ML20090A210
| ML20090A210 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/29/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8407110296 | |
| Download: ML20090A210 (74) | |
Text
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Pubhc Seryce O ITS # E Minc ay1 Gas G Company 80 Park PLua. Newark NJ 07101/ 201430 8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 0/101 Robert L. Mitti General Man +rr Nuclear Assurance and Rsgulation June 29, 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 ITEM STATUS Attachment 1 is a current list which provides a status of the open items identified in Section 1.7 of the Draft Safety Evaluation Report (SER). Items identified as " complete" are those for which PSE&G has provided responses and no confir-mation of status has been received from the staff. We will consider these items closed unless notified otherwise. In order to permit timely resolution of items identified as
" complete" which may not be resolved to the staff's satis-faction, please provide a specific description of the issue ,
which remains to be resolved.
Attachment 2 is a current list which identifies Draft SER Sections not yet provided.
In addition, enclosed for your review and approval (see Attachment 4) are the resolutions to those Draft SER open items listed in Attachment 3.
Should you have any questions or require any additional information on these open items, please contact us.
Very truly yours,
' ~~
8407110296 830629 -
PDR ADOCK 05000354 '
E PDR 00 f
I l Attachments The Ener0y People wu n m no
i Director of Nuclear Reactor Regulation 2 6/29/84 C D. H. Wagner USNRC Licensing Project Manager W. H. Bateman USNRC Senior Resident Inspector FM05 1/2
IATE: 6/29 ATTACHMFNr 1 DSER R. L. MITit I OPD4 SE)CTION A. SCHM:NCER ITEM NUMBER SUB1ECT STATUS LETTER [ATED.
m Sai,d 2.4.5 Wave inpact and runup on service Cmplete 6/1/84 water intake structure 7b 2.4.11.2 Thermal aspects of ultimate heat sink Caplete 6/1/84 9 2.5.4 Soil damping values Cmplete 6/1/84 j 10 2.5.4 Foundation level response spectra Cm pleto 6/1/84 11 2.5.4 Soil shear moduli variation . Cmpleto 6/1/84 12 2.5.4 Combination of soil layer properties Caplete 6/1/84 13 2.5.4 Lab test shear moduli values Cmplete 6/1/84 14 2.5.4 Liquefaction analysis of river bottczn Cmplete 6/1/84 sands 15 2.5.4 Tabulations of shear moduli Cmplete 6/1/84 l
16 2.5.4 Drying and wetting effect on Cmplete 6/1/84 Vincentown 17 2.5.4 Power block settlement nonitoring Cmplete 6/1/84 18 2.5.4 Maximum earth at rest pressure Cmpleto 6/1/84 coefficient 19 2.5.4 Liquefaction analysis for service Caplete 6/1/84 water piping 20 2.5.4 Explanation of observed power block Cmplete 6/1/84 settlement 21 2.5.4 Service water pipe settlement records Cmplete 6/1/84 22 2.5.4 Cofferdam stability Cmplete 6/1/84 23 2.5.4 Clarification of FSAR Tables 2.5.13 cmplete 6/1/84 .
and 2.5.14 !
24 2.5.4 Soil depth nodels for intake Cmplete 6/1/84 structure l
27 2.5.5 Slope stability Cmplete 6/1/84 l
M P84 80/12 1-gs Page 1 of 7
ATTAC19 TINT 1 (Cont'd)
D6ER R. L. MITIL TO OPEN SECTION A. SCHWENCER ITEM NUMBER SUIDECT STA1US IEITER IRTED 30 3.5.1.2 Internally generated missiles (inside Closed 6/1/84 containment) (5/30/84-Aux.Sys.Mtg.)
35 3.6.2 ISI program for pipe welds in Co plete 6/29/84 break exclusion zone 36 3.6.2 Postulated pipe ruptures Complete 6/29/84 41 3.8.2 Steel containment buckling analysis Cm plete 6/1/84 42 3.8.2 Steel containment ultimate capacity Cmplete 6/1/84 analysis 43 3.8.2 SRV/IDCA pool dynamic loads Cmplete 6/1/84 44 3.8.3 ACI 349 deviations for internal Cmplete 6/1/84 structures 45 3.8.4 ACI 349 deviations for Category I Cmplete 6/1/84 structures 46 3.8.5 ACI 349 deviations for foundations Cmplete 6/1/84 47 3.8.6 Base mat response spectra Ccmplete 6/1/84 48 3.8.6 Rocking time histories ccuplete 6/1/84 49 3.8.6 Gross concrete section Ccmplete 6/1/84 50 3.8.6 Vertical floor flexibility response Ccmplete 6/1/84 spectra 53 3.8.6 Design of seismic Category I tanks Ccmplete 6/1/84 54- 3.8.6 Combination of vertical responses Cmplete 6/1/84 55 3.8.6 Torsional stiffness calculation Cmplete 5/1/84 56 3.8.6 Drywell stick model developnent Cmplete 6/1/84 57 3.8.6 Rotational time history inputs Ccmplete 6/1/84 58 3.8.6 "O" reference point for auxiliary Cmplete 6/1/84 l building nodel
, 59 3.8.6 Overturning m ment of reactor Ccmplete 6/1/84 l building foundation mat 60 3.8.6 BSAP element size limitations Ccrpleto 6/1/84 M F04 80/12 2-gs Page 2 of 7 l
t
ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL 10 OPEN SECTION A. SCHWENCER ITEN NUMBER SUlUECT STATUS LETTER IRTED 61 3.8.6 Seismic nodeling of drywell shield Cmpleto 6/1/84 wall 62 3.8.6 Drywell shield wall boundary Caplete 6/1/84 conditions 1
63 3.8.6 Reactor building dome boundary Cmplete 6/1/84 l conditions 64 3.8.6 SSI analysis 12 Hz cutoff frequency Conplete 6/1/84 65 3.8.6 Intake structure crane heavy load Conplete 6/1/84 drop 67 3.8.6 Critical loads calculation for Capleto 6/1/84 reactor building dome 68 3.8.6 Reactor building foundation mat Conplete 6/1/84 contact pressures 69 3.8.6 Factors of safety against sliding and Conplete 6/1/84 overturning of drywell shield wall 70 3.8.6 Seismic shear force distribution in Conplete 6/1/84 cylinder well 71 3.8.6 Overturning of cylinder wall Cmpleto 6/1/84 72 3.8.6 Deep beam design of fuel pool walls Caplete 6/1/84 73 3.8.6 ASHSD damo podel load inputs Cmplete 6/1/84 74 3.8.6 1benado depressurization Conpleto 6/1/84 75 3.8.6 Auxiliary building abnormal pressure Cmplete 6/1/84 76 3.8.6 Tangential shear stresses in drywell Cmpleto 6/1/84 shield wall and the cylinder wall 77 3.8.6 Factor of safety against overturning Conpleto 6/1/84 of intake structuro 78 3.8.6 Dead load calculations Cmplete 6/1/84 79 3.8.6 Post-modification seismic loads for Cmpleto 6/1/84 the torus 80 3.8.6 Torus fluid-structure interactions Conpleto 6/1/84 I
M M4 80/12 3-gs Page 3 of 7 i
ATTACMENT 1 (Cont'd)
D6ER R. L. MITTL 10 OPEN SECTION A. SCHWENCER ITEM NUMBER SUR7ECT STATUS LETTER DATED 81 3.8.6 Seismic displacement of torus Ccmplete 6/1/84 82 3.8.6 Review of seismic Category I tank Complete 6/1/84 design 83 3.8.6 Factors of safety for drywell Cmplete 6/1/84 buckling evaluation 84 3.8.6 Ultimate capacity of containment Ccmplete 6/1/84 (materials) 85 3.8.6 Load combination consistency Ccmplete 6/1/84 88 3.9.1 Stress analysis and elastic-plastic Complete 6/29/84 analysis 89 3.9.2.1 Vibration levels for NSSS piping Ccmplete 6/29/84 systerns !
91 3.9.2.2 Piping supports and anchors Ccmp1ete 6/29/84 92 3.9.2.2 Triple flued-head containment Complete 6/15/84 penetrations 93 3.9.3.1 Inad combinations and allowable Ccmplete 6/29/84 l stress limits 94 3.9.3.2 Design of SWs and SRV discharge Ccmplete 6/29/84 piping 95 3.9.3.2 Fatigue evaluation on SRV piping Cmplete 6/15/84 and IDCA downcomers 96 3.9.3.3 IE Infomation Notice 83-80 cmplete 6/15/84 ,
97 3.9.3.3 Buckling criteria used for cmponent Cmplete 6/29/84 supports 98 3.9.3.3 Design of bolts Ccmplete 6/15/84 99 3.9.5 Stress categories and limits for Ccmplete 6/15/84 core support structures 100a 3.9.6 10CFR50.55a paragraph (g) Ccmplete 6/29/84 I l
102 3.9.6 Imak testing of pressure isolation Ccuplete 6/29/84 !
va*ves 107 4.2 Minimal post-irradiation fuel Ccmplete 6/29/84 surveillance program M P84 80/12 4-gs Pago 4 of 7 L
ATTAOiMENT 1 (Cont'd)
DSER R. L. MITIL 'ID CPEN SECTION A. SOiWENCER ITEM NUMBER SUEk7ECT STATUS !EITER IATED 108 4.2 Gadolina thermal conductivity Cmplete 6/29/84 equation 110b 4.6 runctional design of reactivity Cceplete 6/1/84 control systems lila 5.2.4.3 Preservice inspection grogram Cmplete 6/29/84 !
(ccyponents within reactcr tressure l boundary) 111b 5.2.4.3 Preservice inspection grogram Cmplete 6/29/84 (cmponents within reactor pressure boundary) )
111c 5.2.4.3 Preservice inspection grogram Cmplete 6/29/84 (ccmponents within reactcr tressure '
boundary) 119 6.2 'IMI item II.E.4.1 Cm plete 6/29/84 !
