ML20106J734

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Forwards Tabulation of FSAR Commitments Through Sept 1984 & Corresponding Resolution for Commitments.Response Will Be Incorporated in Amend 8 or 9
ML20106J734
Person / Time
Site: Hope Creek 
Issue date: 10/29/1984
From: Douglas R, Mittl R
Public Service Enterprise Group
To: Schwencer A
Office of Nuclear Reactor Regulation
References
NUDOCS 8411010327
Download: ML20106J734 (47)
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{{#Wiki_filter:' , e ~. - ' Put*c serwce j__ E'ecinc and Gas Cornpany 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 October 29, 1984 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 . Division of Licensing Gentlemen: HOPE CREEK GENERATING STATION DOCKET NO. 50-354 FSAR-COMMITMENT STATUS THROUGH SEPTEMBER 1984 Public Service Electric and Gas Company presently does not plan to issue Amendment No. 8 to the Hope Creek Generating Station Final Safety Analysis Report before November 1, 1984. Accordingly, this letter is provided to document the status of Hope Creek Generating Station responses to NRC requests for additional informe.':. ion which were forecasted to be responded to by September 1984. Attachment I is a tabulation of the Hope Creek Generating Station Final Safety Analysis Report commitments for September 1984, and the corresponding resolution for each commitment. Attachments II through VI provide responses to commitments forecasted to be responded to in September 1984, which will be included in Amendment No. 8 or 9. t 0411010327 841029 ' d)0 ^ PDR ADOCK 05000354 \\ A PDR -s The Energy People 95 4H2 (49) 743

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b ..b IDirectorlof Nuclear-Reactor Regulation 2 10/29/84 Should you have any questions in this regard, please contact us. -Very truly yours, Attachment'I - Hope Creek Generating Station - FSAR Commitment Status through September 1984 - Response to FSAR Section 3.11.2.6 Attachment-II Attachment III - Response to Question 220.15 Attachment IV Response to Question 270.2 . Attachment V - Response to Question 410.38 Attachment VI - Response to SRAI(5) -_ C D. H. Wagner (w/ attach) USNRC. Licensing Project Manager W. H.-Bateman (w/ attach) USNRC Senior Fesident Inspector MP84 123/06 1/2 l

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ATTACHMENT I .Page 1 of 3 HOPE CREEK GENERATING STATION FSAR COMMITMENT STATUS :THROUGH SEPTEMBER 1984 p: lFSAR COMMITMENT LOCATION . COMMITMENT-RESOLUTION

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NRC Generic Letter 83-28 -This commitment concerns Response :1 providing station operating .(PSE&G to.NRC, 3/30/84)' L procedures referenced in the response to NRC Generic Letter 83-28 of 3/30/84. This information will~ be provided in November 1984. FSAR Section 1.14.1.37.2 1This commitment concerns 2. providing setpoints for undervoltage relays and system voltages. This information is provided in Amendment 6 to the HCGS FSAR. 13. FSAR Section 3.11.2.6 This commitment concerns providing a tabulation of all safety-related ~. mechanical' equipment located-in a' harsh environment. This information is provided' in Attachment II and. will be included 'in Amendment 9 to the HCGS FSAR. .4.- ' FSAR' Table-13.1 This commitment concerns providing the resume for the Technical Engineer..This information is provided in letter;.R. L. Mittl (PSE&G) to A..Schwencer (NRC) dated August 15, 1984, and will be - included in Amendment 8 to the HCGS FSAR. 5. . Question / Response. Re: TMI Item I. A.3.1 : This Appendix: canmitment concerns Hope Question'100.6 Creek simulator training. t M P84 164/03 1-gs .w, ~,, .,<,,.-_.m...... _,,.,,--,..~.,.--,-.,-----,..,,_..,...-s--. ._..,..m._y-,

7 Page 2 of 3 FSAR COMMITMENT LOCATION COMMITMENT RESOLUTION y 5. Question / Response Information in Section 13.2, Appendix: which is provided in letter; Ouestion 100.6 R. L. Mittl (PSE&G) to (Cont'd)- A. Schwencer (NRC), dated October 3, 1984, and which will be included in Amend-ment 8 to the HCGS FSAR, indicates that simulator-training is being conducted at.the Susquehanna Training Center until-the Hope Creek simulator is operational. The Hope Creek simulator is scheduled to be. operational in November 1984. 6. Question / Response This commitment concerns Appendix: providing Spent Fuel Rack Ouestion 220.15 analysis, sketches, and mathematical models. This information is provided in Attachment III and will be included in Amendment 8 to the HCGS FSAR. 7. Question / Response This commitment concerns Appendix: providing a preliminary sum-Question 270.2 mary _ report describing the HCGS Environmental Qualifi-cation Program for electri-cal equipment. This summary report is provided in letter; R. L. Mittl'(PSE&G) to A. Schwencer (NRC), dated August 24, 1984. Reference to this information is pro-vided in Attachment IV and-will be included in Amend-ment 8 to the HCGS FSAR. .8. Ouestion/ Response This commitment concerns Appendix: providing Spent Fuel Pool Question 410.38 criticality information. This information is provided in Attachment V and will be included in Amendment 8 to the HCGS FSAR. '9.. Question / Response-This commitment concerns Appendix: testing of SSLIS and AIS Ouestion 421.13a isolation systems. This information is provided in Amendment 7 to the HCGS FSAR. M P84.164/03 2-gs

