ML20057B457
| ML20057B457 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 09/13/1993 |
| From: | Simpkin T COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| RTR-REGGD-01.025, RTR-REGGD-1.025 NUDOCS 9309220060 | |
| Download: ML20057B457 (14) | |
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Commonwsalth Edisen i
C 1400 Opus Place -
O Downers Grove, Illinois 60515 l
i September 13,1993 j
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i Dr. T.E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention:
Document Control Desk
Subject:
Zion Nuclear Power Station Units 1 and 2' Revised Analysis for Fuel Handling Accident in Containment NRC Docket Nos. 50-295 and 50-304 i
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References:
1.
August 25,1993 conference call between T. Simpkin, et al. (CECO) and C. Shiraki, J. Hayes (NRC).
2.
July 10,1991 letter from S.F. Stimac (CECO) to Dr. Thomas E. Murley i
i 3.
March 2,1979 letter from A. Schwencer to Mr. Cordell Reed.
Dr. Murley:
i Enclosed for your review is a reanalysis of Zion Stations Fuel' Handling l
Accident. This reanalysis was performed to correct an input in the analysis pro-vided in Reference 2 in support of Zion's transition to VANTAGE 5 fuel. The error involved the use of a filter efficiency for organic iodine inconsistent'with Regula-i tory Guide 1.25. The revised analysis includes the appropriate filter efficiency for organic iodine species and a reduction of the assumed pre-shutdown power from i
104.5% to 100% of rated power (3250 MWt) to offset the increased offsite dose.
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I As requested by your staff during the Reference 1 conference call, the enclosed information provides the results of the reanalysis of the Fuel Handling Accident (FHA) in Containment for Zion Units 1 and 2. Also included in the at-tachments are relevant licensing documents concerning previous Zion FHA anal-
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Attachmeat A discusses the background and purpose of the reanalysis, key assumptions and inputs, and the results and conclusions.
i Attachment B provides FHA related excerpts from the most recent li-censmg submittal (Reference 2) that involved analysis of the FHA in Containment.
The reference 2 submittal requested an amendment to support the use of Vantage 5 fuel at Zion Station. The results of the reanalysis described in Attachment A 3
therefore supersedes the Attachment B information.
Attachment C provides a copy of the SER for a license amendment (Reference 3) which involved an earlier analysis of the FHA in Containment.
I Attachment D provides a summary of CECO's assessment of the poten-l tial safety impact associated with the revised FHA in Containment analysis based i
j on our understanding of the NRC's acceptance criteria for Zion accident analyses.
I As indicated in the Reference 1 discussions and in the enclosed informa-tion, CECO believes the results of this reanalysis are acceptable and meet the gen-eral acceptance criterion that requires conservatively calculated dose levels te be well below the 10CFR100 limits. In addition, the reanalysis improves consistency with the guidance of Regulatory Guide 1.25.
Please direct any questioris to this office.
i Sincerely, Y*
T. W. Simpkin.
1 Nuclear Licensing Administrator i
1 Attachments:
h A. Background B. Excerpts from Vantage 5 Submittal C. SER for original FHA Analysis 3
D. Assessment of Safety Impact cc: J. B. Martin, Regional Administrator - RIII C. Y. Shiraki, Project Manager - NRR j
J. D. Smith, Senior Resident Inspector - 7sion k:nla t zion t z f ha sulun:2
ATTACHMENT A
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Introduction:==
The Fuel Handling Accident In Containment has been reanalyzed by Westinghouse for Zion Units 1 and 2. This reanalysis was performed because of a deficiency in the previous analysis in the Reload Transition Safety Report (RTSR) for Zion's i
transition from Westinghouse's OFA to VANTAGE 5 fuel which was described in j
Reference 1. Relevant portions of Reference 1 are included in Attachment B. This deficiency was identified in the process of updating the UFSAR Section 15.7.4 aris-ing from VANTAGE 5 implementation. The RTSR analysis assumed 90% filter effi-ciency for both the inorganic and organic iodines as opposed to the 70% filter effi-ciency for organic iodine recommended in R.eg. Guide 1.25.
