ML20100F632
| ML20100F632 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/29/1985 |
| From: | Bailey J GEORGIA POWER CO. |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| GN-570, NUDOCS 8504040385 | |
| Download: ML20100F632 (86) | |
Text
c Georgi: Power Company Rout] 2. Box 299A W;yn;sboro, Georgis 30830 h
Telephone 404 554-9961 t -
404 724-8114 Southern Oompany Services, Inc.
Post Office Box 2625 Birmingham, Alabama 35202
' Telephone 205 870-6011 V
tiePMect March 29, 1985 Director of Nuclear Reactor Regulation File: X7BC35 l
Attention:
Ms. Elinor G. Adensam, Chief Log:
GN-570 Licensing Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555 NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 REQUESTS FOR ADDITIONAL INFORMATION: DSER OPEN ITEMS
Dear Mr. Denton:
Your Staff has requested additional information as part of the VEGP review process. Attached is a listing of the DSER open items, the enclosures where they are addressed and the source of the request. As noted in the remarks, except for open items 98 & 107, the information will be incorporated in Amendment 16.
Sincerely, l
l
.. (4 J. A. Bailey i
Project Licensing Manager JAB /caa Enclosure l
xc:
D. O. Foster R. A. Thomas G. F. Trowbridge, Esquire J. E. Joiner, Esquire C. A. Stangler L. Fowler l
M. A. Miller L. T. Gucwa G. Bockhold l
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l 8504040385 850329 PDR ADOCK 05000424 E
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' Attachment OPEN ITEMS INDEX Open Item Enclosure Remarks 98 A
Response to NRC telephone inquiry, Emergency Operator Procedure Generation Package 86 B
Response to NRC telephone inquiry and NRC March 22, 1985 meeting. The changes noted will be incorporated in FSAR Amendment 16.
105 C
Response provided to questions i
raised at the NRC March 22, i
1985. Revisions supercedes those transmitted by GN-546 dated March 13, 1985. These changes will be incorporated in FSAR Amendment 16.
106 D
Response provided to questions raised (less item 2 on reviewer's notes) at the NRC March 22, 1985 meeting.
Changes are also provided to Section 17.2 to reflect.
GPC corporate changes. These changes will be incorporated in FSAR Amendment 16.
107 E
Response provided to NRC letter dated February 27, 1985 concerning the CRDR.
I I
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- _ _ _ _. ~.... _ _,,,.
Enc.lasare A SER OPEN ITEM #98 - Clarifications:
p MGm 98 Question: Provide the process for determining the required characteristics of instruments & controls.
(i.e.
range of instruments, gallons, etc.).
How do we know that the proper instrumentation is in place when writing related E0Ps?
Response: Since VEGP is under construction, it is not feasible to physically identify the installed equipment. Three other processes are used, however, to determine the proper instrumentation:
- 1) VEGP design
- drawing, i.e.
Mechanical, Electrical Drawings; Instrument qualifications, etc. are used to determine proper instrumention.
2)
Instrument Data Sheets or duplicate instruments installed in the VECP s;.acific siimulator are referenced to determine the range or scale of a specific instrumerit.
- 3) The VEGP Instrument Index is referenced to verify the accuracy of our selection of proper instrumention.
This instrumentation index was specified by Westinghouse for the preliminary E0P Setpoint study.
This study along with the task analysis portion of the Control Room design review will provide final confirmation of the required characteristics of the instrumentation' and controls.
Question: Provide a description of how operators keep track of time in implementing the steps of a time dependent procedure.
Response: On initiation of an event, the time is logged either as a result of normal log book keeping or as a result of completed the required Emergency Plan Notification Forms.
Short term time requirements associated cooldown rate are incorporated into the Proteus Computer, Emergency Response Facility, and Safety Parameters Display System Computers.
Further, the cooldown rate is manually checked every 15 minutes using the Control Room clock or operator's personal time pieces.
This method of time tracking is also utilized for time periods associated with venting reactor head into containment and time until hotleg recirculation.
The computer systems monitor and track the times associated with operations following a transient and thus provide adequate timing records (sequence of event recorders).
VEGP Operators are not required to record the times associated with the steps of a time dependent procedure.
- However, should time tracking problems arise during the E0P validation, the E0Ps will he modified.
Question: What process does an operator use to show that he is on the right step of a procedure?
Response: VEGP utilizes a mechanical marker as a place keeper to indicate the particular page being transitioned from.
Based on our experience in simulator training and E0P pre-validation simulator testing, the practice of checking off steps in E0Ps was found unnecessary.
Further, the two columnar fromat of the E0Ps eleminates the necessity of jumping around on a particular page.
Should operators subsequently experience problems concerning keeping track of the correct step of a procedure, measures will be taken to eliminate this problem.
r EEncic4L4re.
05 d>p a n 1"kC M [2SP VEGE-ESAR-10 10.3.5.4 Chemical Addition VEGP employs an all-volatile treatment (AVT) method to minimize f
general corrosion in the feedwater system, steam generators, and main steam piping.
Ammonia and hydrazine are the two chemicals to be injected into the condensate pump discharge header.
Alkaline canditions reduce the general corrosion rate of ferrous alloys, so ammonia (in the form of ammonia hydrazine) is f-injected to maintain these alkaline conditions.
Although ammonia is volatile and will not concentrate in the steam generator,,it will reach an equilibrium level which will establish an alkaline condition in the steam generator.
Hydrazine is added to scavenge dissolved oxygen present in the feedwater system.
Hydrazine also promotes the formation of a protective magnetite layer on ferrous surfaces and to keep this layer in a reduced state, further inhibiting general corrosion.
10.3.5.5 Action Levels for Abnormal Conditons Prompt and appropriate responses to abnormal chemistry conditions are prudent to assure the long integrity of secondary cycle components.
As such, three action levels have been defined for taking remedial action when monitored parameters are 15 observed and confirmed to be outside the normal operating value.
Normal operating value as it is used here refers to the value of a parameter which is consistent with long term system reliability.
The Vogtle general manager will authorize the action to be taken upon confirmation of one or more chemistry parameters outside normal operator ranges.
In general, these actions will be consistent with action levels described in "PWR" Secondary Water Chemistry Guidelines," EPRI, NP-2704 198g.2 s
Action level 1 is implemented whenever an out-of-normal value is detected.
Maintaining parameter values within the normal range
,uill provide a high degree of assurance that corrosive conditions will be avoided.
Action level 2 is instituted when conditions exist which have been shown to result in some degree of steam generator corrosion during extended full power operation.
Action level 3 is implemented when conditions exist which will result in rapid steam generator corrosion, and continued operation is not advisable.
10.3.5-3 Amend. 15 3/85
i
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VEGP-FSAR-10 TABLE 10.3.5-1 SECONDARY SIDE WATER CHEMISTRY SPECIFICATIONS I..
DURING POWER OPERATION Normal Sample Normal Action Levels I.
Parameter Frecuency Value 1
2 3
Condensate pump discharge
[ ;200 uf Dissolved oxygen, ppb Continuous
$10
>10 or daily
\\
Steam generator blowdown pH Continuous 8.8-9.5
<8.8 or daily
>9.5 Cation mko Conductivity, phe/cm Continuous 50.8
>0.8
>2.0
>7.C
(_
or weekly f5 Sodium, ppb Continuous
- 20
>20
>100
>500 or daily di Chloride, ppb Three times s20
>20
>100 a week Silica, ppb Daily 5300
>300 Gbiface, ppo eekly
$20
>?O w
s Amend. 15 3/85
VEGP-ESAR-10 TABLE 10.3.5-2 STEAM GENERATOR BULK WATER GUIDELINES DURING WET LAYUP f
Initiate Value Prior Parameter Frequency Normal Value Action to Heatup pH 3/ week 9.81-10.5
<9.8 >10.5
>9.0 Hydrazine, ppm 3/ week 75-200
<75 >200 Sodium ppb 3/ week
<1000
>1000
$100 Ghiorida rpb 3/ week
<1000
- 1000
.i100 ^ #
suitat;, ppt 1/ week
<1000
>1000
- 100^s Cafon Condad.viy 3/klcek 4.10.0
> l0.0 4 2.0 44nb %
15 l
Amend. 15 3/85
VEGP-ESAR-10 TABLE 10.3.5-3 f -
STEAM GENERATOR BULK WATER GUIDELINES DURING HEATUP Value Prior Value Prior to Power to Power Normal Normal Initiate Escalation Escalation Parameter Frequency Value Action
>5%
>30%
pH Continuous 28.7
<8.7 28.7 Cation -con-Continuoue 52.0
>2.0 s2.0 s0.8 ductivit9
( f +,,W /cm, '),
g f 'Specifl Continuous
>2.7
/
<ll
' M*^koconductivi y, 9;r c/m O-Dissolved Daily 55
>5 2,
ppb 0
(,
Sodium, ppb Continuous
$100
>100 s100 s20 Chl(iride, Daily s100
>100
$100 s20 15 ppb o
Julfata
-hht siOO
>w0 5100 s2v spb
- _Silic, ppb--dam-y c300
'300 %
s 0614V Amend. 15 3/85
Y Eac.lostara C open T+em tos VEGP-ESAR-1 o
Failure would not directly cause a Condition III or IV event (as defined in ANSI N18.2-1973).
e There is no safety function to mitigate, nor could failure prevent mitigation of, the consequence of a Condition III or IV event.
Failure during or following any Condition IV event e
would result in consequences no more severe than allowed for a Condition III event.
e Routine post-seismic procedures would disclose loss of the safety function.
1.9.30 REGULATORY GUIDE 1.30, AUGUST 1972, QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING OF INSTRUMENTATION AND ELECTRIC EQUIPMENT 1.9.30.1 Regulatory Guide 1.30 Position The requirements for the installation, inspection, and testing of nuclear power plant instrumentation and electric equipment which are included in ANSI N45.2.4-1972, Installation, Inspection, and Testing.
Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations (also designated IEEE Std. 336-1971) are generally acceptable and provide an adequate basis for complying with the pertinent quality assurance requirements of Appendix B to 10 CFR 50, subject to the qualifications in the regulatory guide.
1.9.30.2 VEGP Position The VECP QAP during design and construction conforms to ANSI N45.2.4-1972 and IEEE 336-1971.
The VEGP QAP is described in chapter 17.
The VECP operations QAP conforms with the requirements of ANSI N45.2.4-1972 as it is endorsed by Regulatory Guide 1.30-72 (Revision 0) with the following clarifications,and : < cept ion e.
9 W e h tion: m a.; e l h ;;.
1.
Paragraph 1.5, Referenced Documents.
Guides issued after 1972 have normally stated that when standards are referenced, the NRC position is that as described in the Regulatory Guide which endorses that standard.
1.9-24 Amend. 9 8/84 L
1, vacayaf 1,2, Pv utsdes. VEGP w;ff cmpph se condt%s as s aled in &s pacqmph in acydawe w$
+ke ve se
- {te v1 4a kevi le d e appuul?ie 08de ad cfmda Uos VEGP-FSAR-1 Since Regulatory Guide 1.30-72 does not include such a i
statement, VEGP feels justified'in stating that standards referenced in ANSI N45.2.4-1972 are addressed in appropriate sections of the VEGP FSAR.
19 y)f.
Paragraph 8, Records, requires certain records to be prepared.
VEGP has stated its position on records in section 17.2 and the VEGP position to Regulatory Guide 1.88.
y,g.
For operations phase maintenance and modification activities, VEGP shall control these activities under the operations QAP (sect' ion 17.2).
VEGP shall conform with the regulatory position in that QA programmatic /
administrative requirements (subject to the clarifications below) shall apply to modification activities.
Technical requirements will be made at least to the technical requirements of the FSAR.
9 5 g.
Paragraph 2.3, Procedures and Instructions, will be implemented as set forth in section 17.2 and by conformance with the Technical Specifications and VEGP position to Regulatory Guide 1.33.
h g.
Paragraph 2.4, Results, will be implemented as set forth in section 17.2 and with VEGP position to Regulatory Guide 1.33.
7 [.
ParagrapS 2.5, Measuring and Test Equipment.
VEGP conforms with this paragraph as discussed in section 17.2.
$ f(.
Paragraph 3, Preconstruction Verification.
VEGP conforms with this paragraph as discussed in subsections 17.2.10 and 17.2.11.
9 /.
Paragraph 4, Installation.
VEGP conforms with this 9
I paragraph as discussed in section 17.2.
10 g.
Paragraph 5.1, Inspections, including subparagraphs 5.1.1, 5.1.2, 5.1.3, will be implemented as set forth in section 17.2.
ll J6.
Paragraph 5.2, Tests, including subparagraphs 5.2'.1 '
~
through 5.2.3.
VEGP conforms with this paragraph as discussed in section 17.2.
1.9-25 Amend. 9 8/84
- O
m-VEGP-FSAR-1 l1 }{.
Paragraph 6, Post-Construction Verification.
VEGP 9
conforms with this paragraph as/ discussed in subsections 17.2.10 and 17.2.11.
l 3 )df.
Paragraph 6.2.1, Equipment Tests.
The last paragraph of this section deals with tagging and labeling.
VEGP will follow the requirements for tagging and labeling as set forth in section 17.2 and the VEGP position to Regulatory Guide 1.33.
/jl)4.
Paragraph 7, Data Analysis and Evaluation.
VEGP conforms with this paragraph as discussed in subsections 17.2.10 and 17.2.11.
VEGP shall have procedures for the performance of analyzing test data, but these procedures are not referred to as data 9
processing procedures.
14.
P;;agraph 2.2, Prcrcquisitec.
VECP ;ill confor.T to NEg the condi+4anr 2r stated in thic paragrcph i-n 15 b
accordenca "ith the VECP pccition tahcn te-the Q($
epp,4 2 sim coaom_nma - - a,rds.
1.9.31 REGULATORY GUIDE 1.31, REVISION 3, APRIL 1978, CONTROL OF FERRITE CONTENT IN STAINLESS STEEL WELD METAL 1.9.31.1 Regulatory Guide 1.31 Position This guide describes a method acceptable to the NRC for implementing requirements for the control of welding in fabricating and joining safety-related austenitic stainless steel components and systems in light-water-cooled nuclear power plants.