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10.. - Ouest' ion'/ Response This commitment 1 concerns

^ Appendix:. review of inverters-to re- ~~ LOuestionl430.32 ' quired volt' ge' range..This a --information is provided as response to DSER Open Item No. 258 in letter; R. L. Mittl '(PSE&G) to A. Schwencer-(NRC), dated August 1, 1984, and will be included in Amendment 8 to the HCGS FSAR. l l. j-Question / Response 1This commitment concerns ' Appendix:1 tests and analysis of inver-

-Question 4 30.33.

ters'as' isolation devices. -This information is provided as response to DSER Open Item No. 259. in letter; R. L..Mittl (PSE&G) to A. Schwencer (NRC), dated . October 3, 1984, and will be included in Amendment 8 to the HCGS FSAR. J L12[ Supplemental-Request 'for ThisJcommitment concerns

Additional Information:'

' verifying seismic and dyna- -SRAI (5) mic qualification and -installation of 85-90% of safety-related equipment. This-information is provided in letter; R.'L. Mittl (PSE&G) to A. Schwencer ~. (NRC), datedL October 19, 1984. -Reference to this information is provided in Attachment ~VI and will be included in' Amendment 9 to the HCGS FSAR. -:13. - 1DSER Open I tem No. 103 This commitment concerns updating FSAR Section 3.10 g to show extent of opera-tional testing. This infor-mation will be provided in March 1985. ' 14. - DSER OpenLItem No. 189 This. commitment concerns providing documentation to NRC regarding qualification testing performed on isola-tion systems. This informa-tion will be provided in March 1985. J M ' P84 E 164/03.'3-gs - {

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i HCGS FSAR 8/84 '( environment-is to establish the qualified life by analysis, including the operability requirements during and after the DBE periods, for'the component materials with "significant aging mechanisms" as defined in Section 4.4.1 of IEEE STD-627-1980. Nonmetallic materials with a qualified life greater than 40 years are not considered to be susceptible to significant age degradation. Nonmetallic parts used in mechanical equipment include gaskets, diaphragms, seals, lubricating oil or grease, fluids for hydraulic systems, flexible hoses and packing. Environmental qualification of mechanical equipment is not intended to replace or modify compliance required by adherence to other applicable codes prepared by organizations such as the ASME, AISC and ACI which are the recognized experts in their fields of endeavor.- i A tabulation will be provided by September 1984 listing all safety-related mechanical equipment located in a harsh environment. Nonmetallic subcomponents of this equipment will be t indicated and their qualified status provided. The environments for which this equipment is qualified to operate \\ in are identical to those defined for the electrical equipment qualification program. 3.11.2.7 Oualification Methods for NSSS and Non-NSSS Safety-Related Electrical Equipment 3.11.2.7.1 Margin IEEE-323 and NUREG-0588, Rev. 1, Paragraph 3.(4) are used as a basis for determining margin. The equipment technical i specification for safety-related electrical equipment required to be environmentally qualified include conservative environmental conditions which were derived using environmental parameters which contain conservatisms applied during the derivation of local environmental conditions. The equipment vendor determines what margin must be added to allow for variations in production processes, for inaccuracies in the test equipment and for errors associated with defining satisfactory performance. l The qualification documentation for safety-related equipment will include documented provision that adequate margin has been ( 3.11-8 Amendment 7

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^., TABLE 3.11-4 MECHANICAL EQUIPMENT SELECTED FOR HARSH ~( ENVIRONMENT CUALITIOATION PURCHASE ORDER COMPONENT I.D. NUMBER Relief Valves B21-F013 Safety /'am Isolation Valves M-001 Main St e B21-F022/F028 M-001 M-001 Recirculation Pumps B31-C001 M-001 Recire. System Valves B31-F023/F031 (Suction and Discharge) M-001 Hydraulic Control Units Cll-D001 M-001 CRD Vent Valves Cll-F010/F180 M-001 CRD Drain valves Cll-F0ll/F181 M-001 SLC Pumps C41-C001 M-001 RHR Heat Exchanger Relief Valves Ell-B001 M-001 RHR Pumps E11-C002 M-001 RHR Check Valves Ell-F041/F050 M-001 LPCS Check Valves E21-F006 K-001 LPCS Pump E21-C001 M-001 HPCI Pump E41-C001 M-001 RCIC Pump E51-C001 M-001 NeuXtron Monitoring System C51-J004 Valve Ass'y '(. M-001 RCIC Turbine Assembly E51-C002 P-301(O) Valves P-302(O) Valves P-303A(0) Valves P-306(O) Butterfly Valves P-366(O) Check Valves P-401D Snubbers M-070(O) SACS Pumps M-082(O) Fuel Pool Pumps M-141 Relief Valves M-150(O) Vacuum Relief Valves M-713(O) Centrifugal Fans J-601(O) Control Valves J-605(O) Valves J-703(O) Excess Flow Check Valves J-705(O) Instrument valves J-715(O) Instrument Valves VII-8 ( CECs.az LM3 01

4 l 1 ) ATTACHMENT III i-i k a )- 3 p i 5 s 5 l t b i i a a _. _.. _ _.. _.. _ _ _ _ ~ _,,., _. _ _ - _.. _... _..