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Although both the Fuel Handling Accident scenarios (in Containment and in the Fuel Handling Building) were analyzed for the transition to VANTAGE 5 fuel, the Westinghouse analysis of the Fuel Handling Building event has been superseded.
A new Fuel Handling Accident in the Fuel Handling Building analysis was per-formed by Holtec Inc. in conjunction with the installation of new spent fuel pool storage racks. This analysis was included in the Technical Specification submittal l
approved under Amendment No.142 to License DPR-39 and Amendment No.131 i
l to License DPR-48 issued February 23,1993. The following discussion is therefore l
limited to reanalysis for the Fuel Handling Accident in Containment.
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The calculated doses from this reanalysis are well below the 10 CFR Part 100 lim-its of 25 Rem whole body and 300 REM thyroid. Results of this analysis must be well below the Part 100 limits because they are established for the purpose of siting considerations such as exclusion area determination and not for direct com-parisons to subsequent licensing accident analyses. The 1979 SER for an earlier l
Zion Fuel Handling Accident analysis (Attachment C) considered thyroid doses less l
than 100 Rem to be " Appropriately within the guidelines of 10 CFR 100...". Else-where in the same SER, a conservatively determined value of 138 Rem was also considered to be "well within the guidelines of 10 CFR 100".
In a conference call with the NRC on August 25,1993 (Reference 2), the NRC re-quested that CECO submit the reanalysis assumptions and completed results along with the 1979 SER for an earlier Fuel Handling Accident in Containment analysis, Amendment No. 43 to License DPR-39 and Amendment No. 40 to License DPR-48 (Reference 3). The NRC wished to review these prior to confirming that the results i
j are within the acceptance criteria for the Zion analysis. The Reference 3 SER has been included as Attachment C.
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'Assumptions:
4 The assumptions used in this analysis are summarized in Table 1. There are two significant changes from the previous analysis in Reference 1. The filter efficiency for organic iodine was changed from 90% to 70% to be consistent with Reg. Guide 1.25. To partially offset the decreased filter efficiency, the assumed core thermal power level prior to shutdown was reduced from 3391 MWt to 3250 MWt. This 1
reduced power is consistent with the licensed thermal power limit and Reg. Guide 1.25 statement that "the inventory should be calculated assuming full power opera-tion". The source terms are EOL core source terms taken from UFSAR Table 15.6-32 and are consistent with the analysis in Reference 1. These source term activi-ties were adjusted for assumed core operation at full power with the assembly hav-ing a radial peaking factor of 1.65 and with the assembly decaying 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> fol-lowing shutdown prior to handling. The adjusted terms are given in Table 2. For the analysis, only isotopes that have significant doses after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> were used.
The 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> is significant because Zion Tech. Spec. 3.13.1.A.3 states that irradi-ated fuel will not be moved until the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. In general, the assumptions used in the reanalysis are consistent with those in Reg. Guide 1.25.
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Results:
For the Fuel Handling A:cident in Containment, the site boundary doses are calcu-lated to be 102 Rem thyi aid and 0.6 Rem whole body. These doses are well below the 10 CFR 100 limits. '.'he thyroid dose is somewhat greater than the 93.5 Rem reported in Reference 1 mt is within the ranges found acceptable in the SER for 1
the previous Zion Fuel IJandling Accident Analysis in Containment (Reference 3).
Commonwealth Edison ha ~aluated the potential impact on safety of this revised 1
analysis and has concluded that.t improves consistency with standard regulatory guidance in this area while not significantly impacting safety. A summary of the i
evaluation is provided as Attachment D.
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Future Actions:
Upon NRC review of this analysis and the conclusions regarding Zion's Fuel Han-dling Accident In Containment offsite dose acceptance criteria, the Zion UFSAR will be revised accordingly in the next UFSAR update.
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References:
1.
Letter from S. F. Stimac to Dr. Thomas E. Murley " Zion Station Units 1 and 2 Ap-plication for Amendment to Facility Operating Licenses DPR-39 and DPR-48 and Appendix A, Technical Specifications" dated July 10,1991.
