ML20098C773
| ML20098C773 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/14/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8409270195 | |
| Download: ML20098C773 (97) | |
Text
.
O PS G Company PutAc Sennce Electnc and Gas 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 September 14, 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 AUGUST 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 October 1984.
Accordingly, this letter is provided to document the status of Hope Creek Generating Station responses to NRC requests for additional information which were forecasted to be responded to by August 1984.
Attachment I is a tabulation of the Hope Creek Generating Station Final Safety Analysis Report commitments for August 1984, and the corresponding resolution for each commitment.
Attachments II through VIII provide responses to questions forecasted to be responded to in August 1984, which will be included in Araeniment No. 8.
8409270195 840914 PDR ADOCK 05000354 p#g A
PDR 1 \\
L The Energy People 95 4912 (4W 7 81
m, Director of Nuclear Reactor Regulation 2
9/14/84 Should you-have any questions in this regard, please contact
. us.
Very truly.yours,
(/
Attachment I
- Hope Creek Generating Station - FSAR Commitment Status through August 1984
_. Attachment II
- Response to Question 210.53 Attachment III
- Response to Question 220.21 Attachment.IV
- Response to Question 480.11
- Attachment V
- Response to Question 100.6 (I.C.5)
Attachment VI
- Response to Question 260.15 Attachment VII
- Response to Question 471.14 Attachment'VIII - Response'to FSAR Table 13.1-4 C
D.
H.. Wagner (w/ attach)
USNRC Licensing Project Manager
.W. H..Bateman (w/ attach)
-USNRC Senior Resident Inspector
~
MP84 123/06.1/2D
(l Page 1 of 7 ATTACHMENT I
' HOPE CREEK GENERATING STATION FSAR COMMITMENT STATUS THROUGH AUGUST 1984 FSA'R' COMMITMENT LOCATION' COMMITMENT RESOLUTION 1.
L0uestion/ Response This. commitment concerns providing Appendix:
'FSAR Section 3.6 Tables and Figure Question 210.1 updated to final stress informa-tion..FSAR Tables 3.6-8, 9, 10, 11 have been updated to preliminary stress information and are included
~in Amendment 6 to the HCGS FSAR.
FSAR Table 3.6-5 contains final stress information and is included in Amendment 7.
,FSAR Figure 3.6-34 has been partially completed and is included in Amendment 6.
The pre-liminary tables and figure listed above will be finalized in December 1984.
This revised commitment date will be included in Amendment 8 to the HCGS FSAR..
2.-
' Question / Response This commitment concerns providing Appendix:
the tables and figures listed in Question 210.21 0210.1 and will be resolved as stated above.
3.
Question /ResponFe This commitment concerns providing Appendix:
criteria and analysis results for Question 210.53 stiff. pipe clamps per IE Informa-tion Notice 83-80.
This informa-tion is provided in Attachment II and will be included in Amendment 8 of the HCGS FSAR.
.4.
Question / Response This commitment concerns providing Appendix:
sketches and mathematical models
. Question 220.15 and method of analysis for the Spent Fuel Racks.
This information is provided as response to DSER Item No. 140 in letter; R.
L. Mitti (PSE&G) to A.
Schwencer (NRC),
dated, September 7, 1984, and will
.be included in Amendment 8 to the LHCGS FSAR.
M P84 144/06 1-cag
Page 2 of 7 FSAR COMMITMENT
' LOCATION COMMITMENT RESOLUTION
'5.
Question / Response This commitment concerns-the Appendix:
results of the Soil-Structure Question 220.21 Interaction Analysis for the reac-tor and auxiliary buildings and the service water intake structure.
This information is provided in Attachment III and will be included in Amendment 8 to the HCGS FSAR.
16.
0uestion/ Response This commitment concerns providing Appendix:
Spent Fuel Rack FSAR Figures 9.1-3 Question 281.2 and 9.1-4.
This informatien is provided in Amendment 7 to the HCGS
'FSAR.
7.
FSAR Section This commitment concerns an update 9.1.2.2.2.2 to the Spent Fuel Storage System description.
This information is provided in Amendment 7 to the HCGS FSAR.
8.
Question / Response This commitment concerns providing Appendix:
Spent Fuel Pool criticality inf or-Question 410.38 mation.
This information is pro-vided as response to DSER Item No. 140 in letter; R.
L. Mittl (PSE&G) to A.
Schwencer (NRC),
dated September 7,1984, and will be included in Amendment 8 to the HCGS FSAR.
9.
Question / Response This commitment concerns providing Appendix:
Spent Fuel Rack design details.
Ouestion 410.39
.This information is provided in Amendment 7 to the HCGS FS AR.
10.
Question / Response This commitment concerns testing of Appendix.
isolation systems.
This informa-Question 421.13b tion is provided in Amendment 7 to the HCGS FSAR.
M P84 144/06 2-cag l
u
m.-
Page 3aof 7 ut FSAR COMMITMENT LOCATION-COMMITMENT RESOLUTION
. 11..Ouestion/ Response This commitment concerns-an evalua-tion 'f the effects of high tempe-
.Appen' dix:.
- rature on reference. legs of water o
. Question 421.21 level measuring instruments.
This information is provided as response to -DSER Item No. 202 in letter;.
R.
L.
Mittl (PSE&G) to A. Schwencer (NRC), dated August 3, 1984, and will be included in Amendment 8 to the HCGS FSAR.
n
. 12.
' Question / Response' This commitment concerns 'the ISI
-Appe nd_ix :
program for the main condenser
. Question 430.164
- including the frequency and. extent of inspection.