HCGS FSAR 6 \\ ( I QUESTION 229.15 (SECTION 2.8.4) Provide sketches of the mathematical models used in the design of shnt fuel racks. Describe in detail, the methods of Analysis by witch seismic and other loads are applied to the racks and the pool.

RESPONSE

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I HCGS FSAR 6/84 3.8.4.8.3 Spent Fuel Rack Design l Acceptance Criterion II.4.f requires that the spent fuel racks be which designed in compliance with Appendix D of SRP 3.8.4, requires that construction materials should conform to Section III, Subsection NF of the ASME Code. > /AIS ER.T C= The spent fuel racks are constructed of ASTM A-240 and ASTM A-564 stainless steel. The A-240 and A-564 material specifications are identical to the ASME SA-240 and SA-564 material specifications. All rack steel is supplied with certified material test reports. The rack materials are procured under a Q.A. Program that is intended to comply with: 10CFR50, Appendix B, " Quality Assurance Criteria for a. Nuclear Power Plants and Fuel Reprocessing Plants". b. ANSI /ASME N45.2, " Quality Assurance Program I,g Requirements for Nuclear Facilities", and i~ c. ANSI /ASME NOA-1, " Quality Assurance Program Requirements for Nuclear Power Plants". 3.8.5 FOUNDATIONS f ' Foundations for all Seismic Category I structures and the turbine building and the administration facility, which are non-Seismic Category I structures, are described in this section. i 3.8.5.1 Description of the Foundations i The configuration of the foundation mats for the various structures is shown on Figure 3.8-37. Reinforced concrete mat foundations are provided for all Except for the station service water system (SSWS) structures. intake structure, the sats rest either on the Vincentown Formation or on engineered structural backfill placed on the ( Vincentown Formation. The mat and the lean concrete leveling Amendment 6 3.8-48b

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HCG.5 FSAR 8/84 9.1.2.2.2.2 High Density Spent Fuel Storage Racks High density spent fuel storage racks in the fuel pool store spent fuel transferred from the reactor vessel. Tr3se are top-entry racks. The spent fuel storage racks are of freestanding design and are not-attached to either the fuel pool wall or the fuel pool liner plate. The racks are constructed of stainless steel, and the neutron absorber is Boral. See Figure 9.1-3 for design details of a typical rack and the special rack. e,5 Cf t 06# h h b q C en lA 0 bhe psi h, 05/9/l' 0ll WI ccus bruchkn ah ele Gfeht 5dvoc e r4 clCJ o I \\ l i l l l ( l 9.1-10b Amendment 7 l l

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$l I m C O

o ATTACHMENT IV

}, HCGS FSAR 10/83 O QUESTION 270.2 (SECTION.3.11) Prior to the completion of our review of your license application, it is necessary that we establish that you comply with the Commission's requirements applicable to environmental qualification contained in 10 CFR 50.49 for electrical equipment important to safety; GDC 4, Appendix A,. 10 CFR 50; and Appendix B, 10 CFR 50, Sections III, XI, XVII. As a result of the issuance of Section 50.49 of 10 CFR Part 50, some of tne information requested in SRP3.11 and R.G. 1.70, Section 3.11, is no longer required for staff review. Other new information is required, however, and is defined in this guidance. By utilizing these guidelines to deronstrate compliance with the Commission's regulations, applicants can significantly reduce the need for requests for additional information from the NRC staff. The information required may be submitted in Section 3.11 of the FSAR or in a separate submittel. If the latter approach is chosen, Section 3.11 should reference the information in the environmental qualification program submittal. ~ O %97 describeSin detail the HCGS N j Environmental Qualification Program for electrical equipment. It is the intent of PSE&G to answer NRC questions andamendissy the FSAR to clarify any given position prior to the submittal of th6s -WA ha.) veys/mreport he information N e4 tkt specif in NUREG-0588 as modified y 10 CFR 50.49. W ..n_ 1 l N k d to ;l\\ M M l I l i l O 270.2-1 Amendment 2

F' s ATTACHMENT V l

g.. ;f.