2.
CECO (T. Simpkin et al.) conference call with NRC (C. Shiraki, J. Hayes) on August 25,1993.
l 3.
Letter from A. Schwencer to Mr. Cordell Reed, " Zion Station, Units 1 and 2," SER l
supporting Amendment No. 43 to license No. DPR-39 and Amendment No. 40 to license DPR-48, dated March 2,1979.
4.
D. Malinowski, M. J. Bell, E. R. Duhn, Topical Report " Radiological Consequences of a Fuel Handling Accident", WCAP-7828, December 1971.
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TABLE 1 i
ASSUMPTIONS USED IN CALCULATING THE RADIOLOGICAL CONSEQUENCESOF A FUEL HANDLING ACCIDENT INSIDE CONTAINMENT WITH THE CONTAINMENT / FUEL HANDLING BUILDING EQUIPMENT HATCH OPEN l
Power Level (MWt) 3250*
Lead Rod Burnup (MWD /MTU) 60,000 Number of failed fuel rods 204 j
(1 assembly) 1 Number of assemblies in the core 193 Decay Time before Fuel Movement (hrs) 100 Power Peaking Factor 1.65 i
8 Containment Free Volume (ft )
2.736 x 10 Purge exhaust rate (cfm) 40,000 Time elapsed to Detection of Activity (sec) 31.5 Purge Valve Closure Time (sec) 7 l
Total release through Purge System duration (sec) 39 Fuel-rod inventory released to gap, %
Iodine-131 12 Other Iodine 10
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Krypton-85 30 Xenon 10 Pool decontamination factors For iodines 100**
For noble gases 1
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TABLE 1 (continued)
Fuel Handling Building filter efficiency (%)
For inorganic iodine species 90 For organic iodine species 70*
For noble gases 0
Atmospheric diffusion factor - x/Q (s/m )
9.2 x 10 4 Breathing rate (m /sec) 3.47 x 10 4 a
All activity not released via the Containment Purge System passes through the Fuel Handling Building's ventilation system and charcoal filters prior to release to the environment.
l Other Assumptions per Reg. Guide 1.25 l
These assumptions differ from the assumptions used in Reference 1.
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- Although this number was given in Regulatory Guide 1.25 for 23 feet water depth, it is shown by Westinghouse to be conservative for the Zion 22 foot water depth by the model in R.eference 4.
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i TABLE 2 ACTIVITIES IN THE MAXIMUM RATED POWER ASSEMBLY-WITH AN ASSEMBLY RADIAL PEAKING FACTOR OF 1.65 100 HOURS AFTER SHUTDOWN Isotope Fuel-Cladding Gap Activity (Curies) 4 I-131 6.39 x 10 i
I-133 5.26 x 10 8
Xe-133m 1.15 x 10 4
Xe-133 8.92 x 10 l
Kr-85 2.31 x 10 f
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f ATTACHMENT B PORTIONS OF PROPOSED LICENSE AMENDMENT 139 TO LICENSE NO.
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DPR-39 AND AMENDMENT 128 TO LICENSE NO. DPR-48 RELEVANT TO THE FUEL HANDLING ACCIDENT IN CONTAINAENT i
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l 0.3.3 Rod Eiection Accident i
The Rod Ejection Accident involves a gap release of inventory from a portion l
of the fuel and a release of primary coolant to the containment.
As discussed in Sections D.l.1 and D.l.2 neither the source term or gap fraction are affected by the implementation of VANTAGE 5 fuel; therefore, the radiological consequences of this accident are not affected by the change to VANTAGE 5 fuel.
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D.3.4 f_ugj_Handlina Accident in the Fuel Handlina Buildina t
l The analysis reported in the FSAR uses assumptions that are less conservative than those defined by Regulatory Guide 1.25.
In addition the analysis is inconsistent with the analysis of the fuel handling accident inside containment.
