ML20091N525
| ML20091N525 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/01/1984 |
| From: | Douglas R, Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8406120353 | |
| Download: ML20091N525 (145) | |
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' i 7 Putic Service
}l d-- Qf Electnc and Gas E L.7 Company 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101
- Robert L. Mitti General Manager Nuclear Assurance and Regulation June 1, 1984 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue
~Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:
HOPE CREEK GENERATING STATION DOCKET NO. 50-354 DRAFT SAFETY EVALUATION REPORT OPEN ITEMS Pursuant to your letter dated March 5,1984, which transmitted the Hope Creek Draf t Safety Evaluation Report (SER), enclosed for your review and approval (see Attachment
- 2) are the resolutions to those Draft SER open items listed in Attachment 1.
Should you have any questi ns or require any additional information on these open items, please contact us.
Very truly yours ,
[ C D. H. Wagner (w/a ttach. )
USNRC Licensing Project Manager W. H. Bateman (w/a ttach. ) '
I: USNRC Senior Resident Inspector i'
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l ATTACHMENT 1 SECTION OPEN ITEM NUMBER SUBJECT Sa & d 2.4.5 Wave impact and runup on service water intake structure 7b 2.4.11.2 Thermal aspects of ultimate heat sink 9 2.5.4 Soil damping values 10 2.5.4 Foundation level response spectra 11 2.5.4 Soil shear moduli variation 12 2.5.4 Combination of soil layer properties 13 2.5.4 Lab test shear moduli values 14 2.5.4 Liquefaction analysis of river bottom sands 15 2.5.4 Tabulations of shear moduli i
16 2.5.4 Drying and wetting ef fect on Vincentown 17 2.5.4 Power block settlement monitoring 18 2.5.4 Maximum earth at rest pressure coefficient 19 2.5.4 Liquefaction analysis for service water piping 20 2.5.4 Explanation of observed power block settlement 21 2.5.4 Service water pipe settlement records 22 2.5.4 Cofferdam stability 23 2.5.4 Clarification of FSAR Tables 2.5.13 and 2.5.14 24 2.5.4 Soil depth models for intake structure 27 2.5.5 Slope stability M P84'80/12 1-gs Page 1 of 5
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ATTACHMENT 1 (Cont'd) l SECTION OPEN ITEM NUMBER SUBJECT 30 3.5.1.2 Internally generated missiles (inside containment) 41 3.8.2 Steel containment buckling analysis 42 3.8.2 Steel containment ultimate capacity analysis 43 3.8.2 SRV/LOCA pool dynamic loads 44 3.8.3 ACI 349 deviations for internal structures 45 3.8.4 ACI 349 deviations for Category I structures 1 46 3.8.5 ACI 349 deviations for foundations 47 3.8.6 Base mat response spectra 48 3.8.6 Rocking time histories 49 3.8.6 Gross concrete section 50 3.8.6 Vertical floor flexibility response spectra 53 3.8.6 Design of seismic Category I tanks 54 3.8.6 Combination of vertical responses 55 3.8.6 Torsional stiffness calculation 56 3.8.6 Drywell stick model development 57 3.8.6 Rotational time history inputs
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58 3.8.6 "O" reference point for auxiliary building model
'59 3.8.6 Overturning moment of reactor building foundation mat 60 3.8.6 BSAP element size limitations 61 3.8.6 Seismic modeling of drywell shield wall M P84 80/12 2-gs Page 2 of 5 E
ATTACHMENT 1 (Cont'd)
SECTION OPEN ITEM NUMBER SUBJECT 62 3.8.6 Drywell shield wall boundary conditions 63 3.8.6 Reactor building dome boundary conditions 64 3.8.6 SSI analysis 12 Hz cutoff frequency 65 3.8.6 Intake structure crane heavy load drop 67 3.8.6 Critical loads calculation for reactor building dome 68 3.8.6 Reactor building foundation mat contact pressures 69 3.8.6 Factors of safety against sliding and overturning of drywell shield wall 70 3.8.6 Seismic shear force distribution in cylinder wall 71 3.8.6 Overturning of cylinder wall 72 3.8.6 Deep beam design of fuel pool walls 73 3.8.6 ASHSD dome model load inputs 74 3.8.6 Tornado depressurization 75 3.8.6 Aaxiliary building abnormal pressure 76 3.8.6 Tangential shear stresses in drywell shield wall and the cylinder wall
- 77 3.8.6 Factor of safety against overturning
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of intake structure 78 3.8.6 Dead load calculations l 79 3.8.6 Post-modification seismic loads for the torus 80 3.8.6 Torus fluid-structure interactions 81 3.8.6 Seismic displacement of torus M P84 80/12 3-gs Page 3 of 5 i-
ATTACHMENT 1 (Cont'd)
SECTION OPEN ITEM NUMBER SUBJECT 82 3.8.6 Review of seismic Category I tank design 83 3.8.6 Factors of safety for drywell buckling evaluation 84 3.8.6 Ultimate capacity of containment (materials) 85 3.8.6 Load combination consistency 110b 4.6 Functional design of reactivity control systems 124 6.2.1.5.1 RPV shield annulus analysis 129 6.2.2 Insulation ingestion 152 9.4.4 Radioactivity monitoring elements 154 9.5.1.4.a Metal roof deck construction classificiation 159 9.5.1.5.a Primary and secondary power supplies for fire detection system 161 9.5.1.5.b Fire water valve supervision 162 9.5.1.5.c Deluge valves 163 9.5.1.5.c Manual hose station pipe sizing 164 9.5.1.6.e Remote shutdown panel ventilation 165 9.5.1.6.g Emergency diesel generator day tank protecton 182 15.9.10 TMI-2 Item II.K.3.18 185 7.2.2.2 Trip system sensors and cabling in turbine building 190 7.2.2.7 Regulatory Guide 1.75 192 7.2.2.9 Reactor mode switch 194 7.3.2.2 Standard review plan deviations M P84 80/12 4-gs Page 4 of 5 c.
ATTACHMENT 1 (Cont'd)
SECTION OPEN ITEM NUMBER SUBJECT 197 7.3.2.5 Microprocessor, multiplexer and computer systems 200 7.4.2.2 Remote shutdown system 205 7.5.2.4 Plant process computer system 209 7.7.2.3 Credit for non-safety related systems in Chapter 15 of the FSAR 210 7.7.2.4 Transient analysis recording system 218 9.5.1.1 Fire hazards analysis TS-3 4.4.5 Core flow monitoring for crud ef fects LC-1 4.2 Fuel rod internal pressure criteria i
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ATTACHMENT II s
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, HCGS DSER Open Item No. Sa and d (DSER Section 2.4.5)
WAVE IMPACT AND RUNUP ON SERVICE WATER INTAKE STRUCTURE The applicant has analyzed the wind waves that would traverse plant grade coincident with the PMH surge
__ hydrograph and runup on safety-related f acilities. These calculations were based on the assumption that wind waves would be generated in the Delaware Estuary and progress to the site. As the surge level would begin to rise, resulting from the approaching eye of the postulated hurricane, the wind speed would progressively change direction from the southeast clockwise to the west. Waves encroaching on the southern end of the Island would be depth-limited (i.e., the waves would " feel" bottom and thus become shallow water waves) by plant grade elevation on both the Salem and Hope Creek sites. These depth-limited (shallow water) waves will impact and runup on the southern and western f aces of the safety-related structures in the power block. The applicant has stated that the southern face of the Reactor Building and the Auxiliary Building are designed for a flood protection level of 38.0 ft msl or 3.2 ft above the maximum calculated wave runup height of 34.8 ft.msl and the other exposures of safety-related structures have a flood protection level of 32.0 ft msl or 1 ft above the maximum calculated wave runup height of 31.0 ft msl.
(The staff has requested the applicant to provide additional information on the waves that impact on the river face of service water intake structure. The waves inpacting on this face of the structure are not reduced in height (depth-limited) as those that traverse plant grade.]-Sa As indicated in Section 2.4.1, the applicant states that all accesses to safety-related structures (doors and hatches) are provided with water-tight seals designed to withstand the head of water associated with the flood protection levels. [But, the applicant has not indicated whether the water-tight doors are designed to withstand either the ;
combined loading effects of both static water level and the l dynamic wave impact]-5b or, [as cited in Sections 3.4.1 and j 3.5.1.4 of this report, the impact of a barge propelled by winds and waves associated with a hydrologic event that floods plant grade.]-Sc l
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Item No. Sa and d (Cont'd)
Based upon its analysis according to SRP 2.4.5, the staff concludes that the flood protection level of El. 38.0 ft msl for the southern face of the Reactor Building and Auxiliary Building and El. 32.0 ft msl for the remaining safety-related structures within the power block meets the requirements of Regulatory Guide 1.59. [Until additional information and analysis are available, the staff cannot
__ conclude that the flood protection level of El. 32.0 ft msl for the Servico Water Intake Structure meets the requirements of Regulatory Guide 1.59.]-5d Based on its analysis, the staff cannot conclude that the plant meets the requirements of GDC 2 with respect to the hydrologic aspects of Probable Maximum Surges and Seiche Flooding.
RESPONSE
The requested information for the. service water intake structure has been provided in the responses to the following NRC questions:
Information Provided Question No.
Wave runup elevations 240.8 Wave impact loads 240.9 Plood protection 240.8 and 410.69 M'P84 93 05 02-az
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- DSER Open Item No.-7b (DSER Section 2.4.11.2) e THERMAL ASPECTS OF ULTIMATE HEAT SINK The applicant has analyzed the ability of the service cooling water supply to withstand the effect of such severe
__ natural phenomena as ice blockage, flooding, low water, and thermal aspects of UHS. As indicated in Section 2.4.7, the
. effects of ice blockage would not obstruct the flow to 2 safety-related pumps. Thus the staff concludes that the intake structure and essential service cooling water flow is adequately protected against ice effects. [As indicated in Section 2.4.5, the ability of the service water intake
- - structure to withstand the effects of PMH surge flooding and associated wave runup and overtopping remains an open -
item.]-7b ,
! The applicant reported that the minimum historical low water level at the Reedy Point, Delaware tide station is -8.6 ft mal. The applicant's analysis of the maximum setdown i
5 considered the PMH wind speed of 85 mph (the overland PMH wind speed for the direction resulting in maximum setdown) to be blowing down the estuary coincident with 10%
exceedance low spring astronomical tide of -3.9 ft mal and the associated trough of the 6.0 f t maximum wind wave. The resultant low water level would be -13.0 ft mal. The applicant has stated that -13.0 ft mal is the design basis minimum low water level for service water pumps. Based on its independent analysis, the staff concurs that -13.0 ft
! mal is.an appropriate design basis minimum low water level.
[The applicant has not identified the maximum intake temperature that will allow the plant to safely shut down under normal and emergency conditions as discussed in i Regulatory Guide 1.27 nor the ability of the Delaware River 1
to supply water below this temperature. Until this I
-information is'available, the staff cannot conclude that the plant meets GDC' 44 with respect to the thermal aspects of UHS.]-7a m
' Based upon the evaluation' described above, we conclude the-hydrologic ~ characteristics of the Ultimate Heat Sink meet
-the requirements of'10 CFR Part 100 and 10 CFR Part 100, Appendix A. As' indicated above,'certain_aspets related to flooding' level for the service water intake structure are 4
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e DSER Open Item No. 7b (Cont'd) unresolved. Therefore, the staff cannot conclude that the Ultimate Heat Sink System meets the requirements of General Design Criterion 2 with respect to hydrologic characteristics. In addition, the staff cannot conclude that the Ultimate Heat Sink meets the requirements of GDC 4 e with respect to thermal aspects of the heat transfer systen.
-- RESPONSE For information on the ability'of the service water intake structure to withstand the effects of PMH surge flooding and associated wave runup and overtopping, see the response to DSER Open Item Number Sa and d.
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l pSER Open Item No. 9 ( DSER Section 2.5.4 )
SOIL DAMPING VALUES . ,
On the basis of the applicant's design criteria and construction specifications and the results of the applicant's investigation, laboratory and field tests, and analyses, and the results of the January 1984 audit, the staf f has concluded that the plant
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foundation will be adequate to safely . support the plant structures if the identified open items can be resolved.
RES PONSE This item corresponds to Item A.1 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No.10 ( DSER Section 2.5.4 )
4 FOUNDATION LEVEL RESPONSE SPECTRA ,
on the basis of the applicant's design criteria and construction
, specifications and the results of the applicant's investigation, laboratory and field tests, and analyses, and the results of the January 1984 audit, the staff has concluded that the plant
_ foundation will be adequate to safely support the plant structures if the identified open items can be resolved.
RESPONSE
This item corresponds to Item A.2 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 3 1984, from R. L. Mitti to A. Schwencer.
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' DSER Open Item No.11 ( DSER Se ction 2.5.4) i SOIL SHEAR MODULI YARIATION On the basis of the applicant's design criteria and construction specifications and the results of the applicant's investigation, laboratory and field tests, and analyses, and the results of the
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January 1984 audit, the staf f has concluded that the plant foundation will be adequate to safely support the plant structures if the identified open items can be resolved.
RESPONSE
This item corresponds to Item A.5 from the NRC Structural /Geotechnical meeting of Ja nua ry 10, 1984. A response to this item has been sub-mitted to the NRC by a letter dated April 24, 1984 from R. L. Mitti to A. Schwence r .
, HCGS DSER Open Item No. 12, ( DSER Se ct io n 2. 5. 4 )
COMBINATION OF SOIL IAYER PROPERTIES ,
Cn the basis of the applicant's design criteria and construction specifications and the results of the applicant's investigation, laboratory and field tests, and analyses, and the results of the January 1984 audit, the staff has concluded that the plant
_ . foundation will be adequate to safely support the plant structures if the identified open items can be resolved.
RESPONSE
This item corresponds to Item A.6 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No. 13 (DSER Section 2.5.4)
- I LAB TEST SHEAR MODULI VALUES On the basis of the applicant's design criteria and construction specifications and the results of the applicant's investigation, '
laboratory and field tests, and analyses, and the results of the
-- January 1984 audit, the staff has concluded that the plant foundation will be adequate to safely support the plant structures if the identified open items can be resolved.
RESPONSE
This item corresponds to Item A.8 from the NRC Structural /Geo-
- technical meeting of JL.suary 10, 1984. A response to this item has been submitted to tne NRC by.a letter dated April 24, 1984 from R. L. Mittl to A. Schwe nce r.
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i HCGS DSER Open Item No. 14 ( DSER Section 2.5.4)
LIQUEFACTION ANALYSIS OF RIVER BOTTOM SANDS ,
On the basis of the applicant's design criteria and construction specifications and the results of the applicant's : nvestigation, laboratory and field tests, and analyses, and the results of the January 1984 audit, the staff has concluded that the plant
_, foundation will be adequate to safely support the plant structures if the identified open items can be resolved.
RESPONSE
This item corresponds to Item A.15 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. S chwencer.