1.9.31.2 VEGP Position Conforms to the basic concept of controlling delta ferrite content except for magnetic measurement of the delta ferrite in procedure qualification samples and in production welds.
To meet the intent of the regulatory guide, the control of ferrite content in weld metal is attained by chemical analysis and/or l10 magnetic measurement of the weld metal, as applicable.
g
-Amend.
9 8/84 Amend. 10 9/84 1.9-26 Amend. 15 3/85
~
VEGP-FSAR-1 Welding materials for welding austenitic stainless steel to austenitic stainless may contain 54 to 25 percent delta f
ferrite.
The use of welding materials with a delta ferrite exceeding the recommended Ferrite Number 20 is done in accordance with the regulatory guide since austenitic stainless steel items are not postweld heat treated above 350 F (except during welding) unless they are given a full solution anneal at the material manufacturer's recommended temperature and holding period, followed by water quenching or spraying from the solution heat treating temperature rapidly enough to prevent carbide precipitation.
Control of ferrite content in stainless steel weld metal for NSSS equipment is discussed in paragraph 5.2.3.4.6.
1.9.32 REGULATORY GUIDE 1.32, REVISION 2, FEBRUARY 1977, CRITERIA FOR SAFETY-RELATED ELECTRIC POWER SYSTEMS FOR NUCLEAR POWER PLANTS 1.9.32.1 Regulatory Guide 1.32 Position For the portion of safety-related electric power systems within its scope, the criteria, requirements, and recommendations in IEEE Std.-308-1974 are generally acceptable to the NRC staff and provide an adequate basis for complying with General Design Criteria 17 and 18 of Appendix A to 10 CFR 50 with respect to the design, operation, and testing of electric power systems, subject to the qualifications identified in the guide.
1.9.32.2 VEGP Position Conform.
Refer to comparisons for Regulatory Guides 1.6, 1.9, 1.75, 1.81, and 1.93.
Further discussion is provided in sections 8.1 and 8.3.
1.9.33 REGULATORY GUIDE 1.33, REVISION 2, FEBRUARY 1978, QUALITY ASSURANCE PROGRAM REQUIREMENTS (OPERATION)
-e 4
1.9.33.1 Regulatory Guide 1.33 Position The overall QAP requirements for the operation phase that are included in ANSI N18.7-1976/American Nuclear Society (ANS) 3.2 are acceptable to the NRC and provide an adequate basis for "
complying with the QAP requirements of Appendix'B t6 10 CFR 50, subject to the qualifications.in the guide.
1.9-27 Amend. 9 8/84 gy 4
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VEGP-FSAR-1 1.9.33.2 VEGP Position j
i 15 The.VEGP. operations QAP conforms with this guide, which endorses IMSI N18.7-1976, with the following clarifications:
1.
It is GPC's understanding that ANSI N18.7-1976 applied to the operational phase of plant life.
Section 17.2 of the FSAR defines when the operations QAP becomes effective for plant systems, structures, or components.
2.
ANSI N18.7-1976 identifies other ANSI standards.
GPC addresses its position on those standards / regulatory guides in the appropriate parts of this section.
3.
Paragraph 1, Scope, recommends that this standard apply to activities other than those associated with safety-related equipment, activities, and procedures.
ANSI N18.7-1976 has not fully taken into account the requirements of regulations other than 10 CFR 50.
Conflicts may exist between ANSI N18.7-1976 and those other regulations, such as occupational Safety and Health Administration, 10 CFR 19, 20, 21, 30, 40, 70, 71, 73, and ASME.
Therefore, VEGP shall apply ANSI N18.7-1976 only to those activities determined to be safety related, which are defined as those plant features necessary to assure the integrity of the reactor coolant pressure boundary (RCPB), the capability to shut down the reactor and maintain it in a safely shutdown condition, or the capability to prevent or mitigate the consequences of accidents which would result in offsite exposures comparable to the guideline exposures of 10 CFR 100.
4.
Paragraph 2.2 defines the term quality assurance.
The last sentence of this definition, "It applies to all activities associated with doing a job correctly as well as verifying and documenting the satisfactory completion of the work", is inconsistent with_that o~f ANSI N45.2.10-1974 and 10 CFR 50 Appendix B.
The VEGP definition of quality assurance is consistent with 10 CFR 50 Appendix B.
l9 1.9-28 Amend. 9 8/84 Amend. 15 3/85
~
. ~ - -
VEGP-FSAR-1 personnel qualified in acc,the requirements to have Paragraph 3.4.2 indicated 5.
ordance with ANSI N18.1-1971 and ANSI N45.2.6-1973 3
'fEGF wonfermc eith ^"S! N19.1 in movvidshce w1th ty vr vP positivu Lv Regulat m Guide 1.9; a
_ e contouuo with."."S!
"'5.2.5 in -
che vEGP pudivivu Le Regulatcry C"4d_a-a o i va v.
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Paragraph 5.1 of Program Description.
The fourth sentence in this section required a " Summary Document;" section 17.2 provides a description of the operations QAP.
The plant procedure index lists plant procedures that will address the applicable requirements.
15 Amend.
9 8/84 1.9-29 Amend. 15 3/85
_ _ _'ypy:. /.- '!!I'5
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INSERT The applicability of ANSI N18.1-1971 for qualifying-plant personnel has been addressed in FSAR section 13.1.3; in that section, it is stated that personnel will either meet the minimum eduction and experience recommendations of ANSI N18.1-1971 or will complete a qualification program which will demonstrate their ability to perform their job functions. The 1973 version of ANSI N45.2.6 shall not apply for the VEGP, however, the applicability of ANSI N45.2.6-1978 for qualifying plant personnel has been addressed in FSAR paragraph 1.9.58.2.
1 L
n VEGP-FSAR-1 7.
1 Paragraph 5.2.6, Equipment Control.
VEGP will conform l15 with the-independent verification requirements based on the definition of this' phrase as given under the VEGP position to Regulatory Guide 1.74.
Since GPC sometimes uses descriptive names to designate equipment, the sixth paragraph, second sennence, is replaced with: " Suitable means include identification numbers or other descriptions which are traseable to_ records of the status of inspections and te s'e s. "
The first sentence in the seventh paragraph will be complied with after clarifying operating personnel to mean trained employees assigned to, or under the control of GPC management at an operating nuclear facility.
l15 8.
Paragraph 5.2.7, Maintenance and Modifications,~
discusses retaining documents as specified in section 5.2.12.
VEGP shall retain records as required by Technical Specifications and the VEGP position on Regulatory Guide 1.88 as stated in the FSAR.
9.
Paragraph 5.2.7, Maintenance and Modification.
Since l15 some emergency situations could arise which preclude preplanning of all activities, GPC will conform with an alternate to the first sentence in the second paragraph which reads: "Except in emergency or abnormal operating conditions where immediate actions are required to protect the health and safety of the public, to protect equipment or personnel, or to prevent the deterioration of plant conditions to a possibly unsafe or unstable level, maintenance or modification of equipment shall be preplanned and performed in accordance with written procedures which conform to applicable codes, standards, specifications, 15 and criteria.
Where written procedures would be required and are not used, the activities that were accomplished shall be documented after-the-fact and receive the same degree of' review as if they.had been preplanned."
10.
' Paragraph 5.2.7.1, Maintenance Programs.
VEGP'will l15 i
conform with the requirements of the first sentence of the fifth paragraph, where practical.
This clarification is needed since it is not always possible to promptly determine the cause of~the
~
malfunction.
In all cases, GPC will initiate proceedings to determine the cause, and-will make such determination promptl'y, when practical.
Amend.
9 8/84 1.9-30 Amend. 15 3/85 v
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VEGP-FSAR-1 11.
Paragraph 5.2.7.2, Modifications, discusses ANSI l15 N45.2.11-1974.
VEGP shall perfofm modifications as specified in section 17.2'and in the VEGP position on Regulatory Guide 1.64.
15 12.
Paragraph 5.2.9, Plant Security and Visitor Control, requires certain procedures and controls.
The VEGP position on security is addressed in its position on Regulatory Guide 1.17.
An NRC approved security plan shall be implemented prior to fuel loading.
13.
Paragraph 5.2.10, Housekeeping and Cleanliness l15 Control.
The requirements of this section, beginning with the last sentence of the first paragraph and continuing through the end of the section, will be implemented as described to VEGP conformance to ANSI N45.2.3 and N45.2.1 as described in the FSAR.
9 14.
Paragraph 5.2.11, Corrective Action.
VEGP shall follow l15 the requirement as discussed in subsection 17.2.16.
15 15.
Paragraph 5.2.17, Inspections, requires inspection of modifications and nonroutine maintenance to be conducted in a manner similar to the construction phase.
VEGP will inspect modification and nonroutine' maintenance activities in a manner so as to ensure the reliability and integrity of the item.
Amend.
9 8/84 1.9-31 Amend. 15 3/85 g
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VEGP-FSAR-1 16.
Paragraph 5.2.17, second to the last sentence, l15
" Deviations, their cause, land any..." to be consistent with paragraph 5.2.11, the cause of the condition will be determined for only significant conditions adverse to safety.
15 9
17.
Paragraph 5.3.9 and subsections, Emergency Procedures.
As directed by the NRC, GPC will follow l7 a format for emergency operating procedures in I
accordance with item I.C.1 of NUREG-0737.
15 Exceptions to Regulatory Guide 1.33-1978 are as follows:
1.
Paragraph C.5.e of Regulatory Guide 1.33 (and Section 5.2.13.4 of ANSI N18.7, which it references) will be implemented as discussed in section 17.2.
l l
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7 5/84 Amend.
9 8/04 1.9-32 Amend. 15 3/85 e
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VEGP-FSAR-1 2.
-Fauaymmph C.2.15, nu.1==,'.*.pprcvel 2"A "0" trol of "rcccsurcc.
In the third c;nte..ic ir p *egenpk 7
accideM_an z
ur. a ml umumut"__.,
J fia.md t; includ^
"=n unc::pccted trancient cignificent cpercter erren or g
equipment malfuncti^= 'h4ch
- acul+"
in n - vapn** n hj e 15 s ;e n t. " _._. 31uc c therc in ne clecr guidance for '"h e
- an -
unuouc1 incide"* ie VECP Pcc defined thic ter.- en-that-i t 10 viearly.undcrctccd whuu Lv gutfvum thc applirah]e y&vmudute.ucticus.
9 j[,7(
Paragraph 5.2.16, Measuring and Test Equipment, of ANSI N18.7-1976, which required equipment be suitably marked to indicate calibration status.
Installed process instruments at VEGP are identified by unique instrument numbers.
These instrument numbers are traceable to calibration schedules and calibration records.
These instruments are not tagged or labeled with the date due to next calibration.
j3,4{
Paragraph C.S.g of Regulatory Guide 1.33 will be l15 implemented with the addition of the modifier "normally" after each of the verbs (should) which the Regulatory Guide converts to "shall." It is GPC intent to fully comply with the requirements of this paragraph, and any conditions which do not fully comply will be documented and approved by management personnel.
In these cases, the reason for the exception shall be retained for the same period of time as the affected preoperational tests.
1.9.34 REGULATORY GUIDE 1.34, DECEMBER 1972, CONTROL OF ELECTROSLAG WELD PROPERTIES 1.9.34.1 Regulatory Guide 1.34 Position This guide describes an acceptable method of implementing requirements with regard to the control of weld properties when fabricating electroslag welds for nuclear components made of ferritic-or austenitic materials.
1.9.34.2 VEGP Position Conform.
Refer to paragraph 5.2.3.4.6.
Amend.
9 8/84 1.9-33 Amend. 15 3/85 f.h tro th WS
VEGP-FSAR-1 1.9.35 REGULATORY GUIDE 1.35, REVISION 2,. JANUARY 1976, INSERVICE INSPECTION OF UNGROUTED' TENDONS IN PRESTRESSED CONCRETE CONTAINMENT STRUCTURES 1.9.35.1 Regulatory Guide 1.35 Position This guide describes an acceptable basis for developing an appropriate inservice inspection and surveillance program for ungrouted tendons in prestressed concrete containment structures.
1.9.35.2 VEGP Position Conform.
Refer to subsection 3.8.1 for discussion on this subject.
1.9.36 REGULATORY GUIDE 1.36, FEBRUARY 1973, NON-METALLIC THERMAL INSULATION FOR AUSTENITIC STAINLESS STEEL 1.9.36.1 Regulatory Guide 1.36 Position This guide describes an acceptable method for implementing criteria for the selection and use of nonmetallic thermal insulation to minimize contamination that could promote stress-corrosion cracking in stainless steel components.
1.9.36.2 VEGP Position Conform.
Refer to paragraphs 5.2.3.2.3 and 6.1.1.1.3.
1.9.37 REGULATORY GUIDE 1.37, MARCH 1973, QUALITY ASSURANCE REQUIREMENTS FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER-COOLED NUCLEAR POWER PLANTS 1.9.37.1 Regulatory Guide 1.37 Position The requirements and recommendations for onsite cleaning of materials and components, cleanness control, and preoperational cleaning and layup of water-cooled nuclear power plant fluLd systems that are included in ANSI N45.2.1-1973, Cleaning of-Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants,_ are generally acceptable and Amend. 9 8/84 1.9-34
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VEGP-FSAR-1
)
provide an adequate basis for complying with the pertinent quality assurance requirements of Appendis B to 10 CFR 50, subject to the qualifications identified in the guide.
1.9.37.2 VEGP Position The VEGP QAP during design and construction conforms to ANSI N45.2.1-1973 with the following exceptions and clarifications.
Exceptions are as follows:
1.
The VEGP QAP during design and construction conforms to ANSI N45.2.1-1973 except in regard to installation cleaning.
Carbon steel piping is stored with the end caps removed and without dessicants.
The piping is stored to allow drainage and to prevent entry of rainwater.
Prior to installation the piping is inspected and cleaned if necessary.
Clarifications are as follows:
1.