The fuel handling accident was reanalyzed to utilize assumptions from Regulatory Guide 1.25 except that, in keeping with the discussion in D.l.2 above, the gap fraction for I-131 was increased.from 10 to 12 percent.
Also, instead of using FSAR Table A.3-2 as the accident source term, the EOL core source terms from FSAR Table 14.3.5-5 were used as a basis for the fuel handling accident source term since it provides both a more accurate and a l
more conservative dose analysis.
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The site boundary doses are calculated to be 32.4 rem thyroid and 0.6 rem gamma body.
These doses are well below the 10CFR100 ilmits.
The thyroid dose is substantially greater than that currently reported in the FSAR but the increase is due primarily to the change to encompass the Pegulatory Guide 1.25 methodology.
This is the bounding case for the fuel handling accident in the Fuel Handling Building.
The overall bounding fuel handling accident is that of a dropped fuel assembly inside the containment.
i D.3.5 Fuel Handlina Accident inside Containment l
The radiological consequences were recalculated to take into account the increase in I-131 gap fraction to 12 percent, in agreement with 0.1.2 above.
Also, as for the fuel handling accident in the Fuel Handling Building, the EOL I
core source terms from FSAR Table 14.3.5-5 were used as a basis for the fuel handling accident inside containment source term instead of using the source 1042 l
l term from FSAR Table A.3-2.
Another change in the analysis cas to take crodit for the fact that the containment purge isolation valves close in 7 seconds
~ (Technical Specification 3.9.3) or less instead of the 60 seconds that was assumed in the FSAR analysis.
The doses are calculated to be 93.5 rem thyroid and 0.6 rem gamma body.
These
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t doses are well below the 10CFR100 limits.
The thyroid dose is semewhat l
greater than the 90 rem currently reported in the FSAR but is within the 100 rem acceptance limit specified by the NRC for Zion Station Units 1 & 2.
t D.3.6 Accidental Rglease of Radioactive Gase.1 l
The limiting accident for the release of radioactive gases is that of a gas decay tank failure.
The licensing basis dose analysis for this accident l
assumes that there is 117,000 curies of dose equivalent Xe-133 stored in the tank while the Technical Specifications limit the inventory in any one tank.to 22,000 Ci dose equivalent Xe-133.
The Technical Specification limit will prevent any increase in the dose due to a gas' decay tank failure.
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The other evaluated failures are the volume control tank rupture and the l
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rupture of the seismic class II components in the gaseous radwaste system.
The volume control tank rupture doses are not affected by the VANTAGE 5 fuel use since there would be no significant change to the primary coolant activity levels. The doses due to the rupture of the seismic class II components in the gaseous radwaste system are also not affected by.the use of VANTAGE 5 fuel since the gases present in these components will not be of greater l
concentration than for the gas decay tank which is limited by the Technical j
Specifications.
t D.3.7 Accidental Release of Radioactive Lioulds The only release for which there is a reported analysis of-the radiological consequences is that of the. failure of the primary or secondary water storage tanks.
The quality of water in these tanks is dependent on the amount of processing and cleanup that is utilized.
The calculated radiological consequences of the failure of one of these tanks is thus not affected by the use of VANTAGE fuel.
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l ATTACHMENT C SER SUPPORTING AhENDAENT 40 TO LICENSE NO. DPR-38 AND AAENDAENT 40 TO LICENSE DPR-48 i
DATED MARCH 2,1979 i
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4 UNITED STATES l
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NUCLEAR REGULATORY COMMISSION l
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WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
SUPPORTING AMENDMENT NO. 43 T0' FACILITY OPERATING LICENSE NO. DPR-39 l
SUPPORTING AMENDMENT NO. 40 TO FACILITY OPERATING LICENSE NO. DPR-48 COMMONWEALTH EDIS0N COMPANY l
ZION STATION UNITS 1'AND 2 DOCKET NOS.'50-295 AND~50-304 l
Introduction l
l By letter dated September 21, 1978, Commonwealth Edison _ Company (the licensee) requested a change to Technical Specifications appended to Facility Operating Licenses-DPR-39 and DPR-48 for Zion Station Units 1 and 2, respectively. The proposed amendments would allow the containment equipment hatch of each unit to remain open during refueling-operations. Associated with this proposal is the staff's review of a postulated fuel handling accident inside containment (FHAIC) for which, by letter dated January 14, 1977, we requested that the licensee submit an analysis of that postulated accident.