123 6.2.1.4 Butterfly valve cperation (post Cmpleto 6/29/84 accident) 124 6.2.1.5.1 RW shield annulus analysis Cmplete 6/1/84 125 6.2.1.5.2 Desip drywell head dif ferential Cmplete 6/15/84 i gressure 1 29 6.2.2 Insulation ingestion Cmplete 6/1/84 130 6.2.3 Potential bypass leakage paths Cmplete 6/29/84 132 6.2.4 Containment isolation review Cmplete 6/15/84 1 34 6.2.6 Contairinent leakage testirg Cmplete 6/15/84 1 38 6.6 Preservice inspection grogram for Cmplete 6/29/84 Class 2 and 3 ccmponents 139 6.7 MSIV leakap control system Cmplete 6/29/84 141C 9.1.3 Spent fuel pool cooling and cleanup Cmplete 6/29/84 systm 141g 9.1.3 Spent fuel pool cooling and cleanup Cmplete 6/15/84 syr: tem 142a 9.1.4 Light load handliry systen (related Closed 6/29/84 I to refuelirg) (5/30/84-Aux.Sys.Mtg.)
M P84 80/12 5-ga Page 5 of 7
ATTACtMDFT 1 (Cont'd)
DSER R. L. MITTL 10 OPEN SECTION A. SCHWENCER ITEM NUPSER SUf0FLT STATUS LETTER IRTED 142b 9.1.4 Light load handling system (related Closed 6/29/84
! to refueling) (5/30/84-Aux.Sys.Mtg.)
145 9.2.2 ISI program and functional testing Closed 6/15/84 of safety and turbino auxiliaries (5/30/84-coolng systems Aux.sys.Mtg.)
146 9.2.6 Switches and wiring associated with closed 6/15/84 HPCI/RCIC torus suction (5/30/84-Aux.Sys.Mtg.)
152 9.4.4 Radioactivity monitoring elements Closed 6/1/84 (5/30/84-Aux.Sys.Mtg.)
i 154 9.5.1.4.a Metal roof deck construction Ccaplete 6/1/84 l
classificiation 158 9.5.1.5.a Class B fire detection system Cmplete 6/15/84 159 9.5.1.5.a Primary and secondary power supplies Cmplete 6/1/84 fc.r fire detection system 161 9.5.1.5.b Fire water valve supervision Ccmplete 6/1/84 l
162 9.5.1.5.c Delugo valves ccmpleto 6/1/84
- 163 9.5.1.5.c Manual hose station pipe sizing Ccmplete 6/1/84 l 164 9.5.1.6.e Demote shutdown panel ventilation Ccmplete 6/1/84 165 9.5.1.6.g f>ergency diesel generator day tank Ccapleto 6/1/84 protecton 168- 12.5.2 Equipnent, training, and procedures Ccmplete 6/29/84 for inplant iodine instrunentation l 170 13.5.2 Procedures generation package Ccuplete 6/29/84 sutsnittal 171 13.5.2 'INI Item I.C.1 Crmplete 6/29/84 172 13.5.2 PGP Ccmenitment Ccanplete 6/29/84 173 13.5.2 Procedures covering abnormal releases Cmplete 6/29/84 of radioactivity l
M P64 80/12 6-gs Page 6 of 7 l
1 l ATTACIM NT 1 (Cont'd) l DSER R. L. MITIL 10 ;
I OPDi SECTION A. SCHWENCER [
ITEM NUMBER SUR7ECT STAltJS !EITER [*TED l l l 174 13.5.2 Resolution explanation in PEAR of Cmplete 6/15/84 l 1MI Items I.C.7 and I.C.8 l l
l 181 15.9.5 1MI-2 Item II.K.3.3 cmplete 6/29/84 i
- 182 15.9.10 1NI-2 Item II.K.3.18 Ccuplete 6/1/84 1
! 185 7.2.2.2 Trip system sensors and cabling in Cceplete 6/1/84 c turbine building ;
190 7.2.2.7 Regulatory Guide 1.75 Cmplete 6/1/84 191 7.2.2.8 Scram discharge volune Co plete 6/29/84 l 193 7.2.2.9 Reactor imde switch Cmplete 6/1/84 [
194 7.3.2.2 Standard review plan deviations Cmplete 6/h84
- 197 7.3.2.5 Micrcprocessor, multiplexer and Cmplete 6/1/84 I cmputer systems t 200 7.4.2.2 Remote shutdown system Ccaplete 6/1/84 205 7.5.2.4 Plant Focess ccuputer system Ccmplete 6/1/84 i
209 7.7.2.3 Credit fcr non-safety related systems Ccaplets 6/1/84 in Chapter 1% of the f3AR 210 7.7.2.4 1Yansient analysis recording system C m plete 6/1/84 218 9.5.1.1 Fire hazards analysis Cmplete 6/1/84 T9-3 4.4.5 Core flow monitoring for crud offacts ccupleto 6/1/84 tr-1 4.2 ruel rod internal gessure criteria Cmplete 6/1/84 i
JSigs l M Ps4 80/121-7 gs Page 7 of 7
- ATTACHMENT 2 DATE: 6/29/04 DRAFT SER SECTIONS AND DATES PROVIDED SECTION DATE SECTION DATE
! 3.1 l'
3.2.1 11.4.1 3.2.2 11.4.2 1 5.1 11.5.1
! 5.2.1 11.5.2 6.5.1 13.1.1 1 8.1 13.1.2 8.2.1 13.2.1 8.2.2 13.2.2 8.2.3 13.3.1 8.2.4 13.3.2 8.3.1 13.3.3 4 8.3.2 13.3.4
! 8.4.1 13.4 j 8.4.2 13.5.1 j 8.4.3 15.2.3 8.4.5 15.2.4 8.4.6 15.2.5 8.4.7 15.2.6
- 8.4.8 15.2.7
- 9.5.2 15.2.8 i 9.5.3 15.7.3 i- 9.5.7 17.1 i 9.5.8 17.2 10.1 17.3 10.2 17.4 10.2.3 10.3.2 10.4.1 10.4.2 10.4.3 10.4.4 11.1.1 11.1.2 i 11.2.1 11.2.2
, 11.3.1
- 11.3.2 i
l 3
8 l CTidb i
l MP 84 95/03 01 I
1 i
d ATTACHMENT 3 DSER OPEN SECTION ITEM NUMBER SUBJECT 35 3.6.2 ISI program for pipe welds in break exclusion zone 36 3.6.2 Postulated pipe ruptures 88 3.9.1 Stress analysis and elastic-plastic analy sis 89 3.9.2.1 Vibration levels for NSSS piping systems 91 3.9.2.2 Piping supports and anchors 93 3.9.3.1 Load combinations and allowable stress limits 94 3.9.3.2 Design of SRV's and SRV discharge piping 97 3.9.3.3 Buckling criteria used for component supports 100s 3.9.6 10CFR50.55a paragraph (g) 102 3.9.6 Leak testing of pressure isolation valves 107 4.2 Minimal post-irradiation fuel surveillance program 108 4.2 Gadolina thermal conductivity equation 111A 5.2.4.3 Preservice inspection program (components within reactor pressure boundary) 111H 5.2.4.3 Preservice inspection program (components within reactor pressure boundary)
MP84 80/13 3-db i . _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - - _ . _ _ _ _ _ _ _ _ _ _ - _ - _
t
- DSER ,
OPEN SECTION !
ITEM NUMBER SURJECT ;
111C 5.2.4.3 Preservice inspection program l (components within reactor pressure l l boundary) [
119 6.2 TMI item II.E.4.1 k i
123 6.2.1.4 Butterfly valve operation (post !
accident) t 130 6.2.3 Potential bypass leakage paths f 138 6.6 Preservice inspection program for !
class 2 and 3 components l l
139 6.7 MSIV leakage conttol system 141C 9.1.3 Spent fuel pool cooling and cleanup .
7 system -
14 2a 9.1.4 Light load handling system (related to refueling) 142b 9.1.4 Light load handling system (related
- to refueling) ;
t 168 12.5.2 Equipment, training, and procedures for inplant iodine instrumentation j 170 13.5.2 Procedures generation package l submitttl i r
171 13.5.2 TMI item I.C.1 172 13.5.2 PGP Commitment l t
173 13.5.2 Procedures covering abnormal releases ;
of radioactivity 181 15.9.5 TM1-2 item !!.K.3.3 !
t 191 7.2.2.8 Scram discharge volume i
i i
a MP84 80/13 4-db !
t I
+-_____ _ _ _ _ _ _ _ _ _ _ . -
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ATTAC&iENT 4 I
O e
f
r HCGS DSER Open Item No. 35 (Section 3.6.2)
ISI PROGRAM FOR PIPE WELDS IN BREAK EXCLUSION ZONE Assurance is needed regarding the augmented inservice inspection program for pipe welds in the break exclusion Zone.
RESPONSE
For the information requested above, see the response to Question 210.14.
\
- g. M . P84 9 5/13 4-d h w .
4
i HCGS DSER Open Item No. 36 (Section 3.6.2)
POSTULATED PIPE RUPTURES Additional information is required in several areas dealing with postulated pipe ruptures. Several tables and figures dealing with postulated rupture locations and their associated effects are incomplete. More information on the details of jet impingement and pipe whip analyses is required.
RESPONSE
For the information requested above, see the response to Question 210.21.
MP84-93 09 l-vw L-
HCGS !
DSER Open Item No. 88 (Section 3.9.1) .
L 4
STRESS ANALYSIS AND ELASTIC-PLASTIC ANALYSIS ,
Clarification is needed on experimental stress analysis and )
elastic-plastic analyses. l
RESPONSE
For the information above, see the response to Questions 210.26 and 210.27.
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MP84 93 09 2-vw i
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, -, , . . , . . . _ . . - , , - . . _ . , . _ . . , _ . ---m - - , . _ . . _ _ - . - - - _ - _ _ _ . _ _ . - - . . . _ , , . - _ _ . _ _ _ .-,--..m ..._-
i HCGS DSER Open Item No. 89 (Section 3.9.2.1)
VIBRATION LEVELS FOR NSSS PIPING SYSTEMS Additional information of the criteria to be used for determining acceptability of observed or measured vibration ,
levels for NSSS piping systems needs to be included in the !
FSAR.
i
RESPONSE
i For the information above, see the responses to Questions 210.29 and 210.30. ;
I i
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1 MP84 93 09 3-vw L
HCGS DSER Open Item No. 91 (Section 3.9.2.2)
PIPING SUPPORTS AND ANCHORS Additional information is required or the design of piping supports and anchors which separate seismically designed piping and nonseismic Category I piping.
RESPONSE
For the inf ormation t equested above, see the response to Question 210.34.
MP84 93 09 4-vw
i HCGS DSER Open Item No. 93 (Section 3.9.3.1) j LOAD COMBINATIONS AND ALLOWABLE STRESS LIMITS More information is required on loading combinations, system operating transients and stress limits for each of the following for all classes of construction: vessels, pumps, valves, piping, and supports. Assurance must be provided for the functional capability of ASME Class 1, 2, and 3 piping systems. This is an open item.
