ML20216D845
| ML20216D845 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/30/1999 |
| From: | BROOKHAVEN NATIONAL LABORATORY |
| To: | |
| Shared Package | |
| ML20216D819 | List: |
| References | |
| NUDOCS 9907300165 | |
| Download: ML20216D845 (13) | |
Text
b SUBMITTAL-ONLY SCREENING REVIEW OF THE HOPE CREEK UNIT 1 INDIVIDUAL PLANT EXAMINATION FOR EXTERNAL EVENTS (Seismic Portion)
Rev.1 November 1998 (Finalized April 1999)
Brookhaven National Laboratory 1
54 PDR
1.
INTRODUCTION 1.1 Purpose in response to the NRC issued Supplement 4 to Generic Letter (GL) 88-20, " Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 59.54(f).", Public Service Electric and Gas (PSE&G) Company performed an IPEEE for the Hope Creek Generating Station (ilCGS)
Unit I and submitted the IPEEE results to NRC l Reference 11. Brookhaven National Laboratory (BNL),
as requested by the NRC, performed the submittal-only screening review to verify the technical adequacy of the seismic portion of PSE&G's IPEEE submittal. As a resuh of this review NRC sent a Request for Additional Information (RAI) to PSE&G. PSE&G responded to the RAI in an attachment to a September 21,1998 letter to NRC [ Reference 2]. This Screening Review presents the results and conclusions of the BNL review and evaluation of both the original subminal and the licensee's response to the RAl.
BNL's methodology utilized for the review followed the guidelines provided in the document titled
" Guidance for the Performance of Screening Reviews of Submittals in response to USNRC Generic Letter 88-20, Supplement 4" (Draft, Oct. 24, 1996), as amended by the NRC.
1.2 Background
flope Creek Unit 1 is a General Electric boiling water reactor with a gross electric output of 1118 Mwe, with a Mark I containment. The llCGS was designed in the early 1970's, and started commercial operation in December 1986. The unit has 4 emerg:ncy diesel generatcrs at its disposal.
The llCGS site is located on the east bank of the Delaware River in Lower Alloways Creek Township, Salem County, New Jersey. The Safe Shutdown Earthquake (SSE) for the site is 0.20g, and the plant is binned in the 0.3g focused-scope review category. Soil improvement was conducted on the site to replace the loose hydraulic fill with engineered backfill. Underlying the engineered backfill is the Kirkwood formation, which consists of fine to medium grained sands having blow counts ranging from 20 blow counts per foot (bpf) to 70 bpf. For the seismic design, the free-field ground response spectra from Regulatory Guide 1.60 anchored to 0.2g SSE level was used.
1.3 Licensee's IPEEE Process and Licensee's Insights PSE&G used the PRA method for the seismic IPEEE analysis. All the elements emphasized in NUREG-1407 were considered in the analysis, however relay chatter was excluded from the seismic analysis because the evaluation found no concerns.
Both the EPRI and the revised LLNL hazard curves were used in the analysis. A sensitivity analysis was performed to evaluate the effect of using a hazard curve cutofflower than 1.5g. By extrapolating the LLNL hazard curves 6 rom 1.0g) up to 1.5g, the CDF value was increased by about 61 percent. New floor spectra were obtained based on the median EPRI uniform hazard spectrum (UllS) at 10,000 year return period. A probabilistic approach was used for the soil-structure interaction (SSI) analysis to account for the variabilities in soil and structural properties. The results of the new SSI analysis are provided in the submittal.
I
The walkdown findings are described in detail in the submittal. Based on the walkdown findings, about 90 components were selected for detailed fragility evaluation. A median capacity of 1.5g or HCLPF capacity of 0.5g was used for the final screening of components and structures. All the structurn and block walls (reinforced) were screened out based on this criterion. A total of 17 components, including the generic small LOCA surrogate element, were screened in and included in the system models. Soil failure analyses were performed by Woodward-Clyde to evaluate slope stability, lateral spreading and liquifaction. The IICLPF capacity for potential liquifaction was estimated to be 0.5g. It was concluded that neither lateral spreading nor liquifaction was of concern to the structural integrity of buildings or piping connected to building.
The containment performance was evaluated by addressing various issues including structural integrity, containment isolation, containment bypass, and isolation signal. No unique vulnerabilitas with regard to the containment performance were identified.
