ML20117D750
| ML20117D750 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 04/29/1985 |
| From: | Bailey J GEORGIA POWER CO. |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| GN-588, NUDOCS 8505100207 | |
| Download: ML20117D750 (22) | |
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- Georgia Power Company .
. Ro:na 2, Box 299A
- Wayncsboro, Georgia 30830 Telephone 404 554 9961 404 724-8114 Southern Company Services, Inc.
Post Office Box 2625 Birmingham, Alabama 35202 Telephone 205 87040t t ,
, i April 29, 1985 Director of Nuclear Reactor Regulation File: X7BC35 Attention: Ms. Elinor G. Adensam, Chief Log: GN-588 Licensing Branch #4 .
Division of Licensing-U. S. Nuclear Regulatory Commission Washington, D.C. 20555
- NRC DOCKET NUMBERS 50-424 AND 50-425 l 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.
If your staff requires any additional information, please do not hesitate to contact me.
Sincerely, 9J. A. Bailey
, Project Licensing Manager JAB /sm Enclosure xc: D. O. Foster ;
R. A. Thomas J. E. ' Joiner, Esquire '
B. W. Churchill, Esquire M. A. Miller
!* B. Jones, Esquire '(w/o enclosure) {
L. T. Gucwa G. Bockhold, Jr.
T. Johnson (w/o enclosure) !
'D. C. Teper (w/o enclosure) -
i L. Fowler Vogtle Project File 0205m 8505100207 850429 PDR i\
E ADOCK 05000424 PDR ; i '
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4 4 ,
4 ATTACHMENT
- OPEN ITEMS INDEX
- Open Item Enclosure Remarks 105 A Response to NRC tele-
! conference on 4/22/84 concerning Regulatory Guide 1.94 and position l C.2 of Regulatory Guide 1 1.58. This information will appear in Amendment 17.
51 B Clarification of i GN-514, 4/3/85, 1 submittal and includes j information request i telecopied to SCS on j 4/16/85.
ASB- C Additional information l requested by NRC letter dated 2/14/85. This 7
submittal completes t response to letter l request and will be included in Amendment
- i 16. i j' 24 D Revision to Q210.48 on allowable leakage rate
] for pressure isolation i valves. This change will appear in Amendment 17.
j 97 E Clarification of change l concerning STA's absent
, .from duty for a period of thirty days.
103 ' F Additional information concerning steam I
generator tube rupture j issue. 'This i information will appear j in keendment 16. ,
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0205m
BAl6 L. o.s v A G /)
VEGP-FSAR-1 g 1.9.93.2 VEGP Position 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 Positi_on 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.
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1.9.94.2 VEGP Position
/N.3 ER T* l. 9. q1/, 2 Thc catcnt of conformancc with ANCI N45.2.5 1074 for both-the cperatien21 2nd cerctructier p'acer in diccucced ia-paragraph- 15 3.0.3.0.2.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 1.9.95.1 Regulatory Guide 1.95 Position This guido 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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, Insert 1.9.94.2
) VEGP conforms with requirements of ANSI N45.2.5-1974 as it is endorsed by Regulatory Guide 1.94, Revision 1, for both the construction and operations '
phases with the following clarifications
- 1. Paragraph 1.5, Referenced Documents. !
- This paragrah states that for documents referred to in ANSI N45.2.5-1974, that the specific issue or edition of these documents that
} is required is specified in the Appendix to ANSI N45.2.5. Also, l Paragraph 8 states that when standards referred to in ANSI N45.,2.5-1974 are superseded by an approved revision, then the revision shall apply l
for the VEGP, the specific issue- or edition of documents and standards referenced in ANSI N45.2.5-1974 that shall apply is addressed within the t
appropriate sections of the VEGP FSAR, 1
1 2. Paragraph 3.2, Materials Suitability (3.2.1).
1 This paragraph, Table A, states reinforcement shall be tested for i physical properties per (ASIM A615) ASTM A370. VEGP conforms to this i guidance in that the certification of compliance provided by the vender ensures materials have been tested for physical properties (ASTM A615) j per ASTM A370.
- 3. Paragraph 4.8, In-process Tests on Concrete and Reinforcing Steel.