RESPONSE
Non-NSSS:
The loading combinations for non-NSSS piping and supports are given in Tables 3.9-8 and 3.9-21. Table 3.9-21 has been i revised to include allowable stresses under various plant conditions. In addition, the primary stress limits have been added to Tables 3.9-9 and 3.9-13.
Tables 3.9-10 and 3.9-15 have been revised to include the loading combinations and allowable stress limits for non-NSSS Class 1 valves and Class 2 and 3 valves respectively.
Informaticn on loading combinations, system operating trans-ients, and stress limits for pumps were provided in response to Question 210.52. Information on safety-related vessels designed to the ASME Code are attached in Tables 93-1 and 93-2.
Functional capability of ASME Class 1, 2, and 3 piping sys-tem has been addressed in Question 210.39.
NSSS:
Information on loading combinations, system operating trans-ients, and stress limits for pumps and vessels were provided :
in response to Question 210.52.
M P84 112/02 4-srd
HCGS 1
TABLE 93-1 1
SAFETY-RELATED VESSELS The safety-related hydropneumatic accumulators (STACS) are demonstrated capable of withstanding the following loading conditions and associated loading combinations while stresses remain below the allowable stresses. The design, manufacturer, examination, testing, and inspection of the accumulators is in accordance with the ASME Boiler and Pres-sure Vessel Code,Section III Nuclear Power Plant Compo-nents, Division I for Class 3 components.
PLANT / SYSTEM DESIGN AND I OPERATING SERVICE LOADING ALLOWABLE SERVICE CONDITION LIMITS COMBINATION STRESS LIMIT Design -
PD+OBE ND-3300 Normal A PO+DW+EL ND-3300 Upset B PO+DW+EL+0BE ND-3300 Faulted B PO+DW+EL+SSE ND-3300 Where:
PD = Design Pressure PO = Operating Pressure
. DW = Dead Weight of vessel and contents EL = External Loads due to connected piping i
OBE = Operating Basis Earthquake l
SSE = Safe Shutdown Earthquake DSER OPEN ITEM h l
l M P84 112/03 1-srd
l TABLE 93-2 l l
SAFETY RELATED VESSELS i l
The Safety Related Expansion Tanks and Air Accumulators are demonstrated capable of withstanding the following loading conditions and associated loading combinations while stresses j remain below the allowable stresses. These vessels include the SACS expansion tanks, control area chilled water system {
head tanks.
LOADING LOADING ALLOWABLE l CONDITION COMBINATION STRESSES j Design PD ND-3300 for > 15 psig ND-3800 for ATM ND-3900 for 0-15 psig ;
No rmal PO + DW + EL ND-3300 for > 15 psig -
ND-3800 for A'fM !
ND-3900 for 0-15 psig j Upset PO + DW + EL + OBE Code Case 1607-1 for !
> 15 psig Code case 1657 for :
ATM, 0-15 psig Faulted PO + DW + EL + SSE Code Case 1607-1 for I
> 15 psig !
Code Case 1657 for where: A TM , 0-15 psig ,
I PD = Design Pressure [
PO = Operating Pressure }
DW = Dead Weight of vessel and contents EL = External Loads due to connected piping [
OBE = Operating Basis Earthquake -
l SSE = Sa fe Shutdown Earthquake ,
l :
DSER OPEN ITEM Q.3 l t
l
HCGS FSAR TABLE 3.9-21 Page 1 of 2 DESIGN LOADING COMBINATIONS FOR SUPPORTS FOR ASME B&PV CODE CLASS 1, 2 AND 3 NON-NSSS COMPONENTS A / lou)*Ne Condition Desian Loadino Combinations <a)(a)(a) stress Hydrostatic test (a) HTDW o.F Sy Normal and upset (a) DW + TH + (OBEa + RVCa)sia gggg 3 ,.f,',,2g (b) DW + TH + OBE + RVO (c) DW + TH + FV Aspc scenow.nt, Emergency (a) DW + TH + (OBEa + yya) ia j ff ,, j,, ygy Faulted (a) DW + TH + SSE + RVO A f g e 3 cy,o o 22r, (b) DW + TH + (SSE + RVC2) ia W" #
(c) DW + TH + (SSEa + DBAa)aiz (a) Loads due to OBE, SSE, and DBA include both inertia portion and anchor movement portion when spectra method is used.
i The loads from the inertia portion and anchor movement portion are combined by the SRSS method.
(a) For torus-attached piping, the loading combinations used in evaluating the pipe support loads are those given in the Plant Unique Analysis Application Guide (PUAAG)
. (NEDO-24583-1, October 1979 (Table 5-2 ) ) . -
(3) Definition of symbols used:
HTDW - piping dead weight due to hydrostatic test TH -
reaction at the support due to thermal expansion of the pipe
'DW -
dead weight
- 3) ;
', OBE - operating basis earthquake (s) f E i h RVC -
transient response of the piping system associated l
, with relief valve opesing in a closed system !
E o RVO - transient response of the piping system associated g with relief valve opening in an open system ,
E I
-, , . , . - - , . ~ . - - . - - . - .
- - - - , - - - - - . - - ~ , , . . - _ . . - , . . . - , - = --,__---,_,.,--,,-,-,...-,,--n-,---- . - - - - - - , - - - - , - - , . .
t HCGS FSAR ABLE 3.9-21 (cont) Page 2 of,2 FV - transient response of the piping system associated !
with fast valve closure time less than 5 seconds !
SSE -
DBA - design basis accident (8) c4) he e ss en hial .sa fely - t e/a.f ed c E sp) a ys te m.s a//owa.ble r4eess not to e2c.eed Sy .
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L I
DSER OPEN ITEM 93
HCCS FSAR TABLE 3.9-9 DESIGN CRITERIA FOR ASME B&PV CODE CLASS 1 NON-NSSS PIPING
/ Applicable Code R,rograph pe, m a., y stre s r Condition [>Stcess Li;rit ><a) um ,y 3 Design NB-3221 and NB-3652 /. 5"Sm Normal NB-3222 and NB-3653 /.I Sm Upset NB-3223 and NB-3654
/.8Sm sut not.3k' O m ,,f.f sy J.2fSp b d AO & .9 P* *T~
Emergency NB-3224 and NB-3655 g ,,,j,ysy Faulted NB-3225 and NB-3656 3.o Sm (1) As specified by the ASME B&PV Code,Section III, 1974 through Winter 1974 Addenda, except for the following:
- a. Class 1, 1-inch and smaller piping are designed to ASME Section III, 1975 Summer Addenda, Paragraph NB-3630(d)(I).
- b. Class 1 flanges are designed to ASME Section III, 1979 !
Summer Addenda, Paragraph NB-3658. !
i l c. Class 1 branch connections are designed to ASME Section III, 1979 Summer Addenda, Paragraph NB-3653.1.
(a) Functional capability of essential piping is ensured per l NEDO-21985, September 1978.
I
{
l i
DSER OPEN ITEM k3 l !
(
l TABLE 3.9-13 !
DESIGN CRITERIA FOR ASME B&PV CODE CLASS 2 AND 3 NON-NSSS PIPING !
Condidon / / Strels-Limits (*)h) M ;
Deign, norma)(upsetand T piping con orms to th i ergency equirements f Section I, !
paragraphs -3600 '
Faulted ) The pipi conforms o the !
require ents of AS Code !
Case 1s06 '
D SCRi ~ I / I e ^g ,
(a) As specified by the ASME B&PV Code,Section III, 1974 ddC #'
through Winter 1974 Addenda, except for Class 2 and 3 flanges, which are designed to 1979 Winter Addenda, Paragraph NC and ND-3652.
(2) Functional capability of essential piping is ensured per NEDO-21985, September 1978.
O l
l DSER OPEN ITEM f3 i
. _ _ _ _ . _ _ . _ _ . _ _ _ _ _.. - - _ _ _ . _ _ _ _ _ _ _ _ . _ . - ~ _ _ _ _ . . ._ __ _ .._ _._.
HCGS i
Insert ,
TABLE 3.9-13 APPLICABLE CODE PRIMARY STRESS L I
CONDITION PARAGRAPH (1) (2) LIMITS Design:
Sustained Loads NC, ND-3652.1 1.0Sh .
Occasional Loads NC, ND-3652.2 1.2Sh [
Normal and Upset NC, ND-3652.2 & 1.2Sh 3611 Emergency NC, ND-3611 1.8Sh Faulted Code Case 1606 2.4Sh b
DSER OPEN ITEM f3 ;
i M P84 112/03 3-srd
HCGS FSAR TABLE 3.9-10 DESIGN CRITERIA FOR ASME B&PV CODE CLASS 1 NON-NSSS VALVES !
rNSER T' A nditio[ / Efress Limit /
Des NB-3521(8 ormal and pset NB-32 or NB-350 *)
l (St dard Design ules)
Emer ncy(a) -3526 ,
l aultedca) NB-3527 (1) As specified by the ASME B&PV Code,Section III, 1974 clel ed d. :
through Winter 1974 Addenda.
(a) Where valve function must be ensured (active valves) during emergency or faulted conditions, the specified emergency or faulted conditions for the plant M considered the normal l condition for the valve. L.a r e,
- 3) As required by subsection NB of ASME Section III [
other loads such as thermal transient and thermal gradients may require additional consideration in -
addition to those primary stress-producing loads listed.
- 4) Definition of symbols used:
PD - Design pressure ;
PO - Operating pressure at noted plant condition OBE - Operating Basis Earthquake loads (inertia F) portion) excluding loads from attached piping as SSE - Safe Shutdown Earthquake loads (inertia f
H portion) excluding loads from attached piping g B - Piping end loads at noted plant condition o
5 8
- _ - _ - - . _ _ _ . ~ _ - , _ . , . .
c INSERT.A T3.9-10 Plant Condition Design Loading Stress Limits (1)
Combinations (4)
Design PD The valve shall conform to the
requirement of Normal P0n + Bn Paragraph NB-3500 (Standard Design Rules)
Upset (3) P0u + OBE + Bu NB 3525 Emergency (2) P0e + Be NB 3526 Faulted (2) P0 f + SSE + Bf NB 3527 i
MP 84 112/05 1-mr DSER OPEN ITEM 93 L..