This information is provided in f 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.
- 13. --Question / Response This commitment concerns the IST Appe ndix:
cnd ISI program for the turbine
~0uestion 430.167
' bypass system including the f re-quency. and extent of testing and.
inspection.
This information is provided in letter; R.
L.
Mittl (PSE&G) to A.
Schwencer-(NRC),
dated August 1,19 84, - and will be included in Amendment 8 of the HCGS FSAR.
Il4.
Question / Response This cammitment concerns the trip Appendix:
settings for the plant leak detec-Ouestion 440.10 tion system.. This information will
~
be provided in January 1985.
This revised commitment date will be
~
included in Amendment 8 to the HCGS FSAR.
.f
+
M P84 144/06 3-cag
S A
Page 4 of 7 s
FSARrOOMMITMENT
~ LOCATION-COMMITMENT RESOLUTION 15.
Question / Response This commitment concerns which
~ Appendix: _
evaluation -model is to be used for Question 440.27 the ECCS analysis.
This information.will be provided in July 1985 as. stated..in response to DSER Item No. 136 in letter; R.
L.
Mittl (PSE&G)'to A. Schwencer (NRC), dated August-20, 1984..This revised commitment date will be included in Amendment 8 to the HCGS FSAR.
- 16. - Question / Response
'This commitment concerns updating Appendix:
FSAR Table 3.6-5 per the results of Question 480.11 the pressure temperature _ transient-
~
analyses.
This: information is pro-vided1 1n Attachment IV and will be included in Amendment 8 to the HCGS FSAR.
FSAR Table 3.6-5 is provided in Amendment 7.to the.HCGS FSAR.
-STATUS.OF AUGUST 1984 FSAR COMMITMENTS MADE IN FSAR COMMIT-MENT STATUS LETTER TO NRC (
REFERENCE:
R. ~ L. MITTL (PSE&G)
TO A.
SCHWENCER, DATED AUGUST 13, 1984,-ATTACHMENT I).
- 17. ; Question / Response Re:- TMI -Item I.C.5:
This commit-
' Appendix:.
ment concerns assuring' feedback of Question 100.6.-
operating personnel via proce-dures.
This information-is pro-
- vided ~ in Attachment V and will be included in Amendment 8 to the HCGS
~
FSAR.
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1 Page 5-of 7 4
'FSAR COMMITMENT LOCATION COMMITMENT RESOLUTION 17.
Ouestion/ Response Re:
TMI Item II.B.3:
This commit-Appendix:
ment concerns assuring compliance Question 100.6 of the radioactive gas and liquid (Cont'd) sampling system for shielding and source term requirements.
This information is provided as response to DSER Item No. 148 (see Question 281.15)'in letter; R.
L. Mittl (PSE&G) to A. Schwencer (NRC),
dated September 12, 1984, and will be included in Amendment 8 to'the HCGS'FSAR.
18.
Question / Response This commitment concerns revising
. Appe ndix:
FSAR Section 1.8 to reflect con-Question 260.15 formance'with listed Reg. Guides which - are applicable during opera-tions phase.
This information is provided in Attachment VI and will be included in Amendment 8 to the HCGS FSAR.
.19.'
Question / Response This commitment concerns the mate-Appendix:-
rials monitoring program for the Question 281.14 Spe nt Fuel Pool.
This information is provided as response to DSER Item No.140 in letter; R.
L.
Mittl (PSE&G) to A. Schwencer (NRC),
dated September 7,1984, and will be included in Amendment 8 to the HCGS FS AR.
20.
Question / Response This commitment concerns inf o rma-Appendix:
tion on the Post Accident Sampling
~Ouestion 281.15 System which demonstrates compli-ance with NUREG-0737, Item II. B.3.
This information is provided as response to DSER Item No. 148 in letter; R.
L. Mittl (PSE&G) to A.
Schwencer (NRC), dated Septem-ber 12, 1984, and will be included in Amendment 8 to the HCGS FSAR.
M-P84 144/06 5-cag
~ ~
Page 6 of 7 FSAR COMMITMENT LOCATION COMMITMENT RESOLUTION
- 21.. 0ues tion / Response' This commitment concerns the Appendix:
ability of check valves in the Question 410.91 Equipment and Floor Drain System to maintain a functional pressure boundary.
This information is provided as response to DSER Item N o. 149 in letter; R.
L. Mittl (PSE&G) to A. Schwencer (NRC),
dated July 27, 1984, and will be included in Amendment 8 to the HCGS FSAR.
22.
Question / Response This commitment concerns seismic Appendix:
qualifications of check valves in Question 410.93
' drainage sys tems.
This informa tion is provided as response to DSER Item No.149 in letter; R.
L. Mittl (PSE&G) to A. Schwencer-(NRC),
dated July 27, 1984, and will be included in Amendment 8 to the HCGS FSAR.
23.
Question / Response This commitment concerns reactor
-Appendix:
mode-switch contact misoperations.
Question 421.26 This information is provided in response to DSER Item No. 192 in le tter; R.
L. Mittl (PSE&G) to A.
Schwencer (NRC), dated August 15, 1984, and will be included in Amendment 8 to the HCGS FSAR.
24.
Question / Response This commitment concerns providing Appendix:
the resume for Senior Radiation Question 471.14 Protection Supervisor.
This infor-mation is provided in Attach-ment VII and will be included in Amendment 8 to the HCGS FS AR.