aao840268840 HCGS FSAR 3/34 l ( QUESTION 410.38 (SECTION 9.1.2) insufficient information is provided for review of the crit'icality nf the spent fuel pool. the design bases ath acceptable wTth respect to criticality. the information requiced for the review is promised fe,r later. guch information should include the following: 3 Sufficient structural detail to permit an independent i calculation of tne griticality of the racks. j h. A description of the calculational sethods used along with the results gf the verification o5 the methods. Ihis may be by reference to occuments previously submitted by the organisations doing the analysis. A tabulation of the nominal value of k effective of the g. racks along y,ith the various uncertainties and himses considered in the analysis. g. A tabulation of the reactivity effect of each of the i abnormal lace' .t) situations considerred. RESPQMSZ --!!a.ma :r.!::reti:n f;;.ain f t' c*iticelit ' ef t'" r;:.1 i i L:.-:-:-:, ;r.a.ao, co.s u. a :t=: zm se : = d =512 er { ":;t.__=r ive, annwu1 ne a665a so .: m e1 1 o l Sediw q.i.z.3.3. ha keen tevke/ te klude th e infe6 tePM NV*- l i [ i i l i f I i 410.38-1 Amendment 7 l -++w r . - -e,.--.---,.we-w .-.#---,--wwe,.---,-,-wwww ,,w et-

l HCGS FSAR 3/84 ?- Ab.,. u l 1 The maximum stress in the fully loaded rack in a 1. faulted condition will be provided prior to fuel load. l t. The spent fuel storage racks also have the capability j. of storing control rod guide tubes, control rods, and l defective fuel containers. When the spent fuel is stored in the spaces provided for storir.g the above the l i K,gg does not exceed 0.95. k. Several design features reduce the possibility of heavy ~ objects dropping into the fuel pool. The main and i auxiliary hoists of the reactor building polar crane are single-failure proof. In addition, the main hoist j is physically prevented from traveling in the truncated i segment shown on Figure 9.1-31 by mechanical stops on i the girders of the polar crane. The crane design is discussed in Section 9.1.5. The removable guardrail and the four-inch curb around the refueling cavities further limit the possibility of heavy objects dropping into the fuel pool. i 1. The fuel storage pool has water shielding for the stored spent fuel. Liquid level sensors are installed j to detect a low pool water level. Makeup water is available to ensure that the fuel will not be uncovered should a leak occur. i Since the fuel racks are made of noncombustible i m. material and are stored underwater, there is no The large water volume also i potential fire hazard. protects the spent fuel storage racks from potential i pipe breaks and associated jet impingement loads, DVSW q. I. 2.73 9.1.2.4 Spent Fuel Rack Inservice Inspection I ~ qu Tnsed.5 1 9.1-14 Amendment 7 ) 1 . -. _. _,..,..... _ _. _. _.. _. ~ _ _.. -..__.~._____.__.--.,...__.____....._.,_...__.___...J

e faf6 I HCGS FSAR INSERT A 9.1.2.3.3 CRITICALITY ANALYSIS AND RESULTS The criticality analysis was performed using the input parameters Figure 9.1-20 shows the re'forence contained in Table 9.1-19. geometry gM jfoffp forpticality analysisgw the aery f/s D4,d 7 d nt cr

essembly, the facta gg ces.ft'l sinngles
  • The criticality analysis is based on new fuel with a nominal,No credit is take dWh f flat U-235 enrichment of 3.4 w/o.

burnable poisAn fuel rods which may be present in the fuelas f bel

  • dif fusion Ensory model, CHEETAH-B/CORC-BLADE /PDQ7 as the 49;.#

The analysis includes the various criticality (UAI's) y main working model. safety-related aspects of the rack design, including various The Monte Carlo transport model, sensitivity calculations.is used as the verification model to verify the AMPX/ KENO -IV, reactivity of the nominal rack design. UAI performed similar criticality analyses for Limerick an <{L-accident conditions described in Section 9.1.2.3.1. 1 Table 9.1-20 summarizes the nominal value of K effective of the The racks under normal, abnormal, and accident conditions. various uncertainties and biases considered in the analysis are j also included. ~- + l l .1 ( 8 .,,.-,....+,,-,,--,_n.,,,,.,,,..--,,,,,_,nn,,n_,.w,_,.,_ _m.....,.,,,,,.n,_m._,-,,,

LAI 4-pg t A. gage 2,F 16 H CGS FSAR CALCULATIONAL MODELS This section presents a description of the calculational models and j l the basic assumptions used in this criticality analysis. l The Workinn Model j I The criticality analysis for the Hope Creek BWR spent fuel racks employs the CHEETAH-B/CORC-BLADE /PDQ-7 model as the basic Nis UAI's BWR lattice code engineering tool. CHEETAH-B based on the original LEOPARD code and uses a modified ENDF/B-II 6 cross section library. C IC-BLADE nerates equivalent diffusion theory cross sections for the control blade. The PDQ-7 mgram is the well-known,few-group spatial diffusion f theory code widely used by the industry. The CHEETAH-8/CORC-BLADE /PDQ-7 model, which is also a part of the LEAHS (Lifetime Evaluation and Ar,alysis of Heterogeneous Systems) nuclear analysis I, series of Control Data Corporation, has been extensively tested l through benchmarking calculations of measured criticals as well as l through core physics calculations for several operating power j reactors. i A zero current boundary condition was applied to the four sides of the unit reference storage rack cavity 'f' - 7to produce an infinite array eff ct. The two-dimensional. PDQ-7 calculations l were made for four neutron energy groups, two mesh intervals per l fuel pin, a flat U-235 enrichment description and a zero axial } I buckling to simulate infinite fuel length. r