- This paragraph, Table B, requires in-process compressive strength tests i shall be performed daily on grout in accordance with ASTM C109. VEGP i tests batch plant grout daily in accordance with ASTM C109, while
! non-shrink grout is tested 'in accordance with CRD621-82A prior to its initial use on the job site.
- 4. Paragraph 4.8, In-process Tests on Concrete and Reinforcing Steel.
l This paragraph, Table B, requires that fly ash and pozzolans be checked j for physical properties (ASTM C618) in accordance with ASTM C311 every j 200 tons. VEGP specifications require this testing to be performed every 1000 tons except for loss on ignition and sieve No. 325 which are
, performed every 200 tons.
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! 5. Paragraph 4.9, Mechanical (Caldweld) Spliced Testing.
j (4.9.2) This paragraph requires visual inspection on completed splices <
' shall be performed only af ter the splices have cooled to ambient
- temperatures. The VEGP specification does not require splices'to be ,
l cooled to ambient temperature prior to inspection. . Caldwelds are
< inspected af ter they have cooled such that it will allow the inspection ,
! to be performed without any danger of burns.
i 6. Paragraph 5.5. , Welding.
This paragraph requires inspection of structural steel welding to be performed in accordance with the provisions of AWS D1.1, section 6 i entitled " Structural Welding Code." The visual acceptance criteria is :
{ in accordance with AWS D1.1-75 with the clarifications and modifications [
described on the next page.
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In order to designate the specific set of visual acceptance criteria applicable for a weld, structural steel weld joints within the jurisdiction of AWS D1.1 are classified into the following categories:
o Category A - Structural steel joints which are a part of the main building frame and those joints which connect miscellaneous structural steel with the main building frame.
o Category B - Miscellaneous structural steel joints, not covered in Category A, but provide auxiliary support or framing for systems, components, and equipment (e.g., supports for cable tray and KVAC ductwork, miscellaneous stiffeners and bracing, etc.).
The base metal adjacent to a weld joint of dissimilar categories shall meet the visual inspection acceptance criteria of the category under which that base metal falls.
A. AWS D1.1, paragraph 3.1.4, is clarified as follows:
For category A and B joints, the fillet leg dimension may underrun the nominal fillet size by 1/16 inch provided the undersize length does not exceed 10 percent of the weld length. For flange to web I
joints, the undersize may not be within two (2) flange thicknesses of the weld end.
For Category A and B joints, where an intermittent weld is specified, a continuous weld of the same size is acceptable.
B. AWS D1.1, paragraph 3.6.1, is modified as follows:
For Category A and B joints, the faces fo fillet welds may be slightly convex, flat, or slightly concave as shown in Figure 3.6A, B, and C, with none of the unacceptable profiles shown in Figure 3.6.D. Convexity height may not exceed 1/8 inch.
C. AWS D1.1, paragraph 3.6.4, is modified as follows:
l For Category A joints, undercut not exceeding 1/32 inch for the full l 1ength of the weld is acceptable. For members welded from both l
sides, the cumulative undercut depth includes the sum of both sides if the undercuts on both sides are directly opposite to each other.
l l For Category B joints, undercut not exceeding 1/32 inch for the full l 1ength of the weld is acceptable. Undercut greater than 1/32 inch but not exceeding 1/18 inch is acceptable provided the width is l greater than the depth, and the undercut does not exhibit an acute Intersection at its root. The cumulative length of 1/16 inch l
undercut shall not exceed 50 percent of the total weld length. For members welded from both sides, the cumulative undercut depth or length includes the sum of both sides of the undercuts on both sides are directly opposite to each other.
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t D. AWS D1.1, paragraph 3.6.6, is modified as follows:
For Category A and B - joints, overlap / rollover may not exceed 1/8 inch.
E. . AWS D1.1, paragraph 3.10.1, is clarified as follows:
For Category A and B joints, " Trap" slag on the root and/or end of
' fillet and partial penetration welds is acceptable.
For Category A joints, loose spatter is not acceptable. Tightly adhering spatter is acceptable, if it will not interfere with - -
subsequent NDE (when required).