I HCGS FSAR TABLE 3.9-15 DESIGN CRITERIA FOR ASME B&PV CODE CLASS 2 AND 3 NON-NSSS VALVES
~
/ / / / /
Condition tress Li (*) ;
1 Design nd normal The va e conforms the ;
requ ements of Se ion III, 6 Par graphs NC-350 and ND-3500 pset, eme ency, and aultedC2) e valve conf rms to the ;
requirements f ASME Code ase !
1635-1 ;
1 l
(a) As spe Ified by the ME B&PV Code Section III, 974 ,
thro h Winter 197- Addenda. :
cm) ere valve fun ion must be e sured (active valves) uring emerge y or faulted nditions, th specified emergency or aulted plant onditions ar considered as e normal con tion for the Ive.
/ / / / l L d e t e+ e- l i
g G /2elb -
pages iandA i
DSER OPEN ITEM C3 f
HCGS TABLE 3.9-15 1 of 2 DESIGN CRITERIA FOR ASME B&PV CODE CLASS 2&3 NON-NSSS VALVES Plant Condition Design Loading Stress Limits Combination (1) (4) (1) (2) (3)
Design PD The valve shall conform to the requirements of Section III, 1974 Para-graphs NC-3500 or ND-3500,as applicable Normal POn + Bn Sm < 1.0S (Sm or SL) + Sb i 1.50s Upset POu + OBE + Bu Sm i 1.lS (Sm or SL) + Sb i 1.65S Emergency POe + Be Sm < l.5S (Sm or SL) + Sb i 1.8S Faulted POf + SSE + Bf Sm < 2.0S (Sm or SL) + Sb i 2.4S (1) Definition of symbols:
Sm = General membrane stress SL = Local membrane stress Sb = Bending stress S = Allowable stress DSER OPEN ITD( 93 M P84 112/03 4-srd
HCGS 1
TABLE 3.9-15 2 of 2 t
(1) Definition of symbols (cont'd):
PD = Design pressure PO = Operating pressure at noted plant condition OBE = Operating basis earthquake loads (inertia por-tion) excluding loads from attached piping i SSE = Safe shutdown earthquake loads (inertia portion) excluding loads from attached piping B = Piping end loads.at noted plant condition (2) As specified by the ASME B&PV Code,Section III, 1974, through Winter 1974 Addenda.
(3) Where valve function must be ensured (active v '.ves) during emergency or faulted conditions, the specified emergency or faulted plant conditions are considered as the normal condition for the valve.
(4) As required by subsection NC, ND of ASME Section III, other loads such as thermal transient and thermal gradients may require additional consideration in addi-tion to those primary stress producing loads listed.
DSER OPEN ITEM Gj i
! M P84 112/03 5-srd
HCGS DSER Open Item No. 94 (Section 3.9.3.2)
DESIGN OF SRVs AND SRV DISCHARGE PIPING The staff has reviewed section 3.9.3.3 of the applicant's FSAR with respect to the design and installation, and test-ing criteria applicable to the mounting of pressure relief devices used for the overpressure protection of ASME Class 1, 2, and 3 components. This review, conducted in accord-ance with SRP Section 3.9.3 (NUREG-0800), includes evalua-tion of the applicable loading combinations and stress criteria. The design review extends to consideration of the means provided to accommodate the rapidly applied reaction force when a safety valve or relief valve opens, and the transient fluid-induced loads applied to the piping down-stream of a safety or relief valve in a closed discharge piping system.
The staff requires additional information on the design of safety and relief valves (SRVs) and the main steam SRV discharge piping.
RESPONSE
Information on the main steam SRV discharge piping is provided in response to Question 210.45.
4 M P84 95/12 1-dh
HCGS DSER Open Item No. 9j7 (Section 3.9.3.3)
BUCKLING CRITERIA USED FOR COMPONENT SUPPORTS, The staff's review of Section 3.9.3.4 of the applicant's FSAR relates to the methodology used by the applicant in the design of ASME Class 1, 2, and 3 component supports. The review includes assessment of design and structural integrity of the supports. The review addresses three types of supports: plate and shell, linear, and component standard types. More information regarding the design and construction of ASME Class 1, 2, and 3 component supports is required.
The applicant should provide the buckling criteria used for component supports.
RESPONSE
For the information requested above see response to Question 210.49.
M P84 95/12 2-dh i
[
i HCGS DSER Open Item No. 100a (Section 3.9.6)
10CFR50.55a PARAGRAPH (g)
The applicant must provide a commitment that the inservice :
testing of ASME Class 1, 2, and 3 components will be in !
accordance with the rules of 10CFR50.55a, Paragraph (g) . .
RESPONSE
For the information requested above see FSAR Sections 5.2.4 and 6.6.
I I
e i
M P84 95/12 3-dh I
y - - , , _ - - - - - - - - - , .e,.- -v -.-- --- --- . , -.. - - - , , , - - , - , .-- ,,- ~n,_,,,----,,-,,
HCGS DSER Open Item No. 102 (Section 3.9.6)
LEAK TESTING OF PRESSURE ISOLATION VALVES The applicant has not yet responded to the staff's concern regarding the leak testing of pressure isolation valves.
l
RESPONSE
t For the information requested above, see the response to I Question 210.56. l
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t HOPE CREEK DSER OPEN ITEM RESPONSE DSER OPEN ITEM 107 (Section 4.2)
MINIMAL POST-IRRADIATION FUEL SURVEILLANCE PROGRAM r The applicant must provide a minimal post-irradiation fuel surveillance program consistent with SRP Section 4. 2.II .D. 3. ;
RESPONSE
General Electric and the NRC have negotiated a post-irradia-tion fuel surveillance program that meets the requirements i of SRP 4.2. The NRC has been requiring that certain appli- l cants (Perry, Hanford and Limerick) commit to perform visual l inspections of a prescribed percentage of discharged bundles i each cycle. In a letter dated November 23, 1983, General Electric proposed an alternative program that would transfer much of the burden for these inspections from utilities to General Electric. In a letter dated January 18, 1984, the NRC staff described what would be an acceptable program and requested additional detail from General Electric. In letters dated January 27, 1984, and February 29, 1984, General Electric addressed the NRC questions, and the NRC ,
has verbally agreed to the General Electric program. Public Service Electric and Gas endorses this program for the Hope
- Creek plant.
IB 27 01-bp l I
i MFN-218-83 ,
JSC-072-83 f GENERAL $ ELECTRIC l
NUCLEAR POWER SYSTEMS OM90N l GENERAL ElfCTWC COMPANY
- 175 CURTNER AWNUE e SAN JOSE. CAUFORNLA 95195 !
MC 682, (408) 925-3697 l November 23, 1983 U. S. Nuclear Regulatory Commission 3 Office of Nuclear Reactor Regulation l Washington, D.C. 20555 l
Attention: C. H. Berlinger, Chief Core Performance Branch ,
ii i j, Gentlemen:
SUBJECT:
POST-IRRADIATION FUEL SURVEILLANCE PROGRAM f
..' t
Reference:
Letter, J. S. Charnley (GE) to F. J. Miraglia (NRC), j
" Proposed Revision to GE Licensing Topical Report j NEDE-24011-P-A", February 25, 1983 ,
l The NRC has recently required that newly Itcensed plants adopt a post-frradiation fuel surveillance program that consists essentially of routine visual inspection of discharged fuel at each refueling outage, j The purpose of this letter is to propose the use of the fuel survelliance ;
program described in the attachment, in place of the program required by [
the NRC at newly licensed plants. General Electric believes that its program meets the intent of Section II, Part D, of Standard Review Plan (SRP) 4.2 (NUREG-0800), regarding fuel surveillance. Because of the number of plants coming on-line in the near future that will be affected g l by this issue, GE requests that the NRC expedite consideration of this , i matter. l General Electric Fuel Performance Verification Program l The General Electric fuel performance verification program is described !
in the proprietary attachment to this letter. The attachment is considered .
l proprietary because it contains information which GE customarily maintains ;
l in confidence an( withholds from public disclosure. This information has !
been handled and classified as proprietary as indicated in the affidavit :
provided in the reference letter. We hereby request that this information j be withheld from public disclosure in accordance with the provisions of (
10CFR2.790. l i
DSER OPEN ITEM /d 7 I
.. - - - - . .-. - - - - .- - - - .._ - - _ . . - .-__0
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USNRC GENERAL $ ELECTRIC Page 2 GE Proaram and SRP 4.2 :
Regarding post-irradiation fuel surveillance, SRP 4.2 states that a [
program "should be described for each plant to detect anomalies or ;
confirm expected fuel performance...For a fuel design like that in other j
. operating plants, a minimum acceptable program should include a qualite- ;
tive visual examination of some discharged fuel assemblies from each l refueling."
GE defines expected fuel performance as "the fuel will not fail". !
Failure criteria used in the design process contain conservatisms that i adequately bound conditions that may exist at any plant, and provide , ,
margin to actual fuel failure limits. Additionally, operating limits are !
established such that sufficient margins are maintained to the design l limits during normal operation and transients (in accident analyses, all ;
fuel is conservatively assumed to fail). ;
Expected fuel performance as defined above is confirmed on a generic l basis for s fuel design through the inspection of LTA's, and on a plant- l specific basis through offgas surveillance. The LTA program detects i
! anomalies that may arise, with the added advantage of accomplishing this (
prior to the time that the anomaly might appear in production fuel. As j discussed earlier, a visual examination of some of the discharged fuel i from two early applications of a new fuel design will also be performed, in order to confirm the expected performance of that fuel design. j Discussion of GE Program GE believes that the program it proposes meets or exceeds the intent of l SRP 4.2 and is also more cost effective. The numerous benefits of the GE j program are presented below. ;
l Inspection of LTA's of new designs provides timely, detailed, and useful o information that can be fed back into fuel design, analysis, and manuf acture. ,
LTA's of new designs are usually placed in operation at least a year ' i before in-reactor introduction of production fuel. Prior to irradiation, !
these LTA's may undergo detailed visual, nondestructive,.and dimensional !
characterization. Key measurements may be taken of specific bundle !
features and additional detailed examinations may be performed on specific l fuel rods. Interim examinations may be performed at the end of each !
operating cycle. Upon discharge, a final inspection is performed on the previously characterized fuel rods and final measurements may be taken of key bundle features. As required, more extensive evaluations may be !
performed, including destructive testing. This detailed surveillance of ;
LTA's for new designs provides: (1) early identification of potential !
fuel performance concerns; (2) continuous knowledge of overall fuel .
l 1
DSER OPEN ITEM /C /
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_ . . . _ . _ _ _ . _ _ - _ _ _ _ _ , . _ _ . _ . - ~ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ ~ _ _ _ . _ _ _ _
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, USNRC Page 3 j l
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performance; and (3) systematic acquisition of detailed behavioral data ;
allowing a comparison of predicted versus observed performance character- 1 istics, thus providing feedback into the design process from fuel cperated )
in a commercial reactor. i
- 1 As discussed previously, the detection of fuel failures results in an l investigation into the cause, and corrective actions where appropriate.
l A general visual inspection of the exterior surfaces of a statistically significant number of fuel bundles (24 total - twelve at each of two plants) to confirm the absence of any anomalous behavior at end-of-life discharge for a new fuel design represents ample additional confirmation (
of the design. !
l Secause fewer bundles are examined in greater depth (LTAs) than in the ;
program required by the NRC of newly licensed plants, and because the visual inspections are limited to 24 bundles at end-of-life for a new design rather than at the end of every cycle in perpetuity, the GE !
program leads to a significant reduction in the total costs to utilities, (
while sis.ultaneously providing more valuable data. If a utility were to l contract for the type of visual examination the NRC is proposing the cost l to the utility would be on the order of $60,000 per reload (assuming 12 j bundles are inspected at each outage), in addition to personnel and i
! dechanneling costs. If the utility were to perform the visual inspection :
l itself, the cost in terms of training personnel, procuring proper equipment, !
performing the inspection, and exposing we ?kers to radiation, would also l be substantial. !
