ML20205Q772
| ML20205Q772 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/16/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20205Q771 | List: |
| References | |
| NUDOCS 9904220064 | |
| Download: ML20205Q772 (5) | |
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[paurgk UNIT'.D STATES g
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30866 4001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS.195 AND 203 TO FACILITY OPERATING LICENSE NOS. DPR-51 AND NPF-6 ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT NOS.1 AND 2 DOCKET NOS. 50-313 AND 50-368
1.0 INTRODUCTION
By letter dated June 28,1996 (OCAN069601), as supplemented by letters dated February 23, 1999 (OCAN029905) and March 15,1999 (OCANO39905), Entergy Operations, Inc. (the licensee), submitted a request for changes to the Arkansas Nuclear One (ANO), Units 1 and 2, Technical Specifications (TSs). The TSs currently specify that, during movement of irradiated fuel in containment and core alterations, "the equipment hatch cover shall he in place with a minimum of four bolts securing the cover to the sealing surface." The proposed amendments would revise the TSs to permit the containment equipment hatch to be open during handling of irradiated fuel in containment and during core alterations subject to the condition that capability for quick containment closure be maintained.
The licensee has previously requested similar amendments to permit the air locks to be open during core alterations. Amendment 166 was issued for ANO-2 on September 28,1995, and Amendmeni 184 was issued for ANO-1 on September 26,1996, granting the requested changes pertaining to the air locks. In a letter dated November 19,1996, the Nuclear Regulatory Commission (NRC) staff informed the licensee that the review of the request pertaining to the equipment hatch had been suspended because of rulemaking activities that.
were then underway in the area of controlling shutdown operations. The rulemaking was subsequently canceled and the staff reinitiated its review of the amendment request pertaining to the control of the equipment hatches at ANO during core alterations.
The February 23 and March 15,1999, letters provided clarifying information that did not change the scope of the original application and the initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
Containment barriers are provided for nuclear power plants as the final barrier of the defense-in-depth concept to protect against the uncontrolled release of radioactivity to the environs. The containment function, in combination with other fission product barriers and accident mitigating systems, limits the radiological dose consequences of design-basis transients and accidents to less than the regulatory limits defined by Title 10 of the Code of Federal Reaulations (10 CFR) Part 100.
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- Prior to 1977, the staff did not require licensees to analyze the radiological consequences of a fuel handling accident inside containment (FHA-IC), since the FHA-lC accident consequences were bounded by the "outside containment' FHA case. However, on January 3,1977, Commissioner Victor Gilinsky received a letter from Mr. Robert Pollard that forwarded l
information indicating that fuel handling accidents in containment could have potential l
radiological dose consequences in excess of Part 100 limits if containment isolation and filtration were not available. Subsequently, on January 14,1977, B. Rusche, Director of the Office of Nuclear Reactor Regulation, requested all nuclear power reactor licensees having an operating license (including Arkansas Power & Light Company) to act on Mr. Pollard's request i
to analyze the radiological consequences of an FHA-IC, and provide the results to the staff. As part of its reviews, the staff ensured that TSs included containment and engineered safety feature filtration systems operability requirements during handling of irradiated fuel in i
containment, to the extent necessary to ensure acceptable accident consequences. Prior to this generic action, *acilities' TSs did not typically require that containment integrity be l
maintained during periods when the reactor coolant system temperature was s 210 *F, but only required that containment vent / purge isolation systems be operable. In a July 1981 revision to the Standard Review Plan (SRP), Section 15.7.4 was revised to specifically require the analysis of an FHA-lC, since the guidance of Regulatory Guide 1.25 (March 1972) assumes an accident in the ' fuel handling and storage facility."
The acceptance criteria of SRP 15.7.4, paragraph 11.5, allow a containment to be "open" to the environment during fuel handling, it being expected that the containment will be undergoing ventilation to minimize worker radiation exposure and for worker comfort. However, if fuel handling operations are to be conducted with the containment open, capability for prompt detection of radiation and automrtic isolation is to be provided and reflected in the analysis of FHA radiological consequences. The ANO licensee's proposal would result in containment conditions wherein containment closure would, not meet these criteria (i.e., would not be automatic and immediate). Equipment hatch closure time would require manual actions taking up to possibly 15 minutes.
