ML19320A269
| ML19320A269 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/11/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19320A261 | List: |
| References | |
| NUDOCS 8004210556 | |
| Download: ML19320A269 (10) | |
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t SAFETY EVALUATION BY THE OFFICF. OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 12 TO FACILITY LICENSE NO. DPR-51 ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE - UNIT NO. 1 DOCKET NO. 50-313 1.0 INTRODUCIION By letter dated April 20, 1976, Arkansas Power & Light Company (the licensee) requested an exemption from the requirements of 10 CFR Part 50, Appendix H, Section II.C.2 to. permit operation of the Arkansas i
Nuclear One - Unit No. 1 (ANO-1) facility for the remainder of Cycle i operation with the reactor vessel surveillance specimens removed from the vessel.
By letters dated May 19, June 4, and June 10, 1976, the licensee requested changes to the Technical Specifications appended to Facility License No. DPR-51 for ANO-1.
The changes would:
1..
Revise Specification 4.2.7 to comply with the requirements of Appendix H to 10 CFR Part 50 and to permit operation for the remainder of Cycle 1 with the reactor vessel surveillance specimens removed.
2.
Revise Specifications 3.14 and 4.12 to permit reactor operation using charcoal filters in the Hydrogen Purge System that do not meet the requirements of Amendment No.10 to the facility license issued February 18, 19 76, and 3.
Revise Table 4.1-3, " Minimum Sampling and Analysis Frequency",
j to require an increased frequency of gross radioiodine determinationc '
j during the remainder of Cycle 1 operation and reporting of signifi-l cant increase in gross radiciodine concentration.
i The licensee also submitted a report by letter dated June 1,1976, describing the surveillance specimen holder tube (SSHT) failure at ANO-1, and subsequent inspections, cleanup, and repairs.
This safety evaluation considers the:
' 1.
Proposed exemption to Appendix H, 10 CFR Part 50, Section II.C.2, f
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Proposed Technical Specification changes described above, and 3.
Adequacy of the inspection, cleanup, and repair of the SSHTs.
2.0 DISCUSSION The Arkansas Nuclear One - Unit No. 1 design includes three reactor vessel surveillance specimen holder tubes located adjacent to the reactor inside vessel wall. Each holder tube contains two sur-veillance capsules which hold the specimens to be irradiated in.
accordance with the requirements of the reactor vessel material surveillance program as described in Appendix H to 10 CFR Part 50.
The purpose of the surveillance program is to monitor changes in the fracture toughness properties of ferritic materials in the
- reactor vessel regions resulting from their exposure to neutron irradiation and the thermal enviroament.
The SSHTs (3 inch diameter, 120 mil thickness, type 304 stainless steel) are mounted to the reactor thermal shield by pintles which support, position, and lock the holder tubes in place. An outer shroud tube (31/2 inch diameter,120 mil thickness, type 304 stainless steel) surrounds the upper portion of the SSHT to protect the tube from the impact of coolant flow.
Two surveillance specimens are held at the bottom of each SSHT by a spring loaded push rod assembly.
Spacers are attached to the push rod at four axial locations to position the push rod in the holder tube.
During normal operation, reactor downcomer water flows up the SSHT to cool the surveillance specimens.
Because of observed SSHT failures at other B&W plants, the licensee
'l shut down the ANO-1 facility on March 19, 1976 to inspect the SSHTs. The inspection, performed on March 25 and 26,1976, revealed the following damage to the ANO-1 SSHT.
One holder tube was severed j
5 at the second spacer from the top but the holder tube itself was i
retained in the shroud tube. The lower section of this SSHT was intact and in place as were the associated surveillance capsules.
The other two SSHTs were found to be completely severed.
Portions i
of these' holder tubes and the associated holddown mechanisms (push rod, spacers, and holddown springs) had fallen to the bottom of the reactor vessel.
The lower portion of both of these SSHTs and the associated surveillance capsules also were found to be intact and in place.
Subsequent inspections revealed that wear damage
'had also occurred at the inner surfaces of the shroud tubes of all three SSHTs.
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To preclude the possibility of further SSHT damage occurring during the remainder of Cycle 1 operation, the surveillance specimens, holder tubes, push-rod and holddown assemblies, and portions of the shroud tube will be removed prior to continuation of Cycle 1 operation.
Engineering of new holder tube and holddown assembly design modifications and material procurement will be completed during the remainder of Cycle 1 operation to allow installation of the revised holder tubes prior te the start of Cycle 2 operation.