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9 HCGS DSER Open Item No. 15 (DSER Section 2.5.4)
TABULATIONS OF SHEAR MODULI on the basis of the applicant's design criteria and construction specifications and the results of the
__ applicant's investigation, laboratory and field tests, and analyses, and the results of the January 1984 audit, the staff has concluded that the plant foundation will be adequate to safely support the plant structures if the identified open items can be resolved.
RESPONSE
This item corresponds to Item B.6 from the NRC Structural /Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a i letter dated January 26, 1984, from R. L. Mitti to A.
Schwencer.
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r HCGS DSER Open Item No. 16 ( D6ER Section 2.5. 4 )
DRYING AND WETTING EFFECT ON VINCENTOWN .
On the basis of the applicant's design criteria and construction specifications and the results of the applicant's investigation, laboratory and field tests, and analyses, and the results of the January 1984 audit, the staff has concluded that the plant foundation will be adequate to safely support the plant structures if the identified open items can be resolved.
RESPONSE
This item corresponds to Item B.12 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. Schwencer.
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HCGS l
DSER Open Item No.17 ( DSER Section,2.5.4 )
PO JER BLOCK SETTLEMENT MONITORING .
On the basis of the applicant's design criteria and construction specifications and the results of the applicant's inve s tiga tion ,
laboratory and field tests, and analyses, and the results of
' the January 1984 audit, the staff has concluded that the plant
__ foundation will be adequate to safely support the plant structures if the identified open items can be resolved.
RESPONSE
This item corresponds to Item B.13 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No. 18 ( D6ER Section 2.5.4 )
MAXIMUM EARTH AT REST PRESSURE COEFFICIENT .
The below grade walls of structures were designed to resist both the static and dynamic pressure resulting f rom the surrounding earth and water. The value of the lateral earth pressure
'coef ficient at rest used in the design was 0.43. The dynamic
__ . lateral earth pressure on the below grade wall was determined f rom the results of soil-structure interaction analyses. The procedure uscd to obtain the dynamic lateral earth pressure is in accordance with the state-of-the-art methods required by the Standard Review Plan (NUREG-0800) and is there fore acceptable.
Although the lateral earth pressure at rest is low, during the structural and geotechnical engineering audit held in January 1984, the applicant demonstrated that the below grade walls have the capacity to resist substantially higher lateral earth pressures and will so state this face in a. future amendment to the FSAR.
RESPONSE
Section 2.5.4 has been revised to include a statement that the below grade walls have the capacity to resist lateral earth pressures substantially higher than the actual lateral earth pressures.
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i Several methods were used to compute the settlement of these j structures (References 2.5-115, -116, -117, and -118).. The results of these analyses indicate that the total maximus j settlement under the not loads is estimated to be about one inch
- including the recompression of heave. These analyses were i performed by assuming that the mats were uniformly loaded. The j settlement will be due, for the most part, to elastic l
deformations of the subsoil, a very small fraction being l contributed by the elastic deformation or the lean concrete on j structured backfill. Considering the granular soil type and that the total load of the structures consists mainly of dead load, most of the settlement will have occurred during construction.
As a result, post-construction differential settlement is
- expected to be less than 1/2 inch.
! The areas around the reactor, auxiliary, and turbine building i structures are back-filled to final grade with compacted well-j graded granular soils. The walls of these structures are i designed to resist the lateral pressures of the soils under :
i static and d"namic loadings. The static earth pressures are
- based on "at-rest" conditions, whereas the dynamic earth i pressures are determined based on soil-structure interaction i analysis discussed in Section 3.7.2.5. Figures 2.5-60 and 2.5-61 i provide the earth pressures used
. Addas"/,nerf design bases, A * ) ?
l 2.5.4.10.1.2 Service Water Intake Structure The Service Water Intake Structure, approximately 100 x 120 feet j in plan area, is a safety-related structure. It is located at
, the waterfront and consequently is partially submerged. The i structure will be founded on a mat at elevation +65.5. Tremie concrete will be placed between the base of the mat and the bearing level in the Vincentown sands. The unweathered greenish-gray Vincentown sands considered suitable as a bearing stratum occur at approximately elevation +25 feet in the intake area of i
the site and, borings and initial excavation operations at the
! location of the Service Water Intake structure encountered the unweathered Vincentown at approximately elevation +23 to +29 feet i (Reference 2.5-119). This lower occurrence of the bearing
, stratum in this area was taken into account in the configuration j
and calculated contact stresses of the intake structure.
l The stress relief due to excavation of approximately 70 feet of i submerged soil is expected to be 4000 lbs/fts. However, because i
the total excavation area is only 100 x 120 feet and because sheet piles extend below the excavation level, the elastic rebound is expected to be negligible. About 70% of the removed load will be restored by the time placement of lean concrete is completed at the proposed grade, elevation +65.5 feet. The not load to be imposed by the proposed construction is calculated to be very small because of stress relief and buoyancy effects.
2.5-120 Amendment 3 bMR ohn n;.m se .
HCGS INSERT A Although the static lateral earth pressures given in Figures 2.5-60 are low, the below-grade walls have the capacity to resist substantially higher lateral earth pressures. l h
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DSER Open Item No. 18 M P84 93 05'06-as
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HCGS DSER Open Item No. 1_9_ (DSER Section 2.5.4) i LIQUEFACTION ANALYSIS FOR SERVICE WATER PIPING f
The liquefaction potential was determined by comparing the
- shear stresses induced in the soil by the SSE with the l
-- cyclic shear strength of the soil under' field conditions.
The maximum shear stresses at various points in the foundation soils were obtained f rom previous dynamic j ana lyse s. The cyclic shear strength number for the Vincentown sands was determined through laboratory tests.
The dynamic strength of the soil layers overlying the Vincontown sands (hydraulic fill, river bottom sands, Kirkwood clays, and basal sands) was evaluated on the basis L of static strength tests, index properties, field tests, and correlation with data from literature, in addition to limited dynamic triaxial testing. On the basis of these results, the applicant concluded that only the sandy
! portions of the hydraulic fill may experience SSE-induced liquefaction.
l Because the safety-related structures were surrounded by hydraulic fill, the sliding stability under SSE condition t were further evaluated by the applicant. The applicant l concluded that, because the safety structures were embedded l at least 60 ft in soil and only the upper 30 ft could liquefy, this afforded at least 30 ft of stable soil
- confinement to the power block structure. In addition, the i applicant stated that the nonliquefiable backfill
! surrounding the structure would provide additional
! resistance to sliding. The staff concurs with the applicant's conclusion that the power block structures will be stable under SSR conditions.
However, the applicant has not provided suf ficient information about the sliding stability of the intake structure and the effects of potential liquefaction on the intake structure and the service water pipeline.
RESPONSE
This item corresponds to Item A.2 from the NRC Structural /Geotechnical meeting of January 11, 1984. A response ' to this item has been submitted to the-NRC by a letter ' dated February .17,1984, from R. L. Mitti to 1A. schwencer.
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'. HCGS DSER Open Item No. 20 ( DGER Section 2.5.4 )
EXPLANATION OF OBSERVED POWER BLOCK SETTLEMENT The staf f concurs in the applicant's assessment that the factor i
- of safety against bearing capacity failure is adequate. However, l the measured settlements presented in FSAR Figures 241.25-1 l through 241.25-30 show some erratic movements. The applicant has
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been requested to assess the observed settlements in the powe r l block area and to determine the settlement characteristics along l l the pipeline. ,
e l l RES PONSE l
This item corresponds to Item A.3 from the NRC Structural /
l Geotechnical meeting of January 11, 1984. A response to this l l item has been submitted to the NRC by a letter dated February 17, r l 1984, from R. L. Mitti to A. Schwencer.
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HCGS l
DSER Open Item No. 21 (D6ER Section 2.5.4)
SERVICE WATER FIPE SETTLEMENT RECORDS All safety-related structures as well as the turbine building are founded on lean concrete bearing on structural backfill placed on the denne to very dense sands of the Vincentown fo rma tion . Foundation levels, dimensions, and static loads for the major facilities of the station are listed in FSAR
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Table 2.5-18. The applicant has calculated the factor of safety for bearing capacity to be greater then 3. The calculated set-tienent is about 1 in., including the recompression of heave.
The postconstruction dif ferential settlement is expected to be less than 1/2 in. No settlement estimate is presented along the service water pipeline.
The staf f concurs in the applicant's assessment that the factor of safety against bearing capacity failure is adequate. Howeve r ,
the measured settlements presented in FSAR Figures 241.25-1 t he nisc h 241.25-30 show some erratic movements. The applicant has been requested to determine the settlement characteristics along the pipeline.
RESPONSE
This item corresponds to Item A.4 from the NRC Structural /Geo-technical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated April 24, 1984 from R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No. 22 (DSER Section 2.5.4)
COFFERDAM STABILITY , l On the basis of the applicant's design criteria and construction .
specifications and the results of the applicant's investigation, f laboratory and field tests, and analyses, and the results of the January 1984 audit, the staf f has concluded that the plant foundation will be adequate to safely support the plant struc- .
tures if the identified open items can be resolved. !
RESPONSE
This item corresponds to Item A.5 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated April 24, l 1984, from R. L. Mitti to A. Schwencer.
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HCGS D8ER Open Item No. 23 ( DSER Section 2.5.4)
CLARIFICATION OF FSAR TABLES 2.5-13 and 2.5-14 L 1
On the basis of the applicant's design criteria and construction specifications and the results of the applicant's investigation, i l laboratory and field tests, and analyses, and the results of the January 1984 audit, the staf f has concluded that the plant
, foundation will be adequate to safely support the plant structures i
! if the identified open items can be resolved.
l l RESPONSE f l
This iten corresponds to Item A.6 from the NRC Structural / I Geotechnical meeting of January 11, 1981. A response to this ,
! item has been submitted to the NRC by a letter dated February 17, "
1984, f rom R. L. Mitt 1 to A. Schwencer. !
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. HCGS DSER Open Item No. 24 ( DSER Section 2. 5. 4 )
SOIL DEPTH MODELS FOR INTAKE STRUCTURES On the basis of the applicant's design criteria and construction specifications and the results of the applicant's investigation, laboratory and field tests, and analyses, and the results of the January 1984 audit, the staf f has concluded that t% plant foundation will be adequate to safely support the plant structures i if the identified open items can be resolved .
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RESPONSE
This item corresponds to Item A.13 from the NRC Structural /
- Geotechnical meeting of January 11, 1984. A response to t.nis item has been submitted to the NRC by a letter da ted February 17, l 1984, from R. L. Mittl to A. Schwencer.
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HCGS DSER Open Item No. 27 (DSER Section 2.5.5)
SLOPE STABILITY The applicant stated in the FSAR that there are no natural slopes within the plant boundaries. However , there are slopes along the river bank in the vacinity of the intake structure, and the failure of these slopes could adversely af fect the intake structure. There-fore, the stability assessment of these slopes is required. The applicant, during the January 1984 audit, stated that the design analyses for the river bank slope protection will be provided in April 1984 for NRC review. The staff will provide its evaluation in a future supplement to the SER.
RESPONSE
This item corresponds to Item A. 5 from the NRC Structural /Geo-technical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a le tter da ted Apr il 24, 1984, frain R. L. Mitti to A. Schw ncer.
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HCGS DSER Open Item No. 30 (DSER Section 3.5.1.2) ;
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INTERNALLY GENERATED MISSILES (INSIDE CONTAINMENT) )
Based on our review, we cannot conclude that the design is in conformance with General Design Criterion 4 as it relates to protection against internally generated missiles. We cannot determine that the design of the f acility for :
providing protection from internally generated missiles meets the acceptance criteria of SRP Section 3.5.1.2.
RESPONSE
This item is not an open item per telephone conversation (see attached) between J. M. Ashley (PSE&G) and John Ridgely (NRC-ASB) on March 22, 1984.
JES:dh 6/1/84 M P84 54/07 1-dh v re- V -My y p - e- w ,p eemm ws isW- e-w--*- A g ,= --her
f TELEPHONE NOTES PSE&G Hope Creek Licensing (Bethesda)
Date: March 22, 1984 From: J.M. Ashley To: D. Wag n e r , J . Ridgely (ASB)
Subject:
HCGS DSER Open Items fi Discussion Ashley called to find out what NRC concerns existed with respect to FSAR Sections 3.5.1.2 (Item 30), 9.2.2 (Item 145) and 9.4.4 (Item 152).
Ridgely explained that these items were inadvertently listed as open items in the listing of open items at the front of the DSER. The NRC has no outstanding concerns with the sections.
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HGCS DSER Open Item No. 41 ( DSER Section 3.8. 2 )
STEEL CONTAINMENT BUCKLING ANALYSIS The applicant has been requested to submit information regarding the ultimate capacity analysis of the containment and steel containment buckling analysis. The staff has not received all
' the required information on these two items. The applicant has committed to provide the required information to the staff for review by February 15, 1984. The staff will review and report its resolution of these two items in the Final SER.
RESPONSE
A description of the drywell buckling evaluation has been provided in FSAR Appendix '3E in response to Question 220.11. Additional information has been requested in Item B.1 from the NRC Structural /
Geotechnical meeting of January 12, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. S chwencer.
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HCGS DSER Open Item No. 42 ( DSER Section 3.8.2)
STEEL CONTAINMENT ULTIMATE CAPACITY ANALYSIS The applicant has been requested to submit information regarding the ultimate capacity analysis of the containment and steel containment buckling analysis. The staf f has not received all the required information on these two items. The applicant has committed to provide the required information to the staf f for review by February 15, 1984. The staff will review and report its resolution of these two items in the Final SER.
RESPONSE
A description of the ultimate capacity analysis of the contain-ment has been provided in FSAR Appendix 3I in response to Question 220.22 Additional information was requested in Item B.2 from the NRC Structural /Geotechnical meeting of January 12, 1984. A response to this item was submitted to the NRC by a letter dated Feb rua ry 17, 19 84, f rom R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No. 43 (DSER Section 3.8.2)
.SRV/LOCA POOL DYNAMIC-LOADS With respect .to SRV/LOCA pool dynamic load considerations, the applicant has performed a reevaluation of containment
-designzadequacy based on staff positions provided in NUREG-0661. The applicant has submitted his reevaluation summary report. However, the staff has not completed its
, review. It will report on the resolution of these issue in the Final SER.
RESPONSE
The Plant Unique Analysis Report (PUAR) , which describes the reevaluation of the containment design adequacy based on staff positions provided in NUREG-0661, was submitted to the NRC by a letter dated February 10, 1984, from R. L. Mitti to A. Schwencer.
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. HCGS DSER Open Item No. 44 ( DSER Section 3.8.3 )
ACI 349 DEVIATIONS POR INTERNAL STRUCTURES SRP Section 3.8.3 specifies that the code to be used in the design of concrete internal structures is ACI Std 349 as augmented by RG 1.142. The applicant had been requested to provide in fo rmation
' regarding an assessment and justifications for all deviations of his internal structural design and analysis from the applicable staf f positions as given in SRP Section 3.8.3. The applicant provided the information on January 23, 1984. However , the staf f has not completed its review. It will report on the resolution of this item in the Final SER. Additionally, some of the 39 structural audit action items discussed under Section 3.8.6, as they pertain to this section of the SER, are considered unresolved.