The proposed GE program will allow the NRC to maximize the utilization of
- its resources by eliminating routine, repetitive review. Legitimate L
, , concerns will be easily recognized under the program proposed by GE.
. Summary 4
GE proposes a fuel performance verification program consisting of inspec- !
tion of LTA's, offgas surveillance and visual examination of a limited i but statistically significant number of fuel bundles of two early commercial app 1tcations of new fuel designs. GE believes that this program meets or ;
exceeds the intent of SRP 4.2 regarding fuel surveillance, and in addition i is cost-effective for GE and the utilities as well as the NRC, while i providing timely, detailed, and useful information that will be of i j benefit in enhancing fuel performance. l I
DSER OPEN ITEM /d[
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USnRC Page 4 0000 h IMUIC Please contact W. A. Zarbis (408-925-5070) or myself if you have any questions.
, Very truly yours.
d'%= s J.M. Charnley y
Fuel Licensing Manager Nuclear Safety and Licensing Operation JSC:csc/109091*
cc: L. S. Gifford (GE-Beth)
L. 5. Rubenstein (NRC)
G. G. Sherwood (GE)
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DSER OPEN ITEM /d 7
I ATTACHMENT [
t General Electric Fuel Performance Verification Program j The General Electric fuel performance verification program consists of .
inspection of lead test assemblies (LTA's) for new designs, and offgas :
monitoring of all designs throughout their lifetime. f For new fuel designs, GE will, in addition, agree to conduct a general visual examination of the exterior surfaces of a statistically significant .
number of fuel bundles (12 bundles) upon discharge from each of two early i commercial applications of the new product. The visual examination will j be made using binoculars, borescope, periscope, or TV and will be suffi-cient to meet the objectives presented in SRP 4.2 for visual inspections. (
- The schedule and scope of LTA inspections is contingent on both the ,
availability of the fuel as influenced by plant operation and the expected i value of the information to be obtained. i
' General Electric's LTA's are selectively inspected using one or more of l the following techniques: ;
- 1) Leak detection tests, such as sipping.
- 2) Visual inspection with various aids such as binoculars, ;
- borescope, or periscope, with a photographic record of ;
observations as appropriate. j
- 3) Nondestructive testing of selected fuel rods by ultrasonic and ;
t eddy current test techniques.
- 4) Dimensional measurements of selected fuel rods. l h
Unexpected conditions or abnormalities which may arise are analyzed, and examination of selected fuel rods in hot cell facilities may be undertaken f when the expected value of the information to be obtained warrants this ;
type of examination. Results of this surveillance program will be j updated annually by a GE proprietary letter report. ;
The use of LTA's provides early verification of performance targets as ;
well as early indication of potential performance anomalies. j Specific plant fuel failures are accurately detected by offgas surveil I f lance. Offgas surveillance is performed for all operating plants, and l 1eak detection tests such as sipping are performed by the utilities at !
the end of each cycle, if warranted (based on analysis of the offgas .
surveillance results). Offgas surveillance is a very sensitive measure l of fuel performance, and General Electric fuel failure statistics include l N fuel failures estimated as a result of offgas measurements. These fuel l failure statistics will be updated in the annual letter report. l Q l g - If many fuel failures are detected, an analysis or investigation is ;
e initiated to determine the cause of the failures. In addition to review !
[g of operational parameters such as power history and water chemistry and of GE's current overall fuel experience base, the investigation may j
l include site examinations, and when appropriate, searches of manufac-o g turing records, tests of manufactured spare rods if available, and hot ;
g cell examination of selected irradiated fuel rods.
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i WZ:csc/111184-1 i
I
[ UNITED STATES f NUCLEAR REGULATORY COMMISSION R E.C E I,V E D i U ,
ifASHINGTON,0. C. 20655 )
( **,***. JAN 3 01984 l JAN 1a as4 W 9 V- !
Mr. .R. L. Gridley, Fuel and Services !
Licensing Manager Nuclear Safety & Licensing Operation General Electric Company i 175 Curtner Avenue San Jose. California 95125
Dear Mr. Gridley:
Subject:
Post-Irradiation Fuel Surveillance l
Reference:
Letter from J. S. Charnley (GE) to C. H. Berlinger (NRC).
" Post-Irradiation Fuel Surveillance Program," November 23, 1983. >
Your letter of November 23, 1983 proposes that generic vendor surveillance on lead test assemblies (LTAs) be substituted for routine licensee surveillance to satisfy Section II, Part D of Standard Review Plan 4.2. !
Section II. Part D contains two subparts that are relevant to your request. Subpart 2 describes on-line monitoring and Subpart 3 describes post-irradiation surveillance, i In our view, the licensee offgas surveillance that is mentioned in your letter clearly satisfies Subpart 2 mentioned above. This offgas sur-veillance program has been proposed by all recent BWR operating license ,
applicants and in all cases we have approved it. '
Your letter of November 23, 1983 also= proposes that generic vendor surveillance on LTAs supported by a visual examination of some dis-l charged fuel from two early applications of the new fuel design be i substituted for routine licensee surveillance to satisfy Section II, Part D of Standard Review Plan Section 4.2.
While we agree with the goal of your letter, that is, to provide a 1 better balance between the reduction of the regulatory burden on indi- -
! vidual licensees and the NRC's interests in maintaining the present high level of fuel performance and in identifying potential new anomalies at an early stage, we find the General Electric Company proposal as de- '
scribed in your letter to be inadequate for several reasons.
! We believe that past experience has shown that a surveillance program l which looks only at LTAs and the first two core loadings is not suf- .
l ficient. Fuel problems have occurred with standard fuel designs that i
! have been in service for many years. Some of these problems were due to '
l specific one-of-a-kind problems but other problems have been more ge- !
neric in nature. We consider a small amount of visual surveillance to be important because we are concerned with all types of fuel damage DSER oPEN ITEM /d 7
\
l Mr. R. L. Sridley JAN 181984 f
I I
including that which could affect control rod insertability and accident I doses, as well as those mechanisms that would lead to detectable (i.e..
by offnas) cladding leaks during nomal operation. Many examples of l excess' ve wear, tearing of metal parts, fuel rod defomation and ex- l cessive crud buildup have been observed visually in fuel that showed no ;
evidence of cladding feilure under nomal operating conditions.
We consider these differences in fuel surveillance programs to be ,
reconcilable. General Electric topical report NEDE 24343-P " Experience 6
(
}
With Fuel Through January 1981" describes the GE fuel surveillance program. One aspect of tnis program is stated to be an overall post- l l
1rradiation visual examination of selected fuel bundles. We would i consider this present GE program to be equivalent to that described in the Standard Review Plan if General Electric would: (1) verify that !
this program includes post-irradiation visual inspection of standard !
design fuel bundles which have not been identified as leakers by sipping or other methods and (2) that the current GE fuel surveillance program l for standard fuel designs will continue at its present level of effort. :
i In addition, we have some specific questions on details of the sur-veillance program proposed in your November 23, 1983 letter. The pro- !
gram which is described makes almost all the comitments to the type of surveillance conditional. It is not clear in the letter what these conditions are. For example, the letter states that " prior to irradi- !
ation, these LTAs uma undergo detailed visual, nondestructive and di-mensional characterization." f We believe it is important to discuss and clarify under what circumstances i the conditions would be met so that these inspections would be done. l Also. it is not clear in your letter what threshold of offgas activity j would result in a non-routine inspection of the standard fuel designs. !
This should be discussed and clarified.
It continues to be our position that operating reactor licensees have the final responsibility for the perfomance of the fuel in their , I j
reactors. Although we agree in principle with the GE proposal to lessen the burden on these licensees, if a problem is discovered it is still i the responsibility of the licensees to assure that adequate steps are taken to assure safe operation of the fuel at their facilities. We will also attempt to assure that, should a licensee who is presently a GE customer choose not to continue that relationship, the licensee will subsequently adopt an acceptable fuel surveillance program. l The results of these fuel inspections performed at a licensee's facility
< or perfomed on fuel irradiated at a licensee's facility will be covered ,
j by the reporting requirements of Paragraph 50.73(a)(2)(11) concerning [
the degradation of " principal safety barriers." such as the fuel. [
l DSER OPEN ITEM /47 i
E_ _ . _ _ _ _ _ i
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Gridley JAN I C 1984 g j.
i -
Whilei ht able*ha discussion alternative to the SRP. is we necessary believe your proposal to detemine is a if your pl an acCP J SPProach. !
posith' Sincerely.
. e i m
.x. b3 - e Et ws
% [
4 L. S. Rubenstein. Assistant Director i.
j for Core and Plant Systems Division of Systems Integration. NRR pharnley r
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DSER OPEN ITEM /d 7
GENERAL $ ELECTRIC
'i
, NUCLEAR POWER SYSTEMS DMSiON GENERAL ELECTRIC COMPANY e 175 CURTNER AVENUE e SAN JOSE, C/aWORNIA 95195 68
, 4p8)925-3697 February 29, 1984 JSC-10-84 l i
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C. 20555 l l
Attention: L. S. Rubenstein, Assistant Director !