M P84 144/06 6-cag
Page 7 of 7 FSAR COMMITMENT LOCATION
' COMMITMENT RESOLUTION 25.-
PSAR Table 13.1-4 This commitment concerns providing resumes for Maintenance Manager and Senior Nuclear. Maintenance Supervi-sor.
This information is provided in Attachment VIII and will be included in Amendment 8 to the HCGS FSAR..
i
.M P84 144/06 7-cag
T-ATTACHMENT II
1 HCGS FSAR 6/84 QUEhTION210.53 (SECTION 3.9.3 )
Describe what actions have been taken to address the staf f concerns regarding stiff pipe clamps as described in IE Information Notice 83-80.
RESPONSE
The applications of stif f pipe clamps on HCGirill 5:
iewed based on IE Information Notice 83-80 Section III of the ASME B&PV Code does not provide rules for evaluating stresses due to loadings from nonintegral attachments such as clamps; however, clamp-induced stressesfll 57 valuated by methods consistent with the intent of the Section III of the ASME B&PV Code.
The procedure included the following:
i 1.
Identip e locations of "stif f" clamps installed on ASME Section III Nuclear Class 1 piping systems.
d 2.
Ide nt the types of clamps, the loads acting on the clamp and t bolt pre-load values used in their installation.
I piping stresses due to all loading conditions at the locati of stif clamp also # identified and reviewed.
3.
-Add e primary membrane and bending stresses caused by the t
snubber load being transmitted to the pipe through the clamp to the stresses caused by internal pressure and bending computed by equation 9 of NB-3652.
Clamp-induced stresses i
caused by the constraint of e expansion of the pipe due to the internal pressure "ill,
ded to other secondary and peak stresses by calculating the of fective increases in the j
C1 and K1 stress indices in accordance with NB-3681.
Clamp 4
induced stresses due to dif ferential-temperature and differential-thermal-expansion coefficienty ill decounted i
for by computing the of factive C3 and K3 stress ind$ces.
Clamp-induced stresses on elb ws caused by the constraint of ccounted for by computing the pipe wall ovalizatiopill b[K bending indices.
he fatigue ef fective increases in C2 an 2
from clamp-induced plus other stressesjill _g alculated usage at governing locations.
Although bolt preloads are not addressed under the ASME B&PV Code Calculationp rill b_g o a pipe in damage rules for piping, bolt preloads could result in plastic $made to if a clamp were poorly designed.
e formation ensure that bolt preloads could not result of the pipe walls.
No p ems wen. MSied h k euabdioks dM rJo4Ms dr.scMwl kl aleos.N lAids of Sealin3?I o@de A&ME BjPV code were nd vioWed,f i
210.53-1 Ame ndme nt 6 mih Dw
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a HCGS FSAR 5.m 4-OUESTION 220.21 (SECTION 3.7.2)
Creek soil-structure interaction analysis is performed The H by app ying a specified earthquake input motion at the base of the finite-element model.
The base level input motion is The applicable SRP generated through a deconvolution analysis.
crkteria for soil-structure interaction analysis require that both the half-space and finite-element approaches should be considered.
(For details refer to SRP 3.7.2-11.4).
Public Service should perform necessary analyses to demonstrate conformance to the criteria.
RESPONSE
As described in Amendment 1 of the FSAR (Section 3.7.2.5.2),
seismic soil-structure interaction analysis is performed using both the impedance (half-space) and finite element approaches.
The results of the impedance analysis are used to assess the t
adequacy of.the finite element analysis results.
As discuss during he NRC ructur Audit January
-12 2
)%
1984, a apariso of res' nse spe ra res s from finite-eleme and ha
-space thods I be y vided in ugust 1 for e reac and a iliary uilding and the evice w er i
ake str ure.
t
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Insert A
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Fieures t'to. t! - I lhnu h 110. 21-54 sl>ow a.
cob arison of be, 2
p ferced dampy resjonse specfra.
oblarned inm +ke dess boa hmte element and the impidance ana lyses e ne reaekr buiidiy, auxilia ry butiding and he service water infake sfruefure (swis ).
i Sfor flue reachr and auxiltary uildiys, f!>e yes k b
speciral aetelerahvns a bfathed from the thyedan ce aKal S/S a re enera ll l0wer fhan ikOSe Ohfaihed &n?
Se desiyn basa analysis.
- However, the iyedance
[
. analysa response.specira. are ne'/ cmaplefely enveloped by those o btarved fr,m
+he dessyn ba.sa anaty.ra,
from /. ole 3.5 H e.
espectatty tu +be uen y raye Also, there a re score locaf exceedances av the hayher tio.tI-I
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r4 as 6 en Ft}u,ves uew
( Ftpu.re s Oro L
140. 11-36.
For fbe SwiS 110. CI-37 nr, h
110.11 ),
-th e fedance an4Iy345 th gevierally c.weloped 6
nose res parts e spec tra.
ave g
ob Awed 6,m ee desig basis anal s/5 Af e leua. feen y
. I t +, o fee t Er ener ' elews.6pns, fke t%geda nce i
aw tysis sp ec4<al accelera.hwns cAdeed
+ke.. design basis 3yechal acceleraksns sw some.
k a.e m ra a.g e s.
s p
(,adma.+ely beMeen
(.5 aad Tkese va va 10.0 H 1.
For de reactor bwiday and +he sw/S, seamit shear
(veces and onoments obfansed from he swyedance aulysis exceeded He design basis values.af varicus loca htns winov
+h es e buildi>tf s. Th e a u xt/ra ry batiding seismrc sh ea r force.s and momen h o bfain ed from ae angedance analysis are less ha 4 e destyn ba 515 sb ear 3 aHd on dM enl3.
