ML20216F015
| ML20216F015 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/12/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216F009 | List: |
| References | |
| NUDOCS 9803180223 | |
| Download: ML20216F015 (10) | |
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-4 UNITED STATES
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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. - anat
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOl89TO FACILITY OPERATING LICENSE NO. NPF-6 I
ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT NO. 2 DOCKET NO. 50-368
1.0 INTRODUCTION
By letter dated September 23,1997, Entergy Operations, Inc. (the licensee) submitted a request for changes to the Arkansas Nuclear One, Unit No. 2 (ANO-2) Technical Specifications (TSs).
The requested changes would revise the Reactor Protective System (RPS) and Engineering Safety Actuation System (ESFAS) trip set point and allowable values for steam generator low pressure to support continued plant operation with an increased number of plugged steam generator tubes. The proposed amendment would also relocate the RPS and ESFAS response time tables from the TSs to the Safety Analysis Report as described in Nuclear Regulatory Commission (NRC) Generic Letter (GL) 93-08, " Relocation of Technical Specification Tables of Instrument Response Time Limits," dated December 29,1993.
The letters dated February 27 and March 4,1998, provided clarifying information that did not change the initial proposed no significant hazards censideration determination.
2.0 BACKGROUND
2.1 Relocation of Instrument Response Time Tables (GL 93-08)
Section 50.36 of Title 10 of the Code of Federal Regulations establishes the regulatory requirements for licensees to include technical specifications as part of applications for operating I
licenses. The rule requires that technical specifications include items in five specified categories:
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(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting
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conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. The fundamental purpose of the TSs is to impose those I
conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval.
The relocation of requirements, sech as instrument response time tables, from the TS to the j
Updated Safety Analysis Report (USAR), resulted from NRC staff efforts to develop i
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9803180223 980312 FDR ADOCK 05000368 P
2-improved criteria for delineating those matters that need to be included in TS. The criteria established were included in the final Commission policy statement on TS improvements, published July 22,1993, (58 FR 39132) and were subsequently incorporated into the regulations by an amendment to 10 CFR 50.36, dated July 19,1995 (60 FR 36953).
The Commission's final policy statement recognized, as had previous statements related to the staff's TS improvement program, that implementation of the policy would result in the relocation of existing TS requirements to licensee-controlled documents such as the USAR. The NRC issued GL 93-08 and similar line-item TS improvements in order to improve the content and consistency of TSs and to reduce the licensee and staff resources required to process amendments related to those specifications being relocated from the TS to other licensee documents. Those items relocated to the USAR are controlled in accordance with the requirements of 10 CFR 50.59, " Changes, tests and experiments." Section 50.59 of Title 10 of the Code of Federal Regulations provides criteria to determine when facility or operating changes planned by a licensee require prior Commission approvalin the form of a license amendment.
NRC inspection and enforcement programs also enable the staff to monitor facility changes and licensee adherence to USAR commitments and to take any remedial action that may be appropriate.
2.2 Steam Generator-Low Pressure Setooint E duction ANO-2 has an active damage mechanism affecting the steam generator tubing which requires the repair or the removal of tubes from service when they meet the repair criteria. The unit entered a mid-cycle outage in February 1998, in order to perform inspections of the steam generator tubes and perform plugging of those tubes found to have exceeded the established plugging criteria A reduction in the heat transfer surface area occurs for each plugged steam generator tube and requires an increased differential temperature across those tubes remaining in service in order to support continued operation at the rated thermal power of the reactor core.
The increased differential temperature is achieved by reducing the coolant temperature and steam pressure in the plant's power conversion systems. The lower steam generator pressure anticipated after the mid-cycle outage reduces the operating margin between the full power steam generator pressure and the Plant Protection System (PPS) setpoints based on low steam generator pressure. In order to maintain a comfortable margin between the operating conditions and protection system setpoints, and thereby reduce the occurrence of spurious actuations, the licensee has proposed to reduce the steam generator low pressure setpoints in the TS.
3.0 EVALUATION 3.1 Relocation of instrument Response Time Tables (GL 93-08)
The licensee has proposed changes to TS 3.3.1.1 and TS 3.3.2.1 that remove the references to Tables 3.3-2 and 3.3-5 and delete these tables from the TS. The licensee has also proposed to' relocate TS Figure 3.3-1, *CPC [ Core Protection Calculator) Penalty vs.