This guide applies to onsite cleaning of materials and 9
components and, therefore, not in the direct scope of NSSS supply.
However, controls for cleaning processes during manufacture of NS3S equipment satisfy the objective of ANSI N45.2.1-1973, which is to assure that components delivered to the plant site require only water flushing or rinsing to render them ready for service.
Refer to paragraph 5.2.3.4.1 and subsection 17.1.2.
The VEGP operations QAP conforms with ANSI N45.2.1-1973, as it is endorsed by Regulatory Guide 1.37 (3/73), with the following clarifications:
11 1.
Paragraph 5, Installation Cleaning.
The recommendation that local rusting on corrosion-resistant alloys be removed by mechanical methods is interpreted to mean that local rusting may be removed mechanically, but the use of other removal means is not precluded.
15 Amend.
9 8/84 1.9-35 Amend. 15 3/85 IZmennf(. -ffi?
L
g VEGP-FSAR-1 In addition to the above clarifications, the operations QAP (section 17.2) conforms with Regulatory Gdide 1.37 for modification activities only with the following clarification.
(For operation and maintenance activities, cleanliness will be maintained per VEGP position on Regulatory Guide 1.33.)
1.
Paragraph C.3 of Regulatory Guide 1.37.
The water 9
quality for final flushing of fluid systems and associated components shall meet the requirements of ANSI N45.2.1-1973, but this does.not infer that chromates or.other additives normally in the system water will be added t the flush water.
f, " % Ck($ d K N 1.9.38 REGULATORY GUIDE 1.38, REVISION 2, MAY 1977, QUALITY ASSURANCE REQUIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR POWER PLANTS 1.9.38.1 Regulatory Guide 1.38 Position The requirements for the packaging, shipping, receiving, storage, and-handling of items for water-cooled nuclear power plants that are included in ANSI N45.2.2-1972, Packaging, Shipping,. Receiving, Storage, and Handling of Items for Nuclear Power Plants During the Construction Phase, are acceptable to the NRC staff and, when supplemented by the guidelines identified.in-regulatory position 2, provide an adequate basis for complying with the pertinent quality assurance requirements of Appendix B to 10 CFR 50, subject to the qualifications identified in this guide.
1.9.38.2 VEGP Position The VEGP QAP during design and construction conforms, with the following clarifications:
1.
Brightly or specially colored tape will not be used due to the rigorous flushing program scheduled prior to preoperation.
Tapes and vapor barriers used in 9
packaging processes for NSSS equipment contrast wi.th^.
the material being packaged when such packing materials are commercially available.
2.
Caps and plugs'are used only when required by the specification.
See Regulatory Guide 1.37 co^mparisch.
Tape near a weld may be removed to clean, setup, and inspect surface.
1.9-36 Amend. 9 8/84
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INSERT 1.9-36 2.
Paragraph C.4 of Regulatory Guide 1.37.
This paragraph includes requirements that chemical compounds which are to be used with austenitic stainless steels and nickel-base alloys should not contain chlorides, fluorides, lead, zinc, copper, sulfur, or mercury where such elements are leachable or where they could be released by breakdown of the compounds under expected environmental conditions.
In lieu of this requirement, for expendable consumable materials which are to be used in contact with austenitic stainless steels and nickel-base alloys, VEGP shall establish prescribed maximum levels of water leachable contaminants (chlorides, total halogens, sulfur, etc.)
These levels shall be established so as to prevent any contribution to intergranular cracking or stress-corrosion cracking that could occur.
Criteria for approving the use of a specific type or product of consumable material at VEGP, include the requirements that its water leachable contaminant content is below the prescribed maximum level and that it does not contain these contaminants as basic and essential chemical constituents.
Approved for use consumable materials are delineated in a controlled specification.
Refer to the clarification made to paragraph 3.6 of ANSI N45.2.2 - 1972, FSAR paragraph 1.9.38.2, for a further discussion of the use of consumable materials at VEGP.
o
i VEGP-FSAR-1 3.
The contact prcservative used on the main condenser is not water flushable; it will be cEemically cleaned.
4.
Quality assurance for packaging, shipping, receiving, storage, and handling of NSSS equipment is described in WCAP-8370, Rev. 9A-Amendment 1, Table 17-1.
Refer to chapter 17 for further discussion.
The VEGP operations QAP conforms with this guide, which endorses ANSI N45.2.2-1972, with the following clarifications and exceptions.
Clarifications are as follows:
9 1.
Paragraph 2.5, Measuring and Test Equipment (2.5.2).
VEGP meets the requirements of paragraph 2.5 of ANSI N45.2.2 by providing for calibration and control of 15 appropriate warehouse monitoring instruments under the VEGP planned maintenance program.
The VEGP planned maintenance program provides for calibration and control of all appropriate installed process plant equipment in accordance with the VEGP position to RG l
1.33, paragraph 5.2.16.
l7
!9 2.
Paragraph 3.3, Cleaning (third sentence).
VEGP interprets " documented cleaning methods" to allow generic cleaning procedures to be written which are implemented, as necessary, by trained personnel.
Each particular cleaning operation may not have an individual cleaning procedure, but the generic procedures will specify which methods of cleaning or which types of solvent may be used in a particular application.
3.
Paragraph 3.4, Methods of Preservation (first sentence).
VEGP will conform with these requirements 15 subject to the exception taken to the requirements paragraph 3.2.1 and the definition of the phrase
" deleterious corrosion" to mean that corrosion _which cannot be subsequently removed and which adversely affects form, fit, or function.
- Amend.
7 5/84 Amend.
9 8/84 1.9-37 Amend. 15 3/85
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Sections 4.3, 4.4, and 4.5 of ANSI N45.2.2-1972 titled, respectively, Precautions During Loading and Transit, Identification and Marking, and Shipment from Countries Outside the United States.
VEGP will conform with the requirements of these sections on a case by case basis.
l9 (o /.
Paragraph 5.2, Receiving Inspection Requirements.
l15 Preliminary visual inspection will be performed prior to unloading where practical; however, the receiving inspection of record will be performed in an area and in a manner which does not adversely aftect the h
quality of the item being inspected.
E f Jil.
Paragraph 5.3.1, Acceptable.
Item acceptance status l15 will be indicated by application of tags, stickers, ribbons, or signs.
Storage areas are not designated as accept areas except for bulk items (e.g.,
- rebar, structural steel, aggregate, etc.)
19 h [.
Paragraph 5.7, Documentation.
Receiving inspection IIS records will provide traceability to the item and its status.
Superfluous identification and tagging will not be recorded except when they are the subject of a nonconformance or specifically required by site inspection procedures.
19 M /.
Paragraph 6.2.1, Access to Storage Areas.
Items which l15 I
fall within the Level D classification of the standard will be stored in areas which may be posted to limit access, but other positive controls such as fencing or guards will not normally be provided.
[h [
Paragraph 6.3.3, Storage of Hazardous Mate, rial
'Phe-l15 s
ce n t c r.c c ir rep 1 -a ui " :.hr milca rg. 3 azardous chemicals, paints Sol. vents, and other materials of a like nature suff tc at m a
- naoproved Cabinets or containers vipich are not in clBWe proximity,to 0 656 639te ms e VEGP shan coni 0M +0 +NS SentNe by d6bH N'"p dad M 4r M dems" to be d irlstalled Safe 5b d dd $ ##
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- INSERT Page 1.9-38 4.
Paragraph 3.6, Barrier and Wrap Materials and Desiccants. This section contains requirements that; " Barrier and Wrap Materials shall be non-halogenated when used in direct contact with austenitic stainless steels, shall be noncorrosive, shall not readily support combustion, and shall not be otherwise harmful to the item packaged." In lieu of the require-ment for_this material being nonhalogenated, VEGP shall establish a pre-scribed maximum level for rater leachable halogen content for barrier and wrap materials to be used in contact with austenitic stainless steels.
In general, VEGP shall establish prescribed maximum levels for any water leachable contaminants contained in any consumable materials which are to
.be used in contact with austenitic stainless steels and nickel-base alloys.
The levels for witer leachable contaminants shall be established so as to prevent any contribution to intergranular cracking or stress-corrosion cracking that could occur. At VEGP, only the specific products or types of consumable materials which have undergone an engineering evaluation and have been approved as qualified will be used in contact with austenitic stainless steels. Criteria for approving the use of a specific type or product of consumable material include the requirements that its water leachable contaminant content is below the prescribed maximum level and that it does not contain these contaminants as basic and essential chemi-cal constituents. Approved for use consumable materials are delineated in a controlled specification; while consumable materials specified or supplied by a component vendor that are not covered by this specification shall be used only for the application and in the manner specified by the vendor's drawings, technical manuals, or other official documentation. Refer to the clarification of Regulatory Position C.4 of Regulatory Guide 1.37, March
'73, contained in FSAR paragraph 1.9.37.2 for further discussion of the use of consumable materials at VEGP.
VEGP-FSAR-1 yJ.
Paragraph 6.4.2, Care of, Items..The following l15 d
clarifications are provided for' indicated subparts:
(5)
" Space heaters in electrical equipment shall be energized unless a documented engineering evaluation determines that such space heaters are not required."
9 (6)
" Rotating electrical equipment shall be given~
insulation resistance tests on a scheduled basis unless a documented engineering evaluation or manufacturer's recommendations determine that cach tests are not required."
(7)
Prior to being placed in storage, rotating l15 equipment weighing over approximately 50 lb shall be evaluated by engineering personnel to
.(p gect.h h determine if shaft rotation in storage is g-i" CXusMV6 fdction god required; the results of the evaluation shall coc@sion,
be documented.
If rotation is required, it j
shall be performed at specified intervals, and s%___
documented.
Parts will receive a coating of applicable;)kothatthe lubrication where shaft does not come to r e s_t in'the same dh[
Sh4 t ShqN h6 position occupied prior to rotation?
For long shafts or heavy equipment subject to fotaffd [-Y k hWAb undesirable bowing, shaft orientation after rotation shall be specified and obtained.
l23)(.
Paragraph 7.3, Hoisting Equipment.
Rerating of l15 hoisting equipment will be considered only when absolutely necessary.
Prior to performing any lift above the load rating, the equipment manufacturer should be contacted for his approval and direction.
The manufacturer should be requested to supply a document granting approval for a limited number of lifts at the new rating and any restrictions involved, such as modifications to be made to the equipment, the number of lifts to be made at the new rating, and the test lift load.
At all times, the codes governing rerating of hoisting equipment must be observed.
If rerating of hoisting equipment is necessary and VEGP cannot or does not contact the equipment manufacturer as described above, the test weight used in temporarily rerating hoisting equipment for special lifts will be at least equal to 110 percent of the lift weight.
A dynamic load test over the-full range of the lift using a weight at least equal to 110 l15 percent of the lift, weight will be performed.
E Amend.
9 8/84 1.9-39 Amend. 15 3/85
'lners.
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^
VEGP-FSAR-1 O ik 4 N #0( %
Exceptions are as follows:
1.
Paragraph 3.2.1, Level Items.
As an alternate to the requirements for pa kaging and containerizing items in storage to co trol contaminants (items 4 and 5), VEGP may choose a torage atmosphere which is free g
of harmful contaminants in concentrations that could
,( produce damage to stored items.
Similarly (for item 7), VEGP may delete t. need for caps and plugs with 9
OM i
an appropriate storage atmosphere.
VEGP will protect g h' weld-end preparations -%osed; however, VEGP may delete
/
g ' h e use of caps and plugs for itemy A rcd zu g
My" 4
ois,
M apprg ri e a etcre m w.op w.e ---Prior to inctcil m s C
weld-end preperction, n li -Le inue M ed fo& eni damsge 15 whichay hc ec cccurrcd auMnc c'
Oc.
These clarifications apply to items 4, 5,
or 7 and paragraph 3.4, Methods of Preservation.
1.9.39 REGULATORY GUIDE 1.39, REVISION 2, SEPTEMBER 1977, HOUSEKEEPING REQUIREMENTS FOR WATER-COOLED NUCLEAR POWER PLANTS 1.9.39.1 Regulatory Guide 1.39 Position This guide describes an acceptable method of complying with regulations with regard to housekeeping requirements for the control of work activities, conditions, and environments at water-cooled nuclear power plant sites.
1.9 VEGP Position Con 539.2 ernA o Coa:=r.m.
- w :-
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3g h tc chcpter 17 for furthcr discuccien-1.9.40 REGULATORY GUIDE 1.40, MARCH 1973, QUALIFICATION TESTS OF CONTINUOUS-DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF WATER-COOLED NUCLEAR POWER PLANTS 1.9.40.1 Regulatory Guide 1.40 Position The procedures for conducting qualification tests of continuous-duty motors installed inside the containment of.
water-cooled nuclear power plants which are specified by IEEE Std. 334-1971, IEEE Trial-Use Guide for Type Tests of Amend.
9 8/84 1.9-40 Amend. 15 3/85 l}
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VEGP-FSAR-1 Continuous-Duty Class I Motors Installed Inside the Containment of Nuclear Power Generating Stations, are generally acceptable and provide an adequate basis for complying with the
~.
Amend.
9 8/84 1.9-40a Amend. 15 3/85 t-. r.:
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VEGP-FSAR-1 qualification testing requirements of Criterion III of Appendix B to 10 CFR'50, to verify adequacy of design for service under the most adverse design conditions, subject to the qualifications identified in the guide.
1.9.40.2 VEGP Position To the extent practicable, the procedures for conducting qualification tests specified by IEEE Std. 334-1974 are used to supplement the requirements of IEEE 323-1974 for Class 1E motors inside the containment.
Refer to Regulatory Guide comparison 1.100 and section 3.11.
1.9.41 REGULATORY GUIDE 1.41, MARCH 1973, PPEOPERATIONAL TESTING OF REDUNDANT ONSITE ELECTRIC POWER SYSTEMS TO VERIFY PROPER LOAD GROUP ASSIGNMENTS 1.9.41.1 Regulatory Guide 1.41 Position As part of the initial preoperational testing program, and also after major modifications or repairs to a facility, those onsite electric power systems designed in accordance with Regulatory Guides 1.6 and 1.32 (Safety Guides 6 and 32) should be tested as follows to verify the existence of independence among redundant onsite power sources and their load groups.