Somce Hse swyedance analysis resn/h are nof coylefely enveloped by He desiyi, basis fusk elemen f autysis
- resulis, sanytay s+udres ore conduefed -lo con Srm de adepacy a f de planf desty&e These sampfunj s hulre.S n.
rnclude an evalua hun of fvllowiy fleins :
S +ru c k res Egwymen t Ca ble kny and Hvo4 C syyorb l
Prytyy aJ pGe suyjorts In a ll esses, de tiems reviewed duriy +hese sawy/iy s+udre.S can a ccorarinodale 6e losds resultiny fwm he imyedanee analysis.
Th ere fare, 4e ionyeds. nee analy3&
.tesn/h have ow impact on de p/sn t destyn..
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BCGS FSAR 10/g3 OUESTION 480.11 (SSCTION 6.2.3)
Provide a table listing the high energy lines which pass through the secondary containment and indicate which ones, if any, are For those high energy lines which do covered with guard pipes.
not?have guard pipes, provide results of analyses to demonstrate that the primary and secondary containment structure are capable i
gf withstanding the effects of a high energy pipe rupture occurring inside the secondary containment without loss of j
integrity.
i
RESPONSE
As discussed in Section 3.6.2.4, guard pipe assemblies are not used in HCGS.
Table 3.6-1 provides a listing of all high energy lines in the drywell and reactor building.
The effects of high energy line rupture, inpluding systes description, pipe break location, and a verification of the reactor shutdown capability, For all cases, the structural is provided in Section 3.6.1.2.1.
is l
integrity of the reactor building (secondary containment)
Results of the pressure temperature transient maintained.
analyses till 5?provided in Table 3.6-5,5; 1;;;;;t, ^;0"
.A M4" hggA 1
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480.11-1 Amendment 2
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person in charge of the control room shall be established
=
and limited to persons possessing a current senior reactor operator's license.
The plan shall clearly define the lines of communication and authority for plant management personnel not in direct command of operations, including those who report to stations outside the control room.
Response
The lines of responsibility and authority of the SSS-SRO (as promulgated in the vice president - nuclear letter of September 14, 1982), along with the authority and responsibility of the SSS-SRO (or the individual assigned the control room command function) to limit access, will be contained in Administrative Procedure SA-AP.22-001, Station Organization and
. Responsibilities (available March 1, 1985).
This item is also discussed in the response to Item I.A.1.2.
I.C.5 FEEDBACK OF OPERATING EXPERIENCE Position Each licensee will review its administrative procedures to assure that operating experience from within and outside the organization is continually provided to operators and other operational personnel and is incorporated in training programs.
Response
An integrated nuclear department procedure is being prepared, and will be available by March 1, 1985.
When issued, station
.'1 procedures will be revised to incorporate the prescribed
\\b procedure. Q '
Dcyk4~cmV PevcWe., OP-AP. e a -loS(Q) wuf be twot to g
cuscemiMe. 44ev b 4 0 % % Degv M b-s e t.
[nadd ion, t e Nucl ar Department Trai ng Cente will r vise,
[by J e 1984 the t proced s perta' ing to t s sub' ct t H--
inc de HC Th procedu.s are T 06, P1 Desi Rev w
nd TP 07, Ope t'ional periene eview rog
{Prgram, 4.
/
=
1.10-20 Amendment 4 l
l
)
l 1 '$ Indust y operating experiences in'cluding events occurring within p
our%rganization are reviewed for applicability to Hope Creek by i
Q the Reliability and Assessment Department.
Pertinent information is communicated to the appropriate department for their
{
information and any actions required are tracked until they have I
been satisfactorily completed.
In addition, information is
@i communicated to the Manager - Nuclear Training for incorporating g
new material into the Training Programs.
The activities of the
_ Reliability _and Assessment Department with respect to operating experiencesrare governed by procedure M3-POP-001 " Operating A
Experience Review Program".
.s vendor technical documents describing the operation and maintenance of installed equipment and components associated with Hepe Creek Generating Station shall be controlled in the v
following manner; 1)
When vendor documents are received by disiplines within the Nuclear Department, these documents will be forwarded to the Nuclear Engineering Department for review and approval for inclusion into the vendor Document Control System.
2)
Once approved by the cognizant engineer they will be assigned a unique number and distributed to all user departments,aed i @ aCA L- % nt. M 4 ' 4 9 M M ia^[.
Information on operating experience provided by the NRC through the I & E Bulletins /Information Notices, generic letters and letters on the docket are processed by nuclear licensing and regulation department within the nuclear department.
These letters are distributed to various disciplines within nuclear department for feedback of information.
Response action form is utilized when a respo e or action is required and is monitored through the response cking system M' completion.
h4 In addition, the Nuclear Department Training Center wi-1-L revised 4 ty Ju.c 19 d,
the two procedures pertaining to this subject to include HCGS.
The procedures are TP-30qi, Plant Dc"ign Review Program, and TP-30 Operational Experie nce Revi~ Program.
~
/*
Mc b hl TN sv-a.~p p-
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m ATTACHMENT VI e.
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-N OUESTION 260.15
^
The fourth paragraph of FSAR Section 17.2.2 refers to Section 1.8
'for commitments to Regulatory Guides.
Section 1.8 primariJy addresses Regulatory Guide commitments during design and construction, and the staff review of the FSAR is concerned with Regulatory Guide commitments during the operations phase.