RES PONSE The requested information is included in the response to NRC Question 220.24.
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HCGS DSER Open Item No. 4 5 ( DSER Section 3.8.4 )
ACI 349 DEVIATIONS FOR CATEGORY I STRUCTURES Category I structures other than the containment and its interior structures are all of structural steel and concrete. The struc-tural components consist of slabs, walls, beams, and columns.
"The major code used in the design of concrete Category I struc-tures is ACI Std 381-71. For steel Category I structuros, AISC
" Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings" is used. The applicant had been requested to provide an assessment and justifications of all
- deviations from _the applicable requirements of ACI 349 as augmented by RG 1.142. The applicant provided the inf o rma tion on January 23, 1984. However, the staf f has not completed its j review. It will report on the resolution of this item in the Final SER. Additionally, some of the 39 action items discussed in Section 3.8.6 of this SER pertain to this section, and the items remain to be resolved to the satisf aciton of the staf f.
RESPONSE
The requested information is included in the response to NRC Ouestion 220.26.
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HCGS l DSER Open Item No. 46 (DSER Section 3.8.5)
! ACI 349 DEVIATIONS FOR FOUNDATIONS The design and analysis procedures that were used for these Category I foundations are the same as those approved on previously licensed applications and, in general, are in accordance with procedures delineated in the ACI 318-71.
The various Category I foundations were designed and propor-tioned to remain within limits established by the staf f under the various load combinations. These limits are, in general, based on ACI 318-71 and on the AISC specification for concrete and steel structures, respectively, modified as appropriate for load combinations that are-considered extreme. The applicant had been requested to provide an assessment and justifications of all deviations of this design f rom the applicable requirements of ACI 349 as aug-mented by RG 1.142. The applicant provided the information on January 23, 1984. However, the staff has not completed its review. It will report its resolution of this issue in the Final SER. In the meantime, this item remains open.
Furthermore, some of the action items discussed in Section 3.8.6 of this SER, as they pertain to the foundation design and analysis, should be considered open items and remain to be resolved.
RESPONSE
The rc aested information is included in the response to NRC Question 220.26.
M P84-54/07 3-dh l
HCGS DSER Open Item No. 47 ( DSER Section 3.8. 6)
BASE MAT RESPONSE SPECTRA Frca January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope
' Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any
, additional questions, which may be raised further, will be reported
- in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item A.3 from the NRC Structural /
i Geotechnical meeting of January 10, 1984. A response to this i item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. Schwencer.
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HCGS 1 DSER Open Item No. 4 8 ( DSER Section 3.8. 6)
ROCKING TIME HISTORIES l From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
I As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported i in the Final SER. The resolution of these action items will be needed betore the issuance of the Final SER.
RESPONSE
This item corresponds to Item A.4 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. S chwencer.
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. DSER Open Item No. 4 9 ( DSER Section 3.8.6)
GROSS CONCRETE SECTION From January 10' through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
' The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39
. action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
i This item corresponds to Item A.11 from the NRC Structural /
, Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17,
- 1984, from R. L. Mitti to A. S chwencer.
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HCGS DSER Open Item No. 50 ( DSER Section 3.8. 6)
VERTICAL FLOOR FLEXIBILITY RESPONSE SPECTRA .
From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item A.12 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this
, item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. S chwe ncer .
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i HCGS DSER Open Item No. 53 (DSER Section 3.8.6)
DESIGN OF SEISMIC CATEGORY I TANKS From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit. The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items. The applicant has rubmitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
RESPONSE
This item corresponds to Item A.4 f rom the NRC Structural /
Geotechnical meeting of January 12, 1984. A response to this item has been submitted to the NRC by a letter dated April 24, 1984, from R. L. Mitti to A. Schwencer.
M P84.54/07 4-dh-
o HCGS DSER Open Item No. 54 ( DSER Section 3.8. 6)
COMBINATION OF VERTIC AL RESPONSES From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit. ,
The audit covered each major safety-related structure at the Hope
' Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 act ion - iterc . The staff in in Lhe process of reviewing thece responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item B.5 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. S chwe ncer .
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HCGS DSER Open Item No. 55 ( DSER Section 3.8.6)
TORSIONAL STIFFNESS CALCULATION .
From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
- The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. . The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised f urther, will be reported l in the Final SER. The resolution of these action items will be
- needed before the issuance of the Final SER.
RESPONSE
'This item corresponds to Item B.8 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this l item has been submitted to the NRC by a letter dated January 26, l 1984, from R. L. Mittl to A. Schwencer.
i K51/2-45
. HCGS . 4-DSER Open Item No. 5 6 ( DSER Section 3.8.6)
DRYWELL STICK MODEL DEVELOPMENT From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope
' Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these res ponses . The final resolution of the action items and any additional questions, which may De raised f urther, will be reported in the Final SUR. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item B.9 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mittl to A. S chwencer.
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j' HCGS DSER Open Item No. 57 ( DSER Section 3.8. 6)
ROTATIONAL TIME HISTORY INPUTS .
From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope
' Creek Generating Station.
As a, result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these res ponses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
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!' This item corresponds to Item B.10 from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this
- item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. Schwencer.
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DSER Open Item No. 58 (DSER Section 3.8.6)
"O" REFERENCE POINT FOR AUXILIARY BUILDING MODEL From January 10 through January 12, 1984, the staff met with the applicant and his consult ar;ts to conduct the structural audit. The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items. The applicant has submitted preliminary responses to 22 of the 39 a.ction itees. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
RESPONSE
This item corresponds to Item B.ll from the NRC Structural /
Geotechnical meeting of January 10, 1984. A response to this ittm has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mittl to A. Schwencer.
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HCGS OSER Open Item No. 59 ( DSER Section 3.8.6)
OVERTURNING MOMENT OF REACTOR BUILDING FOUNDATION MAT From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a resdit of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item A.7 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mitti to A. Schwencer.
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DSER Open Item No. 60 (DSER Section 3.8.6)
BSAP ELEMENT SIZE LIMITATIONS .
Frcun January 10 through January 12, 1984, the staf f met with the l applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staf f is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised f urther, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item A.8 from the NRC Structural /
Geotechnical meeting of January ll, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mittl to A. S chwence r .
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HCGS DSER Open Item No. 61 (DSER Section 3.8.6)
SEISMIC MODELING OF DRYWELL SHIELD WALL From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit. The audit covered each _ major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items. The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
RESPONSE
This item corresponds to Item A.9 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mittl to A. Schwencer.
14 P84 54/07 6-dh
. HCGS DSER Open Item No. 62 ( DSER Section 3.8. 6)
DRYWELL SHIELD WALL BOUNDARY CONDITIONS From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these res ponses . The final resolution of the action items and any additional questions, which may be raised eurther, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
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, This item corresponds to Item A.10 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No. 63 ( D6ER Section 3.8. 6)
REACTOR BUILDING DOME BOUNDARY CONDITIONS From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope
- Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item A. ll from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, l
1984, from R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No. 64 (DSER Section 3.8.6)
SSI ANALYSIS 12Hz CUTOFF FREQUENCY From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural
. audit. The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items. The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The
! resolution of these action items will be needed before the l issuance of the final SER.
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RESPONSE
This item corresponda to Item A.12 f rom the NRC Structural /
l Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984, from R. L. Mitti to A. Schwencer.
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. HCGS DSER Open Item No. 6 5 ( DSER Section 3.8.6)
INTAKE STRUCTURE CRANE HEAVY LMD DROP From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
.The audit covered each major safety-related structure at the Hope
' Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staif is in the process of reviewing these res ponses . The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item A.15 from the NRC Structural /
Geotechnical meeting of January 11, 1384. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mitti to A. Schwencer.
K51/2-55
i HCGS DSER Open Item No. 67 (DSER Section 3.8.6)
CRITICAL LOADS CALCULATION FOR REACTOR BUILDING DOME From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope
' Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staf f is in the process of reviewing these res ponses. The final resolution of the action items and any 4
additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item A.17 from the NRC Structural /
, Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26,
- 1984, from R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No. 68 (DSER Section 3.8.6)
REACTOR BUILDING FOUNDATION MAT CONTACT PRESSURES From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit. The audit covered each major safety-related structure at the Hope Creek Generating Station.
^
As a result of the audit, the staff identified 39 action items. The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
RESPONSE
This item corresponds to Item B.1 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mittl to A. Schwencer.
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HCGS DSER Open ' tem No. 69 ( DSER Section 3.8.6)
FACIORS OF SAFETY AGAINST SLIDING AND OVERTURNING OF '
DRYWELL SHIELD WALL l
l Fran January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
'The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the' issuance of the Final SER.
RESPONSE
This item corresponds to Item B.2 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mitti to A. Schwencer.
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. HCGS DSER Open Item No. 70 ( DSER Section 3.8. 6)
SEISMIC SHEAR EURCE DISTRIBUTION IN CYLINDER WALL .
From January 10 through January 12, 1984, the staf f met with the applicant and hic consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item B.3 from the NRC Structural /
Geotechnical meeting of January ll, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, j 1984, from R. L. Mitti to A. Schwencer.
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. HCGS DSER Open Item No. 71 (DSER Section 3.8.6)
OVERTURNING OF CYLINDER WALL From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope
' Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item B.4 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mittl to A. S chwencer.
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A HCGS DSER Open Item No. 72 ( DSER Section 3.8. 6)
DEEP BEAM DESIGN OF FUEL POOL WALLS .
From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
,The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these res ponses . The final resolution of the action items and any additional questions, which may be raised f urther, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item B.5 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mitti to A. S chwe nce r .
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HCGS DSER Open Item No. 73 ( DSER Section 3.8. 6)
ASHSD DOME MODEL LOAD INPUTS i
[ From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-rclated structure at the Hope
" Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
l The applicant has submitted preliminary responses to 22 cf the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This iten corresponds to Item B.6 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mitti to A. S chwencer.
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l HCGS DSER Open Item No. 74 ( DSER Section 3.8. 6) l TORNADO DEPRESSURIZATION ,
From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
,The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these
. res ponse s . The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item B.7 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, from R. L. Mittl to A. S chwencer.
K51/2-63
HCGS DSER Open Item No. 75 ( DSER Section 3.8.6)
AUXILIARY BUILDING ABNORMAL PRESSURE From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope
' Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staf f is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item B.8 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, fram R. L. Mitti to A. Schwencer.
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l HCGS DSER Open Item No. 76 ( DSER Section 3.8. 6)
TANGENTIAL SHEAR STRESSES IN DRYWELL SHIELD WALL AND THE CYLINDRICAL WALL
- From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
' The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these res ponses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the Final SER.
RESPONSE
This item corresponds to Item B.9 from the NRC Structural /
Geotechnical meeting of January 11, 1984. A response to this i
item has been submitted to the NRC by a letter dated January 26, 1984, tram R. L. Mittl to A. S chwencer.
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HCGS DSER Open Item No. 77 (DSER Section 3.8.6)
FACTOR OF SAFETY AGAINST OVERTURNING OF INTAKE STRUCTURE From January 10 through Januray 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the~ staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the
. 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items
- will be needed before the issuance of the final SER.
i RES PONSE This item corresponds to Item B.12 f rom the NRC Structural /Geo-technical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984 from R. L. Mittl to A. Schwencer.
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HCGS DSER Open Item No. 78 (DSER Section 3.8.6)
DEAD LOAD CALCULATIONS From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminnary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
RES PONSE This item corresponds to Item B.13 from the NRC Structural /Geo-technical meeting of January 11, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984 i from R. L. Mitti to A. Schwencer.
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t HCGS DSER Open Item No. 79 (DSER Section 3.8.6)
POST-MODIFICATION SEISMIC LOADS FOR TORUS From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
i RES PONSE 4 This item corresponds to Item A.1 f rom the NRC Structural /Geo-technical meeting of January 12, 1984. A response to this item i has been submitted to the NRC by a letter dated January 26, 1984 I from R. L. Mittl to A. Schwencer.
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HCGS 1
DSER Open Item No. 80 (DSER Section 3.8.6)
TORUS FLUID-STRUCTURE INTERACTIONS From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these
, responses. The final resolution of the action items and any
.i additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
. RESPONSE This item corresponds to Item A.2 from the NRC Structural /Geo-I technical meeting of January 12, 1984. A response to this item u has been submitted to the NRC by a letter dated January 26, 1984 i f rom R. L. Mittl to A. Schwencer.
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HCGS DSER Open Item No. 81 ( DSER Section 3.8. 6)
SEISMIC DISPLACEMENT OF TORUS From January 10 through January 12, 1984, the staf f met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staf f identified 39 action items.
The applicant has submitted preliminary responses to 22 of the action items. The staf f is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised f urther, will be i reported in the Final SER. The resolution of these action action items will be needed before the issuance of the Final SER.
RESPONSE
This iten corresponds to Item A. 3 from the NRC Structural /
Geotechnical meeting of January 12, 1984. A response to this item has been submitted to the NRC by a letter dated January 26, 1984, frca R. L. Mitti to A. Schwencer.
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HCGS DSER Open Item No. 82 (DSER Section 3.8.6)
REVIEW OF SEISMIC CATEGORY I TANK DESIGN From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit. The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items. The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
RESPONSE
This item corresponds to Item A.4 from the NRC Structural /
Geotechnical meeting of January 12, 1984, A response to this item has been submitted . to the NRC by a letter dated April 24, 1984, from R. L. Mitti to A. Schwencer.
M P84 54/07 9-dh k _ . -_ _ _ __ _m _. -
HCGS DSER Open Item No. 83 (DSER Section 3.8.6)
FACTORS OF SAFETY FOR DRYWELL BUCKLING EVALUATION From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER. s RES PONSE This item corresponds to Item B.1 f rom the NRC Structural /Geo-technical meeting of January 12, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984 from R. L. Mittl to A. Schwencer.
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e HCGS DSER Open Item No. 84 (DSER Section 3.8.6)
ULTIMATE CAPACITY OF CONTAINMENT (MATERIALS)
From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
RES PONSE This item corresponds to Item B.2 from the NRC Structural /Geo-technical meeting of January 12, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984 f rom R. L. Mitti to A. Schwencer.
K53/2 L
- HCGS DSER Open Item No. 85 (DSER Section 3.8.6)
LOAD COMBINATION CONSISTENCY From January 10 through January 12, 1984, the staff met with the applicant and his consultants to conduct the structural audit.
The audit covered each major safety-related structure at the Hope Creek Generating Station.
As a result of the audit, the' staff identified 39 action items.
The applicant has submitted preliminary responses to 22 of the 39 action items. The staff is in the process of reviewing these responses. The final resolution of the action items and any additional questions, which may be raised further, will be reported in the Final SER. The resolution of these action items will be needed before the issuance of the final SER.
RES PONS E This item corresponds to Item B.3 f rom the NRC Structural /Geo-technical meeting of January 12, 1984. A response to this item has been submitted to the NRC by a letter dated February 17, 1984 f rom R. L. Mitti to A. Schwencer.