  1. The Verification Model l

i The verification calculation employs the KENO-I AMPX del. t' The basic neutron cross section data comes from the master libraries i of AMPX - a 123 group GAM-THERMOS neutrun library prvpared from j f( ENDF/B version II data. The MITAWL module of the AMPX program is j i I. - = _ _ _. j

115250 fW)bbef [b N l(&$ fE 0 used to perfom a Nordheim integral treatment of the U-238 rer lances accounting for the self-shielding effect. The ~ working library produced by the NITAWL/AMPX module retains the 123 group energy structure and is used directly by KENO-IV. In the 'ENO-IV calculation, the spent fuel rack geometry including each fuel and water rod cell is represented discretely. To simulate the arrangement of a large number of storage rack units, and for a non-leakage condition in the axial directions, a specular reflective condition is applied to all six sides of the reference case storage rack cavity (F';- 7 -.3JL Basic Assumptions To ensure that the analysis follows a conservative approach and confonns to the general guidel,ines of criticality safety analysis ~ ( in Reference W the calculations are perfomed with the following assumptions:4.1-10 1. A flat 3.4 w/o distribution in an 8x8 bundle, with U-234 neglected 2. Fresh fuel, no burnable poison 3. Minor structural members replaced by water, i.e., spacer grids 4. Fresh water 5. Fuel is channeled. t ( I (

psett h are Hcf 16 W f 0""'W. Mccr s FMR s.,...s.u l REFERENCE CASE CALCULATIONS Physical Parameters and the Basic Storage Rack Cavity Geometry The reference storage rack cavity 'J6. das a pitch of l 6.308" + 0.030". The stainless steel canister has a nominal inside . clearance of 6.080 to accomodate 8x8 fuel assently channeled in 0.080" thick Zircaloy-4. Plates of the neutron absorber material Boral, consisting of B C in an aluminum matrix core and clad with 4 an aluminum sheath, are fastened to the outside of the canister. The Boral plate has a nominal total thickness of 95 mils and a minimum 2 B-10 density of 0.028 g/cm. Table #contains the values of the 1-l*l input parameters used in the analy is. The rack must accomodate both channeled and unchanneled fuel. Studies reveal that the channeled fuel in the rack is more, reactive than the unchanneled fuel. Taking the conservative approach, the study here involves channeled fuel (except in the accident condition where the dropped fuel is unchanneled in order to pemit the closest contact between the dropped fuet assembly and the rack). Two small, but non-conservative changes were made to the reference case in order to' facilitate modeling. First, the boral width was set at ~ '4.48" instead 'o'f 4.465". "5 econ'd, the stainless steel' flanges 'used in~ ~ ' ' ~ welding th'e outer w'rapp'er to the inner can were deleted. 'An adjustment waimade using P0Q to account for these differences.

  1. Results of the Reference Case Calculations 5.Hf 9'./-

Using the input data from Table [and Figure (except as noted above), the X,ff values of the reference case at 68'F were cal-culated for the calculational model described " hdun M." ' The msults are: pr.edoas ( W i m- -

InserF k ya9s F eF I6 M fia e h198Las dc(rs

=sAR PDO-7

. FNO-IV 0.9229 0.9306 + 0.0042 k,ff, reference calculation 0.9222 - 0.9390 95% confidence interval .= l l l l I W

[.n5erf~ & ?aya S 0$ 1-B4-4 N gbH Mc Gs FSAR W~ ~

5. i. z. a.3. 2.

psf 688 SENSITIVITY AND TOLERANCE REACTIVITY CALCULATIONS Temperature Effect aing the reference storage rack cavity geometry, the temperature of the fuel and pool water was varied. In addition to the nominal ~ 68'F. 40*F and 212*F were studied and the re'sults of the CHEETAH-B/....._ 9,g.g .....CORC-BLADE /PDQ-7 runs are given on. Tab,la Etataq$3:tadziE;figgady As shown, reactivity decreases continuously as temperature increases 40*F. - ' ' ~ ~ ~~' - ~ ~- "' ~ g Void Effect The effect of boiling (assuming equal voids inside and outs;de of the rack) was studied by varying the voids from 0% to 20% at a temperature of 212*F with the reference geometry. The CHEETAH-B/ 4.f-20 CORC-BLADE /PDQ-7 results are shown in Segamedmuut Tablep. As* indicated, k,ff decreases con'tinuously as the void fraction f increases. l aggE' Pitch Sensitivity The rack design permits the storage cavity pitch to differ from the 6.308" nominal value by _+0.030". The pitch sensitivity calculations of this analysis show the reactivity effect of these tolerance components as well as the reactivity pitch sensitivity by expanding the calculational range from -0.060" to +.030" at.030" intervals. The results, which are gisegueuBN 'i W tabulated in Table bn'dic&te 'that in'the neighbor-hoo'd of the nominal pitch, the pitch' reactivily coefficient is about.15%ak per.030" pitch change. l i e h l, \\ t e 1 [ w ~