With any proposed clarifications or exceptions, provide a commitment in the FSAR to the effect that: "During the operations phase of HCGS, PSE&G commits to comply with the regulatory position in..." the appropriate issue of the Regulatory Guide listed on pages 17.1-26 and 17.1-27 (with RG 1.33 replacing RG 1.28) or NUREG-0800 (Rev. 2 - July'1981).
For systems, components, and structures covered by the ASME Boiler and Pressure Vessel Code Section III (Classes 1, 2 and 3), the code OA requirements should be supplemented by the specific guidance addressed in the regulatory positions of the applicable Regulatory Guides.
(2B3)
RESPONSE
Section 17.2.2 lists regulatory guidance applicable to the OA This list has been revised to include Regulatory Guidesrd g program.
1.116, 1.123, and 1.144.fCNC.ill
components and structures covered by the ASME Boiler end Pressure Vessel Code Section III (classes 1, 2, and 3).
The standard QA program controls apply to Q-Listed code items following receipt 9fQ$gu
/
at the station. L g*p f
Angulh G4 L 13 s m // he. ap f i d' % A s nj E c. L s Sef&
f A C<h t Y o LU Me-5.t e+t.3 $ q vyt 5 1
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260.15-1 Amendment /
l
HCGS FSAR 1/84 1.8.1.63 Conformance to Regulatory Guide 1.63, Revision 2, July 1978: Electric Penetration Assemblies in Containment Structures for Licht-Water-Cooled Nuclear Power Plants l
Although Regulatory Guide 1.63 is not applicable to HCGS, per its implementation section, HCGS complies with the design, qualification, construction, installation, and testing requirements of IEEE 317-1976, as modified by Regulatory Guide 1.63, subject to the clarification in Section 8.1.4.12.
l 1.8.1.64 Conformance to Regulatory Guide 1.64, Revision 2, June 1976: Quality Assurance Requirements for the Design of Nuclear Power Plants Al th;;;5 Dagnl atory C 'id: 1.04 dec; not Opply te MCCS, per its-t 4-t---
-6 ucce --- 1 4 e e mies 4*
AIC.W 5 hi?Yl% (kuf& quj$
j.sy The tr:hitect-engineer indicate. LhaL Limit design verificatier -
ranferr to ANSI 45.2.!! 1074 and 210: that 00 pliane:
r@ ear =A"rae 9 th thio 5Lendard is os &cGifled end inLespretcd by "cvision I
(
ef ": gel-Lucy Guid: 1.50.
Meeever, th Orchit="*-anginaar di4 cerply with Mericica ; in LL L iL ollowcd checkin; cf the nne d ri;n output document by th Originater' rup=""i eer -
See Section 17.2 for further discussion of quality assurance procedures and Section 1.8.2 for the NSSS assessment of this Regulatory Guide.
1.8.1.65 Conformance to Regulatory Guide 1.65, Revision 0, October 1973: Materials and Inspections for Reactor vessel Closure Studs Regulatory Guide 1.65 is not applicable.
See Section 1.8.2 for the NSSS assessment of this Regulatory Guide.
(
6 1.8-36 Amendment [
l
I l
- 6/84 1.8.1.87 Conformance to Reaulatory Guide 1.87, Revision 1, June 1975: Guidance for Construction of C;, ass 1 Comoonents
,n E3evated-Temperature Reactors (Suoolement to ASME Gect:,on III Code Cases 1592, 1593, 1594, 1595, and h596 l
Regulatory Guide 1.87 is not applicable to HCGS.
1.8.1.88 Conformance to Reculatory Guide 1.88, Revision 2, October 1976: Collection, Storace, and Maintenance of Nuclear Power Plant Quality Assurance Records s~s-s During the operations phase, HCGS complies with ANSI N45.2.9-1974, as. modified and interpreted by Regulatory Guide 1.88 /'
' )
During the construction and startup phases, compliance is subject
.s4*cific changes.M bcw %) Ama tothefollowidn 1
a osco (s%
andw s u n - x.rr.y J
2, The architect-engineer indicates th,at.the original HCGS p'roject.
commitment, via the Bechtel nuclear quality assurance manual (NOAM), was to ANSI N45.2.9 (Draft 11, Revision 0, January 17, 1973) rather than to ANSI M45.2.9-1974.
The NQAM was revised to
(
reference the 1974 document, as modified ~and interpreted by the guide, subject to the following specific changes:
o ANSI Section 2.1, Quality Assurance Record System - Add a.
the following sentence at the end of this section: "The procedures shall include control of records required during completion of the work activity."
b.
ANSI Section 2.2.2, Nonpermanent Quality Assurance Records - Revise this section to read:
" Nonpermanent records are those required to show evidence that an activity was performed in accordance with the applicable requirement but need not be retained for the life of the item and do not meet the criteria listed in Section 2.2.1."
ANSI Section 3.2.2, Index - Revise this section to c.
read:
"The quality assurance records shall be listed in an index.
The inder shall include, as a minimum, record retention times and the location of the records within the record system.
The index system used by organizations for the retention of quality assurance 6
1.8-49 Amendment /
=
-.__L_"L..
t HCGS FSAR 6/84 Although Regulatory Guide 1.122 is not applicable to HCGS, per its implementation section, HCGS complies with it.
For further discussion of seismic design, see Sections 3.7 and 3.10.