L For Category B joints, spatter is not a cause for rejection.
F. AWS D1.1, paragraph 4.4, is modified as follows:
-For Category A joints, are strikes shall be blended with the surface.
For Cateogry B Joints, arc strikes are acceptable provided the craters do not contain cracks and the material specification minimum gauge thickness is not violated. !
G. AWS D1.1, paragraph 8.15.1.3, is modified as follows:
Underfill of a weld is not acceptable. A crater is.not considered as an underfill.
For Category A joints, underfilled grove weld craters are acceptable provided the depth of underfill is 1/32 inch or less. Underfilled single-pass fillet weld craters are acceptable provided the crater length is less than 10 percent of the weld length. On multi-pass fillet weld, crater depth of 1/32 inch or less is acceptable.
For Category B joints, underfilled groove weld craters are acceptable provided the' depth of underfill is 1/16 inch or less.
Underfilled single-pass fillet weld craters are acceptable provided ,
the crater length is less than 10 percent of the weld length. On l multi-pass fillet welds, crater depth of 1/16 inch or less are acceptable.
H. AWS D1.1, paragraph 8.15.1.5 , is clarified as follows:
For Category.A joints, the sua diameters of porosity shall not l exceed 3/8 inch in one (1) linear inch of weld and 3/4 inch in 12 linear inches of weld.
For Category B' joints, piping porosity is not a cause for rejection if the major axis is equal to or less than 1/16 inch. For porosity greater than 1/16 inch, the sum of the diameters shall not exceed 3/8 inch in any linear inch of weld nor 3/4 inch in any'12 inches of weld.
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VEGP-FSAR-3 -
C. The extent of compliance with ANSI N45.2.5-1974 is as fcllcWa: DeScR GED iM PA R. A 6MA/0 // /, f. 9 7, 2 ,
. Prep 12 cement inspection ^f ra"" vara fneme s
- 2. spection of concrete placement.
l l 3. Insp tion of concrete curing.
- 4. Inspectio and finishing of re irs.
- 5. Records kept r ANSI N45. 1973.
- 6. Definitions (per SI 5.2.10-1973).
- 7. Concrete quality .d a eptance criteria con-formance with A 318-71.
- 8. Cadweld sp 'ces conformance 'th Regulatory Guide 1. .
- 9. ANSI 45.2-1971, N45.2.2-1972, N45. 3-1973,
- N4 .2.9-1973, and N45.2.10-1973 compl d with as l eferenced. , -)
7 inenactinn eersonnel cualified per ANSI
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N4 5'. 2 . 6- 1973 . A i
3.8.3.6.3 Special Construction Techniques There are no special construction techniques used in the construction of the internal structures.
l 3.8.3.7 Testing and Inservice Inspection Requirements A formal program of testing and inservice surveillance is not required for the internal structures. Tests and inspections for the RCS component supports are discussed in subsection 5.4.14.
3.8.3.8 Standard Review Plan Evaluation l
The Standard Review Plan specifies ACI 349, augmented by Regulatory Guide 1.142, as the acceptable code for design of I concrete structures. The Standard Review Plan also specifies the load combinations that would result from the use of ACI 349, I as modified by Regulatory Guide 1.142. VEGP design is ba' sed on .
ACI 318-71 and is in conformance with the load combinations 3.8.3-12 Amend. 1 11/83 i
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F VEGP-FSAR-1 Clarifications are as follows: gg REPLACE W IT H ATT6c n W^r T-
- 1. Paragraph 1.2 of ANSI N45.2.6-1978, Applicability.
P personnel who approve preoperational, star p, and tes esults and who direct or supervise the onduct of indivi i preoperational, startup, and o rational tests sha e qualified in accordance th the VEGP position to R latory Guide 1.8 in eu of being qualified to ANS N45.2.6 as allo d by Regulatory Position C.1 of Reg tory Gui 1.58 Rev. 1. VEGP personnel who perform h s all meet the requirements of ANST " Recommended Pra e No. SNT-TC-1A" in 15 accordance with regul ry po ' tion C.2 of Regulatory Guide 1.58 Rev. 1. or nuclear erating personnel, VEGP shall appl e requirements o his guide to quality contr inspection personnal; 'ever, for personnel rforming calibration, installa 4on checkou , or routine surveillance, the requi. ments of this ide shall not apply since, as stated in tion 1.' of ANSI N45.2.6, the requirements of this guide re ptional 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 clar:.fication as 1.