The risk quantification was performed by constructing a new seismic event tree (SET) which consists of 11 seismic top events. The seismic damage states (SDS), which correspond to each of the sequences of the SET, were quantified by convoluting the component fragilities with hazard curves using the NUS SEISMIC code. The human error probabilities (llEP) used in the internal event PRA were increased by a factor of ten. The SDS's were then input to the internal event trees to calculate CDF values. The total seismic CDF for the LLNL hazard curves was 3.6E-06/ year (which is a factor of 3.5 less than the internal event CDF of 1.3E-05/ year) and 1.0E-06/ year for the EPRI hazard curves. Five SDSs represent 95% of the total CDF value, with SDS 36 being the largest single contributor at 69% of the total CDF. This SDS represents a seismic-induced failure of all four division 1E 120V AC instrumentation distribution panels.
-2.
REVIEW FINDINGS 2.1 IPEEE Format and Methodology Documentation The submittal appears to be consistent with the guidelines of NUREG-1407. The study addressed all the issues that are emphasized in NUREG-1407, including soit evaluation, plant walkdowns, nonseismic failure, human actions, and containment performance. The level of documentation is considered adequate in most of the areas.
2.2 Seismic Review Team Selection The seismic review team (SRT) consisted of personnel from PSE&G and several consultants, including EQE International, llalliburton NUS (now SCIENTECll) and Woodward-Clyde Consultants. EQE l
performed all the structural analyses including the SSI analyses and fragility evaluation, and Woodward-l Clyde performed soil related analyses. The SRT selection procedure is well documented, and the SRT l
selection appears to meet the NUREG-1407 objectives.
l 2.3 Ilazard Analysis The study used both the EPRI and the revised LLNL seismic hazard curves. A sensitivity analysis was performed to evaluate the effects of using a hazard curve cutofflower than 1.5g. By extrapolating the 2
I LLNL hazard curves from 1.0g up to 1.5g, the CDF value was increased by about 61 percent.
For performing the fragility analysis, the median uniform hazard spectrum (UllS) at a 10.000 year return period, based on the EPRI hazard analysis, was used.
2.4 Component Selection A component list is provided in the submittal which contains about 300 components. The list was developed based on the component list of the internal event PRA model, and by adding passive components, racks, electric panels / cabinets, and the components needed to address various issues such as j
the containment performance and seismic induced fires / floods.
l The component list provided in the submittal appears to be adequate. The " notes" referred to by the identification numbers in the submittal were provided in the R AI responses. After the prescreening, the final component screening was performed using screening criteria of 1.5g median capacity or 0.5g IICLPF capacity.
2.5 Plant Wallulown Approach The seismic walkdowns were conducted based on the guidelines of EPRI NP-6041. The purposes of the walkdowns, according to the submittal, were to prescreen rugged components, to identify component I
failure modes, and to address the required various issues such as spatial interaction, seismically induced I
fires / floods and actuation of fire protection systems. The prescreened components include most of the MOV's and AOV's, horizontal pumps / compressors, wall-mounted small instruments, and all the distribution systems such as piping, cable trays and IIVAC ducts.
A summary of walkdown findings are provided in Table 3-4 of the submittal and the notes to the Table were provided in the RAI responses.
2.6 Fragility Analysis 2.6.1 Structural Response Analysis Probabilistic reismic response analyses were performed by EQE hiternational for the Reactor Building, Auxiliary Building, Turbine Building and service water (SW) Intake Structure. The 10,000 year median EPRI UllS was used as the free-field ground motion. The variabilities in structural and soil properties were considered in generating multiple time histories. The listed coefficients of variation for the variabilities (in pp. 3-20 of the submittal) appear to be reasonable. The calculated floor response spectra are provided in the submittal. A comparison of these spectra with those of the free-field motions indicates that the building responses are significantly reduced. The median PG A's of the free-field motions are about 0.6g, while the median ZPA's in the buildings are about 0.2 to 0.3g. The SSI effects and the spectral shape of the UllS are considered to be the main contributors f +.his response reduction according to the submittal. The submittal provides details of the analysis procedures as well as the analysis results.
3
2.6.2 Structural Fragility Analysis An outline of the structural fragility analysis is provided in the submittal, which indicates that the conventional factor of safety approach was used. Ilased on the screening criterion of 1.5g median capacity, all the buildings were screened out. The results of the calculations are not provided in the submittal but sample calculations were provided la the RAI responses.
i 2.6.3 Component Fragility Analysis The analysis methodology, which is the conventional factor of safety approach. is outlined in the submittal.