By letter dated March 14,1977, the licensee provided that analysis.
Discussion l
The proposed changes to Technical Specification 3.13 would allow the j
equipment hatch.between the containment and the fuel handling building to remain open during all refueling outages. As discussed in the licensee's submittal, the proposed changes would reduce occupational exposures by eliminating the wait time for passing through the equipment hatch during refueling outages. The licensee has estimated that 0.9 to 2.7 man-rems per outage will be saved by leaving the equipment hatch open.
In addition, removal of the hatch during refueling would reduce outage time by a minimum of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per outage by eliminating at least one removal and replacement of the hatch and reducing maintenance on the hatch through less usage.
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i The only potential accidents affected by the proposal would be the fuel handling accidents both inside containment and in the fuel handling building.
The FHAIC was not evaluated during licensing of the Zion Units.
In response to our request, the licensee submitted an analysis of the FHAIC on March 14, 1977. Based on containment air mixing and containment isolation, the licensee concluded that the potential consequences of an FHAIC would be well within the guidelines of 10 CFR Part 100.
The licensee's evaluation, however, did not include the effect of the equipment hatch remaining open after the accident. The licensee's letter of September 21, 1978 addressed the effect of keeping the equipment hatch open during refueling and therefore during an FHAIC. The licensee concluded that, with the hatch open, the consequences of an FHAIC would not exceed those calculated for an accident in the fuel handling building.
Evaluation We have reviewed the licensee's basis for proposing to keep the equipment hatch open during refueling with respect to occupational exposure control. We agree with the licensee that the present Technical Specifications restrict the movement of workers into and out of containment. This restriction on movement will result in a higher occupational exposure compared to that received with the equipment hatch open. We conclude that keeping the equipment hatch open during refueling will help keep occupational exposures "as low as reasonably achievable" (ALARA).
We have reviewed the licensee's proposal to determine if the revised Technical Specifications would decrease the safety margins with respect to postulated accidents. The only postulated accident which would be affected is the fuel handling accident both in the containment (FHAIC) and in the fuel handling building.
We have independently evaluated the potential consequences of an FHAIC.
With the equipment hatch open, both the containment and the fuel handling building should be at the same pressure. As the fuel handling accident would not release a significant amount of energy there should be no driving force to cause exfiltration of the activity from the containment other than through the containment purge system and the auxiliary building ventilation system. The purge inlet system supplies 40,000 cfm of air from 32 ducts surrounding the periphery of the refueling pool. These ducts are angled upward at 45 degrees and should promote 4
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1 i mixing of the activity in the containment. A manipulator crane ventilation system directs air down from the crane toward the pool, thus further promoting mixing. The purge exhaust intake duct is i
I located 17 feet above the operating floor near the containment wall approximately 45 feet from the nearest refueling pool surface. We have estimated that the released activity would uniformly mix in the volume of air between the refueling pool and the purge exhaust duct only on the side of the containment where the duct is located. This represents approximately 4% of the containment free volume.
The air would be drawn into the duct and be directed past the containment isolation valves to the plant vent. An air sample would be drawn from the purge d uct downstream of the purge isolation valve and directed to a radiation detector. Upon detection of high radiation levels the detector would initiate closure of the containment purge supply and' exhaust valves.
In its letter of March 14, 1977, the licensee states i
that it would take approximately 32 seconds from the time the initial activity reaches the outboard isolation valve until initiation of valve closure. The licensee assumed a purge valve closure time of 12 seconds, however, we assumed a more conservative purge valve closure time of 1
60 seconds, as specified in the plant Technical Specifications, resulting in a total release time of 92 seconds. The resulting doses to an individual at the site boundary would be 84 Rem to the thyroid and 0.3 Rem to the whole body.