15 s
le Amend. 9 8/84 1.9-51 Amend. 15 3/85
INSERT Pg. 1.9-51
- 1. 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. Regulatory position C.2 of Reg. Guide 1.58, Rev. 1, states that the 1975 version of SNT-TC-1A is acceptable for the qualification of personnel performing nondestructive examinations.
In lieu of this, VEGP shall use the requirements of the 1980 version of SNT-TC-1A for qualifying personnel performing nondestructive inspection, examination, or testing, and in addition VEGP shall supplement these requirements by replacing the "shoulds" contained in the 1980 version with "shalls" except as follows:
(a) The "should" in the last sentence of Section 4.3 (3) of SNT-TC-1A - 1980, which refers to a NDE level III being qualified to instruct and examine HDT personnel, is not replaced with a "shall". For the VEGP, a NDE level III individual shall be responsible for the training and examination of NDT personnel; while conducting of actual training and examination may be delegated to qualified individuals appointed by the level III individual.
' (b) The "should" in the second sentence of Section 8.4.3 of SNT-TC-1A - 1980 remains as a "should", as allowed by Section 8.4.2. VEGP elects to use a composite grade based on the simple average of the examinations; this shall be prescribed in VEGP's written practice.
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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8 Endesure.
CLARIFICATION OF PREVIOUS VEGP RESPONSE TO NRC STAFF OUESTIONS ON VOGTLE FSAR SECTION 6.4.
CONTROL ROOM HABITABILITY (OPEN ITEM 51)
1.0 BACKGROUND
In a telephone conversation between VEGP and the NRC staff on April ll, 1985, the NRC expressed concern that the assumptions used in VEGP's control room toxic gas hazard evaluation were inconsistent with the recommendations of USNRC Regulatory
., Guide 1.78. Specifically, the staff's technical objections to VEGP's modeling approach appeared to focus on two issues:
- 1. VEGP utilized a toxic gas infiltration model that j took credit for the dilution of C12 gas in a i portion of the control building before infiltrating into the control room. The staff stated that an unpublished NRC document limits the credit that can be taken for dilution of gas in a volume that
- surrounds the control room to a maximum factor of two. VEGP's model provides a control building dilution factor of approximately 100.
- 2. VEGP's model assumes that all of the contaminated outside air entering the control building ic forced in by the control building HVAC intake fans. These fans continue operating following C12 gas detection and control room isolation. The staff's contention is that VEGP should have also assumed additional inleakage through the control building outer walls associated with a 1/8 inch water gauge (w. g.) differential pressure across these walls.
- 2.0 CLARIFICATION OF VEGP's PREVIOUS RESPONSE VEGP's earlier written response to the NRC staff's questions regarding the control room habitability analysis outlined the various sources of conservatism in the VEGP model that we believe resulted in conservatively high predictions of the C12 gas concentration inside the control room. That opinion remains unaltered. However, to more thoroughly address the staff's specific concerns, the following clarifying remarks are provided.
I Issue 1: Dilution of Chlorine Gas Within the Control Building The dilution which occurs as C12 gas passes through rooms in the control building was modeled in exactly the same manner as dilution within the control room. If C12 entering the control room directly from outside air becomes diluted within the free volume of the control room, then the same phenomenon will occur if the gas were introduced into some other room. We do not understand the technical basis underlying the staff's position that dilution within the control building (or any 1
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4 portion thereof) should be limited to a factor of two. We
, surmise that the staff may be referring to the 2-fold dilution l
{ credit allowed by Standard Review Plan Section 6.4 for control l rooms with physically separated outside air intakes. If this is in fact the source of the confusion, the following points ,
should help to clarify the issue: 1 i
- i. 1. VEGP does have two physically separated control roon j_ outside air intakes. They are situated-
] approximately 150 feet apart. Under the postulated worst case C12 release scenario (catastrophic rupture of a 1-ton C12 cylinder 187 meters from j the closest control room outside air intake; wind i blowing directly from the release point towards that 4
intake; Pasquill G stability), one of the control i room outside air intakes would be exposed to peak centerline concentrations of the C12 cloud..