Fragility calculations were performed for about 90 components, and the results are tabulated in Table 3-5 of the submittal. Most of the fragility parameters appear to be reasonable. Ilased on the screening criteria of 1.5g median or 0.5g IICLPF capacity, a total of 17 components were screened in for the system
- analysis, j
2.7 Soil Evaluation j
The liquefaction potential was assessed by Woodward-Clyde Consultants using a probabilistic approach.
j A IICLPF capacity of 0.5g was estimated. According to Fig. 3.10 of the submittal, the lateral spreading due to liquefaction of slopes becomes significantly large at a peak acceleration of about 0.35g. An RAI response stated that the only safety related buried piping was the service water piping in the yard area which is buried in a trench filled with compacted backfill for which Figure 3.11, rather than 3.10, is 1
applicable.
2.8 Relay Chatter Evaluation Approximately 100 potentially low ruggedness relays (LRR) were identified. All these relays were screened out because either (1) the LRR are not associated with safety shutdown or containment performance, (2) the relay chatter is acceptable, or (3) the LRR have high seismic capacity (section 3.1.5.4.3 of the submittal). This evaluation procedure appears to be acceptable. The RAls provided additional information on the relay chatter evaluation.
The licensee did not incorporate relay chatter into the PRA model. Instead, a screening review at the 0.3g l
level (which was the plant's review level earthquake) was performed. This approach would appear to somewhat underestimate the core damage frequency at higher g levels. PRA and seismic margin analysis (SMA) approaches were mixed for different components in the IPEEE seismic evaluation.
l 2.9 Containment Performance l
}
The licensee examined the containment performance from the standpoint of structural integrity, bypass, f
isolation and penetration failures (including hatches).
The containment structure was screened out based on the 1.5g median acceleration capacity criterion.
Regarding isolation failure, all the penetrations were screened out, as were the associated isolation valves, 4
cables, etc. No isolation valves require air to close. Ilowever, instrumentation distribution panels I A(B,C.D)J482 have a median acceleration capacity of 1.03g and a llCLPF value of 0.33g. Failure of these panels is assumed to lead to core damage but will also disable the automatic LOCA and high radiation isolation signals to non-NSSS primary containment isolation valves (PCIVs). Manual actuation of PCIVs is still possible from the control room, and is credited in the analysis. (The licensee states in the containment performance section that the manual recovery was not credited, however, from other parts of the submittal it is apparent that it was - see Section 2 10 below). Therefore, failure of these panels leads to core damage and an early containment failure. The frequency of such an event is 1.6E-7/yr using LLNL hazard curves and 4.6E-8/yr using EPRI curves, or about 5% of the total CDF.
The containment hatches do not depend on compressed air, so ' hat issue is not a concern.
Regarding bypass, all the isolation valves are rugged, and the all the relays were included in the relay chatter study, and no vulnerabilities were found.
The containment pressure suppression and heat removal systems (i.e., the RilR) were found to be rugged by the licensee (median acceleration capacity of at least 1.5g and/or a liCLPF of at least 0.5g), and were screened from further modeling.
In summary, the licensee did a credible job of looking for containment vulnerabilities. The containment was found to be sturdy against seismic damage, except for the effect on the isolation function from the distribution panels, which was modeled with the operator recovery.
2.10 Nonseismic Failures and Iluman Actions The licensee utilized the internal PRA model in the seismic analysis, so that nonseismic failures were automatically included. The internal PRA model seems to be based on the modified / updated IPE model.
A mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the diesel generators was used. No recovery of offsite power is credited.
The results of significant sequences includ,c random failures as well as seismic failures.
The human actions were modeled such that the internal PRA human error probabilities (IIEPs) were raised by a factor of 10. This was necessitated by the quantification scheme (see Section 2.12 below), such that no discrimination could be made with respect to the IIEPs based on the acceleration level of the earthquake. 07e new recovery action llEP was developed just for the seismic IPEEE.
This modeling of the llEPs is simplistic and probably leads to optimistic results. liaving IIEPs on the order of 1.E-4, or 1.E-3, or even 1.E-2 after a large earthquake seems optimistic. Some credit is given for actions that would occur relatively soon after the earthquake, i.e., within 40 minutes to I hour (llEPs on the order of I.E-2 or 1.E-1 are used).