For the purpose of evaluating the ANO application, the staff focused on two features:
. (1) acceptability of manual personnel actions to initiate containment closure, and (2) acceptability, from the standpoint of incremental dose increase, of the additional time delay in closing the containment. The acceptability of dependence on manual operator actions for operation of safety systems is well-recognized for actions required in the 'long-term" phase of an accident (i.e., loss-of-coolant accident post-recirculation switchover). For short-term required actions, the staff accepts limited operator manual actions as substitutes for automatic controls subject to a stringent review of human factors considerations including implementation of appropriate procedures and training.
3.0 EVALUATION The staff undertook discussions with the licensee regarding the human factors considerations l
and reliability considerations involved in the use of manual personnel action to close the l
containment under FHA conditions. The staff found that during outages, the licensee designates a " Reactor Building Coordinator" (RBC) to be responsible for containment evacuation and closure in the event of an FHA-IC or potential loss of decay heat removal. The RBC is trained on the containment evacuation and hatch closure procedure. Allindividuals
6 o qualified for the RBC position are given " dry runs" in hatch closure. Also, unannounced drills are conducted. The RBC position is continuously manned during an outage. The RBC is responsible for air lock closure in addition to all other penetrations. The air lock location is an approximately 3-minute walk from the equipment hatch. The RBC has other duties during an outage and is in direct contact with the outage desk at all times. In the event of an accident, the RBC would be immediately dedicated to the task of evacuating the containment and establishing containment closure. In this configuration, the containment has multiple evacuation pathways, which include the equipment hatch, the personnel air lock, and the escape hatch. By letter dated June 28,1996, the licensee stated that the equipment hatch could be closed within 15 minutes based on the performance of plant personnel during drills.
By letter dated March 15,1999, the licensee committed to closing the equipment hatch within 30 minutes of the determination of the need to evacuate containment. The NRC staff has based its conclusions, in part, on the licensee's 30-minute closure time for the containment equipment hatch. The NRC staff has concluded that the 30-minute closure time is an appropriate time limit to allow for the evacuation of personnel from containment while ensuring that the equipment hatch is in place to provide a barrier in the event of an accident. In addition, based on the number of evacuation pathways, it is reasonable to ensure the closure of the equipment hatch within 30 minutes regardless of the status of the evacuation.
The radiological consequences of an FHA-lO were submitted by the licensee in support of previous amendments (Amendment No.166 (ANO-2) and Amendment No.184 (ANO-1))
pertaining to the control of airlock doors during fuel handling. The staff also performed confirmatory analyses related to possible offsite and control room doses for FHA-lO as part of its evaluations that justified approval of Amendments 184 and 166. The methodology used the guidance of SRP Section 16.7.4. and Regulatory Guides 1.4 and 1.25 and results were found acceptable in the safety evaluation. The fission products from the damaged fuel bundle that had been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> were assumed to be instantaneously released into water covering the fuel and then released from the containment over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period. The thyroid doses calculated by the staff for the exclusion area boundary were 33.0 rem for ANO-2 and 52 rem for ANO-1. The dose acceptance criteria in the SRP for the fuel handling accident are well below the 10 CFR Part 100 criteria. For the limiting dose organ (the thyroid), the well-below dose criterion is 75 rem at the exclusion area boundary. Other doses (whole body and control room) were found to be a smalle, fraction of the acceptance criteria. The analysis model for the FHA IC for Amendments 184 and 166 assumed that all of the radioactive material released to the containment escapes the containment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Given the conservative nature of the calculations, the results are applicable to and support the case for allowing equipment hatch closure following the actual damage of a fuel assembly during an FHA. Based on the proposed requirement to maintain the capability to close the equipment hatch and on the conservative nature of the referenced calculations, the staff finds that the potential radiological consequences of an FHA-IC at ANO-1 or ANO-2 remain well-below the criteria of 10 CFR Part 100. There is also a potentially beneficial effect of an open equipment hatch on radiological consequences of an FHA-IC. The staff finds credible the licensee's statement that the exposure of personnelin containment (and perhaps overall radiological consequences) could be lower as a result of the faster evacuation of containment if an FHA lO occurred when an equipment hatch is available as an escape route.