KS3/2
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'DSER Open Item No. Il0b (DSER Section 4.6) i 4
l FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS F The control rod drive system was reviewed in accordance with
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Section 4.6 of the Standard Review Plan (SRP), NUREG-0800.
- An audit- review of each of the areas listed in the " Areas of f Review" portion of the 'SRP section was performed according to the guidelines provided in the " Review ProcedLres" portion of the SRP section. Conformance with the acceptance criteria formed the basis for our evaluation of the control 4
rod drive' system with respect to the applicable regulations of 10 CFR 50.
[The applicant has not addressed the recommendations of
- NUREG-0803, " Generic Safety Evaluation Report Regarding i Integrity of BWR scram System Piping. "]-110a
! The design does not utilize a CRDS return line to the l reactor pressure vessel. In accordance with NUREG-0619, 4
. "BWR Feedwater Nozzle and Control Rod Drives Return Line Nozzle Cracking," dated November 1980, equalizing valves are
- installed between the cooling water header and exhaust water header, the flow stabilizer loop is routed to the cooling
1 We have reviewed the extend of conformance of the Scram Discharge Volume (SDV) design with the NRC generic study, ' ',, 3
. "BWR Scram Discharge System Safety Evaluation," dated -
l December 1,1980. The design provides two separate SDV
- headers, with an integral instrumented volume (IV) at the i- end of each header, thus providing close hydraulic coupling. Each IV has redundant and diverse level instrumentation (float sensing and pressure sensing) for the scram function attached directly to the IV. Vent and drain
- lines are completely separated and contain redundant vent
!- and drain valves with position indication provided in the
!- main control' room. [With respect to Design Criterion 8, the
! applicant stated that the "SDV Piping is continuously sloped f from its high point to its low point." In order to provide
- a response to Design Criterion 8, the applicant must provide i
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MP84 95 01 1-bp
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HCGS DSER Open Item No. 110b (DSER Section 4.6) (Cont'd) a description of the SDV from the beginning of the SDV to the IV drain. The description should include piping geometry (ie., pitch, line size, orientation).]-110b Except for Design Criterion 8, we conclude that the design of the SDV fully meets the requirements of the above referenced NRC generic SER and is therefore acceptable.
Additionally, the above-described design of the SDV satisfies LRG-II, Item 1-ASB, "BWR Scram Discharge Volume Mod ifica tions . "
Based on our review, we conclude that the functional design of the reactivity control system meets the requirements of General Design Criteria 23, 25, 26, 27, 28, and 29 with respect to demonstrating the ability to reliably control reactivity changes under normal operation, anticipated operational occurrences and accident conditions including single failures, and the guidelines of NUREG-0619 and is, ,
therefore, acceptable. We cannot conclude compliance with the guidelines of NUREG-0803 and the generic document dated December 1, 1980. The functional design of the reactivity control system does not meet the applicable acceptance Criteria of SRP 4.6. We will report resolution of these items in a supplement to this SER.
RESPONSE
FSAR Section 4.6.1.2.4.2(f) has been revised to include a description of the SDV piping.
l MP34 95 01 2-op l
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room. Differential pressure between the reactor vessel and the cooling water header is indicated'in the main control room. Although the drives can function without ,
cooling water, seal life is shortened by long-term exposure to reactor temperatures. The temperature of each drive is indicated and recorded, and excessive temperatures are annunciated in the main control room.
j e. Exhaust water header - The exhaust water header connects to each HCU and provides a low pressure plenum and discharge path for the fluid expelled from the drives during control rod insert and withdraw i operations. The fluid injected into the exhaust water header during rod movements is discharged back up to
! the RPV via reverse flow through the insert exhaust directional solenoid valves of adjoining HCUs. The pressure in the exhaust' water header is, therefore, maintained at essentially reactor pressure. To ensure that the pressure in the exhaust water header is maintained near reactor pressure during the period of l vessel pressurization, redundant pressure equalizing valves connect'the exhaust water header to the cooling ,
water header.
2.mehdime-h
! f. Scram discharge volume - The cram cischarge volume (SDV) consists of two sets of header piping, each of which connects to one-half of the HCOs and drains into ,
hd d'a'"Mscram discharge instrument volume (SDIV) . Each set
.of header piping is sized to receive and contain all i
the water discharged by one-half of the drives during a scram, independent of the SDIV. i M **th a minim sam j Ths, h eaole p:ta.s r f,*p a p VS o ny
,, as par3feat/*P* aa
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The SDIV for each header set is directly connected to the low point of the header piping. The large-diameter pipe of each SDIV thus serves as,a vertical extension of the S.DV. A A Och pipiny conneedson at the bosom ol'the SMV prevedes dreineqe of us 3btV and sh y ysa sfoped drajn
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a 1:ess with e owinteresseT '/S in ch p e r foo t a lop s.
During normal plant operation, the SDV is empty and is vented to the atmosphere through its open vent and drain valves. When a scram occurs, upon a signal from i' the safety circuit, these vent and drain valves are closed to conserve reactor water. Redundant vent and drain valves are provided to ensure against loss of r
reactor coolant from the SDV following a scram. Lights in the main control room indicate the position of these (
l- valves.
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O s t R O P C N tt'c m s t o b
HCGS DSER Open Item No. 124 (DSER Section 6.2.1.5.1)
RPV SHIELD ANNULUS ANALYSIS The applicant's analysis resulted in pressures in the shield annulus that peak at approximately 90 psia in the volumes surrounding the recirculation line break and approximately 100 psia in the volumes surrounding the feedwater line
. break. [The applicant has not provided a graphical presen-tation of the differential pressure (psi) responses as a function of time for a selected number of nodes, as requested.]-124a
[In addition, the applicant has not provided the peak and transient loading on the major components used to establish 124b the adequacy of the supports design. This should include the load forcing functions (e.g., fx(t), fy(t), fz(t)) and transient moments (e.g., Mx(t), My(t), Mz(t)) as resolved about a specific identified coordinate system.] [The applicant also has not provided the projected areas used to 124c calculate these loads. This information was also previously requested.] The staff intends to perform confirmatory analysis using the COMPARE code upon receipt of this.
In fo rma tion.
RESPONSE
The graphical presentation of differential pressure is not required per March 30, 1984 conference call between the NRC and Bechtel. Bechtel noted that the initial containment pressure could be considered constant during the transient and thus differential pressure can be determined by subtracting a c5nstant initial pressure from the already provided graphical presentations of absolute pressure.
It should be noted that Part b is being reconsidered by the NRC and will be provided later if necessary.
The requested projected areas for 'the RPV Shield Annulus Analysis _are provided in the attached table.
i MP84 95 01 3-bp
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VESSEL SHIELD NODE AX AY AX AY 1 5747 1540 6875 1842 2 4207 4207 5033 5033 1 3 1540 5747 1842 6875 l 6875
_ 4 1540 5747 1842 5 4207 4207 5033 5033 6 5747 1540 6875 1842 7 5747 1540 6875 1842 8 4207 4207 5033 5033 9 1540 5747 1842 6875 10 1540 5747 1842 6875 11 4207 4207 5033 5033 12 5747 1540 6875 1842 13 8281 2219 9906 2654 '
14 6062 6062 7252 7252 15 2219 8281 2654 9906 16 2219 8281 2654 9906 17 6062 6062 7252 7252 18 8281 2219 9906 2654 19 8281 2219 9906 2654 20 6062 6062 7252 7252 21 2219 8281 2654 9906 22 2219 8281 2654 9906 23 6062 6062 7252 7252 24 8281 2219 9906 2654 ;
25 6343 1700 7588 2033 26 4644 4644 5555 5555 27 1700 6343 2033 7588 28 1700 6343 2033 7588 29 4644 4644 5555 5555 30 6343 1700 7588 2033 31 6343 1700 7588 2033 32 4644 4644 5555 5555 33 1700 6343 2033 7588 34 1700 6343 2033 7588 35 4644 4644 5555 5555 4
, .. r- w .c 12 4 -/
I VESSEL SHIELD NODE AX AY AX AY 36 6343 1700 7588 2033 '
- 2033 37 6343 1700 7588 38 4644 4644 5555 5555 39 1700 6343 2033 7588 ,
- 40 1700 6343 2033 7588 41 4644 4644 5555 5555 42 6343 1700 7588 2033
- 43 6343 1700 7588 2033 44 4644 4644 5555 5555 45 1700 6343 2033 7588 i 46 1700 6343 2033 7588 47 4644 4644 5555 5555 '
48 6343 1700 7588 2033 49 8347 2237 9985 2676 50 6111 6111 7310 7310 51 2237 8347 2676 9985 52 2237 8347 2676 9985 53 6111 6111 7310 7310 54 8347 2237 9985 2676 55 8347 2237 9985 2676 56 6111 6111 7310 7310 57 2237 8347 2676 9985 50 2237 8347 2676 9985 59 , 6111 6111 7310 7310 60 8347 2237 9985 2676 61 3876 10 38 4636 1242 ,
, 62 2837 2837 3394 3394 i 63 10 38 3876 1242 4636 64 1038 3876 1242 4636 65 2837 2837 3394 3394 C6 3876 1038 4636 1242 67 3876 1038 4636 1242 68 2837 2837 3394 3394 69 1038 3876 1242 4636 70 1038 3876 1242 4636 71 2837 , 2837 3394 3394 72 3876 1038 4636 1242 i
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i HCGS DSER Open Item No. 129_ (DSER Section 6.2.2) i INSULATION INGESTION With respect to insulation debris generation and transport, these issues are insulation types and quantity dependent,
_. and also plant design dependent. HCGS plans to use fiberglass blanket sections with 22 gauge, stainless steel jacketing to insulate structures, equipment and piping within the primary containment. LOCA breaks have the capability to locally destroy (or f ragment) fibrous insulation. The FSAR does not address the question of LOCA generated debris transport at low velocities (i.e., 0.2 -0.4
, ft/sec). The FSAR alleges that it is unlikely that insulation materials would transport to and plug the suction strainers without providing a quantified treatment regarding the minimal amount of insulation destruction and transport thereof. Flow velocities in the vicinity of the RHR suction strainers are suf ficiently high to transport shredded fibrous insulation debris to the suction strainers.
RESPONSE
The information requested above has been provided in the i
report " Evaluation of Drywell Insulation Debris Ef fects on ECCS Pump Performance" provided under separate cover.
Section 6.2.2.2.2 has been revised to reflect this report.
i M P84 93 05 08-as
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l Evaluation of Drywell Insulation Debris Ef fects on ECCS Pump Performance t I
Hope Creek Generating Station Public Service Electric & Gas Company
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Bechtel Power" Corporation San Francisco, CA May 1984 Revision 1 ,
Oste eN w nem oa9 - _ __- _ _ _
1 INTRODUCTION 1.1 PURPOSE There is a concern that the insulation debris created by a high energy line break in the primary containment will collect on the Emergency Core Cooling System (ECCS) suction strainers and impair the pump performance. These debris considerations are part of Unronolved Safety Issue A-43, Containment Emergency Sump Performance.
The Hope Creek Generating Station (HCGS) was ovaluated to dator-mine the maximum quantity of insulation debris that might be generated by a LOCA. The evaluation includes transport of this debris to the ECCS suction strainers and the effect on ECCS pump operation. Only the low pressure coolant injection (LPCI)
! and core spray (CS) pumps are evaluated because only large pipe breaks can generate significant quantities of insulation debris.
Neither the high pressure coolant injection nor the reactor core isolation cooling pumps are able to operate after a large break LOCA.
The only insulation debris in the drywell that might enter the vont pipes, and eventually the suppression pool, are small pieces of shredded fiberglas because of the small openings in the jet deflectors. Whole and torn blankets are assumed to be retained by the drywell internal obstructions and the vent jet deflectors.
1.2
SUMMARY
OF RESULTS It has been determined that the postulated fibrous insulation debris generated by a high energy line break will not jeopardize ECCS pump operation at HCGS. This is based on ovaluation of the transport of the worst caso shredded debris generation in the drywell. The volume of shredded debris and its transport to the ECCS strainers has been conservatively evaluated. The l head loss due to the accumulation of the debris concurrent with the conservative NPSH available conditions does not cause the NPSH available to drop below that required by the ECCS pumps.
The NPSH available to the ECCS pumps is at least 10 f t greator than the required 7.5 ft for the LPCI and 3.5 ft for the core spray pumps. The results of the analysis are shown on Table 3-1.
2 BASES 2.1 HCGS is a GWR with a Mark I containment design.
2.2 The ECCS pumps take suction from the suppression pool water inventory following a LOCA. The suppression pool is located in the torus surrounding the base of the drywell. The fluid f rom a LOCA is released into the drywell and the steam-water-gas mixture is conducted to the supprossion chamber by 1
(su enn trew uy
the eight vent pipes and the vent /downcomer header system in the torus. When the liquid level on the floor of the drywell reaches the elevation of the vent pipe openings this liquid also flows to the suppression pool through the vent system.
The ECCS suction strainers are located above the bottom of the torus in eight different sections. Figures 2-1 through 2-5 show the general arrangement.
2.3 The only thermal insulation used in the drywell is
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Owens-Corning "NUKON", stainless steel jacketed fiberglas.
The fiberglas is totally enclosed in woven fiberglas covers with glass fiber stitching. The blankets are held on with velcro fasteners and protected with 22 gage stainless steel jackets with seismically qualified mechanical latches.
3 METHOD OF ANALYSIS 3.1 IDENTIFICATION OF PIPE RUPTURE LOCATIONS (PRL) 3.1.1 The PRLs for, analysis are based on the locations identified in the pipe break analysis discussed in FSAR Section 3.6 (Reference 444).
3.1.2 The. break locations analysed were chosen in the large diameter lines with greatest potential for generating significant quantities of insulation debris. The PRLs analysed are in the following lines:
- 1. Reactor Recirculation Pump Suction
- 2. Main' Steam Line Feedwater'Line 3.
- 4. RHR Supply Line 3.1. 31 The pipe break is postulated to be circumferential.
' The broken pipe;is not assumed to shadow the jet cone although pipe movement is restricted by whip restraints. This will r' >
result in a conservative estimaticn of the affected insulation.
Slot breaks were not evaluated as the resulting jet would have a smaller zone ofLinfluence and would therefore generate less
' insulation debris, y 3.1.4 Extensive use of pipe' whip restraints and separation of lines eliminates. pipe whip and pipe impact as significant mechanisms for insulation debris generation. The debris generated .by these mechanisms t would be in the form of- the ,
-manufactured whole blankets. This form of debris is retained l
' :in'the drywell and does not affect'the ECCS suction strainers. l l
- 3. 2 . DEBRIS ~ GENERATION .
3.2.1 The pipe break is postulated to produce a-jet from
.each endlof the break. Each jet cone is assumed to expand.at
'2 QseacMaypem[]o9. - - - -- .-. -- -
an angle of 45 degrees f rom the pipe centerline as recommended l in NUREG 0897 (Reference 4.1). The direction of the cone is !
along the original pipe centerline because the pipe movement is limited by the whip restraint.