Core and Plant Systems ,
i Gentlemen:
SUBJECT:
FUEL SURVEILLANCE PROGRAM
References:
- 1) L. S. Rubenstein (NRC) to R. L. Gridley (GE),
" Post-Irradiation Fuel Surveillance," l, January 18, 1984 !
- 2) NEDE-24343-P, " Experience with BWR Fuel Through I January 1981," May 1981 i
- 3) J. S. Charnley (GE) to C. H. Berlinger (NRC), i
" Post-Irradiation Fuel Surveillance Program," :
November 23, 1984 ,.
I This letter provides additional details requested by the NRC on GE's fuel i surveillance program, and replaces our letter of January 27 on this [
subject.
l The fuel surveillance program presented in your letter of January 18 (Reference 1) assures adequate verification of safe fuel performance ;
while still maintaining efficient use of industry resources, and is '
acceptable to General Electric. We would like to take this opportunity {
to provide additional information in order to address the points raised :
- . in your letter. g i
, Reference 1 states that the fuel surveillance program described in i
NEDE-24343 (Reference 2) could be considered equivalent to that described in the Standard Review Plan if GE would: "(1) verify that this program i includes post-irradiation visual inspection'of: standard design fuel [
bundles which have not been identified as leakers by sipping or other t d methods, and (2) that the current GE fuel surveillance program for i standard fuel designs will continue at its present level of effort." !
, , The first item is specifically considered in the GE program. However, ,
g inspection of non-leakers is not performed on a routina basis but only in ;
H cases when information of special interest can be obtained. In these ;
g cases, a total visual examination is performed. For instance, if GE i g desired technical information on a particular subject such as end plug a:
N Q
GENERAL $ ELECTRIC !
Page 2 !
wear or model verification data, then the inspections described in the f first item would be performed. These inspections are performed at a i variety of plants and include plants in which no fuel problems are expected. ;
Regarding the second item above, the GE fuel surveillance program is 1
currently planned to continue at approximately its present level of j
, effort. The number and type of inspections will vary from year to year, of course, depending on offgas measurements and the degree of technical ,
interest as explained in the previous paragraph. l The next point raised in Reference 1 concerns the conditional aspect of i GE's lead test assembly (LTA) program described in Reference 3.
Detailed measurements of LTA's are not performed prior to irradiation in all ;
cases. When the LTA's represent significant design changes, though, such !
as the advanced LTA's in Browns Ferry 3 and Peach Bottom 3, detailed
[
measurements are performed prior to irradiation. In addition, detailed +
examinations are performed at the end of each operating cycle on specific i LTA's and upon discharge of most LTA's, depending on the subsequent j interest in implementing the design change demonstrated in the LTA. t i
The final point raised in Reference 1 addresses the threshold of offgas ;
activity that would result in non-routine inspection of standard fuel i designs. The offgas activity threshold would (a) vary from plant to '
plant, (b) be contingent on the amount of fuel failures predicted from i the increase in offgas, and (c) depend on whether the cause of the '
failures could be identified without performing an examination. Inspec- l tions would generally be performed if the number of failures predicted is ,
on the order of ten bundles, but this number could be more or less [
depending on the surrounding circumstances. For example, if offgas ,
activity approaches technical specification limits and a cause cannot be
. . assessed, fuel inspections could be performed even if the number of fuel , l bundles with failures is judged to be fewer than ten. On the other hand, '
if the cause is assessed - for instance, control blades were withdrawn at l
! power - an inspection would not be performed even if the number of fuel i bundle failures were greater than ten.
[
l We hope that this response provides the clarification required to arrive :
S at a mutually acceptable surveillance program.
f N Very truly yours, M
g . 5. Charnley, Fuel L sing Manager i g Nuclear Safety and Lic nsing Operation
$ J5C:jg/b01231 O y cc: L. 5. Gifford I G. G. Sherwood
DSER OPEN ITEM 108 (SECTION 4.2) 4 GADOLINA THERMAL CONDUCTIVITY EQUATION The gadolinia thermal conductivity equation used in the GESSAR-II fuel centerline melting analysis described in NEDE-24011 was not the same equation submitted and approved in Appendix B to NEDE-23785-1-P. GESSAR-II references NEDE-20943-P (which was withdrawn), which provided a different gadolinia thermal conductivity equation. This raises a concern about the adequacy of GE's gadolinia fuel incipient melting calculations for Hope Creek (in particular, Table 2-4 of NEDE-24011-P). The applicant should confirm the adequacy of Table 2-4 in NEDE-240ll-P or submit updated results for review.
RESPONSE
Discussions with the staff of the NRC Core Performance Branch led to agreement with General Electric that this issue is generic. At the NRC staf f 's request, General Electric recounted this agreement in a February 2, 1984, letter to L. S. Rubenstein. This issue has been resolved for the Perry and Hanford SERs based on prior information identical to that contained in the February 2 letter. The fuel design evaluation and the results described in this letter are also applicable to the Hope Creek fuel.
MP84 95 08 03-az
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GENERAL $ ELECTRIC I
- GENERAL ELECTDC COMPANY
- 175 CURTNER AVENUE e SAN JOSE, CALFORNIA 95195
! MC 682 (408)925-3697 JSC-04-84 l MFN-015-84 l February 2,1984 l
U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation '
l Washington, DC 20555 l Attention: L. 5. Rubenstein Assistant Director Core and Plant Systems Gentlemen: t
SUBJECT:
0VERHEATING OF GADOLINIA FUEL PELLETS f
This letter addresses the confimatory issue pertaining to the over- ;
heating of gadolinfa fuel pellets that has appeared in the Safety !
. Evaluation Reports of individual plants seeking operating licenses. .
,. The NRC has treated this issue as plant-specific and has resolved }
!, the issue on at least two plant dockets based on the information l l- presented in the second paragraph of this letter. However, this issue is [
in fact generic, and GE hereby requests closure of this issue on a generic !
lf basis.
An improved fuel rod thermal-mechanical design code has recently been developed and qualified that utilizes the revised gadolinia fuel themal conductivity relations. This code has been reviewed and approved by the NRC; application of this code is currently being l reviewed by the NRC on the NEDE-24011 docket. This approved code ;
was used to evaluate all GE fuel to be used for new plants. The !
results of the evaluation for fuel centermelting indicate that gadolinfa fuel melting is not expected to occur during normal l steady-state operation or during the largest whole core anticipated operational transient.
Please call me if I can be of any assistance on this matter. ' i' Very truly yours, I i
J. 5. Charnley Fuel Licensing Manager i Nuclear Safety and Licensing Operation '
i l
ec: C. H. Berlinger (NRC)
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L. 5. Gifford (GE) DSER OPEN ITEM /08 R. Lobel (NRC)
G. G. Sherwood (GE) bl406 bk
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HCGS DSER Open Item No. lila and b (Section 5.2.4.3)
PRESERVICE INSPECTION PROGRAM (COMPONENTS WITHIN REACTOR PRESSURE BOUNDARY)
The SER input will be completed after the applicant (a) dockets a complete and acceptable PSI program (b) submits all relief requests with a supporting technical justification.
RESPONSE
The preservice inspection program has been submitted under separate cover (April 13, 1984, letter from R. L. Mittl -
PSE&G to A. Schwencer - NRC). In addition, for the information requested in Item b, see response to Question 250.3.
M P84 95/13 2-dh
HCGS DSER Open Item No. 111c (Section 5.2.4.3)
PRESERVICE INSPECTION PROGRAM (COMPONENTS WITHIN REACTOR PRESSURE BOUNDARY)
The initial ISI program has not been submitted by the Applicant. This program will be evaluated after the applicable ASME Code Edition and Addenda can be determined based on Section 50.55a(b) of 10CFR Part 50, but before inservice inspection commences during the first refueling outage.
RESPONSE
For the information requested above see FSAR Sections 5.2.4 and 6.6.
W M P84 95/12 4-dh
HCGS DSER Open Item No. 119 (Section 6.2)
TMI ITEM II.E.4.1 The HCGS has two redundant hydrogen recombiner packages used for post-accident combustible gas control. The containment penetration associated with the hydrogen recombiner system are a combined design. The hydrogen recombiners are iso-lated by two isolation valves on suction inlet and are located downstream from the purge system isolation valves.
In order to properly evaluate this design, we require infor-mation from the applicant outlined in Section 6.2.4. We will report on the resolution of this matter in a supplement to this report.
RESPONSE
For the information requested above see the response to DSER Open Item 132.
M P84 95/12 5-dh
HCGS DSER Open Item No. 123 (Section 6.2.1.4)
BUTTERFLY VALVE OPERATION (POST ACCIDENT)
Vacuum in the suppression chamber is relieved by a 24-inch vacuum breaker assembly located in each of the two lines between the reactor building and the suppression chamber free space. Each assembly consists of a check-type vacuum relief valve and a pneumatically operated butterfly valve in series. The butterfly valve is located between the contain-ment and the check-type valve. The check-type valves are self-activating and can be remote manually operated from the main control room for testing purposes. The butterfly valves which are normally closed for containment isolation purposes, are activated by differential pressure between the reactor building and the suppression chamber free space.
The butterfly valves can also be remote manually operated from the main control room for tecting purposes. The power failure position of the butterfly valves is the closed posi-tion. The air supply for these valves is a non ESF supply.
We will require the applicant to comment on how these valves can be operated post accident, if this air supply is unavailable.
RESPONSE
The butterfly valves are provided with an accumulator which is designed to ASME Code,Section III, Class 3 require- :
ments. The accumulators are provided with a make-up source from the safety-related primary containment instrument gas ,
- supply system. (See Figure 6.2-29).
I In the event of an air failure, the accumulator will supply the air to activate the valves. Therefore, the valves can be operated post-accident. See revised Section 6.2.1.1.4.1.
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- 2. Flow through the vents is adiabatic.
- 3. The temperature of the suppression chamber ;
atmosphere is equal to the temperature of the suppression pool.