In NUREG-1449, " Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States," Final Report, September 1993, the staff reported the findings of an investigation of shutdown risk. Paragraph 6.9.1 of the report noted that a closed containment
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l 4-can provide a significant reduction in the potential radiological dose associated with loss of decay heat removal and severe core damage. The report also noted that in the event of boiling, workers could be exposed to considerable radiation. In view of these findings, the staff has determined that, if containment penetrations and the reactor coolant system are both open during shutdown operations, the containment should be capable of closure prior to coolant boiling. As discussed in NUREG-1449, an open equipment hatch will be the most limiting containment opening with respect to establishment of containment closure in the event of an outage accident. Therefore, the staff has determined that, with respect to loss of decay heat removal considerations, the equipment hatch should be permitted to be open dving fuel handling in containment subject to the condition that it can be closed prior to crnant boiling.
Paragraph 6.9.2 of NUREG-1449 identifies the considerations involved in rapid equipment hatch closure. They are: (1) radiological and environmental conditions in containment resulting from coolant boiling, (2) number and location of hatch closure bolts, (3) need for and availability of compressed air and electrical power, (4) nearness of required tools, and (5) training and rehearsing of personnel. Each of these five factors are discussed below.
(1) Radioloaical and environmental conditions in containment resultina from coolant boilina:
This item is not a consideration for the proposed amendment, as the staff has determined that the hatch must be capable of being closed prior to boiling.
(2) Number and location of hatch closure bolts: Appendix B of NUREG-1449 provides the details of the staff's Equhment Hatch Survey conducted as part of the NUREG-1449 investigation. It is noted that the ANO hatch design is particularly well-suited for rapid closure in that it uses swing bolting, and no access platform is required. At times, a temporary hatch closure, which provides more rapid closure capability, is installed.
(3) Need for and availability of comoressed air and electrical power: Table B.2 of NUREG-1449 notes that ANO equipment hatch closure does not require services such as compressed air or electrical power. The ANO equipment hatch can be rapidly closed without service power.
(4) Nggrness of reauired tools: NUREG-1449, paragraph 3.3.8, notes that at ANO tools are l
kept in a closed box at the hatch and are clearly labeled for emergency use.
(5) Trainina and rehearsina of oersonnel: NUREG-1449, paragraph 3.3.8, notes that the investigation team found that " Arkansas Nuclear One had a requirement that an equipment hatch be capable of closure within 15 minutes of a loss of RHR [ residual heat removal).
Responsibilities were established for such actions as notifications of loss of RHR, containment evacuation, closure operations, and verifications. Unannounced closure exercises had been conducted." The investigator stated that "few other sites were as well-prepared." (The investigation team visited 11 sites.)
Based on the onsite findings of the NUREG 1449 investigation team, and the information documented in the application, the proposed amendments are acceptable with respect to shutdown risk loss of decay heat removal considerations.
Generic Letter (GL) 88-17. " Loss of Decay Heat Removal," discusses problems with shutdown modes of plant operation, provides recommendations, and requests information from licensees pertaining to (1) prevention of accident initiation, (2) mitigation of accidents before they
5-potentially progress to core damage, and (3) control of radioactive materialif a core damage accident should occur. The guidance provided by GL 8817, including the recommended time limits for closure of containment penetrations, is primarily aimed at reduced inventory "mid-loop" operations, a condition of increased vulnerability to core damage due to a more rapid boiloff.
The proposed amendments would apply primarily to refueling operations during which there is a greatly increased inventory compared to the events described in GL 88-17. The staff has j
previously found that the ANO licensee has implemented plans and procedures to comply with the GL 88-17 guidance related to loss of decay heat removal during shutdown operations. The bases for the staff's acceptance of the licensee's response to GL 88-17 is not affected by the i
changes included in the proposed amendments.
The staff has determined that the proposed amendments are acceptable with respect to the potential dose consequences of a fuel handling accident inside containment. The staff has also determined that the proposed amendments are consistent with precautions considered necessary to minimize the riska associated with a loss of decay heat removal during outage operations. Based on these findings, the staff concludes that the proposed amendments are acceptable.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendments. The State official had no comment.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be 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 amendments involve no significant hazards consideration and there has been no public comment on such finding (61 FR 42280, August 14,1996). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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 public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: W. Long W. Reckley Date: April 16, 1999
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