j Simultaneously, the second intake would see j virtually zero ppm C12 For this scenario, the 1 intake exposed to the peak centerline concentration r i of C12 would isolate. The other intake would not isolate because its chlorine detector would not be
- j. exposed to an activating concentration (2 ppm by j volume) of the gas. -
l 2. If this postulated worst case accident scenario were l
to actually occur, C12 gas entering the control room through the contaminated outside air intake ,
i (prior to isolation) would undergo a two-fold i dilution by the uncontaminated' air from the second i intake. After isolation of the first intake, the i
second intake would continue to draw in 1500 cfm of
{' uncontaminated outside air. The important points to note regarding this sequence of events are as follows:
- a. VEGP's model does not take any credit for the l two-fold dilution of C12 gas that would occur i by virtue of the dual intake configuration.
i such credit would have produced a substantially
{ smaller predicted C12 concentration inside
} the control room at the time when operators are
- assumed to have donned protective breathing i equipment (120 seconds following an alarm from j the chlorine detector).
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- b. Because one outside air intake would continue to deliver 1500 cfm of uncontaminated outside air to the control room following isolation of the contaminated outside air intake, the
- control room would exist at a positive pressure relative to the adjoining rooms within the I control building. Consequently, the not flow of C12 -contaminated air after control room
- isolation would be outward; i.e., from the i
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control room to the control building, rather than the reverse. VEGP's model conservatively assumes a 1/8 inch w.g. differential pressure ;
I driving contaminated air 1R12 the control room ;
i from the adjoining rooms -- exactly the i opposite of what would actually occur.
Additionally, VEGP took no credit for dilution
, of the presumed inleaking air by the 1500 cfm of uncontaminated air supplied by the remaining
.(unisolated) control room outside air intake. l l
Issue 2: Inleakage of Contaminated Outside Air Through the Outer Walls of the Control Building.
Another of the staff's criticisms of the VEGP C12 gas i
- infiltration model is that VEGP omitted consideration of the l additional inleakage of contaminated outside air through the outer walls of the control building caused by a 1/8 inch w.g.
j differential pressure across those walls. Keeping all other ;
i modeling assumptions fixed, adoption of this assumption would i l produce higher predicted control building C12 concentrations j versus time, and correspondingly higher control roon j concentrations.
2
- While we concede that it may have been more realistic to assume
! a 1/8 inch w.g. differential pressure across outer walls of the l control building rather across the control room outer walls, we i continue to believe that our current analysis provides a
, conservative upper bound on C12 concentrations that would
! actually occur within the control room for the postulated worst case onsite C12 gas release. Our reasons are as follows:
1 i 1. At any instant in time, a maximum of only two outer walls of the control building (those facing the oncoming wind) could be under a positive pressure.
! The remaining walls would be under a negative ,
i pressure. Hence, differential pressure could not be t j a driving force causing entry of contaminated l outside air into the control building except on a maximum of two of the control building outer walls.
j 2. The windspeet necessary to create a 1/8 inch w.g.
differential pressure across the outer walls of the j control building is in excess of 7 meters per second
! (~ 15 1/2 miles per hour). Windspeeds of 7 meters per second or greater occur less than 5 percent of j the time at the Vogtle site. At the 0.5 meter per j second windspeed producing the highest predicted 2-minute integrated C12 concentration inside the control room, the differential pressure across the control building outer walls is on the-order of 10-5 inch w.g. At this windspeed, the only t significant mode of entry of contaminated outside '
air into the control building would be vii the
- control building NVAC intake system. , i 4 .
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< 3. ,Any C12 gas entering the control building through !
- - outer wall inleakage would have to pass through multiple layers of rooms. prior to reaching the control room. Dilution would occur within every
- . layer, yielding a negligible addition to the peak i centerline C12 gas concentration assumed to be l supplied directly to rooms adjoining the control j room by the control building HVAC system.
l I t 4. For the postulated worst case C12 release sco-i natio, after isolation dampers in the contaminated i i outside air intake vent for the control room have l- closed, the control room exists at a positive .
