conduct a rebound hammer test for the interior surface of the containment to further. support his inspection finding.
This is acceptable to the staff.
In the area of accelerated prestressing loss, it is kncwn that the loss of prestressing will be greater when the temperature is higher than the original design temperature because both creep in concrete
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and the re. laxation in prestressing tendons 'are known to be function's
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of temperature.
However, the; amount of additional prestressing loss a
H due to the: reported higher temperature is difficult to estimate.
analytically.
During the August 28, 1987. meeting the licensee:showed the staff examples of the records of prestressing tendon surveillance-I which were not included in the JCO.
The prestressing loss derived from these records included 'Le effects of high temperature on:the.
containment.
Based on those-records, the staff was able to judgey that the prestressing force currently believed to be available in/the.
containment is higher than the minimum required design prestressing:
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force and would remain in the same-situation at the next surveillance l
time, which is-scheduled.in the Spring of 1988.
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.l Based primarily on the records'shown to the staff it is concluded ~
that the ANO-1 containment'and its internal structure would be safe' for continued operation.- However, it is requested that'the prestressing tendon surveillance records be officially:sent.to:the staff.' :The staff also requests that it be' informed'at least one month Defore-1 l
the licensee conducts the next prestress tendon surveillanceLso-L that the staff may witness the surveillance work and verify the.
i availability of adequate prestressing force in the' containment-1 structure.
The staff may-require additions' to'the surveillance -
program after it has reviewed the confirmatory information; 1
l With respect to the subject of liner integrity, the licensee indicated that its previous inspections.of. the liners had found no signs:of deterioration, and the conservatism of ANO-1 liner design was docu-mented in the JCO.
The staff agrees with the licensee that the liner-
.I integrity had not and would not be breached due to the high temperature in the containment ~.
Nevertheless, the staff requests the' licensee to provide. additional information'on the liner: integrity by comparing:
the calculated liner strains associated-with the.high temperature-to the design allowable liner strains committed in the FSAR.
With respect tn the effects of high temperature on. internal structures,.
the licensee stated that the internal structures were designed to a higher temperature than they actually had experienced, and, therefore,.
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the internal structures were not adversely affected by the observed high temperature.
l 3.3 Containment Sump Debris 1
The licensee also addressed the issue of containment sump debris and the potential of sump blockage.
By comparing 'the specifications of the coating and insulation used, the licensee found that at a 183 F service temperature, the coating and insulation in containment would not be sufficiently degraded to generate additional debris.
The-staff concurs with the licensee on its conclusion-that there.is
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no significant inoact to the potential sump blockage.
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3.4 Piping and Piping Supports AP&L performea a comprehensive evaluation.of tne effects of elevated-temperatures within the reactor building on safety related piping and piping supports.
Ninety-eight piping systems are identified as possibly being affected by the elevated temperatures. 'The temperature
.of the highest; elevation at which piping was located was 183'F.'
However, most of the reviewed piping was located at elevations with temperatures of 160 F-or less, A line-by-line evaluation was performed on those piping systems and
.five were found to be affected by the elevated temperatures.- All systems were evaluated for piping expansion, stresses, and changes in allowable stress limits for certain stainless steel pipe materials.
The evaluation demonstrated the acceptability of the piping and pipe supports for pressure retaining capability, and for meeting ASME Section III or ANSI B.31.1 Code Stress allowables. The staff finds the evaluation by AP&L of the piping in the reactor building accept-able.
During the discussions, the licensee committed to provide some clarifying information on the piping evaluations. This information is as follows:
Hydrogen Purge lines (HBC-1 & HBB-15) were' reviewed concurrently.
A calculation has been performed but apparently further review i
is being undertaken. The licensee has comitted to provide the conclusion of this review.
For piping HBD-14 and HBD-20, no thermal analysis.was performed for the 2A & 28 cooler nozzle 3.
However, a comitment was made.
to perform an inspection of these nozzles to validate'the
.I approach used for estimating the thermal stress increase in the nozzles of the VCC-2C & 20' coolers. AP&L committed to provide the conclusion of this inspection to the staff.
3.5 Environmental Qualification of Equipment The information provided to the staff included a list of all equipment within the scope of the environmental qualification program and located in the reactor building.