1.
C.1 The plant electric power distribution system, not necessarily-including the switchyard and the startup and auxiliary transformers, should be isolated from the.offsite transmission network.
Preferably, this isolation should be effected by direct actuation of the undervoltage-sensing relays within the onsite system.
2.
C.2 Under the conditions of C.1 above, the onsite
[
electric power system should be functionally tested successively in the various possible combinations of power-sources and load groups with all de and onsite ac power sources for one load group at a time completely disconnected.
Each test should include injection of simulated accident signals, startup of the onsite power source (s) and load group (s) under test, sequencing of loads, and the functional performance of the lohds.
Each test should be of sufficient duration to achieve stable operating ~.-
conditions and thus permit the onset and detection of adverse conditions which could result from improper 1.9-41 Amend. 9 8/84
.:t: ~ r~; ^
~ T= = ~ = -; - L. _. :
VEGP-FSAR-1
. assignment of loads, e.g.,
the lack of forced cooling of a vital device.
3.
C.3 During each test, the dc.and onsite ac buses and related loads not under test should be monitored to verify absence of voltage at these buses and loads.
1.9.41.2 VEGP Position VEGP is committed to follow this regulatory guide.
Refer to
- section 14.2 for further discussion.
- 1.9.42 REGULATORY GUIDE 1.42 Withdrawn.
1.9.43 REGULATORY GUIDE.1.43, MAY 1973, CONTROL OF STAINLESS STEEL WELD CLADDING OF LOW-ALLOY STEEL COMPONENTS 1.9.43.1 Regulatory Guide 1.43 Position This guide describes acceptable methods for implementing requirements with regard to the selection and control of the welding process used for cladding ferritic steel components with austenitic stainless. steel to restrict practices that could result in underclad cracking.
1.9.43.2 VEGP Position Qualification testing is performed on any high-heat input welding process (such as the submerged-arc wide-strip welding process or the. submerged arc 6-wire process) used to clad coarse or fine grained SA-508 Class 2 material.
This test follows the recommendations of this guide.
Production welding is monitored by the fabricator to ensure that essential variables remain within the limits established by the qualification.
If the essential variables exceed the qualification limits, an evaluation is performed to determine if the cladding is acceptable for use.
Where Westinghouse permits the use of submerged-arc strip' process on SA-508 Class 2 material, a two-layer technique is used to minimize
' intergranular cracking.
Refer ~-to paragraph 5.2.3.3.2.
4 1.9-42 Amend. 9 8/84
. =i=.:=.;. -
- = =. :. : =
2._.:. -._.
VEGP-FSAR-1 Clarifications are as follows:
gg 1.
Para raph 1.2 of ANSI N45.2.6-1978, Applicabilit.
VEGP p connel who approve preoperational, sta up, and test resulto nd who direct or supervise th conduct of individual preo rational, startup, and o rational
' tests shall be qu
'fied in accordance Ith the VEGP
' ~ ~. -
~~~~
position to Regulato Guide 1.8 in eu of being k
gCf-qualified to ANSI N45..
as allo d by Regulatory Position C.1 of Regulator ui 1.58 Rev.
1.
VEGP dphgkka personnel who perform NDEs e 1 meet the requirements of ANST " Recommended Pr ice SNT-TC-1A" in 15 accordance with reg ory positio C.2 of Regulatory Guide 1.58 Rev.
For nuclear oper< ing personnel, VEGP shall ap.
the requirements of t
's guide to quality con ol inspection personnal; ho ver, for personne performing calibration, installat' n checko s,
or routine! surveillance, the requi ments of this guide shall not apply since, as stated in ction 1.
of ANSI N45.2.6, the requirements of this qui are tional for these personnel.
m l9 2.
Paragraph 2.5 of ANSI N45.2.6-1978, Physical.
VEGP will implement the requirements of this section with the stipulation that, where no special physical characteristics are required, none will be specified.
The converse is also true; if no special physical requirements are stipulated by VEGP, none are considered necessary.
GPC employees receive an initial physical examination to assure satisfactory physical condition; GPC management shall determine which personnel are required to receive an annual examination.
3.
Paragraph 3 of ANSI N45.2.6-1978, Qualification.
Same clarification as 1.
15 9
Amend.
9 8/84 1.9-51 Amend. 15 3/85
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INSERT Pg. 1.9-51
/, CM7u% Paragraph 1.2 of ANSI N45.2.6 - 1978, Applicability.
VEGP personnel who approve preoperational, startup, and test results and who direct or supervise the conduct of individual preoperational, startup, and operational tasks shall be qualified in accordance with the VEGP position to Reg. Guide 1.8 in lieu of being qualified to ANSI N45.2.6 as allowed by regulatory position C.1 of Reg.
Guide 1.58, Rev.
1.
For Nuclear Operations, VEGP elects to apply the requirements of this guide to quality control inspection personnel.
In lieu of regulatory position C.2 of Reg.
Guide 1.58, Rev. 1, which states that the 1975 version of SNT-TC-1A is acceptable, VEGP shall use SNT-TC-1A-1980 for qualifying personnel performing nondestructive inspection, examination, or testing in accordance with ANST Recommended Practice No. SNT-TC-1A.
The 1980 version shall be used in order to be consistent with the requirements of the ASME Boiler and Pressure Vessel Code.
For personnel performing calibration, installation checkouts, or routine surveillances the requirements of this guide will not be applied, as allowed by Section 1.2 of ANSI N45.2.6 - 1978; personnel performing these functions shall either meet the minimum education and experience recommendations of ANSI N18.1 - 1971 or will complete a qualification program which will demonstrate their ability to perform their job functions.
FSAR Table 13.1.3-1 designates the minimum education and experience recommendations for plant personnel, while FSAR Section 13.2.2 describes the training programs which demonstrate the ability of plant personnel to perform their job functions.
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VEGP-FSAR-1
-1.9.59 REGULATORY GUIDE 1.59, REVISION 2, AUGUST 1977, DESIGN BASIS' FLOODS FOR NUCLEAR POWER-PLANTS 1.9.59.1 Regulatory Guide 1.59 Position This' guide describes the conditions resulting from the worst site-related flood probable at a nuclear power plant that safety-related structures, systems, and components must be designed to withstand and retain capability for cold shutdown and maintenance thereof.
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9.8/84 e
1.9-51a Amend. 15 3/85 l
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VEGP-FSAR-1 1.9.59.2 VEGP Position Conform.
See subsections 2.4.3, 2.4.4, and 3.4.1 for a detailed discussion on flood protection.
1.9.60 REGULATORY GUIDE 1.60, REVISION 1, DECEMBER 1973, DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS 1.9.60.1 Regulatory Guide-1.60 Position This guide describes an acceptable procedure for defining response spectra for seismic design of nuclear power plants.
The Newmark-Blume-Kapur design spectra curves for free field ground accelerations are endorsed.
1.9.60.2 VEGP Position Conform.
Refer to subsection 3.7.1 for discussion on this subject.
1.9.61 REGULATORY GUIDE 1.61, OCTOBER 1973, DAMPING VALUES FOR1 SEISMIC DESIGN OF NUCLEAR POWER PLANTS 1.9.61.1 Regulatory Guide 1.61 Position This guide delineates acceptable damping values to be used in the elastic model dynamic seismic analysis of Seismic Category 1 structures, systems, and components.
1.9.61.2 VEGP Position Conformance.is discussed in subsections 3.7.B.1 and 3.7.N.1.
1.9.62 REGULATORY GUIDE 1.62, OCTOBER 1973, MANUAL' INITIATION OF PROTECTIVE ACTIONS 1.9.62.1 Regulatory Guide 1.62 Position-This guide describes an acceptable method for complying kith the requirements of Section 4.17 of IEEE Std. 279-1971 for.
including the means for manual initiation of protective
~'
actions.
1.9-52 Amend. 9 8/84
- b~:::L =.~T~~r:1 ~ ~~-.- - - - L -, _- ~
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VEGP-FSAR-1 required independent design verification, nor should the independent design verification be construed to dilute or replace the clear responsibility of supervisors for the quality of work performed under their supervision."
3.
C.3 In the first sentence of Section 8 of N45.2.11-1974, the word " effecting" should be inserted before " design changes" for clarification.
- Further, the term " approved design document" should be construed to mean " design output" (Section 1.4) approved by the organization performing the design.
4.
C.4 Sections 4.3, 4.4, and 4.5 of N45.2.11-1974 concern the establishment of procedures for the preparation and control of drawings, specifications, and other design documents.
These sections list typical subjects to be covered by such procedures.
One of the subjects to be covered is "nonconformances."
The NRC staff considers the "nonconformances" listed in these sections to be nonconformances with procedural requirements.
TEus in Section 4.3, item (11), "Nonconformance with drawing requirements," should be construed to mean "Nonconformance with procedures for the preparation and control of drawings;" in Section 4.4, item (7),
"Nonconformance with specification requirements,"
should be construed to mean "Nonconformance with procedures for the preparation and control of specifications;" and in Section 4.5, item (7),
"Nonconformance with design document requirements,"
should be-construed to mean "Nonconformance with procedures for the preparation and control of design documents."
1.9.64.2 VEGP Position i
Alternatives and clarification to the text of ANSI N45.2.11-1974 are contained in WCAP-8370, Rev. 9A, Table 17-1.
C.1 Conform for the design and construction QAP.
VEGP operations QAP conforms with this guide, which endorses l15 ANSI Standard N45.2.11-1974, as described below.
Clarifications are as follows:-
l9 1.
For operations phase modification activities, VEdP shall control these activities under the requirements Amend.
9 8/84 1.9-57 Amend. 15 3/85
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VEGP-FSAR-1 described in section 17.2.
VEGP shall conform with the regulatory position in that quality assurance programmatic / administrative requirements shall apply to these modification activities even.though such requirements may not have been in effect originally.
Technical requirements associated with modifications shall be made at least to technical requirements of the FSAR.
2.
ANSI N45.2.11-1974, Section 11, Audits.
The GPC audit program will be implemented in accordance with the 9
requirement set forth in section 17.2, the Technical Specifications, and the VEGP position on. Regulatory Guide 1.144.
Exceptions are as follows:
1.
Paragraph C.2(1).
For the exceptional circumstances in which the designer's immediate supervisor is the only technically qualified individual available, this review can be conducted by the supervisor, provided that:
a.
The other provisions of the Regulatory Guide are satisfied.
b.
The justification is individually documented and approved in advance by the supervisor's management.
c.
Quality assurance audits cover frequency and effectiveness of the use of supervisors as design.
verifiers to guard against abuse.
The VEGP QAP is described in chapter 17.
1.9.65 REGULATORY GUIDE 1.65, OCTOBER 1973, MATERIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS 1.9.65.1 Regulatory Guide 1.65 Position This guide defines acceptable materials and testing procedures for implementing criteria with regard to reactor vessel closure stud bolting for light-water-cooled reactors.
1.9-58 Amend. 9 8/84 a t ' =.~.2 = : = r 2~-
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VEGP-FSAR-1 1.9.'65.2 VEGP Position VEGP conforms with this guide except for two points.
The use of modified SA-540 Grade B24 material as specified in ASME Boiler and Pressure Vessel Code Case 1605 is not specified in the guide but-is.used by Westinghouse.
The use of this Code Case has been approved by the NRC via Regulatory Guide 1.85.
The maximum limit of 170 ksi ultimate tensile-strength is not explicitly specified by Westinghouse as required by the guide.
Westinghouse-does specify fracture toughness of 45 ft/lb and 25 mils lateral expansion as required by the ASME Code and 10 CFR 50, Appendix G.
These requirements also result in strength levels below the maximum limit, as demonstrated by the actual stud material properties for VEGP which are listed in tables 5.3.1-4 and 5.3.1-5.
1.9.66 REGULATORY GUIDE 1.66 l
Withdrawn.
1.9.67-REGULATORY GUIDE 1.67, OCTOBER 1973, INSTALLATION OF OVERPRESSURE PROTECTION DEVICES 1.9.67.1 Regulatory Guide 1.67 Position This guide describes an acceptable method for the design of piping for safety valve and relief valce stations which have open discharge systems with limited discharge pipes, and which have inlet piping that neither contains u water seal nor is subject to slug flow of water upon discharge of the valves.
1.9.67.2 VEGP Position Conform.
Refer to paragraph 3.9.B.3.
1.9.68 REGULATORY GUIDE 1.68, REVISION 2, AUGUST 1978, INITIAL TEST PROGRAMS FOR WATER-COOLED NUCLEAR POWER PLANTS 1.9.68.1 Regulatory Guide 1.68 Position
-This guide describes the general scopo and depth of initial test programs acceptable to the NRC for light-water-cooled nuclear power plants:
The guido provides a representative 1.9-59 Amend. 9 8/84
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VEGP-FSAR-1 listing of the plant structures, systems, components, and the design features and performance capability tests that should be demonstrated during the initial test program.
The guide also provides information on inspections that will be performed by the NRC and provides guidance on the preparation of procedures for the conduct of initial test programs.
1.9.68.2 VEGP Position Conform as follows, except for Appendix A, Section 5, Subsections V, KK, CC, and MM:
Tests V and MM will not be performed as the results obtained will be similar to the results obtained during a turbine trip from 100 percent power which will be performed.
The closure times for the MSIVs will be verified during hot functional and preoperational testing.
The loss of or bypass of feedwater heaters test (test KK) will not be performed as results will be similar, but less severe than those obtained during the load swing test, section 15 14.2.8.2.27.
The gaseous and liquid radwaste systems (test CC) will be tested as decribed in the gaseous waste processing system preopera-tional test abstract (paragraph 14.2.8.1.48) and the liquid waste processing system preoperational test abstract (paragraph 14.2.8.1.49).
Performance of these tests during the power ascension test phase would produce the same results as testing during the preoperational test phase.