)
4 1.8.1.123 Conformance of Reculatory Guide 1.123, Revision 1, July I
1977: Ouality Assurance Requirements for Control of f
Procurement of Items and Services for Nuclear Power
[
Plants ceiacF 4
J
}
HCGS compiles with Regulatory Guide 1.123 puring construction!and
}
1 startup phases, subject to clarifications stated below. During the item a clarifica ion applies c 'y wi% % gcyM operations phase, p,ww wk nuwm% Tvid I.~5 8-to u n w..s a =ll,p f
I we jn
/g f reddo(s gutMLM L
-Q
/
The architect-engineer indicates that the original HCGS project commitment was to ANSI N45.2.13 (Draft October 1973) rather than to ANSI N45.2.13-1976.
The architect-engineer NOAM has been revised to reference the 1976 document, as modified by the Regulatory Guide, subject to the following specific changes:
Regulatory Guide Section C.2 - This section requires l
i a.
4 the application of elements of the ASME B&PV Code,Section III, Divisions 1 and 2, and Section XI; and ANSI N45.2.13-1976; specifically, those elements not covered by the ASME B&PV Code for procurement of ASME B&PV Code items and services.
The architect-engineer takes exception to the requirement, and has the following alternate position:
The application of the ASME B&PV Code requirements above to the procurement of ASME B&PV Code items and services is adequate, based on the fact that ASME B&PV Code represents the composite knowledge and experience of a large segment of the nuclear industry, that the ASME B&PV Code is constantly being reevaluted for adequacy, that addenda are issued frequently, and that, to our knowledge, historical data do not exist that would indicate that the ASME B&PV Code quality a
assurance requirements, relative to the procurement of ASME B&PV items and services, are inadequate.
f b.
Regulatory Guide Section C.2 - This section of the l \\p regulatory position appears to be inconsistent.
It states that the purchase should verify the implementation of the suppliers corrective action 8
Amendment [
1.8-74 j
HCGS FSAR 8/84 Positions C.1.1.2, C.2.1.2, C.3.1.2, and Table 1 of Regulatory Guide 1.143 require that all material specifications for pressure-retaining components within the radioactive process boundary conform to ASME B&PV Code,Section II.
In addition, they require that piping materials conform to both the ASME and the identical ASTM specification, and they permit substitution of manufacturers' standards, instead of the ASME specification, in the case of pump materials.
Although Regulatory Guide 1.143 does not explicitly address in-line process components, sight flow glasses, Y-strainers, and steam traps procured by the architect-engineer, and the orifice plates and conductivity elements in the NSSS scope of supply do not have certificates of compliance for the materials specified.
Also, the records of shop inspection, required by Table 1, for the Y-strainers and the steam traps are not available from the supplier.
Nevertheless, the quality assurance measures taken provide the reasonable assurance needed to protect the health and safety of the public and that of plant operating personnel.
Position C.1.2.1 requires that the designated high-liquid-level conditions should actuate alarms both locally and in the control room.
For all tanks, a high-liquid-level condition actuates an alarm in the radwaste control room only.
There are no local alarms since the tank rooms are controlled areas and normally unmanned.
Position C.4.3 requires that process lines should not be less than 3/4 inch (nominal).
The crystal 11zer concentrates and slurry waste transfer lines to the extruder / evaporators are 1/2 inch nominal, in order to maintain acceptable flow velocities to y
prevent settling in the lines.
The fluid flowrates are on the order of one (1) GPM as shown in Table 11.4-7 and on Figure
.i
'11.4-9.
1.8.1.144 Conformance to Reaulatory Guide 1.144, Revision 1, l
September 1980: Auditino of Quality Assurance Procrams for Nuclear Power Plants
'i
/d rr rdeias (4^3 c-HCGS co$ $stynli s with R gulatory Guide 1.144/olure'n3 r
p rin3 om comfwtion pk s e, pf, (gj,,:,3 clariAcO ' W b
The architect-engine"r's quality program for safety-related items
!q during the design an. construction phasts meets the requirements
,1 of ANSI N45.2.12-15l: as modified and int erpreted by Regulatory
,i
,j-l) 1.8-115 Amendment 7 ll tt
ATTACHMENT VII
HCGS FSAR 10/83 00ESTION 471.14 (TABLE 13.1-2)
Provide the resumes for the Radiation Protection Engineer and the Senior Radiation Protection Supervisor (see " Areas of Review" 1.A.2 of SRP 12.5 (NUREG-0800)), or provide a schedule for subsitta1.-
BESPONSE Tables 13.1-2 and 13.1-4 have been revised to incorporate the resume for the Radiation Protection Engineer.;3 The resume for the Senior Radiation Protection Engineer rill L;4provided --"-
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TABLE 13.1-4 (cont)
Page 37 of !irt SENIOR RADIATION PROTECTION SUPERVISOR
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Page Ad of M h TABLE 13.1-4 (cont) 89p' 97 SENIOR RADIATION PROTECTION SUPERVISOR Wi1i be p: tid:d by 0 :::b:: 48+-
MAME :
Leo j. Kra}ewsk.
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1 ATTACHMENT VIII 1
,..,-r-,
TABLE 13.1-4 (cont)
Page 14 of l
MAINTENANCE MANAGER Wil be presid:d by June syss-
- EMSGAT~
Amendment 5 l
MAINTENANCE MANAGER 97 NAME:
Peter J. Kudless LICENSES AND CERTIFICATES:
Professional Engineer - New Jersey MILITARY:
1971 U.S. Navy (Active Duty) 19 U.S. Naval Reserve (currently Commander, CEC, 1971 - present USNR-R)
EDUCATION AND TRAINING:
1 196f - 1966 Worcester Polytechnic Institute, BS, Civil Engineering University of Rhode Island, Masters level 1970 - 1971 courses in Statistics, Math, Economics and Law PSE&G OA Orientation for Engineers 1975 ASME, Nuclear Power Engineering PSEEG, CPM of Scheduling 1976.