Effective RTD [Resistence Temperature Detector) Time Constant," which is referenced in a footnote to Table 3.3-2. The licensee has relocated the tables and other information related to specific RPS and ESFAS response time limits to the USAR in the USAR update submitted
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December 9,1997.
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Tables 3.3-2 and 3.3-5 contain the values of the response time limits for the RPS and ESFAS I
instruments. Figure 3.3-1 provides the values for adjustments to the CPC protection functions if the effective RTD time constant exceeds 8.0 seconds. The limiting conditions for operation for the RPS and ESFAS instrumentation specify these systems shall be operable with the response times as specified in these tables. The limits in Tables 3.3-2 and 3.3-5 are the acceptance criteria for the response time tests performed to satisfy the surveillance requirements of TS 4.3.1.1.3 and TS 4.3.2.1.3 for each applicable RPS and ESFAS trip function. These surveillances ensure that the response times of the RPS an' ;SFAS instruments are consistent with the assumptions of the safety analyses performed for design basis accidents and transients.
The changes associated with the implementation of GL 93-08 involve only the relocation of the RPS and ESFAS response time tables, but retair the surveillance requirement to perform response time testing. The USAR contains the acceptance criteria for the required RPS and ESFAS response time surveillances. Because it does not alter the TS requirements to ensure that the response times of the RPS and ESFAS instruments are within their limits, the staff has concluded that relocation of these response time limit tables from the TS to USAR is acceptable.
The staff's determination is based on the fact that the removal of the specific response time tables does not eliminate the requirements for the licensee to ensure that the protection instrumentation is capable of performing its safety function. Although the tables containing the specific response time requirements are relocated from the TSs to the USAR, the licensee must continue to evaluate any changes to response time requirements in accordance with 10 CFR i
50.59. Should the licensee's determination conclude that an unreviewed safety question is involved, due to either (1) an increase in the probability or consequences of accidents or malfunctions of equipment important to safety, (2) the creation of a possibility for an accident or malfunction of a different type than any evaluated previously, or (3) a reduction in the margin of safety, NRC approval and a license amendment would be required prior to implementation of the change.
The staff's review concluded that 10 CFR 50.36 does not require the response time tables to be retained in TSs. Requirements related to the operability, applicability, and surveillance requirements, including performance of testing to ensure response times, for RPS and ESFAS systems are retained due to those systems'importance in mitigating the consequences of an accident. However, the staff determined that the inclusion of specific response time require-ments for the various instrumentation channels and components addressed by GL 93-08 was not required. The response times are considered to be an operational detail related to the licensee's safety analyses and are adequately controlled by the requirements of 10 CFR 50.59. Therefore, the continued processing of license amendments related to revisions of the affected instrument er component response times, where the revisions to those requirements
- not involve an unreviewed safety question under 10 CFR 50.59, would afford no significut benefit with regard to protecting the public health and safety. The staff has verified that the affected TS require-monts have been relocated to the USAR.
In addition to removing the response times from the TS, the licensee is modifying TS Bases Sections 3/4.3.1 and 3/4.3.2 to reflect these changes and to state that RPS and ESFAS response time tables have been relocated to the USAR. These changes are acceptable in that they merely constitute administrative changes required to implement the TS change discussed above.
4 These TS changes are consistent with the guidance provided in GL g3-08 and the TS require-ment of 10 CFR 50.36. The staff has determined that the proposed changes to the TS for ANO 2 are acceptable.
3.2 Steam Generator-Low Pressure Setooint Redui On in order to maintain adequate operating margin between the secondary-side steam generator pressures during normal operation and the PPS steam generator low pressure setpoint (related PPS signals initiate a reactor trip, main steam and feedwater isolations, and emergency feedwater actuation), the licenroe has proposed to reduce the PPS setpoint from 751 psia to 712 psia. The reduction in the protection system setpoints has the potential to affect the plant's response to those transients that rely on the steam generator low-pressure functions to ensure that safety limits or other de sign conditions are not exceeded. The licensee has re-analyzed several design-basis transionts and accidents to ensure that applicable acceptance limits continue to be satisfied if tP a steam generatorlow pressurs setpoints are reduced. The main steam line L 'sk (MSLB) aacident was reanelyzed to determine the effect of the proposed -
change to t <, team geneator low pressure setpoint on the limits associated with the reactor core, the peak pressure ant temperature within containment, and the radiological consequences to the public and control rocen personnel.