3.2.2 Each jet cone is separated into two regions. Region I is the portion from the break to the plane where the jet thrust divided by the jet area is 20 psig. All insulation in Region I is assumed to be shredded into small fragments by the jet impingement forces. Region II is the portion of the jet cone
~~
extending from Region I to the plane where the jet thrust divided by the jet area is equal to 0.5 psig. The insulation in the region is assumed to be dislodged in the as-fabricated form.
3.2.3 The pipe whip restraints and large structural steel members inside the jet cone are assumed to cause " shadowing" of the jet (i.e., insulation in the " shadow" of the member is not assumed to be shre'dded). This is consistent with the criteria for modeling jet impingement forces in FSAR Section 3.6 (Refer-ence 4.4) and SRP 3.6.2 (Reference 4.9).
3.2.4 The stainless steel jacketing on the insulation is assumed to provide no protection of the insulation blankets.
This is a conservative assumption because it is expected that the steel jacketing will provide some protection against shred-ding of the insulation, especially where the jet pressure is between 20 and 60 psig.
3.2.5 A geometric analysis was performed to determine the volume of insulation that would be affected by the selected break locations. The volume of insulation exposed to jet impingement in Region I of the jet cone was quantified. The insulation dislodged in Region II of the jet cone was not quantified because the physical barriors in the drywell dis-cussed in FSAR Section 6.2.2 (Reference 4.4) will prevant the insulation blankets from entering the suppression pool.
3.2.6 The break location generating the largest volume of shredded fibrous insulation is the Main Steam Line. The results of the analysis are provided below:
Main Steam Line Break (Line.D) Insulation M. S. Line D 26"9 27.0 ft3 M. S. Line C 26"9 6. 7 5 f t3 LPCI Line 12"# 12.0 ft3 Recirc Pump Discharge 22"9 4.6 ft3 50.75 f t3 ;
)
'3 io9 QscAoNNsicm - .
The break location is shown in figures 3-1, -2 and -3. The same geometry applies to Main Steam Line A. Main steam lines B and C are less limiting.
3.3 INSULATION DEBRIS TRANSPORT 3.3.1 Short Term Transport 3.3.1.1 The short term is the period during initial blowdown from the postulated pipe break. The blowdown lasts for about
~~
l'.5 minutes for the main steam line break cases. The insulation is transported by the jet force from the break and flow of the spilled fluid to the drywell floor. As indicated earlier, pipe whip and impact are not considered in the analysis.
3.3.1.2 It is conservatively assumed that all of the insulation in Region I of the jet cone is shredded and transported to the drywell floor by the jet forces. In reality, a portion of the shredded insulation would be distributed and retained on the '
structural steel and on the grating and components in the drywell. Only a portion would reach the floor and be available for transport to the torus. The shredded insulation is assumed to be uniformly mixed in the turbulent liquid collecting on the containment floor. When sufficient liquid has collected on the floor to reach the level of the vent pipes, it overflows and is carried to the suppression pool by the vent header. It is conservatively assumed that all of the shredded insulation in the drywell floor pool, except that in the sumps and inside the cylindrical vessel pedestal, is transported to the suppression pool with the overflowing liquid. The area under the vessel pedestal has only one opening at the elevation of the drywell flooding so this volume of water will become stagnant when the equalibrium flooding level in the drywell is reached. The sumps in the drywall are belcw the floor so that after they are filled and the drywell flooding level is above the top of the sump they become stagnant pools, 3.3.2 Long Term Transport 3.2,2.1 The long term is the pcrind starting with the end of the initial blowdown from the postulated pipe break. The transport of insulation is caused by the operation of the ECCS pumps. It is assumed that all of_the LPCI and core spray pumps are operating at their maximum flow rates. This results in conservatively high flow velocities for the transport analysis.
3.3.2.2 It is further assumed that the shredded insulation is uniformly mixed with the suppression pool water at the end of the short term /beginning of the long term. This assumption is based on the even distribution of the blowdown from the down-comers and the turbulent mixing within the suppression pool and 4
b s cA o/c8 trcm 12 9
is consistent with the uniform distribution of insulation in the drywell during the blowdown.
3.3.2.3 At the beginning of the long term the initial blowdown has ended. The discharge from the downcomers during the long term is due to the overflow of the ECC systems from the drywell.
This flow is evenly distributed by the vent header system and results in local turbulence near each downcomer. The contain-ment spray and torus spray modes for the LPCI are used for long term containment cooling following a large break LOCA. The sprays provide an even distribution of fluid returning to the The suppression pool causing only shallow surface turbulence.
bulk of the water in the suppression pool is subject to bulk flow velocities due to the removal of water by the ECCS pumps.
3.3.2.4 The suppression pool has ring girders approximately 2
- feet deep at each mitered joint and at the mid cylinder of each section. The transport test data in NUREG/ CR-2791, (Reference 4.3) indicates that a flow velocity exceeding 0.3 ft/sec is required to entrain fibrous insulation shreds lying on the bottom of the suppression pool. The maximum flow velocity at the bottom of the suppression pool due to the bulk flow near each strainer is less than 0.3 ft/sec. Any insulation debris that sinks to the bottom outside the sections containing the strainers will not be transported to the strainers. The insulation that settles in the section between ring girders containing a strainer ic conservatively assumed to collect on the strainer.
3.3.2.5 The maximum ECCS flow rate is used to determine the flow velocities. This minimizes the time available for settling of the debris and maximizes the flow velocities The evaluation is based on the simultaneous operation of LPCI and core spray
- pumps at their runout flow rates.
! 4 LPCI pumps at 11,000 gpm/ pump 4 core spray pumps at 4,015 gpm/ pump The resulting bulk flow velocity in the region near each strainer '
is 0.037 f t/sec toward each LPCI strainer and 0.014 f t/sec toward each core spray strainer.
i 3.3.2.6 Based on data in NUREG/CR-2791 (Reference 4.3), there l are three types of shredded fibrous debris:
- 1. Debris that immediately sinks i 2. Debris th'at slowly sinks
- 3. Debris.that floats The discussion in NUREG/CR-2982 (Reference 4.2) indicates that the debris absorbs water more_readily and will sink faster in hot (120*F) water than water at ambient temperature.
5 i
'b(bsc A SAOV inre /J F
\
3.3.2.7 Tests perf ormed by Owens-Corning on fibrous "NUKON" f ragments discussed in Topical Report OCF-1, Reference 4.5, indicate that the fragments will readily sink after absorbing water and becoming saturated. The shredded debris that enters the suppression pool will be in contact with hot water. This debris is also mixed with the blowdown water and enters the suppression pool below the water surface. The rate of water absorption is also more rapid when the insulation is hot, which Therefore, is the case for insulation from high temperature lines.
it is expected that most fragments will rapidly become saturated
~
'and settle. Also, because the fragments are thoroughly wetted in transport to the suppression pool, both the floating and slow sinking debris are considered to be slow sinking. Thus, for conservatism no credit is taken for floating debris prevent-ing transport to the strainers.
3.3.2.8 Although there is considerable test data supporting the conclusion made in Section 3.3.2.7, that most "NUKON" fragments will rapidly settle to the bottom of the torus, it is acknowledged there is no specifi'c LOCA test data available describing the post-LOCA buoyancy characteristics of "NUKON".
In the absence of specific LOCA test data for "NUKON", the conservative approach is to assume less than all the "NUKON" fragments rapidly settle. In order to arrive at a credible and conservative factor for the lesser amount of debris that rapidly settles, comparable test data contained in Section 4.7.2 of NUREG/
CR-2791 (Reference 4.3) describing the post-LOCA buoyancy characteristics of mineral wool insulation was used in the analysis. This data is considered conservative because fragments of as-manufactured fiberglas insulation, and "NUKON" in particular, will become wet and sink faster than mineral wool (i.e., fiberglas has a greater tendency to sink compared to mineral wool). This conclusion is based on data contained in NUREG/CR-2982 and Topical Report OCF-1, (References 4.2 and 4.5). There is no reason to believe LOCA effects would change the buoyancy characteristics of "NUKON" debris so that it would be more buoyant than mineral wool. NUREG/CR-2791 (Reference '
4.3) states that 40% to 50% of the fibrous insulation (mineral wool) dislodged by a LOCA can be expected to immediately sink.
The calculation, therefore, used the conservative factor of 40% ,
to determine the amount of "NUKON" fragments that rapidly settle. The remaining 60% of the debris in the suppression pool was assumed to be slow settling.
3.3.2.9 The rapid settling debris is assumed to be dispersed uniformly in the suppression pool. The settling rate for this i
debris is .2 ft/sec based the average sink' rate for saturated I insulation in Topical Report OCF-1 (Reference 4.5). This set-tling rate and the bulk velocity of the suppression pool toward l each strainer were used to determined the volume of rapid settling debris collecting on each strainer.
6 y s u o n w ara m n -
3.3.2.10 The slow-settling debris is assumed to be dispersed
, uniformly in the suppression pool. The settling rate for this debris is .017 ft/sec based on the average sink rate for very small clw ps of insulation in Topical Report OCF-1 (Reference l 4.5). This settling rate and the bulk velocity of the suppres- l sion pool toward each strainer were used to determine the volume of slow settling debris collecting on each strainer.
3.3.2.11 Based on the transport analysis, the maximum expected shredded debris to collect on the RHR strainers is 28.8% of the amount reaching the suppression pool, and that collecting on the core spray strainers is 14.6% of the amount reaching the suppression pool. The total collecting on the strainers is 43.4%
of that in the suppression pool or 29% of the total shredded insulation generated.
3.4- ECCS SUCTION STRAINER HEAD LOSSES DUE TO INSULATION DEBRIS ACCUMULATION ON THE STRAINERS 3.4.1 The volume of insulation debris that will accumulate on each strainer is calculated. Each pump has a separate suction line with a strainer located in the torus as shown on figure 3-1. The design and dimensions of the strainers is shown in figure 3-2. The effective surface area for each LPCI strainer is 15.2 ft2 The2 ef fective surf ace area for each core spray sgrainer,is 5.64 f t The total ECCS strainer area is 83.4 ft .
, 3.4.2 The head loss calculated for accumulation of fibrour insulation debris is based on thickness of a uniform accumulation on the effective surface Reference 4.8 provides a head loss formula developed for a bed of shredded "NUKON" insulation. It is noted that the maximum approach velocity tested in the reference is 0.5 f t/sec, while the strainer approach velocity for HCGS is near 1.5 f t/sec. Review of the data in the reference indicates thatHit is reasonable to assume that the straight line logrithmic relationship ,between head loss and approach velocity can be extended to approach velocities near 1.5 ft/sec.- This was confirmud in discussion with the insula-tion manufacturer. Therefore, the "NUKON" specific head less formula is used. This formula is provided below:
H = 63.8 (tt)I*07 (V)l*73 where H = head loss, f t of water ti = equivalent accumulation thickness, Et v = screen approach velocity, ft/sec 3.4.3- The head loss due to the accumulation of insulation on the. strainers is provided in Table 3-1.
-7; bscR ot*CD srem ja?
Y
3.5 EFFECT OF ACCUMULATION OF INSULATION DEBRIS ON ECCS PUMP NPSH.
3.5.1 The minimum NPSH is available when the suppression pool temperature is 212*F at 24,000 seconds. It is conserva-tively assumed that no noncondensables are added to the torus because of blowdown from the drywell and the noncondensables in the torus remain at the same temperature they were at prior to the blowdown. Therefore, the partial pressure exerted by the noncondensables is the same as existed prior to the LOCA.
3.5.2 Table 3-1 shows that the worst case insulation genera-tion and the resulting accumulation on the strainers results in adequate NPSH for the ECCS pumps. The NPSH available for the LPCI pumps is 19.7 f t and the required NPSH is 7.5 f t. The NPSH available for the core spray pumps is 14.12 ft and the required NPSH is 3.5 ft. The NPSH required is taken from manufacturers certified performance data for the pumps. The required NPSH values in the FSAR and the GE Process Flow Diagrams contain a large safety margin above the requirement given by the pump manufacturers.
3.5.3 Non uniform distribution of the insulation in the torus was examined. The ECCS pumps have adequate NPSH available when 70% of the insulation debris reaching the suppression pool is distributed in one half the pool volume. The NPSH available was determined using the same assumptions as previously.
m K45/14 Osca os u oren' 'a 9
4 REFERENCES 4.1 Serkiz, A.W., " Containment Emergency Sump Performance,"
NUREG-0897 (for comment), NRC, April, 1983.
4.2 Brocard, D.N., " Buoyancy, Transport, and Head Loss of Fibrous Reactor Insulation," NUREG/CR-2982, SAND 82-7205, Alden Research Laboratory, November, 1982.
~~
4.3 Wysocki, J.; Kolbe, R., " Methodology of Evaluation of Insulation Debris Effects," NUREG/CR-2791, SAND 82-7067, Burns & Roe, Inc., September, 1982.
4.4 Final Safety Analysis Report (FSAR), Hope Creek Generating Station, Public Service Electric and Gas Company.
4.5 Owens-Corning Fiberglas Corporation, " Topical Report OCF-1, Nuclear Containment Insulation System, NU'K'ON,"
January, 1979.
- 4. 6 ANSI /ANS-58.2-1980, " Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture".
4.7 ANSI /ANS-58.3-1977 (N182), " Physical Protection for Systems and Components Important to Safety".
4.8 Brocard, D.N., " Transport and Head Loss Tests of Owens-Corning NUKON Fiberglas Insulation, "Alden Research Laboratory, September, 1983.
4.9 NUREG-0800, " Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants", US NRC, July, 1981.
o K45/14 Dss ^ o/cN trcm /2 9
e '
1 F) -
i TABLE 3-1 b Suwary of Calculated ECCS Strainer Head loss
- Due to Accunulation of Fibrous Debris and the Effect on ECCS Punping NPSHA DEBRIS \
VOLLME , MINIMUM NPS% MINIMlM NPS%
ECCS PLMP DEBRIS REACHING DEBRIS DEBRIS WITH CIEAN WI'IH COVERED NPSHR STRAINER FLIM GENERATED STRAINER 'IMICKNESS HEAD ILSS STRAINERS STRAINERS FOR PtMP com ft3 ft3 (1) in. ft ft ft ft MAIN STEAM IKI 11,000 2.47 1.92 22.9 42.6 19.7 7. 5 50.75 4
LINE BREAK CS 4,015 1.21 2.52 30.13 44.25 14.12 3. 5 (1) Value for a single strainer.
(2) Includes piping loss to the punp.
K45/14
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HCGS FSAR 6.1.2.2.1 Effects of Insulation on System Performance Fiberglass blanket sections covered with 22-gauge, 304 stainless steel jacketing insulate structures, equipment, and piping within the primary containment. This form of insulation is not expected to create a debris-clogging problem for containment cooling operation after a LOCA. The vendor has studied the performance
-- - of the materials during a simulated design basis accident (DBA) .
The results of the study have been accepted by the NRC as Topical Report OCF-1, Nuclear Containment Insulation System. As indicated in the study:
- a. The fiberglass insulation will not deteriorate or lose its mechanical integrity during a LOCA. .
- b. The operation of the emergency spray systems will not ,
wash off or dislodge the- insulation from piping, i.
equipment, er structures.
