ML20040A327
| ML20040A327 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 09/10/1976 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Suhrke K BABCOCK & WILCOX CO. |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201200766 | |
| Download: ML20040A327 (14) | |
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Docket tile V. A. Moore NRC PDR R. H. Vollmer L cal PDR M. L. Ernst SEP 101976 LWR #3 File W. P. Gamill D. B. Vassallo W. Mcdonald, MIPC F. J. Williams ELD Docket 110. ST!i 50-561
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M. Rushbrook R. Heineman Babcock & Wilcox Company D. Ross bec:
J. Buchanan ATTil: !1r. Kenneth E. Suhrke J. Knight T. Abernathy Manager, Licensing R. Tedesco
!!uclear Power Generation H. Denton P. O. Box 1260 Lynchburg, Virginia 24505 Gentlemen:
ROUilD 2 POSITIO:!S A:iD REQUEST FOR I*iFOR!iATIC;l As a result of our continuing review of the Babcock 5 Wilcox Standard Safety Analysis Report BSAR-205, your response to certain Regulatory staff positions and requests for information is required. The specific information required is detailed in Enclosure 1.
Regulatory Positions are identified by (RSP) underneath the position number shown in.
We are prepared to meet with you to discuss further any of our positions to assure complete understanding'of the factors at issue and the base's for our positions; however, we do not believe extended or iterative debate would be useful.
In order to maintain our licensing review schedule,'ve need your complete responses to the Enclosure 1 items by October 18, 1976.
Please inform us within seven days after receipt of this letter of the date on which you plan to respond so that we may-revise our schedule if necessary. If you plan to appeal to licensing management on any of these positions, please advise us of your intentions within two weeks.
Please contact us if you have any questions.
Sincerely.
Original Slgned by, 0.~ D. Parr Olan D. Parr, Chief Light Water Reactors Branch flo. 3 Division of Project Management
Enclosure:
Positions and Request for Additional Information
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I SEP 101976 Babcock &LillcoxCompany cc: Uashington Public Power Supply System ATTM: Mr. J. J. Stein Managing Director P. O. Box 968 3000 George Washington May Richland, Washington 99352 Mr. Robert Borsum Bethesda Representative Babcock & Wilcox Nuclear Power Generation Division Suite 5515, 7735 Old Georgetown Road Bethesda, Maryland 20014 B. G. Shultz, Project Engineer Stone f Webster Engineering Corp.
t P. O. 6ax 2325 Goston, Massachusetts 02107
!!r. A. H. t1onteith Ohio Edison Company 47 North Main Street Akron, Ohio 44308 Mr. W. E. Kessler Connonwealth Associates, Inc.
209 East Washington Jackson, Michigan 49201 f
Robert J. Kafin, Esq.
115 Maple Street Glen Falls, New York 12801 i
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'x ENCLOSURE 1 POSITIONS AND REQUESTS FOR ADDITIONAL INFORMATION BSAR-205 DOCKET NO: STN 50-561 110.0 MECHANICAL tau tne r w ir.L 110.38 The rcaponse to Question 110.18 is not completely acceptable. The (3.6.3.1) statement that subsection NF of the ASME Code will be the basis (110.18) for operability tests is unclear, since s.ubsection NF addresses only structural integrity. The staffs prinary concern regarding snubbers is that the structural aspects.of snubber utilization is, fully evaluated. We will require that the following informa, tion concerning snubbers utilized in B-SAR-205 systems be provided in the FSAR.-
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(1). Snubber design specifications.
(2) Description of snubber suppliers performance qualification tests and load tests.
(3) System and component structural analysis showing:
(a) Structural analysis model s.
I (b) Description of the characterization of hydraulic snubber mechanical properties used in the structural analysis s_
including considerations such as(i) differences in tension and compression spring rates, (ii) effect of entrapped air and temperature on fluid properties, (iii) other f actors af f ecting snubber characteristics.
(c ) List load conditions and transients analyzed.
(d) Maximum snubber loads, corresponding piping or component stresses.
(e) Comparison of computed loads and stresses with rated snubber load and stress intensity limits.
(4) Discuss design provisions for accessibility for inservice
" inspection and possible removal for operability testing and repair or replacement of snubbers.
Provide a commitment to include all of the above information in the B-SAR-205 FSAR.
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,l The informatimiin Section 3.9.1.1, " Vibration Operational Test i
In addition to the 110.39 (RSP)
Program" is not completely accep' table.
in the information presented, it is the staff's position that (3.9.1.1)
PSAR, a commitment should be made to conduct preoperational piping vibrational and dynamic effects testing in accordance with NRC Standard Review Plan, Section 3.9.2 on all of the following classes of piping systems if they are within the B-SAR-205 scope of design:
(1) All ASME Class 1, 2 and 3 piping systems.
All high energy piping systems outside containment.
(2)
All Seismic Category 1 portions of moderate energy piping (3) systems outside containment.