We have concluded that the potential consequences of the postulated fuel handling accidents are appropriately within the guidelines of 10 CFR Part 100 and are, therefore, acceptable. Appropriately within the guidelines of 10 CFR Part 100 has been defined as less than 100 Rem to the thyroid. This is based against the 10 CFR Part 100 exposure guidelines. Whole body doses were also examined, but they are not controlling due to decay of the short-lived radioisotopes prior to fuel handling.
We have reviewed the staff's Safety Evaluation Report (SER) of October 6, 1972, for the Zion Station. That SER summarizes the analysis of the postulated fuel handling accident inside the fuel handling building.
The proposed change does not affect the SER analysis since the activity would be exhausted through the ESF grade auxiliary building ventilation system charcoal filters.
In the event that the activity released as a result of a fuel handling accident in the fuel handling building were to pass through the equipment hatch to the containment purge system rather than through the auxiliary building charcoal filters, the resulting offsite doses would be less than those estimated for the FHAIC. This is because the additional air volume in the fuel handling building would result in much greater mixing of the activity than that assumed for the FHAIC analysis. The purge isolation time would remain the same, therefore, the total activity released would be less.
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l r l If all the activity released from the fuel handling accidents were released to the environment with no credit taken for containment isolation, mixing inside containment, or effluent filtration, the thyroid dose to an individual at the site boundary would be 138 Rem.
The whole body dose to the individual at the site boundary would be 0.5 Rem. As this limiting case is well within the guidelines of 10 CFR Part 100, and the expected accident scenarios are as previously described, we conclude that it is not necessary for the licensee to test the mixing within containment or the containment purge or auxiliary building ventilation system with the equipment hatch open.
The assumptions used in calculating the consequences are listed in Table 1.
The low population zone doses would be lower due to a lower atmospheric dispersion factor.
To assure that the auxiliary building charcoal filters will be in operation to mitigate the consequences of a fuel handling accident, we modified the licensee's proposed revision of Technical Specification 3.13.2 to read, "The auxiliary building ventilation system shall be operating in the charcoal filter mode whenever irradiated fuel is being handled." The licensee has agreed to this modification.
E as indicated that dropping a spent fuel assembly A recent study h
into the core during refueling operations may potentially cause damage to more fuel pins than has been assumed for evaluating the Fuel Handling Accident Inside Containment.
This study has indicated that up to all of the fuel pins in two spent fuel assemblies, the one dropped and the one hit, may be damaged because of the embrittlement of fuel cladding material from burnup in the core. This may affect spent fuel assemblies with burnups of less than 5000 MWD /MT.
The probability of the postulated fuel handling accident inside containment is small.
In the several hundred reactor-years of plant operating experience there have been only a few cases of a spent fuel assembly being dropped into the core. None of these dropped assemblies has resulted in measurable releases of activity. The potential damage i
to spent fuel estimated by the recent study was based on an assumption that a spent fuel assembly falls about 14 feet directly onto one other assembly in the core and that the impact results in crushing the fuel pins in both assemblies.
This type of impact is unlikely because the falling assembly would be sub.jected to drag forces in the water which should cause the assembly to skew out of a vertical fall path.
U. N. Singh, " Fuel Assembly Handling Accident Analysis," EG&G Idaho J
Technical Report RE-A-78-227, October 1978.
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. Based on the above, we have concluded that the likelihood of a spent fuel assembly falling into the core and damaging all the fuel pins in t
two assemblies is sufficiently small that refueling inside containment is not a safety concern which requires immediate remedial action.
We have, however, conservatively calculated the potential radiological consequences of a fuel assembly drop onto the reactor core with the l
rupture of all the fuel pins in two fuel assemblies. We have also assumed for this postulated accident that the source term for both I
spent fuel assemblies is that given in Regulatory Guide 1.25. This is conservative because, (1) these two assemblies should not have the i
power peaking factor and clad gap activity recommended. in Regulatory Guide 1.25 and (2) the pool decontamination factor for inorganic iodine should be greater than that recomended in Regulatory Guide 1.25.