Insed N l ? * '# UML a m o.a m.,. HCSS FSRR Effect of Boron The Boral Plates which separate two adjacent fuel assemblies have a nontinal thickness of.095" (consisting of an 73 mil core and 11 mil aluminum sheaths) a nominal width of 4.465" and an overall length of 11 feet 3 inches. The minimum B-10 loading 2 is 0.028 g/cm.'. (a) Boron Width Tolerance The effect of reducing the Boral width was examined. The PDQ-7 calculation for the reference case con-figuration with the Boral width reduced by 0.0625" yielded k,= 0.92641. Hence, the reactivity increases due to the -0.0625" tolerance on Boral width is ( Ak = +0. 00 29... __ _ _..~.-,- (b) Boron Density 2 The boron density was maintained at.028 g/cm for all calculations. ~ This ' areal density is the minimum density allowed by manufacturing design specifjcatiopi.____.. (c) Boral Core Thickness Variation The sensitivity to the Boral core thickness was de-l tarinined by calculations in which the thickness l varied from 61 mils to 80 mils (the aluminum sheaths were i Z.I.T varied. withjn_ toler.an.ce to obtaig.,e worst cas.e-ore _ thickness). The results, tabulated in Table show a continuous increase in reactivity as the core thickness increases. This is due to the fact that the areal density is held constant, so an increase in thickness reduces i '~~ density and.To a. smaWFgree, the boral effectiveness. volumetric l M

M 14 % Trtre+f~k fare 9 e.f 14 nev ni n - ~ n Jwte R.1984m IfC GS FSRll d Dimensional and Positional Tolerances The total Ak bias for dimensional and positional kolerances are calculated from five separate contributions: (1) Pitch Reduction (ii) Boral Width Reduction (iii) Inter-Cavity Spacing Reduction (iv) Off-center Loading (v) Boral Thickness Increase (1) Pitch Reduction.. The effect of reducing the center-to - center spacing of the rack cavities is obtained from time. Table 9I-20: ~ ~' W ;' L-- - %^ -

" " - T '

L ^.- ~ .co1cl [.5 Akj = 0.0015. (ii) Boral Width Reduction.'. The Ak bias due to reducing the Boral width by its tolerance 0.0625" is obtained from GasseppeWW$r) and is Ak2 = 0.0029. Tale 9.t-zo. (iii) Inter-Cavity Spacing Reduction. Any seismic effect that may reduce the separation distance between adjacent cavities j can be detennined from the pitch sensitivity studged duisadv6L' Bringing two adjacent cavities closer by O.048" results in the canisters toging and a reactivity increase Ak j = 0.0023 (from Table passmedynese). Since this reduction is the maximum reduction of pitch possible ~ l in this design, this effect will not be added to item (i), l but will replace it. l (iv) Off-center Loading. The free space existing between a . center fuel assembly and the top casting allows an ass ily to be loaded off-center in a cavity. It was Loh%D x -.

Inse d k f"J" 9 of I'

gghtgy, maag h' CGS FMW shown that this condition causes no adverse reactivity effect since the resulting k,7f for off, centered loading is less than that for properly ccstered assemblies.

(v) Boral Thickness Increase. The worst' case boral core thickness reactivity effect calculated due to manufacturing tolerance ~ stackup (.080") 'is 'obtained nom NNhind is ak =.0001. ~ S The above positive ak contributions are statistically combined to give the total ak bias for mechanical and seismi ancertainties. Ak = / ( k )2 +- (ak )2 +(k)2='0.0037 g 2 5 ) __._-y

l l. .T nsert Appe 10 oV 16 gcCrS F5h8 hv 9.f.2.7.33 y SPECIAL CASES Grappler Drop Accident The accident considered is the inadvertent drop of the assembly grappler used in lifting assemblies within the spent fuel pool. In this accident, the grapnkr is dropped in such a way that assemblies in adjacent rack cavities are displaced such that'they are resting in an off-center loading arrangement. The reactivity effect for this off-center arrangement was, dis-cussid'in 'Sectibn @ (% '(jE) M -_ ; _ L 'T.I 1. 3,J,2 Q Assembly Drop Accident No adverse (a) Single Assembly Dropped on Top of Rack. reactivity effect is expected from dropping a fuel assembly on top of a fully loaded storage rack during fuel' handling because of the large water thickness (-14 inches) existing between the top of the assemblies already inside the cavities and the dropped assembly resting on top of the rack. Moreover the P0Q-7 model assumes an infinite fuel length in the axial direction. (b) Single Assembly Next to Rack. The dropping of an assembly outside the rack is a possible event because of the un-obstructed water area existing between the periphery of the storage racks and the side walls of the pool. A conservative analysis to evaluate this situation is illustrated in Figure f. An assently, presumed to be 9.l-20 W s .- = = = = = = =

p+ -., c-~ Tn. seep & fape I[ op16 a.neviscogr Kne -8,-1984 McGs FSAR dropped during handling, lodge paralled)o an assembly in the outer cavity with no Bora slab separating the two assemblies. The d' ' sped assembly is unchanneled to pemit the closest contact wi the racgged=4MW% 6 -: A i n g ethe shsta The dimensions used are those ~ of the reference case. This arrangement of the dropped fuel assembly with a 31/2 x 3 finite fuel racg is reflected on three sides as indicated in Figure (t$'e~ fourth' side is a zero flux boundary. The k,ff result for this case was 0.9128. The result for the same geometry without the dropped fuel was 0.9064 giving an increase of reactivity of ak = 0.0063 for the above dropped assembly configuration.