ASMC PSE&G, eeM6 Boiler & Pressure vessel Code 1977 General Electric, BWR Design Survey Course PSE&G, Control Valves & Pipe Fittings PSE&G, Welding Inspection W
General Electric, BgR Installation Course j
PSE&G Non-DestructiN Examination f
PSE&G,OA Orientation for Construction Engineers 6
,_m-, -.-.-...-_.,.,.. -,,
-.,.,_,.,+._--__..._t
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/6%
2 Peter J. Kudless EDUCATION AND TRAINING: (cont.)
1978 PSE6G, Hydraulic & Friction Crane Operation, Maintenance & Safety AMR Internation, Project Management 1979 PSE6G, Supervisory Training Program PSE&G, Strategies of Effective Listening 1980 PSE6G, QAD Follow Up Training for Engineering
& Construction Personnel Rutgers University, Advance Management 1981 Training EXPERIENCE:
1971 - Present Public Service Electric and Gas Company Maintenance Manager - HCGS:
Responsible for 6/84 - Present management, direction and control of the work of the Maintenance Department.
Assure conduct of electrical and mechanical maintenance activities is in accordance with facility license, company and government regulations.
Assure maintenance activities are accomplished safely and efficiently by properly trained and qualified personnel.' Assure that maintenance is conducted safely and efficiently during outages to achieve maximum y possible unit availability and reliability.
Develop and control budgets for the Maintenance Department.
Assure a cost effective spare parts inventory.
Act as Vice Chairman and member of the Station operations Review Committee.
12/80 - 6/84 Project Construction Manager, HCGS:
Responsible for monitoring the field construction ef forts of a 110dWWf BWR power Managed a staff in excess of 50 plant.
personnel (Civil, mechanical, electrical, HVAC, cost & scheduling engineers and administrative personnel) who controlled the _ oper.
field construction and support work of d%000 j$-
Also responsible for sit $, security personnel.
contraat.
i
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n97 3
Peter J. Kudless EXPERIENCE: (cont.)
Principke Construction Engineer - HCGS:
11/78 - 11/80 Supervised a staff of 20 Construction-engineers (all disciplines) who monitored and controlled all the field construction effort of 35007 personnel.
Lovde 7/76 - 11/78 Senior Construction Engineer - HCGS:
Supervised a staff of 5 Construction Engineers who monitored and controlled the field construction ef forts of.1500)C personnel in the power block area of the plant. ^- ove v-3/75 - 7/76 Associate Field Representative - Assigned toreMio I the corporate home offices, with ::tirn:1 field assignments, providing staff support (i.e., proce re, specification, drawing review) in ration for the start of the full site construction ef fort on the HCGS.
p.c 8/71 - 7/75 Engineer - Gas Engineering Dept. JMr 2 1/2 years, was senior site representative supervising a staff of 15 construction engineers and administrative personnel who monitored and controlled the sit! construction and support ef forts for a, Sew Spthetic Natural Gas (SNG) plant.
For 1 1/2 years provided home office staff support for construction of the first SNG plant constructed in this country, and a peak sharing Liquified Natural Gas (LNG) storage facility.
l 1966 - 8/71 U.S. Navy - Served on active duty as a Civil Engineer Corps (CEC)Of ficer.
Attended Naval OCS and CEC office (s school.
After commissioning, served at Newport, R.I., and R.I.
Republic of Viet Nam. while at Newport, was assigned as Assistant Operation officer I
for 1 1/2 years and a Resident Officer in Charge of Construction (ROICC) 1 1/2 years at the Navy Public Work Centerg.
In the former
__, gosition was responsible for directing fk,J.y4, Mag futilities, Maintenance and Transportation) of egen.h,ns,$ 3 the Operations Department, which provided the lo.ans services to all the Naval commands i
e l
(approximately 10,000 personnel) at the l
Newport Naval Base.
In the latter position,
l was responsible for administerir.; in ere--
ef en#en a4 p. lee t h
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,------------,~m--,-a-
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--n..
4
/7 Peter J. Kudless -
EXPERIENCES (cont.)
10 million dollars of construction contracts including utility (i.e., boiler repair),
vertical and horizontal construction projects.
I n ile serving 1 year as a ROICC in Viet Nam, s assigned to two different field locations 4%oadministerconstructioncontractsinexcess tof 1 billion dollars.
Type of projects included power plants and utility systems, waterfront, vertical and horizontal construction.
r/
TABLE 13.
.-4 (cont)
Page M of #8 l
M00t*
6% f 97 SENIOR MAINTENANCE SUPERVISOR NAME:
Thomas B. Wysocki EDUCATION:
1958 Greenport High School 1962 Wentworth Institute, Credits toward Associate Degree in Architecture 1962 Franklin Institute, 4
Credits in Math and English t
1973 Delaware Community College, Credits in Computers and Math 1975 Salem Community College, Associated Degree Construction Technology TRAINING 1963 U.S. Navy Machinist Mate A School 1964 U.S. Navy Submarine School 1965 U.S. Navy Nuclear Power School 1966-- 1970 Various U.S. Navy Service Schools on Nuclear Submarine Equipment 1974 Quality Assurance Orientation Course 1977 Quality Assurance Visual Examiners Level I and II Training C6urse 1980 Quality Assurance Liquid Penetrant Method of Nondestructive Examination Course 1980 Quality Assurance Magnetic Particle Method of Nondestructive Examination Course BWR Technology Course 1981 AYst/Onf MVs'erEden}ih*cn kon and J. ode
}9f3 RI3 geene4 &yewiJuf heIding N9f!"