The new HEP modeled recovery from failure of distribution panels I A(B,C,D)J482 (see Contairu.2nt performance above). The failure of these panels disables many automatic actions (e.g., diesel generator loading and sequencing, containment isolation, etc.). It seems that the recovery actions would need to i
occur relatively soon after the seismic event, and some of these recovery actions would be complex. Yet 5
k this action is given a composite 11EP of 0.063, for any earthquake (the Accident Sequence Evaluation Program (ASEP) methodology was used).
It is also not clear if there are other llEPs (besides the 10 " recovery actions" shown in the submittal) which were left at their IPE value.
The licensee did a sensitivity analysis on the llEPs, whereby all the llEPs reverted to their IPE value (except, of course the seismic specific one discussed above). An insignificant reduction in CDF resulted.
This indicates that these llEPs are not important in the model, (or perhaps all the other IIEPs were left at their IPE value). A better sensitivity analysis would have been to set all the llEPs equal to one, especially the short term ones. This would give an indication of how large a potential error could be made with such a simplistic treatment of the llEPs.
In conclusion, while the licensee tried to alter the 11EPs to take into account seismic conditions, the method used has some weaknesses: (1) no distinction is made based on the acceleration level of the earthquake, (2) short term actions were credited, (3) some unrealistic IIEP values resulted, and (3) the sensitivity analysis did not explore the area of interest, i.e., when all llEPs are set to 1.
It should also be noted that the internal events PRA used as a starting point is apparently substantially revised from the IPE (there is a considerable difference in results between the two) and has not been reviewed.
3.11 Seismic Induced Fires / Floods The licensee examined the issue of seismic fires. Ilowever, there is no mention of seismic floods, internal or external (other than in the context of the fire suppression system actuation / rupture). In the external flood section, there is no mention of a dam failure being a difficulty, so seismically induced external floods may not be a problem at this site.
(
The issue of seismic fires was examined from the standpoint of seismically induced fires, seismic actuation of the fire suppression systems and seismic degradation of the fire suppression systems.
In the area of seismically induced fires, the licensee addressed various flammable liquids and gases that are stored and/or used in the plant (hydrogen, fuel oil, lubricating oil, etc.), as well as unanchored non-1 E cabinets in close proximity to IE safety cabinets or safety related equipment. In all of these cases, the licensee found the equipment to be sturdy and/or not near safety equipment.
In the area of fire suppression equipment actuation, the walkdowns found the piping to be rugged and well supported, free of interaction concerns, and none of the low ruggedness relays were associated with the fire suppression equipment.
In the area of fire suppression systems degradatim, the licensee found the systems to be seismically rebust, except for the fire water pump house and the fire water tanks (median acceleration capacity of 0.73g, ilCLPF of 0.26g). Therefore, the fire water system is assumed to fail in a seismic event and no credit is taken for using firewater after a seismic event (the fire analysis also did not credit firewater, according to the submittal). Therefore the licensee found the principal effect of an earthquake, from the standpoint of 6
l seismic fire concerns, would be unavailability of firewater.
The licensee seems to have included most fire related concerns in the seismic walkdown; however, there is no discussion of flooding concerns.
2.12 Logic Models The licensee developed a seismic event tree, which inch. des important seismic failures. This tree is then linked with the rest of the internal events model. The rest of the internal events model seems to be only modified to the extent of increasing certain llEPs for recovery actions.
The seismic event tree seems to take into account important seismic failures of the equipment that was not screened out. Failure modes in this tree include loss of offsite power, loss of room cooling in certain plant areas, loss of certain important panels governing plant response to an initiator, loss of instrumentation buses, loss of de buses, status of the CST, status of containment venting and small LOCAs.
The licensee did include small LOCAs (with a median acceleration capacity of 1.50g and a HCLPF of 0.40g) among the initiators, llowever, there is no discussion of the initiators considered, how the small LOCA fragility values were derived, and why other LOCAs are not important. Nevertheless, it seems that most initiators of concern have been considered.
The seismic event tree seems consistent and to be logically correct. The output of this tree is linked to the internal events event trees. Certain sequences of this tree lead directly to core damage (e.g., unrecovered loss of control room ventilation).