- w _o i w '.12--L

,,.ta M h -. = - m r_ DM p Assembly Moving Between Two Storage Racks The rack structural design does not allow sufficient room to fit a fuel assembly between any two of the high density spent fuel racks. Therefore, the movement of assemblies between racks is precluded. Z --: M44.s>14 cad Ad.iacent to the 1 tact- -m__, -:v _ Specjal constde'rftion was given to the accident cordition of'the.9nc'emnt of two or' thre,e.bundlepi2'WcTei' Trim The = - j M T f !O @ M $ od the pool wall. A comparison ~ ,esn'3Mrized Wy]o tntaraction betMe bun $1esicuts.ide th6,3ack".and the fuet'in the~ rack _ M - - - 1

Insett A P"J' '2 E'

  • Mb I-{ cgs FSRR D une a 1984 y

6. Three Buff 81ssdrranged'in a "T_" ,/ ure 8 shows the PDQ-7 model used for e study. ~ Tab 5 includes the,resulting K-eff 'r he T arrange. nt. For't'his case the rea ivi.ty is not grea p t the rack design lim. of.95. f,/ f T' ee Bundles sh Sne Y.[2 h 6- ',,<#^ / 4 y-F.i ,/ mode'l used for the s dy. ['/ g6're 9 shows the PTable 5 includes' thg ulting K-eff for the linear / arrangement; For;this ca e the reactidy is not I greater than the rack desig itmit of.95. I ,f The K-eff o8 Table 45s lower t 'n'tha't of th ack with noassembfies cent 1b it. Th reason for this is thatatjf ra'ck is uncouhed from the undles by the e)l $ inches of, wafer separating theIn the case of1the linear bundle arrangement, the b dies outside d the rack are less reactise than the,,,r_ k itself. e bundles In the c4se of the "I" bundle.. arrangement, outside ' lire indN$eactive than' thecky 6. 3 Two Bundles ..g - Sin e tw'o bGndles adjacent to the rack is less active than three bundles adjacent to the rack, this c y ~ " = Q..'. - - (l,{ 3,"3,N4 New Fuel Storage in the Soent Fuel Racks The feasibility of storage of fresh fuel in the high density spent fuel racks was analyzed. Storage of new fuel in the mist, partly flooded, and dry conditions are addressed below. ( W

truarf k nye i3 &l6 f t{c (,5 FShit 25% Mist Condition The storage of new fuel of unifonn 3.4'w/o U-235 enrichment in the high density spent fuel rack in a 25% aqueous mist environment was analyzi with the KENO model feefer *^ Pr @The resultin~g k, and 95% confidence interval are shown below: 95% Confidence Interval 25% Mist .63751 0054 .6267 .6483 Dry Condition UAI experience in the analysis of poisoned rack criticality indicates that the fully flooded rack configuration is the most reactive with reacitivy decreasing with a decrease in moderator density. The 25% mist condition analysis confirms this as shown below. For this reason a dry condition analysis was not perfonned since it too will be less reactive than the flooded condition. 95% Confidence Interval Moderator Density = 3 Reference Case: 1.00 g/cm .93061 0042 .9223.9389 3 25% Mist Condition:0.25 g/cm .63751 0054 .6267.6483 Partly Flooded Condition The totally flooded condition as analyzed in the reference case is,more reactive than that of the partly flooded condition. q'f.2.3.35 M Special Spent Fuel Rack Storage ,?;"" .pte uvi. m. m +"-' %. a2 M, 5x6 non-borated special rack is to be installed in the ege:S spent fuel pool. Storage of control rods, control rod guide tubes and defective fuel is provided for by this special rack. This rack was analyzed for storage of ruptured fuel as shown in Figure CSpecial rack input parameters are summarized in Tabl% ct.HT 9,l-RI .as.

f aget A

  • )& NYtVlb f

dcGSFsnn The storage of ruptured fuel is a more reactive evaluation than that of control rods or contn)1 rod guide tubes. M Storage of Ruptured Fuel in the Fully Flooded Special Rack The storage of ruptured fuel assemblies within defective fuel ' storage containers inserted into the special ' - ! = Sntdack was analyzed using the CHEETAH-B/PDQ-7 diffusion theory model. The case was analyzed as an infinite array in ' order to simulate storage of.. r -Muptured fuel assemblies in the special rack. The resulting K,ff for this case was.6589. Considering that this K,ff accounts for no radial or axial leakage, the reactivity for the storage of fuel in the special ( rack is well beiow the design limit K;ff of.95. Storage of undamaged fuel within the special rack . is less reactive'than storage of damaged fuel. This is due to the fact that in the ruptured fuel case, the defective fuel storage container displaces water. For this reason, the stcrage of undamaged fuel was not analyzed. q StoAge of Ruptured EveTJin the Orained ecial Rae p The accid beiny considereddlere. s the unlikely __ ey that the spent f pool is/ drained while J' ruptured fuel remairts tored i the special rack / Under these con t ions, the ck woul /be ined of wa ra placed with r. The y tive fu s age contained woul , remain i .s . n.