~
- c. Only those segments of insulation that are subjected to the violent forces of a component rupture, jet impingement, or pipe whip could be expected to become 4
potential clogging debris. .Y0 ::: th:n 50 blanket --
- tiene e' the centain.T. cat insulatien ins;ntery :: ^-
- ted t: i: :: :ffected in i pertul:ted LOC.'.. ~_
- d. " :n i,f :::: ef the hier'et :::ti::: ::: t :::; rted t-the pression chamber ring header and from- ther the sup sion pool, the fiberglass will not g the suction str es. If whole blanket sect should i lodge on a stra , the material is ous and will not impede the suct f i. ' . 51 .he insulation will
- NM--* sink,. and since the stri er zles are offset up from
, the suppression chamber . the insulation should not come to rest on straine let area. If some of the blanket ons should be s dded during the :
LOCA, the i idual fibers will not_bi together or to stra surfaces due to the inert natur the-ma al. Therefore, fiberglass insulation doe t i
..;;titut; a petential etcainec clegging v6vbi--. j lt. If some fibers do pass through the- 0.125-inch strainer mesh, the study has shown that. pump function and spray nozzle performance are not affected. Particles of this I size or smaller will not impair the safe ~ function of 6.1-30 .
O
INSERT d The.only path for insulation to enter the suppression pool is through the vent pipes and the downcomer ring header.
The- jet deflectors prevent debris from entering the vent pipes directly. Floor grating, structural steel, and components in the drywell will retain insulation debris and prevent it from reaching the floor or the vent pipes. Only a small portion of the insulaton debris generated will
__ actually be available for transport to the suppression pool.- The openings at the jet deflectors will prevent all but small fragments from entering the vent pipes.
Much of the insulation debris that is transported to the suppression pool is fast settling, the remainder settles more slowly. After the initial blowdown the insulation debris that reaches the suppression pool begins to settle to the bottom. The flow velocity created by the ECCS pump
. operation for the bulk of the suppression pool is very low, therefore, only a portion of the insulation debris in the suppression pool will collect on the ECCS strainers. The strainers are located above the bottom of the suppression pool and the velocities generated at the bottom of the suppression pool are not sufficient to reentrain insulation that has settled except very near the strainers.
An evaluation of the transport and accumulation of postulated insulation debris was performed in accordance with the guidance of NUREG 0897 (issued for comment). This evaluation was transmitted under separate cover (R. L.
Mittl, PSE&G, to A. Schwencer, NRC, dated May 15, 1984).
.The flow restriction caused by the insulation accumulation on the strainers in this analysis does not adversely effect operation of the ECCS pumps.
t l
1 l
14 P84 93 05-09-az bset ored Jrtm M 1
. HCGS l
DSER Open Item No. 152 ( DSER Section 9.4.4 )
RADIOACTIVITY MONITORING ELEMENTS Turbine enclosure ventilation system RSS PONSE
__ This item is not an open item per telephone conversation between J. M. Ashley ( PSE&G) and John Ridgely (NRC-ASB) on March - 22, 1984.
K3/9
l TELEPHONE NOTES PSE&G Hope Creek Licensing (Bethesda)
Date: March 22, 1984 From: J.M. Ashley To: D. Wagner, J. Ridgely (ASB)
Subject:
HCGS DSER Open Items Discussion Ashley called to find out what NRC concerns existed with respect to FSAR Sections 3.5.1.2 (Item 30), 9.2.2 (Item 145) and 9.4.4 (Item 152).
Ridgely explained that these items were inadvertently listed l as open items in the listing of open items at the front of the l
DSER. The NRC has no outstanding concerns with the sections.
I i
- n-bsts onw irem s .:~a .
, . - ~ - . ._ _ _ _
- HCGS l DSER Open Item No.154 ( DSER Section 9. 5.1. 4.a )
METAL ROOF DECK CONSTRUCTION CLASSIFICATION Metal roof deck construction is noncombustible , but is not listed as " acceptable for fire" in the UL Building Materials Directory and, therefore, is not consistent with Section C.S.a(10 )
The staf f will require the applicant to of BTP CMEB 9.5-1.
provide metal roof deck construction that is classed " acceptable for fire" in the UL Building Materials Directory or that meets the criteria for Class 1 roof deck systems in the FM system approval guide.
RESPONSE
Metal roof deck construction for HCGS meets the criteria for Class 1 roof deck systems outlined .by Factory Mutual's systems approval quide. Therefore, HCGS complies with Section C.S.a(10) of Branch Technical Position CMEB 9.5-1.
FSAR Section 9.5.1.1.7 has been revised to address compliance with the above requirement.
9 e
l 1
K3/9 )
b HCGS FSAR 1/84 9.5.1.1.6 Cable Spreading Room .
The cable spreading room (CSR) is separated from other areas of the plant by 3-hour-rated fire barriers. Fire detection and an automatic preaction sprinkler system are provided. Cabling to
__the remote shutdown panel room is independent of the CSR and provides the necessary means to attain a safe shutdown if the CSR is lost.
9.5.1.1.7 Building Materials Selection Interior walls, partitions, structural components, thermal insulation materials, and radiation shielding materials are noncombustible. Areas containing systems or equipment required for safe shutdown are either unfinished or finished with noncombustiale materials.
4 Suspended acoustical ceiling panels are Underwriters Laboratories, Inc (UL)-listed and have a flame spread, fuel l contribution and smoke development rating of 25 or less. *-~ ~ )
j Suspended ceiling supports are noncombustible.
k ltated a.s Cla s s I.
-c--
I Metal deck roof construction ^="k is noncombustible and h:: ::ir. forced l :::.;::t; si:b :... :.: :: for= "cch. -~~,
i by Fae.to3 1%ho.1 CFm) 5 sices 3 n.er rava.1 S u.' d e..
9.5.1.1.8 Protection from Transformer Fires i
Medium and low voltage-amperage transformers located indoors are
- dry and air-cooled. Oil-filled medium voltage-amperage transformers (main and station service transformers) are located i
outdoors near the turbine building and circulating water pump structure. The turbine building and circulating water pump structure contain nonsafety-related systems.
l All main and station service transformers are provided with individual water spray systems and are separated from each other
- by a 1-hour fire barrier. Each transformer has a collection dike
( and drainage outlet for collecting transformer oil spills and l
fire suppression system water and draining it to the oily waste j drainage system.
i l
l' f .
9.5-4 Amendment 4 s aw ma am at L
4
, HCGS DSER Open Item No. 159 (DSER Section 9.5.1.5.a)
PRIMARY AND SECONDARY POWER SUPPLIES FOR FIRE DETECTION SYSTEM Primary and secondary power supplies for the detection system in accordance with NFPA 72D, which the staf f references in Section C.6.a (6) of its guidelines, have not been provided.
The staff will require that primary and secondary power
'7 hupplies for the fire detection system satisfy the provisions of Section 2220 of NFPA 72D.
RES PONSE As indicated in Section 9.5.1.2.15, the fire detection system is supplied from an inverter system which has batteries and SDG-backed MCCs as power supplies. Section 9.5.1.6.16 has been revised to provide discussion on compliance with NFPA 72D requirements.
s I
K3/9
HCGS FSAR 1/84 removal of a detection device from a detector circuits, and power failure. If any of the above problems occur, a fire detection system trouble is annunciated locally and in the main control room. Plant operation will periodically test the system for proper functioning, similar to inservice testing of other plant systems. j l
At HCGS, the fire and smoke detection system is in compliance with NFPA 72D except that the operation and supervision of the system is not the sole function of the plant operator. The plant operator's duties cover operation of the generating station and monitoring and supervising the fire protection systems.
9.5.1.6.15 Paragraph C.6.a.(3) l J
Paragraph C.6.a.(3) requires that the fire detectors be installed in accordance with NFPA 72E.
I At HCGS, the location of early warning fire and smoke detectors
- was determined and performed under the direction of a registered fire protection engineer. The location of the fire and smoke i detectors complies with the guidelines of NFPA 72E except for the i location of ionization and photoelectric detectors in high-bay areas. The detectors are not located in each bay formed by deep i beams. NFPA 72E allows detector locations to be determined based
, on engineering judgement considering ceiling shape, ceiling surfaces, ceiling height, configuration of contents, combustible characteristic and ventilation.
I' At locations in areas where composite construction is used, the diffusion of combustion particulates throughout the compartment volume produced during the incipient and smoldering stages of the 1 fire will negate the effect of beam depth and result in j acceptable levels of detection coverage.
~
l 9.5.1.6.16 Paragraph C.6.a.(6) l posse ParagraphC.6.a.(6)requiresprimaryandsecondaryksuppliesbe providedfor{electricallyoperatedcontrolvalvesconformingto NFPA 72D. the fire ddecis'en s y s t.em o.n d i
[ .3:n s EAT- A ---*
d 9.5-52 Amendment 4 j i
hscR othrN stem tf?
, - - .--- .- , _ __ ,. , , . - - - . - , - - . _ _ , , - - . _ , - _ . , _ _ - .,.n,,n_n, - , -.c
l HCGS INSERT A At HCGS, the fire detection system is supplied with uninterruptible 120 volt ac power fed from an inverter type system which has three power supplies. The normal or primary power supply is from an offsite source and the alternate or secondary power supply is from a 4-hour station battery supply. A third power supply serves as backup to the primary and secondary power supplies. In addition, both the primary and backup power supplies are connected to buses which are backed by standby diesel generators (SDGM). The buses arc disconnected from the SDGs during a LOCA event; however, the buses can be reconnected to the SDGs under administrative control. Figure 8.3-11, Sheet 3, is the single line diagram of the power supplies to the fire detection system equipment. Therefore, the fire detection system is furnished with power supplies which meet the NFPA 72D requirements.
I MP84 95 01 4-bp Dsed OH4 agin /19 L_ J
HCGS 1
DSER Open Item No. 161 (DSER Section 9.5.1.5.b)
FIRE WATER VALVE SUPERVISION Supervision has not been provided for all valves in the fire protection water supply system in accordance with NFPA 26.
To meet staff guidelines in the Section C.6.c of BTP CMED
. 9.5-1, the type of valve supervised and the frequency at which its position is verified should be as listed.
RESPCNSE All valves in the fire protection water supply system are supervised in accordance with NFPA 26 except locked valves which are inspected to verify valve condition. FSAR Section 9.5.1.2.3.4 has been revised to indicate this.
2 9
MP84 95 01 5-bp E ._ .. ._. .
i 2
- - 1
- The outdoor, underground yard loop was designed in accordance i with NFPA 24. The yard loop consists of 12-inch diameter cement mortar-lined ducti'e iron pipe that extends around the power block. Post-indi or valves are provided for sectional control.
Two-way hydc7nts, cu . rolled by individual curb box valves, are
'j installed on the yard loop at maximum intervals of 250 feet. A hose house is provided for each hydrant and equipped with j 200 feet of hose, fittings, and accessories in accordance with j NFPA 24.
9.5.1.2.3.4 Water Supply for Automatic and Manual l Sprinkler / Spray Systems
!. Automatic and manual sprinkler / spray systems headers are i connected to the in-plant loops that are fed from the main l underground fire protection water piping or yard loop by two i separate supplies. The in-plant loops are 8-inch and 10-inch
! lines. Since the in-plant loops are fed by two separate j' supplies, they are considered an extension of the main underground yard loop. Automatic and manual sprinkler / spray
! systems and hose standpipe systems serving a single safety-related area have takeoffs from an in-plant loop, separated by f sectional control valves. The header arrangement is such that, by manual positioning of the sectional valves, no single piping ;
failure can impair both the primary and backup fire protection L provided for a single area. ,
l AC power supply for sprinkler / spray system contr'ol panels
!' including the fire status panel in the control room is provided :
! from a non-Class IE inverter. The inverter is fed by non-l' Class IE batteries and non-Class IE motor control centers (MCCs) r backed by standby diesel generators (SDGs). Loss of normal ac j power will not prevent the panels or systers from operating.
1 An outside screw and yoke (OS&Y) gate valve for each sprinkler ,
j and deluge system is located adjacent to the system control or l [
alarm valve. The branch connection into the building is provided f with a' post-indicator valve at the connection to the yard loop.
l Each sprinkler and deluge system is provided with local water f flow alarms and remote annunciation in the main control room.
ZAMLA7~--*
9.5.1.2.4 Wet Pipe Sprinkler Systems Wet. pipe sprinkler systems are provided for the plant areas i listed in Table 9.5-2. The density coverage and installation for l l
9.5-18 Amendment 4 as@ ' OArd GCm H* I
_= = _ _ . . _ . . . _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _
t HCGS l
SER Open Item No. 161 INSERT Control and sectionalizing valves in the fire protection
- water system are either electrically supervised or administrative 1y controlled in accordance with NFPA 26. The valves that are electrically supervised are those valves that control the water suppression system and the valves in the. fire pump suction and discharge lines located in the fire pamp house. These valves are shown on Figures 9.5-13 through 9.5-16 and 9.5-18. The electrically supervised valves are provided with normally open contacts that close in the event of valve movement. The electrical supervision
- signal is indicated on the local control panels and i i
registers as a system trouble on the fire protection status
! panel in the main control room.
The valves that will be administrative 1y controlled are the post indicator valves in the yard area that provide sectional control of the fire main loop and fire water
- supply lines branching into various buildings, the sectional i valves in the in-plant loop and supply piping, and the j valves that control the water supply to standpipe and hose
, systems. These valves are padlocked in the open or closed position so that-they cannot be inadvertently operated. The control valves for the standpipe and hose system in the reactor building and intake structure are normally closed to maintain these systems in a dry condition.
Valves are either electrically supervised, or locked and inspected monthly. documentation recording this inspection will'be made.
I ti MP84 95 01 6-bp
HCGS DSER Open Item No. 162 (DSER Section 9.5.1.5.c)
DELUGE VALVES The applicant is not providing approved deluge valves for the deluge systems. This is not in accordance with NFPA 13 which the staff references in its guidelines. The staff will require the applicant to provide deluge valves approved by a nationally recognized testing laboratory as components for fire protection systems, as specified by NFPA 13, which the staff references in Section C.6.c of BTP CMEB 9.5-1.
RESPONSE
The deluge valves for the HCGS water spray, protection and deluge system are Viking Corporation Model D-5 water control valves which are UL 11std per the 1983 UL Fire Protection Equipment Directory. Sections 9.5.1.2.5, 9.5.1.2.6, 9.5.1.2.7 and 9.5.1.2.8 have been revised to reflect this.
Section 9.5.1.2.7 has been revised to clarify that the electric-motor-operated valves in the preaction water spray systems are not deluge valves.
MP84 95 01 7-bp
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HCGS FSAR 1/84 the sprinkler systems are in accordance with NFPA 13. Each l sprinkler system is provided with an alarm check valve or flow switch that annunciates in the main control room. OSEY gate valves serving as shutoff valves to automatic sprinkler systems are supervised with any problems annunciated in the main control room.