3.2.1 Reactor Core and Reactor Coolant System Response The licensee evaluated those Chapter 15 analyses that are potentially affected by the proposed reduction in the steam generator low pressure setpoints. The transients and accidents evaluated are (1) excessive heat removal due to a secondary system malfunction, (2) mcin steam line break, and (3) main feedwater line bresk. These analyses also considered a proposed reduction in RCS flow that may be caused b/ the expected increase in the number of plugged steam generator tubes. A separate TS amendment and safety evaluation (that contains the applicable parts of the following evaluation) are being issued for the reduction in minimum RCS flow rate.
3.2.1.1 Increase in Steam Flow Event An excess steam demand (ESD) event !s caused by a failure of the main steam system that results in an increase in steam flow from the steam generator. In the presence of a negative moderator temperature coefficient, the event results in an increase in core power and a reduction in DNBR.' The system response tu the event is dependent on the rate of heat transfe.' through the steam generators. The reduced heat transfer area resulting from steam generator tube plugging will slow down the cooiing of the RCS primary system. The reduced initial RCS flow tends to increase the rate of primary cooldown for a given rate of heat transfer. During the transients, the CPC will trip the reactor to avoid violation of DNER safety limit. To assure that the CPC can accurately sense the cooldown associated with the event, the licensee performed a CPC transient filter analysis for Cycle 13. In the analysis, the limiting conditions (design minimum flow reduced by 10% anc ne reduction in steam generator heat transfer area) identified by the licensee's sensitivity studies, were assumed for the ESD event. The results show that the minimum acceptable thermal margin to the DNBR limit in the SAR case remains available. Since
. the results of the existing CPC transler.1 filter analysis verify that CPC trip functions are conservative and demonstrate that the SAR case remains the bounding case. the staff concludes that the effects of a reduction in RCS flow and steam generator tube plugging are appropriately considered for the ESD event.
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The licensee also assessed the impact of a lower steam generator low pressure setpoint of 620 psia (reduced from 678 psia in the current SAR analysis) on the ESD event. The event assessed by the licensee is an inadvertent opening of atmospheric dump valves (ADV) event, previously identified as the limiting ESD case. The licensee's assessment shows that a lower steam generator low pressure setpoint delays isolation of the affected steam generator with an opened t
ADV and results in a 10% ir, crease in the amount of steam release compared to the SAR case.
However, the resulting total mass is well within those considered for the main steam line break (MSLD) event. With a greater sf sm release, the overcooling effect of the MSLB resuis in limiting core conditions that bouna the ESD avent. Since the results of MSLB analysis (discussed in section 3.2.1.2) show that the minimum DNBRs re greater than the DNBR safety i
'imit, the licensee stated, and the staff agrees, that the results of the ESD event with a lower steam generator low pressure setpoint can meet the DNBR safety limit. Therefore, the staff concludes that the effects of reduced RCS flow rate and a reduction in the steam generator low pressure setpolni are appropriately considered for the ESD event.
3.2.1.2 Steam Line Break l
The licensee reanalyzed the MSLB event with consideration of the effects of a reduction in RCS flow and a decrease in the steam generatorlow pressure setpoint to close the main steam i
isolation valves (MSIVs). The analys;s was performed with the NRC-al: proved codas: CENTS l
for calculations of the system response, ROCS /HERMITE for calculations of the reactivity feedback and peaking factors for hot rods, and HRISE for the DNBR calculations. The licensee j
used RELAPS/ MOD 3 to calculate the feedwater flow for the MSLB at hot full power conditions.
RELAP5/ MOD 3 is not an approved code for licensing calculations. At the staff's request, the l
licensee provided the feedwater flow rates calculated with the RELAPS/ MOD 3 code and l
compared them with the flow rates calculated with the HSTA code, a code used in the approved MSLB analysis to support the licensing amendments for ANO-2. The comparison shows that l
RELAP5/ MOD 3 predicts a Ngher flow rate throughout the transient. The use of higher feedwater j
flow rates increases the occooling effects and is conservative. The staff has determined that the use, in the SLB analysis, of feeuwater flow rates that are higher than those calculated by HSTA is conservative and is therefore acceptable. Th'e SLB analysis is, therefore, adequate and acceptable for ANO-2. The staff notes, however, that this action does not approve the use of l
RELAP5/ MOD 3 computer code for this or any other licensing analysis for ANO-2. Future use of REPAPS/ MOD 3 for licensing applications should be preceded oy staff review and approval of the code and the its specific application.