1.9.68.3 Regulatory Guide 1.68.2, Revision 1, July 1978, Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants 1.9.68.3.1 Regulatory Guide 1.68.2 Position This guide describes an initial startup test program acceptable to the NRC for demonstrating hot shutdown capability and the potential for cold shutdown from outside the control room.
1.9.68.3.2 VEGP Position Conform; the initial startup test program is described in chapter 14.
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Amend. 9 8/84 1.9-60 Amend. 15 3/85
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VEGP-FSAR-1 1.9.68.4 Regulatory Guide 1.68.3, April 1982, Preoperational Testing of Instrument and Control Air Systems 1.9.68.4.1 Regulatory Guide 1.68.3 Position This guide describes a method acceptable to the NRC for verifying that instrument.and control air systems and the loads they supply will operate properly at normal system pressures and to assure the operability of functions important to safety in the event that system pressure is lost, reduced below normal operating level, or increased above the design pressure of the air system components to the upstream-safety valve accumulation pressure.
1 1.9-60a Amend.
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VEGP-FSAR-1 1.9.68.4.2 VEGP Position The instrument air system has no safety design basis as discussed in subsection 9.3.1.
VEGP conforms with this guide with the understanding that the provisions of position C.8 are satisfied as follows:
The ability of the instrument air system to perform its design function will be demonstrated during the instrument air preoperational test described in chapter 14.
Monitoring of the response of each safety-related pneumatic valve upon loss of air occurs during construction acceptance tests for each valve and is a prerequisite test for the preoperational test of the system.
In performing this testing, the air pressure that will 15 be supplied will be equivalent to the air pressure supplied by the instrument air system during normal plant operation, and it will be demonstrated that each valve responds properly (assumes its fail-safe position) for both a simulated sudden loss of air anc' for a gradual loss of air pressure.
Since it is verified, on an individual basis, that each safety-related pneumatically operated valve will assume its fail-safe position, performance of a large-scale loss-of-air test encompassing several branches of the instrument air system is not necessary to verify correct valve response.
1.9.69 REGULATORY GUIDE 1.69, DECEMBER 1973, CONCRETE RADIATION SHIELDS FOR NUCLEAR POWER PLANTS 1.9.69.1 Regulatory Guide 1.69 Position This guide endorses ANSI N101.6-1972 which addresses the design and construction of concrete radiation shields.
1.9.69.2 VEGP Position Not applicable since VEGP uses conventional concrete for shielding, not concrete shields addressed in ANSI N101.6-1972.
1Property "ANSI code" (as page type) with input value "ANSI N101.6-1972.</br></br>1" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..9.70 REGULATORY GUIDE 1.70, REVIS702t 2.,
NOVEMBER 1978, STANDARD FORMAT AND CONTENT c' SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS 1.9.70.1 Regulatory Guide 1.70 Position The purpose of the FSAR is to inform the NRC of the nature of the plant, the~ plans-for its use, and the safety evaluations Amend. 9 8/84 1.9-61 Amerd. 15 3/85 2
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i VEGP-FSAR-1 that have been performed to evaluate'whether the plant can be operated without undue risk to the!! health /and safety of the -
public.
The FSAR is the principal document for the applicant to provide'this information.
The purpose of_this guide is to indicate.the-information-to be provided in the FSAR and to-establish a uniform format acceptable to the NRC for presenting this information.
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VEGP-FSAR-1 1.9.70.2 VEGP Position ij
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Conform as discussed in subsection'1.1.6.
1.9.71 REGULATORY GUIDE 1.71,-DECEMBER 1973, WELDER QUALIFICATION.FOR AREAS OF LIMITED ACCESSIBILITY 1.9.71.1 Regulatory Guide 1.71 Position This guide describes a method acceptable to the NRC for implementing requirements-with regard to the control of welding-for nuclear components.
1.9.71.2 VEGP Position This guide provides guidelines above and beyond requirements of
All welder qualification at VEGP is in conformance with ASME Section IX.
Few welds of limited accessibility are expected to-be encountered.
Reasonable engineering judgment will be used to determine if performance qualification is necessary under simulated access conditions
- for any specific case.
Westinghouse practice does not require qualification or requalification of welders for areas of limited accessibility as described by the guide and has provided welds of high quality.
Limited accessibility qualification or requalification, which are additional to ASME Section III and IX requirements,'is an unduly restrictive requirement for shop fabrication, where the welders' physical position relative to the welds is controlled and does not present any significant problems.
In addition, shop welds of limited accessibility are repetitive due to multiple production of similar components,
-and'such welding is closely supervised.
Refer to section 5.2.3 for further discussion.
1.9.72 REGULATORY GUIDE 1,72, REVISION 2, NOVEMBER 1978, SPRAY POND PIPING MADE FROM FIBERGLASS-REINFORCED THERMO-SETTING RESIN 1.9.72.1 Regulatory Guide 1.72 Position
-This guide describes a method acceptable to the NRC -for designing, fabricating, and testing fiberglass-reinforced thermo-setting resin piping for spray pond applications.
1.9-62 Amend. 9 8/84 W
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VEGP-FSAR-1 Subdivision 5.6 of ANSI N45.2.9-1974.
The National Fire Protection Association (NFPAf No. 232-1975,
" Standard for the Protection of Records," also contains provisions for records protection equipment and records handling techniques that provide protection from the hazards of fire.
This standard,.within its scope of coverage, is considered by the NRC staff to provide an acceptable alternative to the fire protection provisions listed in Subdivision 5.6 of N45.2.9-1974.
When NFPA No. 232-1975 is used, quality assurance records should be classified as NFPA Class 1 records (NFPA No. 23-1975, Chapter 5, Section 5222).
[g k cyMvHS-1.9.88.2 VEGP Position 5(,
The VEGP QAP conforms with the requirements of ANSI N45.2.9-1974) 15
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- Rylatq 6ulde I.88, Rev 2,.
1.9.89 REGULATORY GUIDE 1.89, NOVEMBER 1974, QUALIFICATION OF CLASS 1E EQUIPMENT FOR NUCLEAR POWER PLANTS 1.9.89.1 Regulatory Guide 1.89 Position The procedures described in IEEE Std. 323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, dated February 28, 1974, for qualifying Class 1E equipment for service in light-water-cooled and gas-cooled nuclear power plants are generally acceptable and provide an adequate basis for complying with design verification requirements of Criterion III of Appendix B to 10 CFR 50 to verify adequacy of design under the most adverse design conditions subject to the following:
1.
C.1 Reference is made in IEEE Std. 323-1974, Section 2, 6.3.2(5), and 6.3.5, to IEEE Std. 344-1971, Guide for Seismic Qualification of Class 1 Electric Equipment for Nuclear Power Generating Stations.
The; specific applicability or acceptability of IEEE Std. 344 will be covered separately in other regulatory guides, where appropriate.
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Amend.
9 8/84 1.9-71 Amend. 15 3/85
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VEGP-FSAR-1 2.
C.2 The radiological source term,.for qualification tests in a nuclear radiation environment should be based.on the same source term as that used in Regulatory Guide 1.7 (Safety Guide 7, March 10, 1971) for boiling water reactors (BWRs) and PWRs.
An equivalent source term (i.e.,
100 percent of the noble gases, 50 percent of the halogens, and 1 percent of the remaining solids developed from maximum full-power operation of the core) should be used for high temperature gas-cooled reactors (HTGRs).
The containment. size should be taken into account in each case.
For exposed organic materials, calculations should take into account both beta and gamma radiation.
1.9.89.2 VEGP Position Conform.
See section 3.11.B for information on environmental conditions and design bases for mechanical, instrumentation, and electrical safety-related equipment.
For NSSS equipment, Westinghouse conforms to IEEE Std. 323-1974 by implementation of the final NRC approved version of WCAP-8587.
1.
C.1 See Regulatory Guide 1.100 comparison.
2.
C.2 Conform.
1.9.90 REGULATORY GUIDE 1.90, REVISION 1, AUGUST 1977, INSERVICE _ INSPECTION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES WITH GROUTED TENDONS 1.9.90.1 Regulatory Guide 1.90 Position This_ guide describes bases acceptable to the NRC for developing
(
an appropriate surveillance program for prestressed concrete containment structures with grouted tendons.
1.9.90.2 VEGP Position i
This guide is not applicable since VEGP does not use grouted tendons.
u 1.9.91 REGULATORY GUIDE 1.91, REVISION 1, FEBRUARY 1978, EVALUATIONS OF EXPLOSIONS POSTULATED TO OCCUR ON l
TRANSPORTATION ROUTES _NEAR NUCLEAR POWER-PLANTS 1.9-72 Amend. 9 8/84
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This guide describes methods acceptable.to the NRC for determining whether the-risk of damage due to an explosion on a nearby transportation route is sufficiently high to warrant a-detailed investigation.
1.9.91.2 VEGP Position Conform.
Refer to subsection 2.2.3 for discussion on this subject.
1.9.92-REGULATORY GUIDE 1.92, REVISION 1, FEBRUARY 1976, COMBINING MODAL RESPONSES AND SPATIAL COMPONENTS IN SEISMIC REPONSE ANALYSIS 1.9.92.1 Regulatory Guide 1.92 Position This' guide describes the procedures to be used for combining modal responses of individual modes and the combination of effects.due to the three independent spatial components of an earthquake in seismic analyses of nuclear power plant structures, systems, and components.
y l
1.9.92.2 VEGP Position-Conform with the exception that Westinghouse uses an alternative method of combining modal responses to satisfy Regulatory Guide 1.92, Revision 1, as described in paragraph 3.7.N.2.7.
Refer.to sections 3.7.B and 3.7.N for discussion on this subject.
1.9.93 REGULATORY GUIDE 1.93, DECEMBER 1974, AVAILABILITY OF ELECTRIC POWER SUPPLIES 1.9.93.1 Regulatory Guide 1.93 Position
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This guide describes. operating procedures and restrictions acceptable to the NRC which should be implemented if the available. electric power sources are less than the limiting -
conditions for operation (LCO).
1.9-73 Amend. 9 8/84
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VEGP-FSAR-1 1.9.93.2 VEGP Position
/_
l VEGP will conform with this guide by implementing,the appropriate NRC approved standard Technical Specifications.
Refer to the Technical Specifications for.further discussion.
1.9.94 REGULATORY GUIDE 1.94, REVISION 1, APRIL 1976, QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF STRUCTURAL CONCRETE AND STRUCTURAL STEEL DURING THE CONSTRUCTION PHASE OF NUCLEAR POWER PLANTS 1.9.94.1 Regulatory Guide 1.94 Position This guide describes a method acceptable to the NRC for complying with the quality assurance requirements for installation, inspection, and testing of structural concrete and structural steel during the construction phase of nuclear power plants.
This guide endorses ANSI N45.2.5-1974 as generally acceptable to the NRC as a basis for complying with Appendix B to 10 CFR 50.
1.9.94.2 VEGP Position The extent of conformance with ANSI N45.2.5-1974 for both the 15 operational and construction phases is discussed in paragraph 3.8.3.6.2.C.
Refer to Regulatory Guide 1.55 comparison for a discussion of the standards being used in the placement of concrete in Category 1 structures.
1.9.95 REGULATORY GUIDE 1.95, REVISION 1, JANUARY 1977, l3 PROTECTION OF NUCLEAR POWER PLANT CONTROL ROOM OPERATORS AGAINST AN ACCIDENTAL CHLORINE RELEASE i
1.9.95.1 Regulatory Guide 1.95 Position l
This guide describes design features and procedures that are acceptable to the NRC for the protection of nuclear plant control room operators against an accidental chlorine release.
- Amend. 3 1/84 Amend. 9 8/84 1.9-74 Amend. 15 3/85
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~1.9.114.2.VEGP Position q
a Conform.
Refer to chapter 18 and section 13.5.
1.9.115 REGULATORY GUIDE 1.115, REVISION 1, JULY 1977, PROTECTION AGAINST LOW TRAJECTORY TURBINE MISSILES 1.9.115.1 Regulatory Guide 1.115 Position This guide describes methods acceptable to the NRC for protecting safety-related structures, systems, and components against low-trajectory missiles resulting from turbine failure.
1.9.115.2 VEGP Position Conformance is discussed in paragraph 3.5.1.3.
1.9.116 REGULATORY GUIDE 1.116, REVISION 0-R, JUNE 1976, QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF MECHANICAL EQUIPMENT AND SYSTEMS 1.9.116.1 Regulatory Guid 1.116 Position This guide endorses ANSI N45.2.8-1975 which describes a method acceptable to the NRC for complying with regulations with regard to quality assurance l requirements for installation, inspection, and testing of 5echanical equipment and systems for water-cooled nuclear power plants.
1.9.116.2 VEGP Position Conform for the design and construction QAP.
The VEGP QAP is described in chapter 17.
The VEGP operations QAP conforms with the requirements of ANSI N45.2.8-1975 as it is endorsed by this guide with the following clarifications,;..d cac ptiere.
Elarincatione arc :: f elle crea 1.
Paragraph 2.7, Personnel Qualifications.
VEGP has ~
addressed this requirement in its posit' ion to Regulatory Guide 1.58..
1.9-85 Amend. 9 8/84
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1 VEGP-FSAR-1 2.
Paragraph 7, Records, will be implemented in accordance with the VEGP! position to Regulatory Guide 1.88.
3.
For modification activities, VEGP shall control these activities under the operations QAP as described in section 17.2.
Technical requirements associated with modifications shall be the original requirements or better (e.g.,
code requirements, material properties, design margins, manufacturing processes, and inspection requirements).
4.
Paragraph 1.4, Definitions in this Standard, and definitions which are included in ANSI N45.2.10 will be used as clarified in VEGP position to Regulatory Guide 1.74.
P 5.
Paragraph 2.2, Procedures and Instructions, will be implemented as set forth in section 17.2 and by conformance with the Technical Specifications and VEGP position to Regulatory Guide 1.33.
6.
Paragraph 2.3, Results, will be implemented as set forth in section 17.2 and by VEGP position to Regulatory Guide 1.33.
7.
Paragraph 2.4, Cleaning, will be implemented as set forth in the VEGP position to Regulatory Guide 1.37.