Most seismic failures of concern have been considered in the submittal: piping failures, anchorage failures, relays (screened out), electrical cabinets, building failures (screened out), battery racks, IIVAC ducts, electrical trays, etc. In section 5.8, the licensee even considers the core damage frequency due to detritus raised in the river during the seismic event and clogging the service water intake screens (there have been severe problems with detritus clogging the SW screens at HCGS). The resulting CDF is between 5.2E-7/yr and 9.2E-7/yr, a significant fraction of the seismic CDF (but not included in that number) but below "the seismic included loss of service water, v hich is 1.33E-6 per year", according to the submittal. (This hst statement is confusing, as all SW components are screened out and were apparently not modeled).
No dependency matrix is provided in this section.
There is no discussion of correlated failures, but it seems that a perfect correlation was assumed (e.g., all 4 distribution panels fail).
No surrogate element was used to substitute for screened components. The licensee used a higher screening criterion than most (a HCLPF of 0.5g) and states that parametric studies show negligible contribution from screened components. Many important components are screened out, or are not used in the seismic event tree. There is no discussion as to why certain screened in components are not used in the seismic event tree, nor what would be the contribution from components not used and/or screened l
out (including relays). There is no discussion of instrument air and service water systems, or the diesel generators, although some of their components apparently were screened in during the walkdown 7
k screening.
The treatment of IIEPs in the model is optimistic. Only two HEPs are called out in the seismic event tree, i.e., are made independent of the internal PRA model. These are operator recovery of long term room cooling in the switchgear room, and operator shutdown from the remote shutdown panel. The other operator actions were left (by necessity) in the internal events model, which made it impossible to distinguish between different acceleration levels for such IIEPs (i.e.. creating acceleration dependent fragilities for the human actions), due to the quantification scheme employed. The quantification scheme called for convoluting the seismic hazard curve with the quantification of the seismic event tree, and then multiplying the result by the conditional core damage probability (modified by the changed 11EPs and by the failed events from the seismic event tree) from the internal events model. This scheme makes for simpler quantification but requires treatment of HEPs in the gross manner explained above. It also makes it impossible to separae the seismic CDF contribution from different seismic acceleration hins.
In summary, the licensee seems to have done a generally comprehensive and credible job with respect to the logic models, with the weaknesses noted above in the area of initiating events, screening, and the quantification scheme used for (and treatment of) HEPs.
2.13 Accident Frequency Estimate The seismic CDF using the LLNL hazard curves is 3.6E-6/yr, and using the EPRI hazard curves is 1.0E-6/yr.
The licensee performed several sensitivity analyses. In one, the LLNL hazard curves were extrapolated to 1.5g (from 1.0g maximum). The CDF rose by 61 % (f rom 3.6E-6/yr ta 5.8E-6/yr).
j Another sensitivity analysis regarding IIEPs (reduced to the IPE levels) was mentioned above. Negligible CDF reduction resulted (although a few of the sequences did decrease substantially). A more informative sensitivity analysis would have been to raise all the HEPs to 1.
The third sensitivity dealt with the ability of operators to recover from a loss of a certain panel, disabling IE instrumentation in the control room. No recovery was assumed in the model. When 50% success probability of shutdown without IE instrumentation is assumed, the CDF was reduced by 33%. Giving this action a 90% probability of success resulted in a 61 % reduction in the CDF. It should be noted that this manual control of equipment without instrumentation is possible, but is not proceduralized.
No importance analysis or uncertainty analysis was performed.
2.14 Dominant Contributors The licensee lists the dominant sequences. There is no breakdown of contribution by acceleration level bin, due to the quantification method employed.
The dominant sequences, accounting for 95" of the seismic CDF, are:
1)
A seismic failure of all 4 divisions of IE 120 V AC instrumentation distribution panels 1 8
4-A(B,C,D)J481. Leads directly to core damage. CDF of 2.5E-6/yr,69% of the total CDF.
2)
.A seismic induced failure of IE power to all four 125V DC distribution panels 1
' A(B,C,D)D417, -Leads directly to core damage. CDF of 4.4E-7/yr,12% of the total CDF.
3)
A seismic loss of off-site power (LOOP) and seismic failure of high pressure injection, with subsequent random failures (reactor depressurization and diesel generator failures dominant). CDF of 1.9E-7/yr,5% of the total CDF.
4)
A seismic failure of all four divisions of 1E 120V AC instrumentation distribution panels 1 A(B,C D)J482, with a failure of manual recovery of automatic functions. CDF of 1.6E-7/yr,4.5% of the total CDF.