The licensee performed analyses for 4 double-ended guillotine MSLB ( with break sizes of 6.357 2
ft ) cases in order tc determine the limiting cases for approaching the fuel design limits. The 4 cases analyzed are:
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A large MSLB during full power (HFP) conditions in combination with a single failure, loss of offsite power and a stuck CEA.
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Case 1 with offsite power a silable.
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A large MSLB during zero power (HZP) conditions in comoination with a single failure, loss of offsite power and a stuck CEA.
4.
Case 3 with offsite power available.
!' To maximize the overcooling effen, the licensee made the following assumptions: (1) the highest actuation pressure for a safety injuction actuation signal (SIAS) was assumed to delay the injection borated water to the RCS, (2) the cooldown of the RCS was terminated when the affected steam generator blew dry, (3) a CPC low DNBR trip was credited for the loss of offsite power cases and the setpoint was based on 96.5% of the RCP design speed, (4) a low steam generator pressure was assumed at 620 psia to trip the reactor and to actuate the steam generator low pressure signal that closed the main steam isolation valves (MSIVs), main feedwater isolation valves (MFIVs), and back-up MFIVs, (5) the most negative moderator j
temperature and Doppler coefficienb were used to maximize the reactivity addition resulting from i
the cooldown effect, (6) two emergenc/ eedwater pumps were assumed to be available to f
l maximize the cooling potential of the EFW tystem, and (7) the boron from the safety injection tanks was not credited.
l For single failure considerations, the analyses assumed that for the loss of offsite ac power
- cases, one emergency diesel generator (EDG) failed to start. The failure of an EDG resulted in i
the failure of one high-pressure safety injection (HPSI) pump and the MFIVs to close. For the HFP case with ac power available, a bus fast transfer failure was identified as the worst single failure. Tne single failure resulted in the failure of the back-up MFIVs and a HPSI pump. For the HZP case with ac powcr avsRable, a single failure of a HPSI train was assumed.
l The analyses show that t'ns HFP cases remain subcritical throughout the post trip event and that j
i the HZP cases show a retum-to-criticality that is bounded by the SAR results. The calculated l
DNBRs for all cases are greater than the DNBR safety limit and, thus ensure that no fuel failure l
will occur. Since the licensee used NRC-approved codes for analyses, the values used for input i
parameters are conservative, and the results show that the minimum calculated DNBRs are I
greater than the acceptable safety limit to assure fuel integrity, the staff concludes that the analyses are acceptable.
3.2.1.3 Feedwater Line Break i
The licensee performed sensitivity studies of a 10% reduction in the RCS flow and 30% steam
~ generator tube plugging on the feedwater line break (FLB) analysis p esented in the SAR. The results show that changes in initial RCS flow have minimal effects on the FLB analysis, and that the cases without assumed steam generator tube plugging result in a slightly higher peak P.CS pressure. Since a minimum design RCS flow rate without steam generator tube plugging are i
assumed in the SAR case, the licensee's sensitivity studies demonstrates that the SAR case remains conservative for the FLB analysis.
To assess the effect of a lower steam generator low pressure setpoint ( 620 psia) to close the MSIVs during the events, the licensee reanalyzed the feedwater line break (FLB) event with loss of ac power, which is the limiting case identified in the SAR.
The licensee performed FLB analyses for various break sizes with the approved CENTS code and identified that the break of 0.24 fta resulted in the highest peak RCS pressure. To maximize the calculated peak RCS oressure, the licensee made the following assumptions: (1) the least negative Doppler coefficient corresponding to the BOC core was used to maximize the power increase, (2) the initial plant conditions were assumed to be during full power operailon with a loss of offsite power at the time the reactor trip breakers open, (3) a conservative CEA insertion j
7 curve corresponding to the axial power shape of +0.6 ASI was assumed, (4) a steam generator low pressuro signal was assumed at 620 psia to actuate the MSIVs with a closure time of 3 seconds, (5) the blowdown of saturated liquid from the affected steam generator was assumed.