- t titi g);esej 69 ene Mininy Amendment 5 l
TABLE 13.1-4 (cont)
Page FE of St-l EXPERIENCE 1970 - Present Public Service Electric and Gas Company 1981 - Present Senior maintenance supervisor for Hope Creek Generating Station:
Responsible for planning assignments, assigning manpower, and establishing work priorities.
Involved with the maintenance department establishment, spare parts acquisition, system and facility design reviews, procedure writing, and preparing for the startup test program.
1975 - 1981 Maintenance supervisor for Salem Generating Station:
Responsible for the direct planning and execution of maintenance activities.
Direct supervisor of bargaining unit work force 1972 - 1975 Startup test engineer for Salem Generating Station:
Duties included developing, writing, and directing startup procedures.-
Prepared test reports and evaluated test.
results 1970 - 1972 Junior staff assistant for corporate office:
Assisted the production department performance engineer in collecting data and preparing reports 1965 - 1970 Machinist mate, U.S. Navy:
Served as a mechanical repairman / operator on S5W pressurized water reactor.
Responsibilities included repairing, testing, and operation of the reactor and associated systems Amendment 5 l
i V
W TABLE 13.1-4 (cont)
Page & of ST l
T$U SENIOR NUCLEAR MAINTENANCE SUPERVISOR lb wm s g e.neee sv =..e e, w t
P i
Amendment 5 l
% )lt J3.Id (caf pyc. gs M rg rudCLM p
gJ SENIOR MAINTENANCE SUPERVISOR NAME:
MARK SHEDLOCK EDUCATION:
i 1971 - Colonia High School 1975 - Union Junior College A.A. Degree in Engineering 1977 - Rutgers University - College of Engineering B.S.E.E. in Electrical Engineering, Magna Cum Laude TRAINING:
1977 - Nestinghouse P.W.R. Course 1977 - QAD Course #25 'for P.C.D. Personnel 1977 - Non-Destructive Examination Course 1978 - 23rd Annual Appalachian Underground Corrosion Short Course 1978 - Seminar on State of the Art Power Plant Construction 1978 - QAD Course #5 for Orientation Training for Engineers k
1978 - QAD Course #27 for Reg. Guide 1.58 & ANSI N45.2.6 1979 - CCM-5 Construction Backcharges/ Subcontracts Course 1979 - CCM-3 Construction Scheduling Course 1979 - CCM-2 Construction Accountability Course 1979 - CCM-1 Overall Construction Process Course
~
1984 - QA Training for Hope Creek Start-up and Test Personnel EXPERIENCE:
1977 to Present - Public Service Electric & Gas Company, i
1983 to Present - Senior Maintenance Supervisor for Hope Creek Generating Station: Responsible for planning assignments, assigning manpower and establishing l
work priorities for maintenance activities and i
start-up testing activities.
Involved with the Maintenance Department establishment, system and i
facility acceptance, acquisition of department tools and test equipment & procedure review and approval.
I 4
O.
~..-. m
./
ae EXPERIENCE (cont. ) :
1979 to 1983 - Resident Electrical Engineer for Salem Generating Station, responsible for resolution of electrical operating problems to maintain unit reliability; develop specifications for subcontra't work pack-c ages; modification to system design to suit station conditions; review new design for constructability and material needs; site fire protection coordinator; interface with maintenance contractor and home office Engineering regarding new work items and NRC commitments; member - Salem Emergency Response Team.
May 1979 to
- Public Service Electric and Gas Company Nov. 1979 Hope Creek Generating Station Construction Engineer responsbile for monitoring the progress and quality of electrical, control and instrumentation activities in the power block; procurement review for permanent plant and con-struction equipment and materials; subcontract administration encompassing specification review, bid invitation, contract award and progress to closeout.
July 1977 to - Public Service Electric and Gas Company May 1979 Salem Generating Station - Unit No. 2 Acting Lead Construction Engineer - Electrical responsible for integration of discipline activities; review and approve Engineering changes; direct admin-1,gstration of Electrical subcontractor work packages; review and revise work package specifications; provide functional direction to Field Engineers and Designers.
Identify and resolve all construction problems; certi-fled Start-up/ Test Engineer - Level II.
-,,.,-._,-,-,,,,_,_._,---,,_-.-,-,-----n-,
i
{t s
GY TABLE 13.1-4 (cont)
Page 45 v 52 l
1s597 1975 - 1981 Instrument supervisor - Salem Generating Station:
Supervised I&C technicians in all aspects ~of instrumentation system maintenance;$t? sting, and calibration.
Functioned as department planning coordinator for initial startup of Unit 2 and two refueling outages for Unit 1 1972 - 1975 IEC technician'f e-Salem Generating Station:
Performed maintenance, calibration, and
~
testing activities on instrumentation and control systems 1965 - 1972 Electronics technician for U.S. Navy:
Performed maintencnce and testing of reactor plant' instrumentation systems V
.t.
s Readtor operator:
Daily operations of submarine S5W reactor plant b
' Engineering watch supervisor:
Supervised r
activities of technicians, machinists, and j
electricians in the varied aspects of g
submarine reactor plant operation and t
maintenance t
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