I pressure relative to adjoining rooms within the j control building (see the earlier discussion).
i Thus, C12 infiltration into the control room would i j effectively cease upon isolation. The C12 concentration inside the control room at the time of 2
isolation would be approximately 1 x 10-3 ppa. i l The concentration would then begin to decline.
l i 3.0
SUMMARY
OF VEGP's POSITION j VEGP takes exception to the NRC position that the results of i our current analysis are nonconservative and non-conforming to i Regulatory Guide 1.78. Ne do not understand the technical
- basis for the NRC staff's apparent position that no more than a
! two-fold dilution credit can be taken for initial dilution of
! C12 within the control building. In addition, we do not i believe'that our analysis results are nonconservative because !
4 we did not assume the-existence of a 1/8 inch w.g. differential '
pressure across all outer walls of the control building. At :
l the 0.5 meter per second windspeed which maximizes the i
predicted C12. concentration within' the control room, the ;
contribution of control building outer wall inleakage is clearly negligible.
At higher windspeeds, where a measurable positive differential ,
pressure could exist across walls of'the control ~ building-(but !
on no more than two outer walls at any one time), the control room would still be at a positive pressure relative to the
- adjoining rooms, resulting in essentially sero infiltration. !
i considering these factors and the numerous conservatisms in our !
) analysis (e.g., no credit for the dual outside air intake l configuration as allowed by Standard Review Plan Section 6.4), -
we believe that we have adequately demonstrated conformance ;
} with.the USNRC Regulatory Guides 1.70 and 1.95 by providing an i acceptable alternative. ,
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/ VEGP-FSAR-Q Question 410.55 In FSAR subsection 9.4.9, you state that the piping penetration
( ventilation system will maintain the concrete surrounding the piping restraints for the main steam and feedwater systems below 200'F. Verify that the ambient air temperatures in these areas, including the valve rooms, will be maintained at a low enough temperature to allow personnel to inspect equipment during normal plant operation. If there is another heaving, 4
ventilation, and air conditioning (HVAC) system that performs this function, identify the system.
Response
During normal plant operation, the maximum temperature has been calculated to be ll5'E in the main steam /feedwater isolation valve areas. This is adequate to allow personnel entry for inspection during normal plant operation. No other HVAC system performs this function.
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Q410.55-1 Amend. 9 8/84
INSERT TO Q410.55 The piping penetration ventilation system is not safety-related because loss of this system would not limit operation of the I plant or constitute a safety hazard. The ventilation system is required to keep the long term concrete temperature below 200*F. This is not a safety-related function.
An analysis has been performed for loss of the ventilation system and for continued normal operation of the plant. The analysis indicates that the main steam isolation valve room
- ambient temperature will not exceed the equipment environmental
! qualification temperature of 126*F. The room would be cooled by natural circulation instead of forced circulation from the fans. Low flow alarms have been provided in the control room to
- alert the operator to the loss of forced ventilation.
The temperature of the concrete surrounding the main steam line i
and main feedwater line restraints would increase due to the heat transmitted from the main steam and main feedwater pipe
! restraints. ACI Standard 349-76 permits concrete temperatures up to 350*F for any short-term period. A finite element heat j transfer analysis was performed to design the forced ventilation i system. This calculation also envelopes the natural convection j condition with no forced ventilation, indicating that the concrete temperature would not exceed 320*F.
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HCt. O.S UA 6 D t
VEGP-FSAR-Q In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more 4 valves provide isolation, only two of the valves need to be leak tested, i
i Provide a list of all pressure isolation valves included in your testing program along with four sets of piping and instrument i diagrams which describe your reactor coolant system pressure !
isolation valves. Also discuss in de*. ail how your leak testing 1 program will conform to the above staff position.