The licensee has requested a change to Technical Specification 4.2.7 to update this specification to comply with 10 CFR Part 50, Appendix H, Section II.C.3.C and to ' reflect removal of the reactor vessel surveillance specimens during the remainder of Cycle 1 operation.
The existing Specification 4.2.7 is based on a withdrawal schedule recommended in ASTM-E-185-70 and was adopted prior to publication of Appendix H to 10 CFR Part 50.
The lic asee also has requested a waiver of the requirements of Technical Specifications 3.14.1.b and 4.12.2.a (involving Hydrogen Purge System charcoal filters) to permit reactor operation with l
filters that do not fully =cet the requirements of.the existing Technical Specifications.
The licensee's June 10, 1976 request for a change to the sampling frequency and reporting requirements for the gross radiciodine determination (Table 4.1-3) was made to meet NRC staff requirements.
3.0 EVALUATION 3.1 Exemption to Appendix H, 10 CFR Part 50 As required by Paragraph II.C.2 of Appendix H -to 10 CFR Part 50, i
the surveillance capsules of ANO-1 are positioned during reactor operation so that the neutron flux received by the specimens is at least as high, but not more than three times as high as that i
received by the vessel inne surface. A recent calculation by Babcock and Wilcox Company 1, applicable to ANO-1 and acceptable to the staff, indicates that the neutron flux is 2.4 times greater
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,a 4-at the specimen location than at the reactor vessel wall at 1/4 wall thickness (1/4 t).
To date, Cycle 1 operation of ANO-1 accumulated approximately 0.93 EFPY* of actual-exposure
.to the reactor vessel wall at 1/4t; therefore, the specimens accumulated approximately 2.23 EFPY (0.93 x 2.4) of equivalent irradiation.
The total planned Cycle 1 operation would accumulate 1.37 EFPY of actual exposure to the reactor vessel at 1/4t.
1herefore, the specimens removed af ter 0.93 EFPY of Cycle 1 operation have already received an irradiation equivalent to 2.23 EFPY which is more irradiation than the vessel wall at 1/4t would accumulate during all of Cycle 1 operation (1.37 EFPY).
The irradiation affects accumulated during the 0.93 EFPY of Cycle 1 operation would not ba altered in the specimens stored for the remainder of Cycle 1 operation.
On the basis of the above considerations, we have concluded that removal of the surveillance capsules during the remainder of Cycle 1 operation is acceptable and that the surveillance specimens not subjected to destructive testing after the 0.93 EFPY of Cycle 1 operation may be stored until the beginning of Cycle 2 to permit redesign of the capsule holders.
3.2 Proposed Technical Specification 4.2.7
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The licensee has proposed.a surveillance specimen withdrawal schedule at specimen exposures of 3, 9.5,16, and 22.5 EFPY referenced to 1/4 reactor vessel wall thickness. Provisions an also included for updating Technical Specification 3.1.2 (talating to reactor heatup, cooldow'n, and pressure-temperature limits) based on the results of' specimen tests. We have reviewed this proposal'and determined that it meets the requirements of Appendix H to 10 CFR Part 50, Section II.C.3.C and therefore is acceptable.
3.3 Proposed Specification Regarding Hydrogen Purge System Charcoal Filters The licensee's June 4,1976 letter requested a waiver of the requirements of Technical Specifications 3.14.1.b and 4.12.2a
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which require a carbon sample analysis of the charcoal filters 3
.in the Hydrogen Purge Systeu (HPS).
These requirements were
. established by Amendment No.- 10 to DPR-51 issued February 18, 1976.
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'That amendment required initial inspections of the filters to be l
completed within 90 days of issuance of the amendmeut (i.e., by j
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' May 18, 1976).
Specification 3.14.1.b established the carbon
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sample analysis criterion of 3,90% rad 00 active methyl iodide removal when tested at an air 3elocity within 20% of system design floa, 0.15 to 0.5 mgm/m inlet-nethyl iodide con-0 centration, >70% relative humidity (R. I.) and > 193 F.
The licensee proposed to use replacement filters currently on hand which have shown >99% methg1 iodine-131 removal tested at rated air flow, 50 mgm/m inlet non-radioactive methyl iodide concentration, 70% R.H. and 150*F.