8.
Paragraph 2.5, Receiving, Storage, and Handling, will be implemented as set forth in the VEGP position to Regulatory Guide 1.38.
9.
Paragraph 2.6, Housekeeping, will be implemented as set forth in the VEGP position to Regulatory Guide 1.39.
10.
Paragraph 6, Data Analysis and Evaluation.
Where required the plant shall have procedures for the performance of analyzing test data, but these procedures are not referred to as data processing 9
procedures.
15 Amend. 9 8/84 1.9-86 Amend. 15 3/85 mic'$ -]h Hl Y -- -
s 6-A VEGP-FSAR-1 1.9.123 REGULATORY GUIDE 1.123, REVISION,1, JULY 1977, QUALITY ASSURANCE REQUIREMENTS FOR CONTROL OF PROCUREMENT OF ITEMS AND SERVICES FOR NUCLEAR
' POWER PLANTS 1.9.1'23.1
' Regulatory Guide 1.123 Position
. The requirements that are included in ANSI N45.2.13-1976 for control.of. procurement of items and services for nuclear power plants are acceptable to the NRC staff and provide an adequate basis for complying with the pertinent quality assurance requirements of Appendix B to 10 CFR 50, subject to the
. qualifications identified in the guide.
- 1.9.123.2 VEGP Position
- VEGP conforms during the design and construction QAP except for components purchased from the NSSS vendor which conform except i
for regulatory position C.6b.
The NSSS vendor routinely identifies notification points in procurement documents when applicable.
Such points are not always identified in pre-and
.pos -t award meetings.
However, the required notification / hold
- points-are specified by changes to the procurement documents in-a reasonable time prior to their being accomplished to allow the purchaser the opportunity to witness the event.
- Alternatives and clarifications to the text of ANSI l9 N45.2.13-1976 are' contained in the text of WCAP-8370, Rev.
9A-table 17-1.
The VEGP'QAP is' described in chapter 17.
4 The VEGP operations QAP conforms with the requirements of ANSI N45.2.13-1976 as it is endorsed by this guide with the following clarifications:
r-1.
Paragraph 1.3, Definitions.
With two exceptions, procurement document and QAP-requirements, definitions in this standard and the definitions which are included in the VEGP position to Regulatory 2
Guide 1.74 will be used.
The two exceptions are defined in the VEGP positoin to Regulatory Guide 1.74.
2.
Paragraph 1.2.2, Purchaser's Responsibilities.
Item c is one of the options which may Le used by VEGP to assure quality; however, any of the options given in
~
10 CFR 50, Appendix B, Criterion VII, as implemente'd by section 17.2, may also be used. ~
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Paragraphs 3.2.3, 3.2.4, and 3.2.6.
VEGP does not consider that these paragraphs or vendor qualifications apply for the procurement of off-the-shelf items.
Off-the-shelf items (which include original as well as spare and replacements) are commercial grade items which are:
a.
Not subject to design or specification requirements that are unique to facilities or activities licensed by the Nuclear Regulatory Commission.
b.
Used in applications other than facilities or activities licensed by the NRC.
c.
Ordered from the manufacturer, distributor, supplier, or retailer on the basis of the manufacturer's catalog or product description.
4.
Paragraph 3.3 requires procurement documents to be reviewed prior to bid or award of contract.
The quality assurance review of procurement documents is satisfied through review of the applicable procurement specification and purchase requisition prior to bid or award of contract.
15 5.
Paragraph 4.2, Selection Measures, outlines certain methods acceptable for the selection of suppliers.
GPC's historv of using similar methods has proven adequate in the procurement of items; therefore, VEGP wishes to replace paragraph 4.2(a), (b), and (c) with the following selection methods:
a.
The supplier's quality assurance capabilities as '
determined by a direct survey of his facilities and personnel, and the implementation of his QAP.
b.
Evaluating the supplier's history of providing a product which performs satisfactorily in actual use.
One or more of the following information
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shall be evaluated:
Amend.
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Experience of users of identical or similar products of the same prospective supplier.
(2)
GPC records that have been accumulated in connection with previous procurement actions and product operating experience.
Historical data should be representative of the supplier's current capability.
If there has been no recent experience with the supplier or if he is a new supplier, the prospective supplier shall be requested to submit information on a similar item or service.for evidence of his current capabilities.
(3)
Evaluating the supplier's current quality records supported by documented qualitative and quantitative information which can be objectively evaluated.
This would include review and evaluation of the supplier's QAP manual and procedures, as appropriate, to ensure that the applicable requirements of 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, are met.
(4)
Verification that the supplier holds an active certificate of authorization from ASME to supply or manufacture materials or the items described in the purchase requisition.
(5)
A supplier may be considered acceptable, without a survey, to supply off-the-shelf items.
An inspection shall be performed to 0\\p assure that the correct item was received 0((kh b
and no damage exists.
A suppl r may be considered acceptable to suppl "Q" quality
- commcalty)without a survey.
S items procured m st be produced to a standard and must have defined ratings, such as pressure, temperature /l voltage.
The procurement of 15 such items shall be based upon a documented engineering evaluation.
At receipt, an inspection shall be performed by quality control to verify compliance to the description and that it has been approved for use by an evaluation.
(6)
Verification that the supplier is listed in the current _ Coordinating Agency for Supplier Amend.
9 8/84 1.9-93 Amend. 15 3/85
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VEGP-FSAR-1 Evaluation (CASE) register.
However, the audit report whi6h form 6d the basis for listing the supplier.in the CASE register must be obtained and reviewed for applicability to the procurement.
All deficiencies which could degrade the procured item must be resolved prior to the procurement.
This review shall be documented and, together with the audit, report, be retained.
i Amend.
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1.9-93b Amend. 15 3/85
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VEGP-FSAR-1 (7)
Verification that the supplier's QAP is acceptable to the NRC under the Licensee Contractor and Vendor Inspection Program and has been satisfactorily implemented as evidenced by a confirming letter from NRC-IE.
Alternately, the acceptability of the supplier's QAP and its implementation may be determined by reviewing the NRC's inspection report (s) for applicability to the procurement.
All such reviews shall be documented, and together with the NRC's inspection report (s) retained.
8.
Paragraph 5.2 shall be applicable only for new procurement; it shall not be applicable for spares or replacements parts that do not change the original design.
l15 "aragrnph ; 3. - Pranunrd Evaluutivn.
CFC will vuufenTi
^
aith an alternate paragraph.:hich readc:
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unucual circ"mntances (e.c _~raplacemcat gutLu are ~~
-neudud_Lv.greclude tha dcvelvement or some unsare oT*
unduaAcubiu vuudi Livu at a nucluut fuvility),
a- -
prercard eval"'*4mn of the cupplice ck,11 he _.
performuu ao desciibcd in VCOE eva2Liva hera4" 15 l
Cl kOT Paragraph 6.2, Planning and Coordination.
GPC will conform with the exception that the NSSS vendor routinely identifies notification points in procurement documents when applicable.
Such points-are not always identified in pre-and post-award I
meetings.
However, t,he required notification / hold Amend.
9 8/84 1.9-94 Amend. 15 3/85 JJ/rg-
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VEGP-FSAR-1 points are specified by changes to the procurement documents in a reasonsbleitime prior to their being accomplished to allow therpurchaser the opportunity to witness the event.
l() )dI.
Paragraph 7.5, Personnel Qualification.
Refer to VEGP Position to Regulatory Guide 1.58 (Clarification 1).'
l ( )df.
Paragraph 8 provides guidance for purchaser review and disposition of vendor nonconformances.
GPC, as purchaser, satisfies this requirement by requiring, as a minimum, deviations to specifications that cannot be brought into conformance with specification 15 requirements, prior to shipment of the material to be submitted to GPC for approval.
Such deviations, when approved by the purchaser, are required to be submitted along with shipment of the material.
Amend.
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Regulatory Position C.3 indicates that the purchaser should verify the implementation of the supplier's corrective action systems when such a system is required, but this verification need not be included I
as part of the purchaser's corrective action i
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Amend.
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i 1.9-96 Amend. 15 3/85
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VEGP-FSAR-1 1.9.132 REGULATORY GUIDE 1.132, REVISION 1, MARCH 1979, SITE INVESTIGATIONS FOR FOUNDATIONS OF NUCLEAR POWER PLANTS 1.9.132.1 Regulatory Guide 1.132 Position Paragraph C of the guide addresses site investigations for foundations.
1.9.132.2 VEGP Position VEGP site investigation conforms with the requirements of this regulatory guide.
Refer to section 2.5 for discussion on this subject.
l10 1.9.133 REGULATORY CUIDE 1.133, REVISION 1, MAY 1981, LOOSE-PART DETECTION PROGRAM FOR THE PRIMARY SYSTEM OF 3
LIGHT-WATER-COOLED REACTORS 1.9.133.1 Regulatory Guide 1.133 Position This guide describes a method acceptable to the NRC for implementing requirements with respect to detecting a potentially safety-related loose part in light-water-cooled reactors during normal operation.
1.9.133.2 VEGP Position Refer to subsection 4.4.6.4 for a discussion of the digital metal impact monitoring system (DMIMS) which is the VEGP loose part monito g system.
VEGP conforms to Regulatory Guide LO 1.133, with ollowing clarifications to Provision C.6.
Upon receipt of an alarm, VEGP will investigate the alarm to confirm 15 if a loose part exists.
An engineering evaluation of confirmed 13 loose parts will be performed to determine whether a reportable condition has occurred as described in 10 CFR 50.72 and 10 CFR 50.73.
VEGP shall follow the requirements of 10 CFR 50.72 and 10 CFR 50.73 for providing prompt notification and followup reporting of the confirmation of a loose part.
Amend. 10 9/84 Amend. 13 1/85 1.9-103 Amend. 15 3/85
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VEGP-FSAR-1 (This page has intentionally been left blank.)
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.VEGP-FSAR-1 Containment Isolation Provisions for Fluid Systems, are generally acceptable and provide an adequate basis for complying-with the pertinent containment isolation requirements of Appendix A to 10 CFR 50, subject to the qualifications.
identified in the guide.
1.9.141.2' VEGP Position
-VEGP conforms as discussed in subsection 6.2.4.
1.9.142 REGULATORY GUIDE 1.142,' OCTOBER 1981, REVISION 1, SAFETY-RELATED CONCRETE STRUCTURES FOR NUCLEAR POWER PLANTS (OTHER THAN REACTOR VESSELS AND CONTAINMENTS) 1.9.142.1 Regulatory Guide 1.142 Position This guide endorses the procedures and requirements described in American Concrete Institute (ACI) 349-76 subject to the
! qualifications provided in this guide.
1.9.142.2 VEGP Position ACI 318-71 is used in lieu of ACI 349-76.
Refer to subsection 3.8.4 for discussion on this subject.
1.9.143 REGULATORY GUIDE 1.143, REVISION 1, OCTOBER 1979, DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND COMPONENTS INSTALLED IN LIGHT-WATER-COOLED NUCLEAR POWER PLANTS 1.9.143.1 Regulatory Guide 1.143 Position 1
l This guide furnishes design guidance acceptable to the NRC regarding seismic and quality group classification and quality assurance provisions for radioactive waste management systems, structures, and components.
1.9.143.2 VEGP Position Conform, with the following clarifications:
~
e Radioactive waste management systems, structures, and 5
components are classified in table 3.2.2-1.
1.9-109 Amend. 9 8/84
- b:: ~:
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T L~~
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- VEGP-FSAR-1 15 e
ACI 318-71 is.used for design of concrete structures in lieu of ACI 318-77.
See section 11.4 for further discussion.
1.9'.144 REGULATORY GUIDE 1.144, REVISION 1, SEPTEMBER 1980, AUDITING OF QUALITY ASSURANCE PROGRAMS FOR NUCLEAR POWER PLANTS-1.9.144.1 Regulatory Guide 1.144 Position The requirements that are included in ANSI /ASME N45.2.12-1977 for auditing QAPs for nuclear power plants are acceptable to the NRC staff and provide an adequate basis for complying with the pertinent quality assurance requirements of Appendix B to 10 CFR-50, subject to the qualifications identified in the guide.-
- 1.9.144.2 VEGP Position VEGP conforms with Regulatory Guide 1.144 with the following clarification.
VEGP does not conform to the latest revisions of the following ANSI standards:
ANSI N45.2, ANSI N45.2.9, and ANSI N45.2.10.
VEGP conforms to ANSI N45.2-1971, ANSI N45.2.9-1974 and ANSI'N45.2.10-1973.
Conformance to Regulatory 7 l15 Guides.l.28, 1.74, and 1.88 is indicated in this section.
The VEGP quality assurance program is described in chapter 17.
1.9.145 REGULATORY-GUIDE 1.145, AUGUST 1979, ATMOSPHERIC DISPERSION MODELS FOR POTENTIAL ACCIDENT CONSEQUENCE ASSESSMENTS AT NUCLEAR POWER PLANTS 4
1.9.145.1 Regulatory Guide 1.145 Position t
This guide identifies acceptable methods for:
e Calculating atmospheric relative concentration (x/Q) values.
e Determining x/Q values on an overall site basis.
e Determining x/Q values _on a directional basis.
[
Amend.
5 4/84 l
Amend.
7 5/84 l
Amend. 15 3/85 Amend.
9 8/84 1.9-110
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ese iam
- - =
- ed am
1 i
VEGP-FSAR-1 e.
-Choosing.x/Q values to be used in evaluations of the
- types of events described in'Pegulatory Guides 1.3 and 1.4.
1.9.145.2 VEGP Position
~
Conform.
Refer to subsection'2.3.4.
1.8.146.fREGULATORY GUIDE 1.146, AUGUST 1980, QUALIFICATION OF QUALITY ASSURANCE PROGRAM AUDIT PERSONNEL FOR NUCLEAR POWER PLANTS 1.8.146.1 Regulatory Guide 1.146 Position This guide describes a method acceptable to the NRC for complying with regulations with regard to qualification of.QAP audit personne1'for nuclear power plants.