! 4. No credit is taken for heat losses to the drywell l wall, suppression chamber walls, and internal structures.
l 6.2.1.1.4 Negative Pressure Design Evaluation j
'6. 2.7 .1. 4 .1 Containment Vacuum Relief Valves l i
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The containment is designed to withstand an external-to-internal l
- differential pressure of 3 psi. To ensure that this design limit l is not exceeded, vacuum relief valves are provided to limit the !
i inward pressure loading on the drywell and suppression chamber
- l walls to no more than 2.5 psi. l l
Vacuum in the drywell is relieved by eight 24-inch vacuum relief l valves located on the vent header of the drywell-to-suppression
) chamber vent system. These valves are self-actuating, check-( type, that can also be remote-manually operated from the main ;
control room for testing purposes. The vacuum relief valves t between the drywell and the suppression chamber are sized to provide a total flow area of no less than approximately one-sixteenth of the net vent system cross-sectional area. ;
- 1 I
Vacuum in the suppression chamber is relieved by a 24-inch vacuum l breaker assembly located in each of two lines between the reactor i building and the suppression chamber free space. Each assembly I consists of a check-type vacuum relief valve and a pneumatically operated butterfly valve mounted in series, with the butterfly valve located between the containment and the ' check-type valve. ;
The check-type valves are self-actuating and can be remote- !
manually operated from the main control room for testing t purposes. The butterfly valves, which are normally closed for containment isolation purposes, are actuated by differential pressure between the reactor building and the suppression chamber free space. The butterfly valves can also be remote-manually operated from the main control room for testing purposes. The l controls and instrumentation for each butterfly valve are powered !
from different Class 1E electrical channels to ensure that l f
6.2-20 DSER OPEN ITEM /M fC) &b O O} L ;
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' failure of a singla electrical channel does not disable more than one vacuum breaker assembly. Each vacuum breaker assembly is sized on the basis of the flow of air from the reactor building required to limit the containment collapse pressure to within ,
2.5 psi. The maximum containment depressurization rate is a function of the containment spray flow rate and temperature and !
the assumed initial conditions of the containment atmosphere.
Low spray temperatures and containment atmospheric conditions that yield the minimum numbers of contained noncondensable moles of gas are assumed for conservatism. ;
The containment vacuum relief valves are qualified to Seismic Category I criteria and are designed and manufactured in i accordance with the requirements of the ASME B&PV Code, (
Section III, Class 2. The valves and appurtenances'are designed l to operate at a maximum pressure and temperature of 62 psig and l 3400F, respectively, concurrent with a maximum relative humidity !
of 100%. During such environmental conditions, the valves open fully within 1 second, with a 0.25 psi differential pressure existing across the valve. Each valve is equipped with redundant valve-position limit switches, which are suitably sensitive to provide main control room indication of valve closure to a ;
tolerance of 0.01 inch.
6.2.1.1.4.2 Containment Depressurization Evaluation Negative pressure differentials (negative corresponding to an ,
inward loading) across the drywell walls are caused by the rapid j depressurization of the drywell. Events that cause l depressurization in the drywell are:
- a. Cooling cycles j
- b. Inadvertent containment spray actuation during normal ;
operation t
- c. Steam condensation following RCS pipe ruptures with ;
inadvertent containment spray actuation. l
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Cooling cycles result in minor pressure transients in the drywell, which occur slowly and are controlled by h' eating and i ventilating equipment. Inadvertent spray actuation during normal l operation results in a more significant pressure transient and becomes important in sizing the suppression chamber-to-reactor .
building vacuum breaker assemblies. Steam condensation following i RCS pipe ruptures with inadvertent containment spray actuation i t
within the drywell results in the most severe pressure '
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HCGS Insert The normal air supply for these valve actuators is from the instrument air system. To assure these butterfly valves can operate post-accident, they are provided with an accumulator which is designed to ASME Code,Section III, Class 3 '
requirements. The accumulators are provided with a make-up source from the safety-related primary containment instrument gas supply system. (See Figure 6.2-29).
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DSER OPEN ITEM /83 i
M P84 112/02 3-srd !-
HCGS DSER Open Item No. 130 (Section 6.2.3)
POTENTIAL BYPASS LEAKAGE PATHS Although the primary containment is enclosed by the second-ary containment, there are systems that penetrate both the primary and secondary containment boundaries, creating potential paths through which radioactive material in the primary containment could bypass the filtration, recircula-tion, ventilation system. The criteria by which potential bypass leakage paths are determined are the BTP CSB 6-3,
" Determination of Bypass Leakage Paths in Dual Containment Plants." These criteria include specific requirements fdr barriers - such as water sealing systems, leakage control systems, and closed systems employed to process or preclude bypass leakage. Utilizing these criteria the applicant has identified in FSAR Table 6.2-15 those line's penetrating the primary containment that are potential reactor building bypass leakage paths, and the bypass leakage barrier (s) that will prevent bypass leakage. Since the applicant has not fully responded to our concerns regarding the Containment Isolation System (Section 6.2.4), we are unable to complete our review of the potential bypass leakage paths. We will report on this matter in a supplement to this SER. .
RESPONSE
For the information requested above see the response to DSER Open Item No. 132.
M P84 112/01 1-srd
HCGS DSER Open Item No. 138 (Section 6.6)
PRESERVICE INSPECTION PROGRAM FOR CLASS 2 AND 3 COMPONENTS The complete evaluation of the PSI program will be presented in a supplement to the SER after the applicant submits the required examination information and identifies all plant specific areas where ASME Code Section XI requirements cannot be met and provides a supporting technical justification.
RESPONSE
For the information requested above, see the response to DSER Open Item 111a and b.
4 M P84 95/13 1-dh
HCGS DSER Open Item No. 139 (Section 6.7)
MSIV LEAKAGE CONTROL SYSTEM The MSIVLCS is protected from the dynamic effects associated with the LOCA, the only pipe break and event where this system is required to operate. However, insufficient information has been provided in the FSAR to allow us to conclude that the components of each subsystem are protected by separation and barriers against internally and externally generated missiles such that their function will not be impaired under postulated LOCA conditions. Thus, we cannot conclude that the requirements of General Design Criterion 4, " Environmental and Missile Design Bases," and the guidelines of Regulatory Guide 1.96, Positions C.2 and C.4, are satisfied.
RESPONSE
Section 6.7.3.1 has been revised to discuss the effects of internally and externally generated missiles on the MSIV sealing system.
The effects of single active failures (including one MSIV failure to close) are provided in Section 6.7.3.2.
M P84 95/14 3-dh
HCGS FSAR 6.7.2.4 Equipment Required The following equipment / components are provided:
- a. Piping - Process piping is carbon steel pipe throughout and it is designed and constructed to ASME B&PV Code, as discussed in Section 3.2.
- b. Valves - Motor-operated, air-operated, relief, and check valves
- c. Instrumentation - The requirements and criteria for the MSIV sealing system instrumentation are discussed in Chapter 7.
The remainder of the piping and components are discussed in Section 9.3.6.
6.7.3 SYSTEM EVALUATION An evaluation of the capability of the main steam isolation valve (MSIV) sealing system to control the release of radioactivity from the MSIVs following a loss-of-coolant accident (LOCA) has been conducted. The results of this evaluation are presented in the following sections.
6.7.3.1 Functional Protection Features The equipment in the two independent subsystems (inboard and outboard) are physically separated. The equipment is designed to operate under the expected environmental conditions appropriate '
to the equipment location. #internej) ,
pedede s y separa.+ ion f" c.sased b4 egus!
and barrier From 1 ga;kre lsec Se d* *T**
- Le minimiz;"
( ~~~~TheMSIVsealingsystemequipmentissch[nged5eh[5
>the* exposure of the system components tq missile pipe breaks, and jet forces caused by the LOCA event. Equipment is located in the reactor building, hence the effects of the design basis recirculation line break would not impact the system ability to function. Furthermore the primary containment instrument gas system equipment that supplies gas to the MSIV sealing systein is located in the reactor building outside the steam tunnel, and pstula:/ col ex+e r-n a./ m o s s ,'/}es DSER OPEN ITEM /3 f 6.7-7 -
HCGS i
DSER Open Item No. 141C (Section 9.1.3)
SPENT FUEL POOL COOLING AND CLEANUP SYSTEM The applicant has not committed to include the portions of the cooling and cleanup systems which are not normally operating in the inservice inspection and periodic func-tional testing programs as decribed in Sections 6.6 and 3.6.6 of the SRP. The Applicant has not specified the freq-uency of the testing. Thus, the requirements e, General Design Criterion 45, " Inspection of Cooling Water Systems,"
and 46, " Testing of Cooling Water Systems," are not !
satisfied.
RESPONSE ,
l The spent fuel cooling system does not perform a specific function in shutting down the reactor or in mitigating the consequences of an accident; therefore, does not meet the ,
criteria for being included in ASME B&PV Code Section XI testing requirements.
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DSER OPEN ITEM 142a AND b (SECTION 9.1.4)
LIGHT LOAD HANDLING SYSTEM (Related To Refueling)
Redundant interlocks anc: limit switches have not been provided to prevent accidental collision with pool walls.
The applicant must prcvide these redundant interlock and limit switches or provide the results of an analysis which shows that the effects of a fuel bundle colliding with the pool wall is bounded by the fuel handling accident analysis in Chapter 15 of the PSAR.
Based on the above, we cannot conclude that the requirements of General Design Criteria 61, " Fuel Storage and Handling and Radioactivity Control" and 62, " Prevention of Criticality in Fuel Storage and Handling" and the guidelines of Regulatory Guide 1.13, Position C.3 with respect to prevention of unacceptable radioactivity releases and criticalfty accidents are satisfied.
RESPONSE
Strict administrative and procedural controls will assure that a collision of a fuel bundle with the pool wall will not occur. An analysis has shown that a postulated, accidental fuel-bundle collision with the pool wall cannot I result in more mechanical damage to the bundle hardware or the fuel or result in more fission product release than i could result from the postulated drop of a fuel bundle over the reactor core. The analysis of the fuel drop accident, described in Section 15.7.4, shows that the resulting fission product release would be within the guidelines of 10CFR100. Since the consequences of the wall collision ,
accident would be bounded by those of the fuel drop accident, re dundant interlocks and limit switches are not required.
MP84 95 08 01-az
DSER OPEN ITEM 168 (SECTION 12.5.2)
EQUIPMENT, TRAINING, AND PROCEDURES FOR INPLANT IODINE INSTRUMENTATION.
The applicant will utilize portable ventilation systems equipped with HEPA filters, or HEPA and sorbant filters, to minimize airborne contamination in highly contaminated areas. Continuous air monitors will be used to monitor airborne concentrations at specific work locations.Section III.D.3.3 of NUREG-0737 states that each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the f acility where plant personnel may be present during an accident. The applicant will provide a description of the equipment, training, and procedures to comply with Section III .D.3.3 by June 1, 1985. This is an open item.