5)
A seismic LOOP with random failures (dominated by emergency diesel generator (EDG) failures), CDF of 1.3E-7/yr. 3.5 % of the total CDF.
The dominant sequences are consistent with the information in the submittal, 2.15 Unresolved Safety Issues (USIs) and Generic Safety Issues (GSIs)
GSI-131 Potential Seismic Interaction involvine the Movable'In-Core Flux Mannine System Used in Westinchoose Plants i
GSI-131 is not applicable to this plant.
USI A-45 Shutdown Decay lleat Removal Reauirements Shutdown decay heat removal requirement were addressed in Section 3.2.1 of the submittal, and the licensee concluded that the decay heat removal function at the llCGS is seismically robust.
GSI-156 Systematic Evaluation Prouram (SEP)
GSI-156 is not applicable to this plant.
GSI-172 Multiple System Resnonse Prouram (MSRP)
GI-172 issues were addressed as follows:
)
The effects of fire protection system actuation was addressed in Section 4.8.1.2 of the submittal, where it is concluded that seismic actuation of fire suppression systems does not pose a significant risk of flood or a significant likelihood of disabling safety related equipment.
Seismic / fire interactions were addressed in Section 4.8.1 of the submittal. No credible failures were found.
Ilydrogen line ruptures were addressed on p.4-60 of the submittal. No concerns were found.
Seismic-induced flooding was not addressed.
9 1
- r 4
Seismic-induced spatial and functional interactions were addressed as part of the walkdown procedures discussed in Section 3.1.2.2 of the submittal.
Seismic-induced relay chatter is discussed in Section 2.8 of this review report and Section 3.1.5.4 of the submittal.
Failures related to human errors were addressed in Section 3.1.5.3.2 of the submittal.
2.16 Vulnerabilities / Plant Improvements Defining a plant vulnerability as "a scenario which contributes inordinately to the llCGS core damage frequency," the seismic IPEEE study found no fundamental weakness or vulnerability. No specific plant improvements with regard to the seismic events were proposed in the submittal.
3.0 OVERALL EVALUATION AND CONCLUSIONS The submittal appears to be consistent with the guidelines of NUREG-1407 in applying seismic PRA methodologies. The study addressed most of the major issues that are emphasized in NUREG-1407, including plant walkdown, relay chatter, liquefaction, nonseismic ' failure, human action, recent developments in seismic hazard evaluation, and containment performance, f
The completeness of the documentation appears to be adequate.
The treatment of the seismic hazard curves seems to be adequate. The study used both the EPRI and the revised LLNL hazard curves, and a sensitivity study was performed for hazard curve cutoff.
The walkdown procedure is described in the submittal in detail, and is considered to be adequate.
The evaluations for relay chatter, potential soil liquefaction, and containment performance appear adequate.
Description of the logic model was comprehensive and self-consistent.
The seismic-fire interaction description appears to be adequate.
1 Several minor weaknesses were identified:
Large and medium LOCAs are screened out without explanation. There is no discussion of the process used to arrive at the small LOCA fragilities.
Seismically induced internal and external flooding was not discussed.
Disposition of screened components in the logic model was not well explained.
While the licensee tried to alter the IIEPs to take into account seismic conditions, the method used has some weaknesses: (1) no distinction is made based on the acceleration level of the earthquake, (2) short term actions were credited, (3) some unrealistic IIEP values resulted, and (3) the sensitivity analysis did not explore the area of interest, i.e., all llEPs are set to 1.
10 J
7-i Overall, the licensee appears to have satisfied the objectives outlined in the Generic Letter with respect to the IPEEE.
4.0 REFERENCES
{l]
Hope Creek Generating Station Individual Plant Examination for External Events, Attachment to Letter dated July 31,1997 from Louis Storz, Sr. Vice President - Nuclear Operations - Public Service Electric and Gas Company, to USNRC.
[2]
Response to Request for Additional information Regarding Generic Letter 88-20, Supplement 4, 1 lope Creek Generating Station, Facility Operating License NPF-57, Doc Let No. 50-354, Attachment to Letter dated September 21,1998 from Louis Storz, Sr. Vice President - Nuclear Operations -
Public Service Electric and Gas Company, to USNRC.
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I HOPE CREEK GENERATING STATION INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE)
TECHNICAL EVALUATION REPORT FIRES 4
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