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! Response l
A list of pressure isolation valves and their function is
- provided in table 210.48-1. The piping and instrumentation j diagrams which show the layout of these valves are shown in j figures 210.48-1 through 210.48-5. The valves listed in table 210.48-1 correspond to those valves discussed in paragraphs l 5.2.5.2.1.A through 5.2.5.2.1.E of the FSAR with the exception
, of check valves 1204-026 through 1204-029 and 1204-013. As j discussed in paragraph 5.2.5.2.1.E of the FSAR, leakage past -
- valves 1204-026 through 1204-029 and 1204-013 will be isolated i by motor-operated valves HV-8801 A/B and manual valve 1204-007.
l Leakage past valves HV-8801 A/B and manual valve 1204-007 is not j; possible since the valve inlets are maintained at a pressure -
r greater than reactor coolant system pressure by the operating j charging pump. $
The valves listed in table 210.48-1 will be included in the VEGP !
! Technical Specifications with Limiting Conditions for Operation '
] which will require system isolation or shutdown in the' event i
! that leakage limits are not met. A draft version of tho VEGP j Technical Specifications is scheduled for cubmittal to the i Nuclear Regulatory Commission (NRC) at least 12 months prior to j j fuel load.
i Periodic leak testing of each valve listed in table 210.48-1
! will be performed after each refueling outage; prior to J
returning the valve to service following maintenance, repair, or I replacement work on the valve; and for those systems rated at less than 50 percent of reactor coolant system design pressure i following valve actuation due to automatic or manual action or j flow through the valve.
! Cha;4ng._the maximum allowable leak rate for each pre 4 isolation valve, voGT -- os to use an - eak rate of
- 1/2 gal / min for each inch . "
ith a maximum e l
! upper limit M N fm n. The justification for t !^ Q l kk&i w LolloWs. '
6l g '
I N s e R 7' A I l l Q210.48-2 Amend. 12 12/84 l .
F (
Insert "A" Concerning the maximum allowable leak rate for pressure isolation valves, VEGP's leak rate limiting conditions for operation will be. equal to or less than one gym for each pressure isolation valve. Pursuant to favorable resolution by the Committee to Review Generic Response's (CRGR) the maximum allowable leak rate for pressure isolation valves will be 1/2 spo per inch of nominal valve size with a maximum upper limit of 5 spm.
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<- VEGP-FSAR-Q a study which was sponsored by the NRC staff (EG A. I Rep t EGG-NTAP-6175, February 1983, In Service L Testi of Primary Pressure Isolation Valves, R A.
Livings n), it was concluded that allowable ak rates based on Ive size were superior to a' sin allowable value-becau a single allowable value i oses an unjustified pe alty on larger valves w out providing information on p tential valve degra tion. Also, the larger valves must e repaired in ace which subjects plant personnel to r iation exp ure in order to meet an overly conservative tandar .
In addition, an indexing c terion to account for gross increases in leakage fr on test to a later test, as found in the ASME cod , paragr h IWV-3427 (b) is a direct indicator of otential va e degradation. Since such an-indexing iterion will be sed by Georgia Power Company, is will provide at ast as good, if not better, a . indication of valve det ioration as the 1-gal / min c terion.
1 B. The ori nal 1-gal / min. criterion was based- a very cons ative estimate of the pressure relief s ~ tem ca city for a plant. The'l-gal / min criteria is ot an
{- dicator of imminent accelerated deterioration or potential valve failure.
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f i
i
'Q210.48-3 Amend. 12 12/84
EAlc L O.s u tts 5 -
VEGP-FSAR-13 Approximate Curriculum Outline Duration Electrical, pressure vessel, and 3 days piping codes and standards (including nondestructive testing review)
G. Incumbents and New Employees I Personnel with experience that exceeds NRC commitments may fill a position in the electrical career path provided that the maintenance superintendent certifies that the employee's experience qualifications exceed the position requirements. The training department may also accept prior training or experience to fill specific course requirements. ,
13.2.2.1.5 Shift Technical' Advisor Training Program ,;_
A. Education Requirements Shift technical advisors will have a bachelor's degree in a scientific or engineering discipline. ,
B. Training Program l
The candidate holds or has held an NRC senior reactor operators license for that type of reactor, or the candidate completes a Georgia Power Company (GPC) ,
~s hift technical advisor training program described in table 13.2.2-1. In either case, the shift technical I
advisor shall receive specific training in the response and analysis of the plant transients and accidents. ,
C. Experience Requirements [ec The candidate will have 1 year of power plant .
experience and will have performed reactor operator or senior reactor operator duties for that type of reactor, or the candidate will receive 1 month of on- '
the-job training as an extra shift technical advisor.