The licensee has indicated that:
- 1. - the on-hand filters meet all requirements of Specifications 3.14 and 4.12 except the carbon sample analysis criteria, described above, 2.
carbon sample tests of the on-hand filters would require complete disassembly of the filters because of their design (normally the carbon sample tests for these filters are performed before manufacture of the filter), and 3.
replacement filters which meet all the requirements j
of the Technica,1 Specifications have been ordered l
and will be available by the end of Cycle 1 operation.
i The licensee concluded in the June 4, 1976 letter that the difference in test requirements would not affect the performance of the filter units.
We, however, do not fully agree with that position.
The test parameters vary with regard to the temperature of the test and the inlet ion concentration.. The lower temperature test performed 3
on the on-hand filters would result in a less conservative i
test than the existing Technical Specifications require j
whereas the higher inlet concentration would result in a more conservative test than that. required by the existing 1
Technical Specifications.
The on-hand filter test criterion i
(i.e., 3,99% methyl iodine-131 removal) is more conservative l
than the 90% radioactive methyl iciide removal required by the. existing Technical Specifications.
Considering these i
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facts collectively, we have concluded that the health and safety of the public would not be endangered by permitting i
.the licersee to use the on-hand filters; however, because the test criterion and parameters cannot be compared l
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...a quantitatively, we would permit use of the on-hand filters only until the next refueling shutdon.1-(presently scheduled for October, 1976). We have discussed our conclusions with 1'
.the licensee's staff and the licensee had agreed with our recommended changes to the June 4,1976 request; namely, that the initial tests of the HPS filters be deferred until the end of Cycle 1 operation.
The licensee has agreed to install fully qualified filters at the end of Cycle 1 operation.
3.4 Inspections, Cleanup, and Repairs The licensee has completed a t'horough inspection and cleanup program as described in the " Surveillance Specimen Holder Tube Repair Report", dated June 1,1976.
Visual inspections by remote television cam' era were performed or. the reactor vessel interior wall and hottom surfaces, the core guide lugs and incore guide tubes, the reactor internals (including assembly bolts and locking caps), and cose.upport asse=bly (including distribution plate, lower grid forging, incore guide tube support plate, and flow distribution head). Both video.and ultrasonic test equipment were used to inspect the SSHT pintles and ' shroud tubes.
In the course of the inspections, debris from the failed SSHTs was removed from the vessel and internals.
Large pieces were removed by grappling; smaller pieces were removed by flushing and vacuum cleaning.
In addition all of the intact portions of the SSHTs were removed (including the surveillance specimens) and the shroud tubes were cut and removed in areas where signi-ficant wear had occurred (wall thickness <100 mils).
The licensee reports that only minor and insignificant damage (primarily small scratches and dents) was observed on the surfaces of the reactor vessel and core internals.
The licensee also reported that the "as left" condition of the shroud tubes had been evaluated and that no safety or structural problem will result from continued operation.
The licensee performed an underwater video inspection on all 177 fuel assemblies which were in.the reactor at the time of the March 19, 1976 shutdown.
The inspection included the " bottom grillage of the lower end fitting plus the bottom 2.75 inch
,section of fuel rods between the grillage and lower spacer grid."
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ion because the licensee concluded that the filtering action ofl the lower end fitting
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and the lowest spacer grid would prevent recoverable debris from passing through tais region. Allobsefveddebrisinthelower end fitting grillage and spacer grid was removed during this 4
inspection.
The licensee indicated that only pieces smaller than 1/8" could pass through the lower region and into the active portion of the fuel assembly.
Such pieces would either pass through or be lodged in the fuel assembly. The licensee also inspected two fuel assemblies over the entire length of the assembly. These assemblies had large pieces of debris entrained in the lower end fitting. No indication of debris was observed in the upper region of the bundle.
The licensee found no evidence of physical damage to the fuel assemblies i -
during.the incpection.
The licensee also inspected the upper tube sheet of both once-through steam generators for pos1 *ble. damage or debris The. tube sheets were found to be free of febris and in excellent condition.
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.All debris removed during the above inspection was collected I
and weighed.
The weight was compared with the calculated l
weight of the same material determined from the as-built i
drawings of the SSHTs.
The material removed was 4 lbs.
less than the calculated dry weight and 6 lbs. less than the calculated wet weight.
These differences are'less than 1%
of the total material recovered.
The licensee has evaluated the possible effects on fuel assemblies of small pieces of debris which may st'ill remain l
in the fuel upon return to operation and concluded that i
. operation of the reactor may be safely resumed.
This 9
j conclusion is based on the design of the fuel, the results of the licensee's inspections (which showed no indication-i of damage), and the recent operatin'g history of the reactor.