In addition, the results of a re-analysis of all equipment on the list were provided.
That informa-tion provided the basis for the discussion with the licensee.
The major thrust of the reanalysis was to determine if the-qualified life of the equipment of interest has been exceeded as a result of the elevated temperature in the reactor building.
The licensee's reanalysis demonstrated that, although the
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I i elevated temperatures resulted.in a shorter qualified life for a j
significant number of environmentally qualified components, none of the qualified lives of those components had been exceeded.
However, the qualified life of 'one component (a pressure transmitter model 1153ADSRB, manufactured by Rosemount) will' expire approximately 100 days from the date of the reanalysis.
The licensee stated that this component will be replaced during the mid-cycle outage of j
October 1987.
The shortest remaining qualified-life for any of the
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other components was 1.3 years.. All components in this Category will j
be replaced during the refueling outage scheduled for September 1988.
(1.08 years from the date of the reanalysis).
As a result of the review of the informatio L provided by the if censee, j
the staff finds the reanalysis acceptable and concurs with the replacement schedule.
To maintain compliance with 10 CFR 50.49, the.
licensee must update the ANO-1 EQ program.to reflect the results of the reanalysis.
i 3.6 Instrumentation inside the Reactor Buildino The staff was concerned that the elevated ambient temperatures inside the reactor building would change instrumentation drif t characteristics, i
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As a result of this concern, AP&L performed a systematic review to evaluate the potential impact of the elevated reactor tuilding j
temperature on both safety related and nonsafety-related instruments-tion including all the input variables associated with the integrat-ed control system /Non-nuclear instrumentation (ICS/NNI). AP&L provid-l ed a component specific listing of all instrumentation within the re-actor building.
This included transmitters, detectors, (RTDs, TDs, etc.), gages, switches, and signal: conditioners. The list orovided j
the primary design information about the components:and included:
q Component Tag Number j
Component Description l
Manufacturer /Model Number l
1 Component Location (including elevation) l 1
IEEE System Identifier j
QA Category (Safety Related, Non-Safety Related, Fire l)
Protection Related)
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EQ Determination The Design Temperature Rating and the Component Temperature Error I
were primarily obtained from vendor supplied documentation. When such data was not available, vendor representatives were contacted.
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L i_i____.._____._______..___________________
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c, 1 to provide assistance and additional informatier.
The component
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I specific ambient temperatures.were then obtained by evaluating the location of the devices against the' reactor building temperature profiles and rounding up the temperature-to the next highest decade.
The design temperature ratings were compared.to.the ambient tempera-tures and any incremental error introduced by the elevated temperature y
was assessed. This also included loop error analysis if warranted.
The impact of the elevated temperatures _on cabling systems and components was analyzed generically with respect to its effect on various ins;rumentation loops.-
The puduct of this effort was a comprehensive data file of all instrumentation located in the reactor building. This data. file was, the basis for the evaluation of the elevated ambient temperature impact on instrumentation.
Based on the review and discussions with licensee, the staff concluded that the instrumentation accuracy was mostly unaffected by the higher temperature with over'650 instruments having no significant additional-measurement error.
For the relatively few instruments which did encounter minor error ~ changes, the temperature changes were.found to not prevent the instruments from meeting their_ functional requirements.
AP&L stated'that there was sufficient' margin in the ANO-1 setpoint methodology to account for the increased drift caused bylthe temperature differential.
Staff review of the information led.to the conclusion that the setpoints were unaffected by the increased. temperatures and that con.inued operation was acceptable. 'However, as a confirmatory item, the licensee was requested to submit additional details' regarding the few instruments where accuracy was~ affectec by the higher tempera-tures. These details, as a minimum, should include-the channel identification, the total amount of accuracy that was affected by the higher temperatures, the original temperature drif t, the new tempera-ture drift, the original margin, and the new margin.
3.7 Non-E0 Electrical Equipment inside Reactor Building The licensee also reviewea electrical equipment, motors, contactors, relays, etc., for adverse impacts resulting from the elevated temperatures. The approach used was similar to that described in Section 3.6.