Response
A description of the equipment, training, and procedures for accurately determining the airborne iodine concentration in areas within the HCGS plant where personnel may be present during an accident will be provided by June 1, 1985. These procedures and associated training will meet the intent of Section III.D.3.3 of NUREG-0737.
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HCGS DSER Open Item No. 170 (Section 13.5.2)
PROCEDURES GENERATION PACKAGE SUBMITTAL In Attachment D to a letter from R. L. Mittl to the Director, NRR, dated April 15, 1983, and in the PSAR Amend-ment 4 clarification to the letter, the applicant has com-mitted to implement Supplement 1 to NUREG-0737 (Generic Letter 82-33) and has committed to submit its procedures generation package (PGP) in January 1985. The PGP will be based on the BWR Emergency Procedure Guidelines prepared by General Electric and the BWR Owners Group which have been reviewed by the staff and approved by Generic Letter 83-05, dated February 8, 1983.
RESPONSE
As noted in the Draft SER it is the intent of PSE&G to sub-mit the HCGS procedures generation package in January 1985.
M P84 112/01 2-srd
. s HCGS l
DSER Open Item No. 171 (Section 13.5.2)
TMI-2 ITEM I.C.1 The staff will review the PGP for compliance with Supplement 1 to NUREG-0737. Our review must be completed prior to issuance of the operating license and will be addressed in a supplement to this Safety Evaluation Report. Until the review is completed, Task Action Plan Item I.C.1 is con-sidered open.
RESPONSE
The procedures generation package will be submitted to the NRC for review in January 1985.
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M P84 112/01 3-srd
HCGS DSER Open Item No. 172 (Section 13.5.2)
PGP COMMITMENT The applicant should commit to the following: 1) the PGP [
will be submitted to NRC three months prior to start of operator training, 2) all proposed operating and maintenance .
procedures will be completed at least three months prior to fuel loading, and 3) procedures will be available for review in advanced draf t form at least six months prior to fuel t loading. It is the staff position that procedures must be completed in sufficient time to ensure operator and appropriate plant staff familiarization. The PSAR should be '
modified to describe how adequate operator and plant staff familiarization will be assured.
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RESPONSE
FSAR Sections 1.10, 13.5.2 and 13.5.2.1 have been revised to provide the requested information.
P M P84 95/14 1-dh L
l activities will be listed with a brief overview of l their scope. This procedure will be deleted at the !
start of the first refueling. i In addition to these station administrative procedures, operationally oriented administrative procedures provide guidelines for the operations senior shift supervisors and their i shift crews, as well as procedures for night order book usage and !
control. Operations administrative procedures meet the 4 requirements of 10 CFR 50.54(i), (j), (1), and (m). ,
Figure 13.5-1 indicates the main control room area designated as ,
"at the controls," the area restricted to licensed personnel and l the limitations of the reactor operator while manipulating the j controls.
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[13.5.2 OPERATI G AND MA TENANCE P EDURES (
The perating and maint ance pro dures meet t e relevant .
I re irements as discus ed in Sec on 1.8.
l l I?.5.2.1 Main CorWrol Room eratina Pro dures !
l The f lowing ategories elineate t se procedur that are I !
l ontrol room perf rmed pr arily wi. n the main !
13.5.2.1.1 Operating Instructions j 1
Operating instructions are provided for startup, normal, manual, and automatic modes of operation of each system or subsystem .
related to plant safety. Detailed checkoff lists are included, !
where appropriate, within each procedure. These lists prescribe the proper valve lineup or switch position for the addressed mode 4 of operation, j 13.5.2.1.2 Overall Plant Operating Procedures Overall plant operating procedures provide instructions for f integrated plant operations. Checkoff lists are used for confirming completion of major steps in the proper sequence. l DSER OPEN ITEM /72 13.5-9 [
_. _ _ . . _ . . _ . . . _ _ . - _ _ . . . _ _ _ _ , _ , , _ _ _ . _ _ _ _ _ . _ ~ _ . . . , . , , , _ _ _ . . _ . _
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13.5.2 OPERATING AND MAINTENANCE PROCEDURE The operating and maintenance procedures meet the relevant requirements as discussed in Section 1.8.
It is planned that most operating and maintenance [
procedures will be completed at least three months l prior to fuel load and will be available for review ' (
in advance draft form at least six months prior to i fuel load. This will provide sufficient lead time ,
to ensure that plant personnel can become f amiliar !
with them. Where practical the preoperational j testing phase will be used to demonstrate tho -
adequacy of the operating procedures. !
13.5.2.1 MAIN CONTROL ROOM OPERATING PROEDURES i
The following categories delineate those procedures (
that are performed primarily within the main -
control room. Operator familiarization with these procedures is acquired though initial, requalifi-cation and replacement training programs. Further-more, these procedures will be utilized in simulator training.
DSER Open Item 172 M P84 95/14 2-dh
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Response l See Section 13.1 for discussion of the PSE&G and HCGS I organizations.
! The safety review group reports directly to the general manager -
nuclear support as discussed in Section 13.4.4 and shown
- on j J
Figure 13.1-8. t
- !.C.1 SHORT-TERM ACCIDENT ANALYSIS AND PROCEDURE REVIEW j Position -
In our letters of September 13 and 27, October 10 and 30, and November 9, 1979, we required licensees of operating plants, i
. applicants for operating licenses, and licensees of plants under l construction to perform analyses of transients and accidents, !
prepare emergency procedure guidelines, upgrade emergency procedures, and to conduct operator retraining (see also Item !
Emergency procedures are required to be I.A.2.1 of this report).
consistent with the actions necessary to cope with the transients and accidents analysed. Analyses of transients and accidents were to be completed in early 1980, and implementation of procedures and retraining were to be completed 3 months after i i emergency procedure guidelines were established; however, some j difficulty in completing these requicements has been experienced. '
i Clarification of the scope of the task and appropriate schedule revisions were included in NUREG-0737, Item I.C.I.
Pending staff approval of the revised analysis and guidelines, !
l the staff will continue the pilot monitoring of emergency ,
.- procedures described in Item I.C.8 (NUREG-0660). The adequacy of e l the boiling water reactor vendor's guidelines will be identified
- to each near-term operating licensee during the emergency j procedure review.
gesponse All emergency procedures will be written following the guidelines L
' of INPO 82-017, Emergency Operating Procedure Writing Guideline, and the guidelines of the BWR Owners Group-Emergency Procedures ;
Committee, as long as the guidelines do not contradict existing l NRC directives. These procedures will be available March 1, l
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t.o , // pro s n de a <>s in ,m u m o f 3 mo n t h s pr,'o r io Lhe l o
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HCGS i
DSER Open Item No. 173 (Section 13.5.2) .
PROCEDURES COVERING ABNORMAL RELEASES OF RADIOACTIVITY A procedure or procedures covering abnormal releases of radioactivity should be included among the available procedures.
RESPONSE
The following procedures will be developed to address abnormal releases of radioactivity:
- 1. Abnormal operating procedure titled " Abnormal :
Release of Radioactivity."
- 2. Emergency operating procedure titled " Radioactivity Release Control" (this procedure will be developed from BWROG emergency procedure guidelines).
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DSER OPEN ITEM 181 (Section 15.9.5)
TMI-2 ITEM II.K.3.3 I
Response given in Section 1.10 of the FSAR is not accept-able. The staff requires a detailed response explaining how f
PSE&G is planning to comply with the requirements of Item II.K.3.3.
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Response
HCGS will report any failure of a safety relief valve to open or close when called upon, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by phone, confirmed the first working day following the event by tele-graph (or similar transmission) and followed up with a writ-ten report in two weeks. This written report will be in the i form of a Licensee Event Report.
The PSE&G HCGS annual report to the NRC will list each ;
safety relief valve which is challenged during the year and ;
will include the number of times each is challenged. -
These reporting requirements will be included in the HCGS Technical Specification. ,
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M P84 95/07 1-cag l
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l HCGS FSAR l c. Increase in drywell sump level. !
- II.K.2 COMMISSION ORDERS ON BABCOCK & WILCOX PLANTS !
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4 Response ;
i These requirements are not applicable to HCGS. I l
- II.K.3 FINAL RECOMMENDATIONS OF B&O TASK FORCE ,
- II.K.3.1 INSTALLATION AND TESTING OF AUTOMATIC PORV ISOLATION SYSTEM .-
Response
f This requirement is not applicable to HCGS. !
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! e II.K.3.2 REPORT ON OVERALL SAFETY EFFECT OF PORV ISOLATION SYSTEM !
Response !
' r This requirement is not applicable to HCGS.
, Position
- [
Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report. This requirement is to be met before fuel load.
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DSER OPEN ITEM /$[
1.10-72 1
Response '
grg 3__Afhzeheed y rss=G will uvmply wa6u whw 6wyvissmsula vi ihis ilsm v6ies te I f.:: I::d.
- II.K.3.5 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING j I
- Response l l
l l This requirement is not applicable to HCGS.
t e II.K.3.7 EVALUATION OF PORV OPENING PROBABILITY DURING OVERPRESSURE TRANSIENT I
Response i This requirement is not applicable to HCGS.
I e II.K.3.9 PROPORTIONAL INTEGRAL DERIVATIVE (PID) CONTROLLER MODIFICATION ,
Response
This requirement is not applicable to HCGS. i e -II .K.3.10 PROPOSED ANTICIPATORY TRIP MODIFICATION l l
Response
This requirement is not applicable to HCGS.
DSER OPEN ITD4 /g/ _
Response
HCGS will report any failure of a safety relief valve to open or close when called upon, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by phone, confirmed the first working ' day following the event by tele-graph (or similar transraission) and followed up with a writ-ten report in two weeks. This written report will be in the form of a Licensee Event Report.
i The PSE&G HCGS annual report to the NRC will list each safety relief valve which is challenged during the year and i will include the number of times each is challenged.
These reporting requirements will be included in the HCGS ,
Technical Specification. I i
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i M P84 95/07 2-cag DSER OPEN ITEM / g/
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i DSER Open Item No. 191 (Section 7.2.2.8)
The applicant is required to revise FSAR Figure 7.2-1 to show the correct SDV level instrumentation design.
RESPONSE
For the information requested above, see the response to Question 421.14.
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M P84 112/01 5-srd .
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