The calculated potential radiological consequences at the exclusion area boundary for the complete rupture of fuel pins in two assemblies are twice the values given in Table 2.
These conservatively calculated potential consequences are within the guidelines of 10 CFR Part 100; consequently, we have concluded that the potential consequences of 3
this postulated accident are acceptable.
Based on the above considerations, we conclude that the proposed change to keep the containment equipment hatch open during refueling does not involve a significant hazard to the public and, therefore, is acceptable.
Environmental Considerations l
We have determined that the amendments do not authorize a change in effluent types or an increase in total amounts of effluents nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
i Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration (2)
Q 6-l there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) j such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
3 Dated: March 2,1979 l
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i Table 1 Assumptions Used in Calculating the Radiological Consequences of a Fuel Handling Accident Inside Containment *
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Power Level (Mwt) 3250 Fuel Exposure Time (yrs) 3 Equivalent Number of assemblies 4
damaged 1
Number of assemolies in core 193 Decay Time before Fuel Movement (hrs) 100 Power Peaking Factor 1.65 3
4 Volume of air activity mixed in (ft )
9.54 x 10 4
j Purge exhaust rate (cfm) 4 x 10 Time Elapsed to Detection of Activity (sec) 32 Purge Valve Closure Time (sec) 60
+
Fraction of activity released from con-j tainment 0.64 Fraction of activity released from spent i
fuel area building 0.36 Auxiliary building filter efficiency (%)
7 (fromOctober6,1972SER) 90 Exclusion Area Boundary y /Q (s,3) 5.1 x 10 4
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j Other Assumptions per Regulatory Guide 1.25 Table 2 Radiological Consequences of Postulated Fuel Handling Accident Inside Containment Doses (Rem)
Thyroid Whole Body Equipment hatch closed 84 0.3 Equipment hatch open 90 0.5 oContainment/ Auxiliary Building equipment hatch open.
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l ATTACHMENT D 4
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SUMMARY
OF POTENTIAL SAFETY IMPACT Commonwealth Edison has evaluated the potential impact on safety of the revised -
1 Fuel Handling Accident in Containment analysis as summarized below.
i Analysis Changes:
l The reanalysis applies two changes from the assumptions used in the previous i
j analysis submitted with the proposed license amendment of July 10, 1991: the fil-
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ter efficiency used for organic iodine species and the assumed power level for pre-ceding operation.
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The filter efficiency assumed for organic iodine in the 1991 analysis was in-i consistent with the Regulatory Guide 1.25. The primary reason this revised analy-sis was performed was to change the filter efficiency for organic iodine from 90% to i
70% as stated in Regulatory Guide 1.25.
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The assumed power level was reduced from 3391 MWt to 3250 MWt which is the licensed thermal power rating for the Zion Units.' This is consistent with the
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Regulatory Guide 1.25 assumption of operation at maximum rated power.
j Analysis Results:
l The doses are calculated to be 102 rem thyroid and 0.6 rem whole body. These j
doses are well below the 10 CFR Part 100 limits.
Safety Impact:
This analysis change will not result in an increase in the probability of occurrence of the accident. Neither decreasing the assumed efficiency of the filters in the Fu-i el Handling Building nor reducing the assumed operational power makes it more i
probable that a fuel assembly would be dropped and damaged during handling.
i This analysis change will not result in a significant increase in the consequences of j
the accident. The doses are well below the 10 CFR Part 100 limits.
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ATTACIIMENT D
SUMMARY
OF POTENTIAL SAFETY IMPACT l
l This analysis change will not create the possibility of an accident of a different type than any previously evaluated accident in the UFSAR. This change in analysis l
will not modify any component or system or operating procedure, nor will it change l
the operation of equipment in the station.
This change does not significantly affect the margin of safety as defined in the Bas-es for any Technical Specification. The revised results are well below the 10 CFR Part 100 limits. The reanalysis provides improved consistency with the standard regulatory guidance (Reg. Guide 1.25) for determining the impact of this type of event.
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