' Wet pipe sprinkler system operation is initiated upon a rise in ambient temperature to the melting point of fusible links on sealed sprinkler heads, thus causing the spray heads to open.
l The flow of water through an alarm check valve or flow switch energizes a local alarm and registers an alarm condition on the
- fire monitor panel in the main control room. Once initiated, the l wet sprinkler system operation is terminated manually by shutting l
either a gate valve external to the hazard or a post-indicator valve outdoors.
! 9.5.1.2.5 Water Spray Systems I
l Water spray systems are provided for the plant areas or equipment
- listed in Table 9.5-2.
i i
UL-/a O
! The water spray systems have directional solid cone spray nozzles
! or perforated pipe. The water flow is controlled by deluge i valves. A system alarm and a valve position alarm for supervised j OS&Y gate valves for each spray system are provided in the main i control room. Spray densities and installation complies with
! NFPA 13 and 15.
l:
! Operation of the automatic spray systems is initiated by a i temperature sensor. This sensor detects a rapid rise in ambient
! temperature and/or attainment of a fixed high temperature and f releases a tripping device to open the deluge valve, thus l- supplying water under pressure to the open spray nozzles.
i Actuation of a sensor also initiates a local alarm, and registers the alarm condition on the fire protection status panel in the main control room, independently of water flow in toe system.
Water flow in the system initiates a local alarm and registers
.the system-actuated condition on the fire protection status panel in the main control room independent of the detection alarm.
l i Manual release of the deluge valve tripping device also initiates local and remote water flow alarms. System operation is
- terminated by manually closing a gate valve external to the hazard area.
i
, 9.5-19 Amendment 4
. Dse:ROMN'*'?*_ __,- _ _,-_--_____ __ _ _______-
l I
i HCGS FSAR 1/84 Operation of the manual water spray systems is initiated by a pushbutton on the local panel and opening a normally closed OS&Y gate valve. A temperature sensor detects a fixed high temperature and registers the alarm condition on the local ,
control panel and in the main control room. The system is octivated by operating a pushbutton on the local control panel and opening the OS&Y gate valve. Water flow in the system initiates a local alarm and registers the system-actuated condition in the main control room independent of the detection alarm. System operation is terminated by manually closing the OS&Y gate valve external to the hazard area.
l 9.5.1.2.6 Deluge Systems l Deluge systems are provided for the diesel fuel tank rooms as listed in Table 9.5-2. l yL-lasted The deluge syst ms have open sprinkler heads. Water flow is controlled by a deluge valve actuated by a local manual switch, i and a normally closed OS&Y gate valve. A system alarm., and a valve position alarm on supervised OS&Y gate valves for each l deluge system is provided in the main control room. The density coverage and installation for the systems are in accordance with NFPA 13.
Water flow in the system initiates a local alarm and registers the system-actuated condition on the fire protection status panel in the main control room. ;
1 System operation is terminated by manually closing a gate valve external to the hazard area.
9.5.1.2.7 Manual Preaction Water Spray Systems l Manual preaction water spray systems serve the reactor building. l Specific equipment covered is listed in Table 9.5-2.
Individual hazards are protected by fixed water spray nozzles on dry piping at atmospheric pressure. Each individual system is controlled by an electric-motor-operated valve. rvir; :: : " -
i l delt;r velve.*-Eac dels;;'falve is connected to a common header
! Eystem within the reactor building. The header is pressurized to
- l. 20 psig with air. Water supply to the header is controlled by a Y eleefric inobor-operated 9.5-20 Amendment 4 l 'DscR ortN !!Em__ _ ' _ ' " _ _ _ _ _ . . _ _
1 l
HCGS FSAR 1/84 UL-/ Nied i olelug e )
A preaction valve assembly with a building penetration having one normally open isolation valve. Fixed temperature line-type heat detectors are used and any fire condition is annunciated in the main control room. Alarms are provided for low air pressure and for the closed cendition of supervised OS&Y gate valves. Water flow is detected by a pressure switch downstream of the motor-operated valve and is annunciated in the main control room.
Design and inatallation comply with NFPA 13 and 15.
The manual preaction water spray system operation is initiated by l either a fixed high temperature thermostat, or a manual switch in the main control room, which actuates the preaction valve and charges the system with water to the inlet of each individual
. hazard valve. No water is discharged through the closed hazard valve at this time. Any fire condition is annunciated in the main control room.
i High temperature due to fire condition at any individual hazard activates the local fixed high-high temperature thermostat and annuniates a fire condition locally and in the main control room.
The hazard valve (motor-operated gate valve) is opened manually by a pushbutton on the local control panel and water is discharged onto the hazard.
1 When the fire is controlled and the environment is cooled to a temperature below the thermostat, the fire alarms at the local i control panel are silenced and the fire indicating lights in the main control room are turned off. The discharge of water may then be stopped by a manual pushbutton on the local control panel
) which closes the hazard valve (motor-operated gate valve). The ball-type drip valve at the low point of the open piping system automatically drains the system downstream of the hazard valve into the radwaste drainage system. The main header remains ,
pressurized. ;
! The system is capable of being reset and returned to normal i status without entering the hazard area as follows:
- a. The water spray portion of the system is reset and
, drained automatically.
l
- b. The fire main gate valve is manually closed and the normally closed drain valve is manually opened, draining the piping downstream of the preaction valve assembly into the radwaste drainage system.
9.5-21 Amendment 4 1
I ( # #
- c. The preaction valve is manually reset, the header is l pressurized with supervisory air, and when the fire main valve is manually opened, the entire system is returned to full-service status.
9.5.1.2.8 Preaction Sprinkler Systems Preaction sprinkler systems are providedAfor plant areas or equipment as listed in Table 9.5-2. \#
I c U L -lis+ed f (gela0 g)
Preaction sprinkler system operation is initiated by heat actuating devices located in the hazard area, which actuates the preactiongvalve and charges the system with water up to the closed, fusible link sprinkler heads. No water is discharged to the hazard area at this time. Any fire condition is annunciated in the main control room. High temperature due to fire condition melts one or more of the fusible link sprinkler heads and water discharges onto the hazard.
When the fire is controlled, water discharge is terminated by manually closing the fire main gate valve, and the normally closed test and drain valves are opened, draining the system.
Used sprinkler heads are replaced, the preaction valve is manually reset, and the header is pressurized with supervisory air.
9.5.1.2.9 Wet Standpipes and Hose Stations Wet standpipes for fire hoses are designed in accordance with NFPA 14. Standpipes are installed adjacent to stairwells, exits, End other points in all normally accessible areas in plant buildings. Four-inch standpipes are provided for three or more hose connections, and 3-inch standpipes are provided for one or two hose connections. The standpipe hose connect ~ ions are
.squipped with 1-1/2-inch hose valves and 75 or 100 feet of 1-1/2-inch woven jacket lined hose with spray nozzles.
Wet standpipes are maintained in a dry condition in the reactor building and the intake structure.
Adjustable spray nozzles with shutoff capabilities, UL-listed for Class C fires, are provided.
L 9.5-22 Amendment 4
?KL*?ETM*" T" - . . . . . - _ - - , .-..- .-- . - - - - -. ---
HCGS DSER Open Item No. 163 (DSER Section 9.5.1.5.c)
MANUAL HOSE STATION PIPE SIZING Manual hose stations are located throughout the plant in accordance with NFPA 14. Three-inch-diameter piping is used to serve up to two hose stations in some areas. This does not meet staf f guidelines. The staff will require the applicant to provide 4-in. diameter piping consistent with the guidelines in Section C.6.c(4) of BTP CMEB 9.5-1.
RESPONSE
At HCGS the pipe size for the wet standby system meets NFPA 14 requirements. Also, as stated in FSAR Section 9.5.1.6.21, all standpipe connections to the in-plant loop are 4-inch diameter and feed multiple hose connections.
Except for one instance, branches off the standpipes that feed one or two hose connections are 3-inch diameter. The as-built plant configuration has one 3-inch standpipe connection, with 3 hose connections to it. These 3 hose stations are not all on the same floor and all 3 hose sta-tions could not be used to fight the same fire. This 3 inch branch has been evaluated and found acceptable to meet NFPA 14 flow and pressure requirements. As stated in FSAR Section 9.5.1.6.19, the fire water supply can provide water at the required flow and pressure to supply any hydrauli-cally designed sprinkler or deluge system and all hoses which can be used to fight the same fire.
FSAR Section 9.5.1.2.9 has been revised to clarify the design of HCGS wet standpipe system.
MP84 56/11 1-db
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- c. The preaction valve is manually reset, the header is pressurized with supervisory air, and when the fire main valve is manually opened, the entire system is returned to full-service status.
4
$ 9.5.1.2.8 Preaction Sprinkler Systems Preaction sprinkler systems are provided for plant areas or equipment a3 listed in Table 9.5-2.
Preaction sprinkler system operation is initiated by heat actuating devices located in the hazard area, which actuates the preaction. valve and charges the system with water up to the closed, fusible link sprinkler heads. No water is discharged to the hazard area at this time. Any fire condition is annunciated in the main control room. High temperature due to fire condition melts one or more of the fusible link sprinkler heads and water discharges onto the hazard.
4 i
! When the fire is controlled, water discharge is terminated by manually closing the fire main gate valve, and the normally closed test and drain valves are opened, draining the system.
Used sprinkler heads are replaced, the preaction valve is manually reset, and the header is pressurized with supervisory air.
9.5.1.2.9 Wet Standpipes and Hose Stations Wet standpipes for fire hoses are designed)I~ ,in accordance tW36W with h NFPA 14. Standpipes are installed adjacent to stairwells, exits, and other points in all normally accessible areas in plant buildings. Four-inch standpipes are provided for three or more hose connections, and 3-inch standpipes are provided for one or two hose connectiong The standpipe hose connect' ions are equipped with 1.-1/2 pinch hose valves and 75 or 100 feet of __
1-1/2-inch woven jac ket lined hose with spray nozzles.
-acnscAT B r/scATC-+
Wet standpipes are maintained in a dry _ condition in the reactor building and the intake structure.
. Adjustable spray nozzles with shutoff capabilities, UL-listed for l Class C fires, are provided.
l I'
t 9.5-22 Amendment 4 b3 c A O PC^! M'" ' O
HCGS FSAR 4/84 At HCGS, each sprinkler and deluge system is provided with an OS&Y gate valve adjacent to the system automatic control or alarm valve. The branch connection into the building is provided with
- a. post indicator valve at the connection to the fire main loop.
Each sprinkler and deluge system is provided with local water flow alarms (using pressure or flow switch) and remote ,
annunciation in the main control room. l 1
t An OS&Y gate valve is provided at each branch (off the in-plant fire main loop) supplying the sprinkler, deluge or standpipe system.
The standpipe systems are not provided with water flow alarms.
Waterflow in the standpipe systems is indicated by pump running annunciation in the main control room without automatic system actuation annunciation.
9.5.1.6.21 Paragraph C 6.c.(4) i Paragraph C.6.c.(4) requires individual standpipes be at least 4 inches in diameter for multiple hose connections and 2 1/2 inches in diameter for single hose connections.
At HCGS, all standpipe connections to the in-plant loop are 4 inch diameter for standpipes feeding multiple hose connections.
See Figures 9.5-13 through 9.5-18. But branches off the standpipes, that feed two or less hose connections, are 3 inches in diamete4 As stated in Section 9.5.1.6.19, the fire water supply can Tprovide water at the required flow and pressure to supply anyisprinkler or deluge system and all the hoses which can be used tof fight the same fire.
N .rM3fM D 9.5.1.6.22 Paragraph C.6.c.(4)
Paragraph C.6.c.(4) requires that provisions be made to supply water at least to standpipes and hose connections for manual firefighting in areas containing equipment required for safe plant shutdown in the event of a safe shutdown earthquake (SSE).
The firewater piping serving such hose stations should be analyzed for SSE loading, and should be provided with supports to
- ensure system pressure integrity.
l 9.5-55 Amendment 5 l
Dse.R ot%/ tre m is: _ - _ _ _ . - - - _ .
l l
HCGS DSER Open Item No. 163 Insert A' to provide 65 psig at the topmost outlet of the hose standpipe system with 100 gpm flowing from the outlet.
Insert B except in one instance where 3 hoses are connected to a 3 inch branch. (These 3 hose stations are not on the same floor and cannot be used to fight the same fire.)
Insert C The fire water supply can provide water at the required flow and pressure to supply any hydraulically designed sprinkler or deluge system and all the hose streams which can be brought to bear on the same fire. See Sections 9.5.1.6.19 and 9.5.1.6.21.
Insert D except in one instance where 3 hoses are connected to a 3 inch branch. These 3 hose stations are not all on the same floor and could not be used to fight the same fire. This 3 inch branch has been evaluated and found acceptable to meet NFPA 14 pressure and flow requirements.
1 6
4 MP84 56/11 2-db.
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r HCGS DSER Open Item 164 ( DSER Section 9.5.1.6.e)
REMOTE SHUTDOWN PANEL VENTILATION The remote shutdown panel room is supplied by an HVAC system that also supplies the control room. A single fire could disable both areas.
To meet our guidelines in Section C.7.f of BTP CMEB 9.5-1, we will require the applicant to provide a ventilation system for the remote shutdown panel that is isolated from the main control room.
RESPON3E The remote shutdown panel (RSP) room and the control room are served by independent HVAC systems as described in FSAR Sections 9.4.3.1.3 and 9.4.3.2.1 respectively .
The RSP room and its associated HVAC unit are separated by a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire-rated barrier. A 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire-rated floor between the main control room HVAC and RSP room HVAC supply units meets the criteria for separation outlined in Branch Technical Position CMEB 9.5.1. A single fire will not affect both the MCR and RSP rooms.
l MP84 56/11 3-db l
HCGS DSER Open Item No. 165 (DSER Section 9.5.1.6.g)
EMERGENCY DIESEL GENERATOR DAY TANK PROTECTION A 550-gal. fuel oil day tank is-provided in each diesel generator room. No enclosure or dike is provided for the day tanks. This is not consistent with staff guidelines.
The staf f will require that the applicant protect the day tanks in accordance with its guidelines in Section C.7.i of BTP CMEB 9.5-1.
RESPONSE
A dike has been provided for each day tank. Section 9.5.1.2.26 has been revised to indicate this.
MP84 56/11 4-db
i HCGS FSAR 1/84 provided in the vicinity of battery rooms. The " low air flow" remote alarm in the control room registers fan failure conditions. Automatic fire detectors alarm locally, and alarm and annunciate in the main control room.
i 9.5.1.2.25 Turbine Lubrication and Control Oil Storage and
- Use Areas Fire Protection i
, The turbine lubrication and control oil storage and use areas are
! remote from areas containing safety-related systems, and are
- separated by 3-hour fire barriers with Class A fire doors.
I The condenser area, secondary condensate pump area, RFPT lube oil
. reservoir and purifier rooms, and RFPT rooms are protected with
! wet pipe sprinkler systems. The lube oil storage room and the main turbine lube oil reservoir and purifier room are protected by a water spray system.
i 9.5.1.2.26 Diesel Generator Area Fire Protection
! The diesel generators are separated from each other and other
, areas of the plant by 3-hour fire barriers with Class A fire
- doors.