.[nseef- & ctye iF of~l6 gy f gune],.1984 m f ( G5 PS U 4, I. 2.. U. 6 l g

SUMMARY

AND CONCLUSION The final result as calculated by both the working model (CHEETAH-B/CORCBLADE/ l . PDQ-7) and the verification model (AMPX/ KENO-IV) is sumarized in this section and compared to the NRC regulation k,77 limit of 0.950. The " Reference Case" referred in this report uses the nominal dimensions 9.1-2 I of given. in Figure f an. Table without the dimensional and material d. tolerances inciuded. g Resr'ts of the Transoort Monte Carlo (AMPX/ KENO-IV) Verification i Calc Jlations and the Calculational Bias k,ff, Reference Case M.l.2. 3.3 ( 9

0. 306 t 0;0042 ~~ -

Benchmark bias. Ak -0_.001 .9296 t 0.0042 95% Confidence Interval k,ff - 0.9212 < 0.9380 j The bias of the KENO-IV vs. measurement is based on criticality experiments performed with fixed neutron poisons U. These experi-ments were chosen because they approach the fuel storage rack configur-ation in that they used fixed poison plates between fuel rod clusters. The result of the benchmark calculations was that the KENO-IV results I f were 0.001Ak above the mea,sured value. This demonstrates a negative I bias of 0.001Ak. l g Stsmary of Results 0.9296 t 0.0042 k,ff, adjusted (KENON) Dimensional and Positional Tolerance, Ak 0.0037 (PDQ N) PDQ correction for non-conservative assumptions in the reference case, 0.0006 Ak (PDQ h ) Dropped Asserely. Ak (PDQM) 0.0063 _Q

.o e-J g $ g&t~ l'l fofe (6 ori* UM M P ~ '.9402 1 0.0042 O Final K,ff 95% Confidence Interval 0.9318 - 0.9486 9.T".S 0.00 Design Limit, k,ff - C.% SD The final k,ff value (0.9486) includes all the design specification tolerances, the postulation of a dropped fuel assently, the model bias, and the 95% confidence interval from the KENO calculations. However, the negative reactivity effect (- 0.5% ak) due to the presence of U-234 and the parasitic structure materials (i.e., spacer grids) in each assen61y was not included. s e 9 l I l (

p-e' O ATTACHMENT VI 1 l i I l 1

HCGS FSAR 8/84 (5) Eauipment Ciualifications: Seismic and Environmental - Seismic and. Dynamic Qualification and Qualification of Mechanical Equipment reviews-consist of two elements: a review of the FSAR, and a detailed onsite audit. The information required for the review is included as l and its attachments. The information will r not be required until the first calendar quarter of i 1984.. An o k

RESPONSE

A+ least [ I It i; c.rected th;t 85 t; 00 percent of the safety related l r equipment will be qualified (seismic and dynamic qualification) ( and_ installed by4th: third grrrter :f 1984. 3,A' status summary I list has been provided (letter from R.L. Mitt 1, PSEEG to <B A.~ Schwencer, NRC, dateds ;1, 7, 1984) for all safety related f d i equipment-for the SORT and PVORT audits in anticipation of the - ~~ NRC audits being conducted the third quarter Of 1000. -A parti =1- -; tate. euan:ry-liet h== b - cubmitted gr.dct seperet 00'fer '!:tt:r frar R. L. Mitt!,'PCric &n A Och-enuws, NRC, dated April 13 ???t}. ( k & m -Hg of b 1985'. The format of the status summary list follows the sample provided i in enclosure 9 of the NRC acceptance review letter dated June 23, 1983, with the following modification. The list will be is l assembled on a purchase order basis, with individual tag number [ identified. The items on the list-will-have system designators t identified. g SQRT and PVORT forms for all safety related equipment will be available for NRC review during the audit. The PVORT forms will contain information which is currently available per the purchase specifications for the subject components. e Information on seismic and dynamic qualification test programs for non-NSSS supplied components has been submitted under separate cover (letter from R.L. Mitt 1, PSE&G, to M.A. Schwencer, I-NRC,' dated October 5, 1983). The seismic and dynamic L qualification test program for NSSS supplied components has been L completed. r Seismic and environmental quelification of the containment vent f and purge valves is discussed in Section 6.2.5.2.2. i .However, additional seismic testing is scheduled to be done under i the NSSS environmental qualification program. This schedule has j been submitted under separate cover. i i i

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