1.8.146.2 VEGP' Position-Conform.
The QAP is discussed in chapter 17.
1.9.147 REGULATORY GUIDE 1.147, REVISION 2, FEBRUARY 1982, INSERVICE INSPECTION CODE CASE ACCEPTABILITY, ASME SECTION XI,. DIVISION 1 1.9.147.1 Regulatory Guide 1.147 Position This regulatory guide-lists those Section XI ASME code cases that are generally acceptable to the NRC for implementation in the ISI of light-water-cooled nuclear power plants.
1.9.147.2-VEGP Position Conform.
Refer to section 6.6 for further discussion.
_ l.9-111 Amend. 9 8/84-
- :.::.=:..: = = r : =.- -
VEGP-FSAR-1 1.9.148 REGULATORY GUIDE 1.148, MARCH 1981, FUNCTIONAL SPECIFICATION FOR ACTIVE VALVE ASSEMBLIES IN SYSTEMS IMPORTANT TO SAFETY IN. NUCLEAR POWER PLANTS 1.9.148.1 Regulatory Guide 1.148 Position This guide delineates a procedure acceptable-to the NRC for implementing regulations with respect to the detailed specification of information pertinent-to defining operating requirements for active valve assemblies in light-water-cooled
. nuclear power plants.
1.9.148.2 VEGP Position Conformance is addressed in table 3.9.B.3-10.
15 1.9.149 REGULATORY GUIDE 1.149, APRIL 1981, NUCLEAR POWER PLANT SIMULATORS FOR USE IN OPERATOR TRAINING 1.9.149.1 Regulatory Guide 1.149 Position
'This regulatory guide describes a method acceptable to the NRC for specifying the functional requirements of a nuclear power plant simulator to be used for operator training.
1.9.149.2 VEGP Position Conform, except with regard to Section 5.4(3) of ANSI /ANS 3.5-1981.
GPC will conduct periodic simulator performance testing in response to plant changes.which affect training.
Since digital software does not drift or change, retesting of
. verified simulator ~ response will not be conducted.
Through the use of startup test data, operator observations supported by plant transient charts and plant change notices, the VEGP
' simulator will be modified and tested to match plant response.
See' subsection'13.2.1 for additional discussion.
Amend. 9 8/84 1.9-112 Amend. 15 3/85
~.
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D INSERT Response to Q260.62. A.6 Of3n Itam IC6 The inspection and testing requirements (including maintenance requirements) for the containment building polar bridge crane are discussed in FSAR Section 9.1.5.
INSERT Response to Q260.61 A.7 FSAR Section 17.2.18,
" Audits",
lists the areas to be audited.
The accident-related meteorological collection equipment operations, calibration, maintenance and reliability will be included in the audit plan for environmental monitoring, plant chemistry and environmental technical specifications.
INSERT Response to Q260.61 A.8 Radiation Protection Systems, including necessary equipment and supplies will be included in the audit plan for Health Physics and Radiation Protection. This audit will verify that the equipment is present, functional, and calibrated; that there are appropriate procedures, they are being used, and they are working effectively; that involved personnel are adequately trained.
INSERT Response to Q260.62 B6 and 7 The inspection and testing requirements (includirg maintenance requirements) for refueling machine and fuel transfer system are discussed in FSAR Section 9.1.4.
" 'n
- = tion, x; ;; int;n n;; p;; form;d n the c;fuc? #n; xdine o..J Tsci i...n
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INSERT Response to Q260.61 C.2 The effectiveness of plant shielding area radiation survey results, and frequency of area radiation surveys will be included in the audit plan for Health Physics and Radiation Protection.
Temporary shielding to reduce personnel exposure in high radiation areas will also be included in the audit plan for Health Physics and Radiation Protection.
The application of additional shielding (temporary or permanent), including seismic and stress design considerations will be included in the audit plan for design change and plant modification control.
INSERT Response to Q260.61 C.18 The Iodine Moni toring System Internal to the plant will be included in the audit plan for Health Physics and Radiation Protection.
The Iodine Monitoring System External to the plant, including Stack Gas Monitoring Systems and Liquid Effluent Discharge Monitoring systems will be included in the audit plans for Environmental Monitoring, Plant Chemistry, and Environmental Technical Specification and the audit plan for the emergency plan and procedures, as applicable.
This audit will verify that the equipment is present, functional, and calibrated; that there are appropriate procedures, they are being used, and they are working effectively; that involved personnel are adequately trained.
TABLE 3.'2.2-1 -(SilEET 82 OF.97)
-(d)
(f)
. (g)
(J)
(a)
(b)
(c)
- VECP, (e)
Codes and Principal Principal System tocation Sou rce o f - Qua l i ty Sa fe ty Se i sm ic Standa rd s Construc- (h).. (i)
Environ-Sarety mental
( k)-
and cpoponents Un i_t I uni t;_g _Supp ly_ _G roup Class Cateqqry posigna_. tor 11o_n.,_ Code. 9-Jul (tglyptd Dgpigna gor Cg_mments ROD CONIROL POWLR SYSi t M 1.
NA 1
1 E
mfg '
Y Y
switchgear 2.
Other swi tch-W NA 6
2 E
mfg N
N gear FULL LENGTH ROD CONTROL SYSTEM 1.
Rod control W
NA 6
1 J
mfg
' f4 gg equipment ROD POSillON INDICAll0N SYSTEM 1.
Rod position W
NA 6
1
.J mfg N
N instebeenLaLion k
RADIAllON MoNiinftlNC SYSilM (9
1.
Safety-related W
NA 1
1 J
mfg Y
Y portions 2.
Nonsafety-W NA 6
1 J
mfg N
N
,y related, seismic (n
Category 1 portions 8
3.
Other portions W
NA 6
2 J
mfg N
N
[SF ACIUATION SYSTIM 1.
All portions W
NA i
1 J
mfg Y
Y REACIOR INSTRUMENIAlloN 1.
All portions W
NA 1
1 J
mfg Y
Y inputting to reactor pro-tection
- p 2.
Other portions W
NA 6
2 J
mfg N
N 3
O REACTOR CONTROL SYSTEM D
f 1.
Protection-re-W NA 1
1 J
mfg Y
Y lated portions cc 2.
Other portions W
NA 6
2 J
mrg N
N 4\\
co
.s=
r UEGP-FSAR-3 TABLE 3.2.2-1 (SHEET 97 OF 97)
Position CMEB 9.5-1, Appendix A, attached to Nuclear l3 Regulatory Commission (NRC) Standard Review Plan 9.5.1.
w.
The quality assurance program to be applied to radioactive waste management systems is described in Regulatory Guide 1.143.
x.
The Seismic Category 1 fire protection standpipe system serves no safety function but is classified as project 3
class 313 to ensure the implementation of a Seismic Category 1, ASME III-3 design and installation.
7 ENERAL NOTES 1.
For systems under the Westinghouse scope of supply, all piping and all manual valves 2 in..
and smaller are supplied by Bechtel, except for the reactor coolant loop piping, the
{
pressurizer surge line, the pressurizer relief piping 1
Complex, reactor vessel bottom mounted instrument tubing, I
reactor vessel head vent piping to refueling disconnect i
flange, and reactor vessel seal leak detection leakoff I
appurtenance.
2.
Hangers and supports for Seismic Category 1 systems and components are designed as Seismic Category 1.
In general hangers and supports for Seismic Category 2 piping, cable tray, and ducting in Seismic Category 1 buildings are designed to maintain their structural integrity under the 12 postulated earthquake conditions; however, exceptions to this requirement are permitted when it is demonstrated that j
their failure will not adversely affect adjacent Seismic Category 1 equipment or systems.
3.
All "Q" listed coatings are assigned a project f
classification of 02C.
Q listed coatings are not l
seismically qualified but will not fail in a manner that l
would compromise the function of safety-related equipment
/
in the event of an earthquake since they are applied to i
Seismic Category 1 structures.
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Amend.3 1/84 Amend.12 12/84
VEGP-FSAR-17 responsibilities are accomplished through the organization shown in figure 17.2.1-1 and are discussed in the following paragraphs.
GPC's executive vice president-power supply regularly assesses the scope, adequacy, and compliance of the OQAP to 10 CFR 50, Appendix B, through frequent meetings with the general manager-quality assurance and radiological health and safety (GMQA), review of safety review board activities, and assessment of quality assurance audits.
An annual assessment of the effectiveness of the OQAP is performed by the general manager-quality assurance, and results are reported to the executive vice president-power supply.
17.2.1.1.1 Safety Review Board The safety review board shall be appointed by the executive vice president-power supply for the purpose of advising him in matters related to nuclear power plant safety.
Board members shall have competence in various fields of engineering technology and be f amiliar with nuclear safety, environmental, and regulatory requirements.
The safety review board has access to the advice and services of technical specialists within GPC 8
and outside expertise as necessary.
Board member responsibilities and authorities are described in the plant Technical Specifications.
- de<f 046)
+keSe\\;ocd6se<+ 09 17.2.1.1.2 Nuclear Operations The senior vice president-nuclear power,gand the,vice president and general manager-nuclear operations are responsible to the executive vice president-power supply for the implementation of the quality assurance program for plant operations at all nuclear power generating plants in the GPC system.
The general manager-Vogtle nuclear operations reports to the vice president and general manager-nuclear operations, through th manager-rucic;. vpu - - --.a.
-Thz ---^; r nuciaar operation a o o m.c the rerpensibilitic; of-tic: pre ident and 9eumial-rc.onagcr numim;r operati:n in hi absence.
The manager-nuclear planning and control reports to the vice president and general manager-nuclear operations.
Some of the responsibilities of this position are to provide long range planning and scheduling of maintenance work to be performed at GPC nuclear plants /[E3't provide the long range manpower plan for GPC nuclear plants.
g A669Pd4'\\(8 W
s 17.2.1-3 Amend. 8 7/84 L
hk hYRYQ fI 6 lH89 0 VEGP-ESAR-17 The manager-nuclear training reports to the vice pr sident and general manager-nuclear operations.
A few of the responsibilities of this position are to provide the GPC nuclear plants with training programs to ensure compiliance with andAInstitute Nuclear Regulatory Commission (NRC) ra "{ *is e
of Nuclear Power Operations standards
.C t
.sure that nuclear 8
o operations personnel have the education, training, and skills to safely and efficiently operate and maintain the plants.
The manager-nuclear engineering and chief nuclear engineer is responsible to the vice president and general manager-nuclear operations for day-to-day monitoring of plant activities, special projects as required, licensing support, and interfacing with appropriate companies and organizations in the areas of nuclear fuel management, procurement, and reprocessing.
17.2.1.2 Plant Organization The plant staff (as described in chapter 13) will perform safety-related activities in accordance with written, approved procedures.
Quality assurance requirements will be included in detailed plant procedures.
The superintendents and supervisors in the VEGP organization will be responsible for implementation of the OQAP for activities under their purview.
The general manager-Vogtle nuclear operations will regularly assess the workload of all departments involved in the OQAP to ensure that a suf ficient number of personnel are available for complete and efficient implementation.
17.2.1.2.1 General Manager-Vogtle Nuclear Operations end-A :istent riant " mager;c) D 0 p. j C C g.y [,'j g e
The general manager-Vogtle nuclear operations is responsible to the vice president and general manager-nuclear operations for 8
all activities at the VEGP, including implementation of the OQAP requirements with the exception to controls that are assigned to the quality assurance department.
The general manager-Vogtle nuclear operations is also responsible for the safe, reliable, yj{
and efficient operation of VEGP.
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' 17.2.1.2.g3 Quality Control Supervisor NY,Y[Nb 3,
The quality control supervisor is responsible to the general manager-Vogtle nuclear operations for administration and implementation of an effective quality control inspection program at VEGP.
Quality control specialists report to the quality control supervisor.
The quality control supervisor, or his representative, is involved in day-to-day safety-related 17.2.1-4 Amend. 8 7/83
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VEGP-FSAR-17 activities.
These activities include work planning meetings, plant review board meetings, and routine staff meetings.
8 Quality control personnel have procedural authority to stop work and to control further processing, use, or installation of nonconforming items.
17.2.1.2.g'fsuperintendentofMaintenance The superintendent of maintenance is responsible for effective implementation of the OQAP of applicable mechanical, electrical, and inst:cument and controls maintenance of plant equipment, systems, or structures.
17.2.1.2.g5 Superintendent of Operations The superintendent of operations has the responsibility to ensure that the plant is operated in accordance with approved procedures and license requirements.
17.2.1.2./[pSuperintendentofPlantEngineeringServices The superintendent of plant engineering services is responsible for providing engineering-related technical services in support of VEGP operations.
He shall assist, as required, in the procurement of equipment, materials, and services and shall coordinate review, approval, and closeout of design changes.
17.2.1.2.p'r7 Superintendent of Administration The superintendent of administration has the responsibility of maintaining the plant documentation files, procedure and change logs, and distribution control for plant-originated procedures, plant review board minutes, and correspondence.
17.2.1.2.#$ Materials Supervisor The materials supervisor is responsible for material and equipment control and material requisitioning to maintain plant
. stock levels.
He is also responsible for receiving, handling, and storing materials.
17.2.1.2./GSuperintendentofNuclearTraining The superintendent of nuclear training is responsible for the development and implementation of training programs for the 17.2.1-5 Amend. 8 7/84 L
VEGP-FSAR-17 plant staff.
He ensures that the VEGP training programs are adequate to provide qualified personnel.
17.2.1.2.f' Superintendent of Regulatory Compliance The superintendent of regulatory compliance will advise plant management on matters concerning compliance with requirements of operating license, Technical Specifications, approved plant procedures, Security Plan, Emergency Plan, OQAP, and applicable federal, state, and local regulations.
17.2.1.2.p6 Superintendent of Health Physics and Chemistry 8
The superintendent of health physics and chemistry is responsible for the health physics and chemistry program.
He is also responsible for the as-low-as-reasonably-achievable prcgram at VEGP.