' One 550-gallon capacity diesel generator fuel oil day tank is located in each diesel generator room. An automatic fixed carbon dioxide total flooding system is provided in each diesel i
generator room. Manual water hose stations are provided as a backup fire suppression system. Photoelectric and infrared detectors that alarm locally and annunciate in the main control room are provided in each diesel generator room.
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f Each diesel generator room drains via normally closed isolation valves to a common drainage sump pump basin that has a sump pump capable of discharging 100 gpm. ,
l The normal ventilation system can be used for manual smoke venting. Each supply and return duct is provided with ETL-operated fire dampers. l t
4 9.5-36 Amendment 4 ysgd offs rTEW /W
_ . _ _ _ . _ - - . _ _ . . _ - _ , , _ . _ ~ _ _ _ _ . . _ . _ _ . . , _ _ . . . _
HCGS SER Open Item No. 165 Insert A Each fuel oil day tank is provided with a dike which surrounds the floor area ur. der the tank. The dike area is also below the equipment access grating that surrounds the diesel generator. Three sides of the dike are made of a 6 inch channel that is bolted to the floor with a neoprene oil-resistant gasket. The fourth side is the east wall of the diesel generator room. The floor area within the dike is sloped to a sump area located in the middle of the dike area. No drain is provided to drain the dike area or the sump. However, the dike has sufficient capacity to hold 110 percent of the contents of the fuel oil day tank.
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MP84 56/11 5-db 1
HCGS DSER Open Item No. 182 ( D6ER Section 15.9.10)
TMI-2 ITEM I I . K 3.18 l The applicant should specify which option they are planning to implement. Either option 2 or option 4 is acceptable to the ;
staff.
RESPONSE
Option 4 will be implemented at HCGS, as indicated by our response to Question 421.12.
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, HCGS DSER Open Item No. 185 ( DSER Se ction 7. 2. 2. 2 )
TRIP SYSTEM SENSORS AND CABLING IN TURBINE BUILDING 1 The applicant is required to describe the separation utilized between redundant channels listed below and demonstrate that the design can withstand the ef fects of missiles, HELB and seismic events in a way that is consistent with satisfying the safety analysis described in FSAR Chapter 15.
IDENTIFICATION DESCRIPTION LOCATION SB-PT-N052A Main (Turbine) Stop valve Turbine Building <
Closure and Turbine Con-trol Valve Fast Closure Trips Bypass SB-PT-N052B Main (Turbine) Stop Valve Turbine Building Closure and Turbine Con-trol Valve Fast Closure Trips Bypass SB-PT-N052C Main (Turbine) Stop Valve Turbine Building closure and Turbine Con-trol Valve Fast Closure Trips Bypass j SB-PT-N052D Main (Turbine) Stop Valve Turbine Building Closure and Turbine Con-trol Valve Fast Closure Trips Bypass SM-PT-N076A MSIV*-Low Steam Line Turbine Building .
Pressure Trip (PCRVICS)**
SM-PT-N076B MSIY*-Low Steam Line Turbine Building Pressure Trip (PCRVICS)**
SM-PT-N076C MSIV*-Low Steam Line Turbine Building Pressure Grip (PCRVICS)**
SM-PT-N076D MSIV*-Low Steam Line Turbine Building Pressure Trip (PCRVICS)**
SM-PT-N075A MSIV*-Low Condenser Turbine Building Vacuum Trip (PCRVICS)**
SM-PT-N0758 MSIY*-Low Condenser Turbine Building Vacuum Trip (PCRVICS)**
185-1
l HCGS DSER Open Item No. 185 ( Con t' d )
MSIV*-Low condenser Turbine Building SM-PT-N075C Vacuum Trip (PCRVICS)**
MSIV*-Low condenser Turbine Building SM-PT-N075D Vacuum Trip (PCRVICS) * *
- PCRVICS = Primary Containment and Reactor Vessel Isolation Control System
RESPONSE
The response to Question 421.17 addresses the concerns of RPS sensors located in non-seismic structures (turbine building).
4 185-2
HCGS DSER Open Item No.190 ( DSER Se ction 7. 2.2.7)
REGULATORY GUIDE 1.75 We asked the applicant to provide an overview of the plant elec-trical distribution system with emphasis on the reactor protection system (i.e., reactor trip, engineered safety features actuation and supporting features) instrumentation including the se n so r s ,
logic, and actuation relay power supplies and divisional separation as a background for addressing FSAR Chapter 7 concerns.
In a meeting, the applicant provided a response to this question.
The staf f reviewed this response and found it to be acceptable if reference is made to figures and a table of the FSAR are revised.
We require that the applicant aug me nt this response to make re fe r-ence to Figures 8.3-8, 8.3-9, 8.3-11, and 8.3-13 and revise F ig ur e 7. 2-1 of the FS A R.
RESPONSE
The response to Question 421.9 provides the requested information
! concerning an overview of the plant electrical distribution system.
1 Figure 7.2-1 was cavised in response to Question 421.14.
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l HCGS DSER Open Item No. 192 (DSER Section 7.2.2.9)
REACTOR MODE SWITCH We require the applicant to augment the response concerning the reactor mode switch to indicate that HCGS has responded to IE Notice 83-42 and that the modified mode switch will be installed prior to fuel load. In addition, we require the applicant to clarify the response regarding the use of operational procedures as the primary method of controlling rod movement during refueling.
RESPONSE
The response to Question 421.26 addresses the concerns about the reactor mode switch installed at HCGS.
MP84 56/11 6-db m
F HCGS l
DSER Open Item No. 194 ( DSER Se ction 7.3.2.2)
STANDARD REVIEW PLAN DEVIATIONS The staf f has reviewed the applicant's response concerning SRP deviations and has concluded that they are acceptable with the exception of the ESF equipment area cooling system and the SSEAVS.
For these systems, the applicant is required to provide additional justification or show system applicability to the Standard Review Pla n Table 7.1.
RESPONSE
The response to Question 421.2 provides the justification for any deviations between HCGS control systems design and SRP Table 7-1 requirements.
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r-l HCGS DSER Open Item No. 197 ( DSER Se ction 7.3.2.5)
MICROPROCESSOR, MULTIPLEXER AND COMPUTER SYSTEMS We require the applicant to expand the response regarding the Bailey 862 modules and to provide a typical set of drawings and the instruction manuals for the Bailey Model 862.
RESPONSE
The response to Question 421.6 provides the requested information concerning the reliability of the Bailey 862 equipment. Typical drawings and Bailey 862 instruction manuals were provided to the NRC as additional documents during the Ja nuary 13 ICSB meetinas l
F
. HCGS
, DSER Open Item No. 200 ( DSER Se ction 7.4. 2.2)
REMOTE SHUTDOWN SYSTEM The applicant is required to confirm that the HCGS design meets the staf f's guidance for remote shutdown capability.
RESPONSE
The response to Question 421.38 identifies how the HCGS remote shutdown systems design meets the NRC staf f's guidances fo r remote shutdown capability.
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i HCGS l
o DSER Open Item No. 205 ( DSER Se ction 7. 5. 2.4)
PLANT PROCESS COMPUTER SYSTEM Pending final revisions to FSAR Sections 7.5 and 7.7. These revisions should provide clarification of the safety categori-zation of the information systems addressed in these sections.
RESPONSE
The response to Question 421.55 addresses the concerns regarding the plant process computer system.
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e HCGS DSER Open Item No. 209 ( DSER Se ction 7.7. 2.3)
\
CREDIT FOR NON-SAFETY RELATED SYSTEMS IN CHAPTER 15 OF THE FSAR The peak vessel pressures resulting from the analyses of the transients without taking credit for nonsafety-related struc-tures, systems, and components are bounded by the peak pressure limit of the overpressure protection system as described in the Hope Creek FSAR.
The staff is reviewing the applicant's response relating to this concern and will report its finding in a future SER.
RESPONSE
The response to Question 421.54 identifies which of the nonsafety-grade systems / components that may be actuated during the course of anticipated operational occurrences (transients) are included in the Te chnical Specifications.
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C HCGS DSER Open Item No. 210 ( DSER Se ction 7.7. 2.4)
TRANSIENT ANALYSIS RECORDING SYSTEM The applicant is required to document the test result details regarding the GETARS I remote multiplexer unit and its asso-ciated electrical isolation. The staff is presently reviewing the information provided by the applicant in response to our request for additional information and will document the results of this review in a future safety evaluation report (SER).
RESPONSE
The response to Question 421.49 fully describes the transient i analysis recording system being used at HCGS.
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HCGS DSER Open Item No. 218 (DSER Section 9.5.1.1)
FIRE HAZARDS ANALYSIS i
i GDC #3 requires: " Fire fighting systems shall be designed to assure that rupture or inadvertent operation does not significantly impair the safety capability of those struc-tures, systems, and components." To satisfy this require-l ment, the applicant has designed components required for hot
- shutdown so that rupture or inadvertent operation of fire
! suppression systems will not adversely af fect the opera-
! bility of these components. Where necessary , appropriate protection is provided to prevent impingement of water spray
, on components required for hot shutdown. Redundant trains l of components that are susceptible to damage from water spray are physically separated so that manual fire suppres-sion activities will not adversely af fect the operability of components not involved in the postulated fire. However, the staf f is concerned that the mechanism by which fire and i fire fighting systems may cause the simultaneous failure of redundant or diverse trains has not been adequately con-
! sidered in the design. The staff will require that the
! applicant identify such mechanisms that were considered in j his fire hazards analysis and the measures taken to preclude the fire or fire-suppressant-induced failure of redundant or diverse safety trains.
, RESPONSE In response to Appendix R and IE Notice 83-41, each suppres-sion system covering safety related areas was reviewed for spurious actuation either by seismic induced error or operator error. Those systems with closed heads were found L
t acceptable as is, since a second, failure of one or more heads would be required to discharge water. This applied to the cable spread room at elevation 77 f t. and the intake structure service water pump rooms. The automatic CO2 systems cover the control equipment mezzanine at elevation 117 f t.-6 in., the diesel generator fuel oil tanks and the diesel generators. Spurious actuation of any diesel generator CO2 system will cause a trip of that DG set but j since each DG set is separated, this will not prevent safe shutdown utilizing the remaining diesel generators or
! offsite power. Spurious actuation of the control equipment mezzanine CO2 system will not af fect the cable therein or MP84 56/11 7-db t
I als-t
f safe shutdown from the main control room or the remote shutdwon panel room, since the mezzanine only contains cable.
The open head type systems were changed to manual opera-tion. These are the FRVS charcoal filter system, main control room emergency charcoal filter system, and the diesel fuel oil tank room systems. The deluge valves are manually actuated by pushbutton on the local panel. The OS&Y gate valve is also kept closed to prevent spurious actuation. Please refer to Section 9A.4.1.2.2.
Drainage and flooding caused by automatic or manual fire fighting has been considered and will not prevent safe shutdown. Please refer to Section 9.5.1.1.9 and the reply to Question 640.9.
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MP84 56/11 8-db I sis 4 L
OUESTION 640.9 (SECTION 14.2.12)
Modify FSAR Subsection 14.2.12.1.29 (KC-Fire Protection - Deluge) to provide assurance that:
- 1. Upon automatic sprinkler actuation, adequate drainage in the affected spaces is provided to preclude flooding (including
__ expected hand-held hose volume).
- 2. A walk-down of plant equipment is conducted to identify potential incidences where the actuation of fire suppression systems could cause damage to or inoperability of systems important to safety.
See IE Information Notice 83-41: Actuation of Fire Suppression System Causing Inoperability of Safety-Related Equipment, June 22, 1983.
RESPONSE
The :::;lt: Of ::: finding; en ;d;;;;;, ;f tr.: d::ir,;;; to "I;22 *; ;F;iI' i;; ;I "t i;I "i I ";
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I Section 14.2.12.1.29.b has been revised to include a prerequisite walkdown of the fire protection system to identify potential areas where the fire protection system could cause damage.
Section 14.2.12.1.29.b has boon revised to address the provinion to drain areas where automatic sprinkler actuation might affect I l safo shutdown equipment. ,,s/
1 D3t8 0/WW' /npe 4/g 640.9-1 ,
Amendment 5 )
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- 2. The system responds to simulated fire signals.
- 3. The refrigeration system operates to maintain pressure and temperature as specified by the manufacturer's technical instruction manual.
14.2.12.1.29 KC-Fire Protection - Deluge
- a. Objective The test objective is to verify the capability of.the fire protection system to deliver water to the sprinkler system, pre-action and deluge systems, hose stations, and hydrants at rated pressure and flow.
- b. Prerequisites
- 1. Component tests have been completed and approved.
- 2. System instrumentation has been calibrated and approved.
- 3. AC and de power are available.
- 4. The diesel fire pump local fuel oil storage tank is in service.
- 5. Adequate fire protection water supply is l available.
- 6. A walkdown has been performed to identify components or areas that may be susceptible to damage due to actuction of the deluge system.
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- c. Test Method
- 1. All valves, controls, alarms, interlocks, and .
logic are checked for proper operation.
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- 2. Normal system flow paths are verified. '
1 Qeg opew wesn ai9- 14.2-78 Amendment 4 o,
3
Insert A
- 7. Floor drains have been provided to remove the expected fire fighting water flow from automatic sprinkler
_ systems, hand hose lines, etc. Temporary build up of water in the affected spaces will not flood safe shutdown equipment.
I c OSCA OPcN /TE m 41/ 8
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y-HOPE CREEK DSER OPEN ITEM RESPONSE TECHNICAL SPECIFICATION ITEM 4.4.5 (TS-3):
Core Flow Monitoring for Crud Ef fects Crud deposition causes gradual flow reductions in some light water reactor cores. However, measurement of core flow by jet pump pressure drop and core plate pressure drop will provide adequate indication of such flow reductions, if they should occur. Technical Specifications will require that the core flow be checked at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to detect flow reduction.
RESPONSE
Crud deposition is assumed in the General Electric methods used for the design of fuel and for calculating pressure drop. This assumption has no significant impact on the CPR results. Thus, crud deposition is conservatively accounted for when predicting fuel performance. The build up of crud occurs very slowly, especially in the early years of fuel life, therefore, such a short inspection frequency is not justifiable.
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I MP84 56/11 9-db L.___
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HOPE CREEK DSER OPEN ITEM RESPONSE LICENSE CONDITION ITEM 4.2 (LC-1): Fuel Rod Internal Pressure Criterion The applicant must demonstrate t. hat a fuel rod internal pressure criterion which allows the internal fuel rod pressure to exceed system pressure will not (1) lead to fuel system damage during normal operation and A00s, (2) prevent control rod insertion when required, (3) result in an under-estimate of the number of fuel failures in, or radiological consequences of, postulated accidents or (4) lead to loss of coolable geometry.
RESPONSE
General Electric has proposed an alternative internal pressure criterion that would satisfy the SRP requirement and would resolve this issue i generically. The NRC staff is presently reviewing General Electric's I
proposal as part of its review of an amendment to GESTAR II (NEDE-24011).
f Completion of that review is expected by April 1984, and General Electric will pursue this matter to a complete resolution.
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