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- In particular, the-licensee has observed no significant
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increase 'in iodine activity in the primary -coolant prior j
. to shutdown even though the holder tubes were in the damaged condition.
This indicates that no fuel damage had occurred 3
l as a result of the debris that was in the reactor vessel.
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4 4 The NRC staff has reviewed the licensee's inspection, cleanup, and repair operations described in the June 1,1976 report. We have concluded that the actions taken by the licensee constitute an acceptable program and that further inspection of the facility (including the fuel assemblies) is not required.
It is our opinion that the small amount of material remaining within the vessel, internals, and' fuel is of such small size that feasible inspection techniques to search for these materials would not be effective.
The NRC staff has also reviewed the licensee's analysis of the effects of debris which may remain in the primary coolant system.
Sir.ce all the larger pieces of the disintegrated holder tubes were removed from the reactor vessel during the cleanup operation, it is very unlikely that a blockage of flow passages could occur.
However, 'the presence of small fragments in the primary coolant could cause some problems.
Some of these particles could pass through the end fitting grillage and the lower spacer grid of the fuel assemblies and be lodged in the spacer grid adjacent to the active fuel section of the rod.
Under certain circumstances flow induced vibrations could cause some fretting action which could lead to damage of the fuel claddings.
Some pieces of the debris could also enter the control rod guide tubes and cause some interference with the control rod operation.
These two possibilities were considered by the staff.
The probability of any of th'ese events occurring is proportional to the concentration of the particles in the coolant.
This i
concentration is estimated to be about 1-2 particles per cubic foot of coolant for 1/8" diameter particles.
The quantity of debris circulating in the primary coolant system i
would be continuously decreased as the purification system would remove the-particles from the coolant.
The half-life of the purification system in the Arkansas Nuclear One - Unit No.1 plant is about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, hence in a relatively short time the concentration of particles would be significantly reduced.
The probability that the particles would deposit i
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in the core is, therefore, relatively small and it would con-tinuously decrease during the operation of the plant.
It is the staff's opinion that the possibility of some of the particles being deposited in the core cannot be completely ignored. Frequent monitoring of primary coolant iodine activity would provide indication of fuel failures caused by fretting.
The staff has reviewed the licensee's primary coolant surveillance program described in the June 1, 1976 report and found it accep table. This program, however, differs from the existing Technical Specification requirements.
The staff, therefore, would require continued frequent monitoring of the radiciodine 1
in the coolant and prompt reporting on any significant change in the steady state radioiodine concentration while the reactor is operating.
The licensee was informed of these requirements on June 9,1976, and, by letter dated June 10, 1976, the licensee proposed
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Technical Specifications to change Table 4.1-3, " Minimum Sampling and Analysis Frequency", to require determination of primary coolant gross radioiodine concentrations three times per week and to require notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the steady state gross radiciodine concentration increases t
by a factor of ten or more.
This increased monitoring and reporting program would be required only until the end of Cycle 1 operation since cladding damage due to fretting, if it is to occur, should have occurred by that time.
The likelihood of this wear mechanism also would be continuously diminished by the removal of the debris by the purification syp. tem. We have reviewed this request and concluded that it meets our requirements and therefore is acceptable.
The probability of interference of control rod operation by debris is very remote since the coolant entering the control rod guide tube has to pass through a 1/8" diameter orifice while the clearance between the control rod and the guide tube is approximately 1/16".
Particles less than 1/16" would pass through.
Thus, the only particles which =ight lead to interference with control rod motion are between 1/8" and 1/16".
Because i
of the relatively small concentration of particles (of all sizes) in the coolant, the continued reduction of particle concentration
.by the purification system, and the fact that control rod motion i
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we have concluded that interference with control rod motion is very unlikely. Blockage of the control rod guide tube
'trifice is not considered likely for the same reasons.
The staff has concluded that, with the incorporation of the modified radiciodine monitoring program described above, the operation of the Arkansas Nuclear One - Unic No.1 plant could be resumed without hazard to the health and safety of the public.
4.0 ENVIRONMENTAL CONSIDERATION
S We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result 1.
any significant environmental impact.
laving made this determinatica, we have further concluded that the amend-ment involves an action which is insignificant from the standpoint i
of environmental impact and pursuant to 10 CFR 851.5(d)(4) that an environmental statement, negative declaration, or environmental impact appraisal freed not.be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) because the changes do not involve a significant
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Date:
JUN 11 1976 e
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