(6) the tolerance for the safety valves and secondary safety valves was assumed to be +3% of the setpoints, and (7) the decay heat was maximized by assuming an equilibrium core.
The initial pressure and initial steam generator inventories were selec,ed such that the low steam generator water level trip in the intact steam generator and the high pressurizer pressure trip occurred simultaneously with the dryout of the affected steam generator. The sensitivity study showed that this assumption resulted in a maximum peak RCS pressure after the trip.
The results of the reanalyses show that the peak RCS pressure is 2730.1 psia which is less than 110% of the design pressure. Since the licensee used NRC-approved codes for the analysis, the values used for input parameters are conservative, and the results show that the peak calculated RCS pressure is within the acceptance criteria of 110% of the design pressure, the staff concludes that the reanalyses are acceptable.
3.2.1.4 Conformance to SER Conditions I
As a result of the findings from the NRC's Maine Yankee Lessons Learned Task Force, the staff requested that the licenses verify conformance to conditions stated in NRC safety evaluation reports (SERs) associated with the topical reports (TRs) reviewed and approved by the NRC staff for referencing in licensing applications. In response to the staff's request, the licensee and the fuel vendor, ABB-CE, evaluated their compliance with the conditions specified in the staff's SERs for those topical reports referenced in the submittels pertaining to the proposed TS changes (reduction in minimum RCS flow rete and reductioa in the steam Cenerator low pressure setpoint). The licensee's evaluation determined that the SEF: conditions associated with the referenced topical reports had been satisfied. Accordingly, the staff finds that the licensee has adequately addressed the request to ensure that previously approved topical reports are used consistent with the limitations established as a condition of NRC approval.
3.2.2 Containment Response A lower steam generator pressure setpoint may delay the receipt of reactor trip and main steam and feedwater isolation signals following a main steam line break. Such delays would result in an increase in the mass and energy released into the conta!nment following a MSLS. The licensee analyzed a spectrum of break sizes, power levelc, and initial conditions within the containment building. Based upon the time at which the setpoint would be reached, the licensee assumed conservative satpoints that accounted for instrument uncertainties for either normal conditions or harsh environments. The licensee calculated mass / energy releases into the containment using a combination of ths RELAFSMOD3.1 for feedwater flowlermalpy data and the SGN-ill code for blowdown mass and energy values. The SGN-ill output was input to the COPPATTA containment code to predict the containment response to the MSLS.
The analysis techniques and assumptions are generally the same as those currently described in the USAR. A study performec, as part of this analysis determined that the limiting single failure related to the containment response to a MSLB is the temporary loss of a vital electrical bus due to failure of the fast transfer mechanisms. This failure results ;n delays in (1) a containment fan
8 cooler to start, (2) a containment spray pump to start, and (3) the closure of the back-up main feedwater isolation valves, until startup of the associated emergency diesel generator. Based on the mass and energy data from SGN-Ill and the containment modeling within the COPPATTA code, the peak containment conditions following a MSLB were determined to be 53.0 psig and 423 'F. These values are less than the containment design pressure of 54 psig and the current j
MSLB peak temperature estimate of 426 'F.
Following the analysis of the various MSLB scenarios for different break sizes and initial conditions, an error was discovered in the assumed maximum break size. The maximum pipe 2
break area had bcen assumed to be 6.19 ft which is less than the maximum installed area given 2
a guilloti.e break in the main steam line (6.357 ft ). The analysis was repeated with the larger break area considering the limiting single failure, an increase in the assumed service water temperature to 120 *F to be consistent with current USAR assumption, and e correction (addition) of heat sink data. In order to offset the increase in the mass and energy release due to the larger break area, a steam generator tube piugging level of 10% was assumed in the revised analysis (resulting in a lower initial steam pressure and reduced heat transfer from the reactor coolant system during the MSLB accident). Given that the plugging levelin the ANO-2 steam generators currently exceeds 13%, the staff finds that the 10% assumption is conservative and can be used to offset the needed changes in break area and service water temperature.