When equipment ratings were exceeced, maintenance' history was revieweo to determine if there were problems attributed to the higher temperatures.
The licensee stated that there are five RCP auxiliary pump motors that are exposed to ambient conditions outside_their rated temperature limits, but are intermittent in operation,'or only operate during outages. Although a review of the motor design and duty cycles has determined that a 40 year life for the insulation (class "B") can be expected, the preventive maintenance-program inc_ludes pump and motor scheduled replacement every six' years. -The licensee should confirm that the insulati3ns in the replacement pumps will be up graded to ciass "F" as well as the insulation _in any other replacement motors in the Reactor Building.
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.. The-staff concluded that there is no significant impact on the: safe operatior of.the plant with the non-EQ electrical. equipment l
operatinc at elevated ambient temperatures as detailed in the AP&L l
submittal. However AP&L was requested to confirm in writing that J
.The Maximum Ambient temperature referred to in. Appendix II.D.2-2 of the JCO is the Ambient' Design Limit temperature.
For rotor replacements in the Reactor. Building the new electric '
i moters will.have Class "F" insulation. -
3.8 Comitments and Confirmatory Items This section summarizes' licensee commitment and confirmatory items listed in Sections 3.1 through 3.7.
Some of the items came about during discussions between the licensee and staff. Others are i
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commitments made by the licensee in the JC0 and repeated in this SE for emphasis.
It should be noted that all comitments made in the -
l JC0, whether repeated in the SE or not, are expected to be fulfilled j
by the licensee.
The suninary list of commitments and confirmatory items is as follows:
Corcuct rebound hammer test for interior surface of the l
containment.
I Serd prestressing surveillance records to the staff.
Licensee to inform the staff one month before it conducts the t
next prestress tendon surveillance.
Provide additional information on the liner integrity by comparing the calculated liner strains asociated with the high temperature to the design allowable strains committed in the'FSAR.
Hydrogen Purge Lines (HBC-1, HBC-2,.HBB-15) were reviewed concurrently. A calculation has been performed but apparently further review is being undertaken.
The licensee is to provide the conclusions of this review.
For piping lines HBD-14 and. HBD-20, no thermal analysis was performed for the 2A & 2B cooler nozzles.
However a commitment was j
made to perform an inspection of these nozzles to validate the approach used for estimating the thennal stress increase in the nozzles of the VDC-2C & 2D coolers.
AP&L ccmmitted to provide the results of this inspection-to the staff.
Replace Rosemount transmitter during mid-cycle outage.
During the Fall 1988 refueling outage, replace all components with a remaining life of 1.3 years.
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. Submit additional details regarding the few' instruments where accuracy was affected by. higher temperature.
Include at least:
channel identification amount of accuracy affected by the higher temperature
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the original temperature drift
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the new temperature drift
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the original margin the new margin
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Confirm that the Maximum Ambient temperature referred to'in Appendix II.D.2-2 of the JC0Lis the Ambient Design Limit 1
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-temperature.
Confirm that' for motor replacement in the Reactor Building,'new -
electric motors will have a Class "F". insulation.
Licensee to conduct performance tests during the mid-cycle outage J
on building coolers to verify that the coolers have not degraded.
1 In order to maintain' compliance with 10 CFR 50.49, licensee is to update EQ program to reflect the results of the reanalysis.
4.0 CONCLUSION
Based on the-review of the licensee's Justification for: Continued Opera-tion (JCO) and the discussions held on August 28, 1987, the staff finds the licensee's evaluation and analyses of the affectc of elevated reactor i
building temperatures to be acceptable.
Therefore, subject' to fulfillment of the commitments included in the ' licensee's JC0 and those additional' 1
ones listed in Section 3.8 of this evaluation, the staff concludes that the ANO-1 rector building, internal structures, and components contained' therein are safe for continued operation.
It is noted that the licensee has a study underway which will include development of a long term action plan for the reduction of the ' ambient.
reactor building temperatures during operation. The plan'is expected to be completed in January.1988. The staff requests.that the plan be submitted l
for review.
Dated: Qctober 15, 1987 Principal Contributors:
M. Hartzman C. Li C. Liang J. Ma J. Mauck S. Saba H. Walker 1
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