& 1 M o rt l'1. 2..).2,I 2 Snpeden/ed el-59lneeny Liaw1 17.2.1.2.p{k Plant Review Board g jg,1,l,'A,13 Peace 6Mgd $d75fW 66 N eM W -
The plant review board shall be comprisea o respoHsible plant department personnel and shall advise the plant manager on matters pertaining to safety-related activities.
17.2.1.3 Quality Assurance Department The quality assurance department, under the direction of the GMQA, verifies implementation of the OQAP.
The qualifications for the GPC quality assurance department personnel meet the requirements of Regulatory Guide 1.146, Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants, as described in section 1.9.
The minimum qualifications for the GMQA and the VEGP quality assurance manager are that they hold an engineering or equivalent degree and have a minimum of 5 years experience in the areas of engineering, field construction, or plant operation.
Two of these 5 years must be in the field of quality assurance, of which at least 6 months must be in the field of 8
nuclear quality assurance experience.
The size of the quality assurance organization is based on meeting the cor.mitment for audit coverage required by the Technical Specifications and on experience gained at Hatch Nuclear Plant.
The approximate size of the site quality assurance organization is 10 technical people for Unit 1 17.2.1-6 Amend. 8 7/84 L
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The CPC poucr cupe-y enginccring cnd c;rviccc dcp2rtnent, under the directicr of the vice-prccident cnd chief engineer poucr cupply engineering 2nd cervices -ic rc rosa;Lic for providing angi eering ea *^cFnic21 cupport t f M 'vericuc pcuer cupply crganicatienc.
During th^ apa*^ tion vf VCOP, the pcucr-cupply enginee;ing cnd cerviccc departr.cnt ic recr^reible for *ke man 2genent of dccign engineering cupport, the qu21 fic2tien of
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u 3 SCS is the architect engineer for 7EGP during opera : ion.
Ghe 9Pe pcucr cupply engineering and cervicec der" *+ cr <
Vog t 1 c_
project m2n27er scrius ao the intcrf2ce betueen the CPC nu-Lear powe; departncnt end SCE for coordin2" ^n additi e nj[cti er
'nd dire cf euw_..smiing cupport provided by 300.
-Ir SCS provides quality assurance support, including audits, supplier qualification and surveillances, and engineering procedure reviews.
SCS also administers for VEGP the contract for 8
engineering services provided by BPC.
Activities within the SCS work scope are governed by the SCS VEGP Operational Support Policy and Procedures Manual and the SCS Engineering Policy and Procedures Manual.
These procedures are reviewed and concurred with by the SCS quality assurance organization.
The GPC GMQA performs or causes to be performed audits of these functions.
BPC is under' contract to SCS to provide architect / engineering services.
The work scope includes plant design, development of purchase recommendations for equipment and materials, administration of purchase orders resulting from SCS-developed purchase recommendations such as the nuclear steam supply I
system, and support of the SCS supplier surveillance functions l
by providing procurement surveillance services for selected Q-list items.
Activities within the BPC work scope are governed by the BPC VEGP Nuclear Quality Assurance Department Procedures Manual and Vogtle Project Engineering Procedures Manual.
The GPC GMQA performs or causes to be performed audits of these procedures and functions.
17.2.1.5 Vogtle Project The vice president and general manager-Vogtle Project is responsible to the senior vice presidegt-ugl g
r for implementation of the OQAP for perform 1n cohs Y n
activities at en operating nucle 2r pcuer p_ w VE GP anits, 0379V 17.2.1-10 Amend. 8 7/84
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VEGP-FSAR-17 17.2.4 PROCUREMENT DOCUMENT CONTROL The quality requirements for the control of documents prepared by Georgia Power Company or its designated agents, for safety-related components, materials, and services are consistent with the provisions of Regulatory Guide 1.123, Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants, as described in section 1.9.
The vice president and general manager-nuclear operations is responsible for procurement planning of VEGP.
To satisfy the quality assurance requirements for procurement document control, the following specific requirements shall be implemented.
Measures shall be established to ensure:
A.
Procedures are established delineating the sequence of actions to be accomplished in the preparation, review, approval, and control of procurement documents.
The general manager-Vogtle nuclear operations, in conjunction with the vice president-engineering and 8
p f g 4*,4 services, is responsible for the preparation, review, approval, and control of procurement documents.
B.
A review and concurrence of the adequacy of quality assurance requirements stated in procurement documents is performed by qualified personnel knowledgeable in the quality assurance requirements.
Plant procedures will define training requirements for ensuring knowledge of quality assurance requirements.
This review is to determine that all quality assurance requirements are correctly stated, can be inspected and controlled, and provide adequate acceptance and rejection criteria and that the procurement document has been prepared in accordance with quality assurance program requirements.
C.
The person performing the review will be an individual independent of the procurement document originator.
Documented evidence of the review and approval of procurement documents is provided and available for i
verification.
D.
Procurement documents identify those 10 CFR 50, Appendix B, requirements that must be complied with and described in the supplier's quality assurance program.
This quality assurance program or portions thereof shall be reviewed and concurred with by the general manager-Vogtle nuclear operations, in 3
conjunction with the vice president-engineering and COANeM 0kN A services, and qualified personnel knowledgeable 17.2.4-1 Amend. 8 7/84 L
VEGP-FSAR-17 Review of documentation recuired by the purchase crder l-
$ t x 1, f 4 6 g e # c W 6 n 1 6 rs f is performed by p 1' I
m vle w S e c h;0 vio B.
Inspections, tests, and other specified records attesting to the acceptance of materials, equipment, and components are completed and available at VEGP prior to installation or use.
C.
Materials, equipment, and components are inspected and judged acceptable in accordance with predetermined inspection instructions prior to installation or use.
Items accepted and released are identified as to their inspection status prior to forwarding them to a controlled s orage area or releasing them for installation or further work.
6 Nonconforming items are clearly identified, controlled, and segregated where practical, until proper disposition is made.
For commercial (off the shelf) items where specific quality assurance controls appropriate for nuclear applications cannot be imposed in a practical manner, special quality verification requirements shall be established to provide the necessary assurance of the item.
The quality assurance department will audit the control of purchased materials, equipment, and services to ensure proper implementation of the requirements.
See subsection 17.2.18 for a description of the quality assurance audit system.
0385V 17.2.7-3 Amend. 8 7/84
VEGP-ESAR-17 and test equipment and the criteria for determining when a test or inspection is to be performed are contained in the established plant procedures.
The quality control supervisor is responsible to plant management for administering and implementing tests and inspections assigned to the quality control department.
He is also responsible for analyzing the results of applicable inspections performed at VEGP.
Inspections are performed by quality control specialists (qualifications discussed in subsection 13.1.3) who are independent of the individuals performing the activity being inspected.
Plant management ensures that quality control specialists and any other personnel performing inspection activities are qualified, and qualification records are documented and kept current.
The 8
training and qualification of these personnel will be verified through written examinations, proficiency testing, or oral examination.
Documented evidence of qualification for these personnel will indicate the function which the individual is qualified to perform.
Proficiency of these personnel is maintained by retraining and reexamining in accordance with applicable codes, standards, and/or procedures.
The criteria for determining the size of the quality control organization is based on known or anticipated tasks which require quality control inspection and other functions based on experience gained at Hatch Nuclear Plant.
The approximate number of technical personnel planned for the ality control organization during normal plant operation is which may be augmented by contractor personnel during outages 18 l
The quality control supervisor and quality control specialists have written stop-work authority, including the authority to prevent equipment or systems from being returned to service if l
the activity was not performed in accordance with an approved l
procedure, specification, or drawing.
If specified inspection hold points / witness points, requiring witnessing or inspecting by an inspector and beyond which work is not to proceed without inspector approval, are necessary, the specific hold points will be indicated in the work procedure.
If at these checkpoints tha activity is found to be unsatisfactory, further processing of the activity is suspended until the problem is resolved.
il' Procedures requiring inspection criteria will be reviewed by l
qualified personnel from quality control prior to performance of 3
l, the work.
This review will include a check for the need for
_J l,
inspection, identification of inspection personnel, and
'{
documentation of inspection results.
Further, this review will ensure that inspection requirements, methods, and acceptance i
criteria have been identified.
Quality control involvement in day-to-day work planning meetings and staff meetings should l
17.2.10-2 Amend. 8 7/84 i
VEGP-FSAR-17 17.2.12 CONTROL OF MEASURING AND TEST EQUIPMENT Measures for the control of measuring and test equipment are consistent with the position of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), as described in section 1.9.
Provisions for the control of applicable measuring and test equipment require that:
A.
Procedures shall be established which describe the calibration technique, calibration frequency, maintenance and control of measuring and test instruments, tools, gauges, fixtures, reference standards, transfer standards, and nondestructive test equipment to be used in the measurement and inspection of safety-related components, systems, and structures.
B.
Measurement and test equipment shall be uniquely identified and have traceability to calibration test data.
C.
Measuring and test instruments shall be calibrated and maintained at specific intervals, based on the required accuracy, purpose, degree of usage, stability characteristics, and other conditions affecting the measurement.
D.
Measuring and test equipment shall be calibrated on or before the designated due date or remove] f rom service.
un t.;l e o re ective. o.ction ca.n b e t.a. ken.
E.
When measuring and test equipment is found to be out of calibration, an investigation shall be conducted and documented to determine the validity of previous inspections performed and the acceptability of those items previously inspected.
Procedures are established to ensure that measurements made with measuring and test equipment found to be out of ca.libration will be evaluated to determine the need for reinspection.
' Calibrating instruments shall have known, valid F.
relationships to a nationally recognized standard or natvesl phy sica.1 co nsta.nt. If no na.tionst, st ands.cd er na.to ra.1 physica.1 constant exists the be. sis o f ca. libra. tion will be clocu mented j
a.nd ayprov e4 by the respan.s*We dspar 6 meat..soperIn du -
8 dent ae a hi hee ievel ofon**A3*'"u t 3
G.
Records will be maintained which indicate the complete status of all items under the calibration system and Amend. 7 5/84 17.2.12-1 Amend. 8 7/84 e
r
~
VEGP-FSAR-17 reflect the last and future calibration dates or the last calibration date and frequency, if applicable.
H.
Calibration facilities used for calibrating sensitive and close tolerance measuring and test equipment shall provide an environment that is suf ficiently controlled to allow the measuring device to be evaluated and calibrated to its required accuracy.
Reference standards will have an uncertainty (error) requirement of no more than one-fourth of the tolerance of the measuring and test equipment being calibrated; a greater uncertainty will be acceptable and egoipment and standards not meeting this requirement will be documented and approved by the responsible department superintendent or by a higher level of management.
Comparison standards used in calibration of reference standards will have a tolerance egoat to or sma.Ilar tha.n the esf erence stadned a nd will be rescen.ble to she st,e.nd a eds haased in she Ha.tlana.1 Burea.o o f S tanda rds,. Qe calihta[l ort SfandahAS R5in8 COMfaf160'1 SfendatdS Will be SWffoMtd by of te4ctenct :
cepoefs and dam _ sheefS,/
%G.gett Fication The superintendent or mainte~ nance is re W nsible to the general manager-Vogtle nuclear operations for developing, approving, and implementing procedures and instructions to establish a control and calibration program.
He also approves calibration procedures.
The administrative procedures that control the calibration program are reviewed by the plant review board and approved by the general manager-Vogtle nuclear operations.
The quality control department is responsible for certifying the control of measuring and test equipment for compliance to plant procedures.
Quality control will verify control of measuring and test equipment during inspections of work activities.
Effectiveness of the program is ensured through periodic audits performed under the quality assurance audit system that is described in subsection 17.2.18.
0390V Amend. 7 5/84 17.2.12-2 Amend. 8 7/84
l Eu.bsur3 E
Cpac,14am 167 Open Item 107 - CRDR PROGRAM PLAN CLARIFICATIONS.
1.
Modifications to the generic ERGS in order to develop plant specific E0Ps will be documented in the VEGP step document (Procedure No.
11894-C) as part of the E0P development.
In the task analysis, a task data form will be generated and the task will be analyzed for all plant specific E0P steps.
2.
VEGP has no E0Ps that are not covered by the ERGS.
3.
All operator information and controls used to accomplish the tasks outlined in the ERGS /EOPs will be documented and analyzed. The re-quired function to assure safe plant conditions are derived from the ERGS at the task level. The functional requirement for control room equipment is to support the performance of those required tasks.
Re-quired characteristic of specific control room equipment are inherent in the task elements. The task elements needed to perform each task must be associated with specific equipment, and cannot be developed independent of the actual control room in a design review. The spe-cific requirements can vary with each task element and are therefore being evaluated at that level by one or more CRDR team members during the control room walkthru-instrument verification. Applicable require-ments such as range, accuracy, and trending capability are evaluated for each task element and any deficency documented as a Human Engineer-ing Discrepancy on the Task Data Form.
4.
While the required tasks in the TASK ANALYSIS are developed from the Westinghouse Emergency Response Guidelines, the task elements in each task are developed from the corresponding plant specific Emergency Operating Procedures.
In this way the plant specific means to accom-plish each task is identified in the task analysis. All EOPs are de-veloped from ERGS.
5.
Control Room Survey items not incorporated in our checklists are docu-mented in INPO 83-042 appendix B-H.
The Control Room Design Review Team will review those items and document how they have been addressed in the CRDR or why they were inappropriate or not applicable. Human Engineering Discrepancies will be developed for all applicable or ap-propriate guidelines which are not met.
I
.(
I
.o OTHER REVIEWER CONCERNS - REPLY 1.
Georgia Power is committed to achieving a well designed control room.
The CRDR is the p imary accountability of the review team leader and all other personnel are available as needed to complete the CRDR.
2.
All CRDR procedures, evaluations, and recommendations are reviewed and approved by all review team members or alternates.
3.
The CRDR team considers the impacts of any control room changes in the development of HED resolutions.
4.
The CRDR team will document review of improvement verifications.
5.
The diverse nature of potential HED's precludes the prior development of rigid criteria for evaluation and resolution.
The multi discipline CRDR team, will draw on their collective experience to evaluate problems and develop optimum solutions.
These decisions are developed by concensors and documented in CRDR records.
.