D. Requalification Training for Shift Technical Advisors Shift technical advisors will attend the same i
requalification program as NRC-licensed operators.
"cr: n: n:t ::tively perfer.ing th: chift technical m j 2dviser functiene fer a peried Of ^_ :: nth: er lenger j, L ch ll, prier te see"-ing reepeneihiliti-e of *k- !
pe_itien, ce .~4n4.,,- n c.4u. v.4n4ng en onnur. ew y l
jAf3ggg7 A 13.2.2-11 Amend. 16 4/85 l
W$ $N
Insert "A" Shift Technical Advisors which do not stand shift for a period of 30 days or longer will receive on-the-job instruction from the duty STA concerning plant changes which have occured during their absences.
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a Y
9
a
- l VEGP-FSAR-13 !
I cre : gni :nt of f: ility/precedur: change: that eccurred during their abeanrae
~
Persons not performing the shift technical advisor function for a period of 6 months or longer shall, prior to assuming the responsibilities of the position, undergo an individual requalification program.
13.2.2.1.6 Nonlicensed Operator Training Program A. Initial Training After the start of fuel load, all personnel assigned to perform independent plant equipment manipulations will either complete this initial training program, be qualified to the shift technical advisor level or certified to the senior reactor operator level, or have experience which is equivalent to the following program.
Approximate il Curriculum Outline Duration General employee badge and health 6-10 h'Ud physics training j Industrial safety 4h .
New employee fire training 2h i Nuclear power plant fundamentals 1 week l Power plant components 1 week ,
VEGP systems 4 weeks !
On-the-job training 1 week l
B. Continuing Training
- i After completing initial qualifications, the nonlicensed operator will complete qualification on the plant systems on which he was not initially qualified. Normally, the nonlicensed operator will qualify on all systems outside the control room and containment during the individual's first 3 years in the plant operations department.
C. Annual Requalification Training or Exemption Testing ,
Nonlicensed operators will complete annual requalification training or exemption testing to make
- a. Duration depends on whether the individual attends additional radiation worker training as described in paragraph 13.2.2.1.8.A.
" :nd.
2 1/2 W Amend. 4 2/84 13.2.2-12 Amend. 16 4/85
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Enc.d65LWe Fx VEGP-FSAR-Q Question 440.131 Figures 15.6.3-1 and 15.6.3-4 show a differential pressure of about 1000 psi between the primary and faulted steam generator at 30 min. Figure 15.6.3-11 shows an increasing water volume due to the break flowrate as shown in figure 15.6.3-9 at 30 min. Paragraph 15.6.3.2.1 states, however, that leakage flow through the ruptured tube is assumed to be terminated within 30 min of the initiation of the event. Unless these parameters show discontinuous behavior at 30 min, it would appear that the assumption and the figures are in conflict. Please resolve this.
If leakage flow is terminated at 30 min, how is it accomplished? Any equipment used should be listed in table 15.0.9-1 and qualified.
If the leakage flow is not terminated at 30 min and since the flow through the steam generator safety valve has approached a non-zero asymptote at this time, it would appear that additional radioactive material will be released to the atmosphere. In this event, you will need to reanalyze the radiological, consequences.
Response
A subgroup of the Westinghouse Owners Group (WOG) has been formed to generically address several of the steam ge'nerator ~
1 tube rupture licensing issues which have been raised for NTOL plants.
Southern Company Services. Georgia Power Company is represented in the .
December 1984 in response to steam generator tube ruptureThe appendixes are scheduled for this WCAP. . Two WO to radiological (scheduled for consequences April 1985) of a steam generator tube ruptuThe re . first ,'
is a r overfill (scheduled for July 1985).and the second is a response to SC
<' generic program, recci'2*4^= ef Following completion of the m=====L apc. w HI impienen+th:::+ke ::n:: n: will be generic. Stud)/, recomen4%s o f the
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Q440.131-1 Amend. 9 8/84'