Changes were also assumed in the response time of the containment spray trains by assuming i
2 an earlier delivery of t, pray to the containment than was assumed in the 6.19 ft case. The revised assumptions of a faster containment spray response removed margin that the licensee had added for future considerations but remained bounded by the response times currently in TS Table 3.3-5. The revised analysis resulted in a peck containment pressure of 52.3 psig which i
remains below the containment design pressure of 54.0 psig. The staff finds that the licensee has adequately demonstrated that the proposed reduction in the steam generator-low pressure setpoint will not result in exceeding the design limits of the containment following the rupture of a main steam line.
TS Figure 3.6-1, " Containment Intemal Pressure vs. Average Air Temperature,is developed using the limiting initial conditions for analyses associated with (1) the loss of coolant accident evaluation performed per 10 CFR 50.46, (2) the containment design negative pressure differeritial of Spsid (potentially caused by an inadvertent actuation of containment spray), and (3) the containment design pressure of 54 psig following either a loss of coolant accident or secondary-side high energy line break. The licensee presented the results of a series of analyses releted to containment initial conditions (pressure, temperature, and relative humidity) for the revised main steam line bretak conditions. The new analyses, ccmbined with previous evaluations of the sensitivity of the containment pressure response to initial conditions for the loss of coolant accident and the limiting initial conditions for the loss of coolant evaluation model and negative pressure differentiallimits, resulted in minor changes to the locus of points defining the area of acceptable operation in Figure 3.6-1. The staff finds the changes 5 the figure are consistent with the analyses presented by the licensee and are acceptable.
3.2.3 Assessment of Radioloaical Consecuences The licensee perform 3d an assessment of the radiological dose consequences of a MSLB accident in support of the proposed change in the steam generator low pressure setpoints. That a
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assessment was based upon a primary to secondary leakags of 1.0 gpm (300 gpd) allowed by TS 3.4.6.2. The licensee assumed that the 1.00 gpm leakage was divided into 0.5 gpm to the faulted steam generator and 0.5 to the intact steam generators. The licensee fn/ad the radiological dose consequences acceptable, assuming allowable activity levels in the primary coolant of 60 pCi/g dose equivalent I for a pre-existing spike condition and 1.0 pCilg dose equivalent 'l for the accident-initiated spike condition.
The staff has independently calculated the doses resulting from a MSLB accident using the methodology in Standard Review Plan (SRP) 15.1.5, Appendix A. The staff performed two i
separate assessments. The first assessment was based upon a pre-existing iodine spike activity l
level of 60 pCi/g of dose equivalent l in the primary coole at. The second assessment was based upon an accident-initiated iodine spike. Both assessmerits utilizsd dose conversion factors listed in TlD 14844, talculation of Distance Factors for Power and Test Reactor Sites."
For the accident initiated spike assessment, the staff assumed that the accident initiated an increase in the release rate oilodine from the fuel by a factor of 500 over the release rate to maintain an activity level, of 1 pCi/g of dose equivalent l in the p'imary coolant.
For each assessment, the staff calculated doses for individuals located at the Exclusion Area Boundary (EAB) and at the Low-Population Zone (LPZ). The contrcl room operators thyroid dose was also calculated. The staff also reviewed the licensee's description of the revised MSLB accident analysis and the postulated dose results. The results of the staff's independent calculations described above were used to confinn the acceptability of the licensee's analysis methodology Based on comparisons of results, the staff found the licensee's analysis to be appropriate.
The staff has concluded, based upon the considerations above, that the proposed change to the Technical Specifications is acceptable. The staff has rietermined that reasonable assurance exists that, in the event of a postulated MSLB, the doses to persons at the EAB and LPZ would continue to be well within 10 CFR Part 100 doce guidelines, and that the postulated control room operator doses would continue to be less than the criteria in the SRP and 10 CFR Ptrt 50,
' Appendix A, GDC ig.
4.0 FTATE CONSULTATION in accordance with the Commission's regulations, the Arkansas State official was notified of the j
proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a f acility component
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located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may t's released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment i: wolves no significant hazards consideration, and there has been no public comment on such finding (63 FR 4311). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no
e e environmentalimpact statement or environmertal assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the.oublic will not be endangered by operation in the proposed manner, (2) such activities voll be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: S. Sun, NRR/SRXB M. Blumberg, NRR/PERB W. Reckley, NRR/PO41 Date: March 12, 1998 i
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