110.19, 110.20 and 110.24 are related to se'ismic Questions 110.40 (RSP) qualification of mechanical and electrical equipment and Responses to these questions in Amendment 2 (3.9.1, instrumentation.
to the B-SAR-205 PSAR are not entirely in accord with currentiEEE-344-3.9.2, The responses cite several documents, 3.10)
NRC requirements.
71, IEEE-382-72, and topical report BAW-10082, which are not 410.19, i
5.20) currently acceptable references.
~. 10.21, It is recognized that much B & W designed equipment has been 110.24) previously qualified by methods and to standards which may or The principal items may not be acceptable by current standards.
of concern are those vital appurtenancer necessary for the actuation and continued operation of safety related pumps and The operability of such valves during accident conditions.
appurtenances as electrical switching gear, motors and valve operators subjected to complex oscillatory motions is dif ficult to verify analytically. Consequently, some such components mayhavetoberetestedtocurrentseismiclqualificationstandards specified in section 3.10 of the NRC standard review plan.
Our position is that you should provide, in the BSAR 205 document, one of the following:
3.9.1, 3.9.2 and 3.10 of the BSAR-205 to be Revised sections entirely consistent with the corresponding sections of the (1)
NRC Standard Review Plan or commitment to conform to the final generic resolution by the NRC seismic qualification task group and B & W of out-CE)
A standing quettions pertaining to the seismic qualification of B & W equipment.
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i 110.41 The response to Question 110.24 in Amendment 2 to the PSAR is (3.9.2.4) not, completely acceptable as a response to Question 110.21.
It is the staffs' position that the manufacturers of all ASME Class 1, 2 and 3 active pumps and valves must be required to demonstrat that the pump or valve will 'perate normally when subjected to all loads and other environmentaA conditions associated with a faulted condition. These' loading conditions should be clearly defined to the pump or valve manufacturer. Prov.ide a specific commitment to this pos'ition in Section 3.9.2.4 of the PSAR.
110.42 It is the staffs' position that the applicant should verify that (3.9.2) all ASME Class 1, 2 and 3 systems, components and supports which (5.2.1) are required (1) to assure safe shutdown of the plant or (2) to prevent or mitigate the consequences of an accident and which are designed to emergency and faulted condition stress limits will.
function as designed. The stress limits listed in the various sections of the PSAR assures the structural integrity of the various systems, but does not necessarily guarantee their functional integrity. Provide your commitment in the PSAR that the stress limits used in the design of all of the above noted systems, components and supports will not result in inelastic deformations l
which would prevent the normal operation or function of the item.
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-.41.0 MATERIALS ENGINEERING - MATERIALS INTEGRITY
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121.4-Your reply to request No.121.1 is 'not satisfactory. Although (RSP)
Regulatory Guide 1.14 Revision 1 bears the overprint "For 4
(Reg. Guide Coment" it is an " approved" and not an " interim" document
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i 1.14) and should be addressed in BSAR 205.
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130.0 STRUCTURAL. ENGINEERING 130.24 Your response to request number 132.22 concerning decoupling (3.7.2 RSP) criteria for subsystem is incomplete. It is the staff's position' that the decoupling criteria for subsystem should be expressed in tems of mass ratio and frequency ratto, as stipulated in NRC Standard Review Plan 3.7.2-II.3.b.
Provide appropriate commitments to the SRP criteria, or describe and justify any exceptions taken to the SRP criteria.
Provide a commitment to review the BOP design information and drawings 130.25 (3.8 RSP Interface) for compliance with your.NSSS design criteria.
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I 211.0 CLASSIFICATION, CODES AND STANDARDS 211.18 In your response to 211.13, valves V4, V2A and V2B on Figure 9,3-1, sheet 2 are apparently incorrectly classified as Quality Group C components. Please correct a.s necessary.
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230.0
- Reactor Fuel 231.0 The response to lat-round question 231.1 is incomplete because l
it does not provide the requested information on experimenta l
231.30 loads.
con'firmation of the fuel handling and shipping de (4.2.1.1) been confirmed experimentally.
I The response to lat-round question 231.1 is 231 31 Please cite the current design limits for these phenomena, (4.2.1.1) outline the on-going or planned R&D prograss which should
- limits, yield confirmatory information on the specific cesignDiscuss how d and present fall-back positions.
l accounted for in the summation of stresses in the fue 31.2).
assembly (as suggested in the response to question 2 i
i Please The response to 1st-round question 231.4 lacks detail.
describe how the s'pecified coolant temperature limits and 231.32 i
associated cladding loading are used in the fuel rod fat gue (4.2.1.2)
Show by means of an example how the coolant tem-d to perature limits and associate._, cladding loading are use analysis.
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" identify the conservative conditions for input-to the stress f
4 analysis," as asserted in the response to question 231..
The response to 1st-round question 231.5 requires amplificatio regarding (1) the " conservative models" said to be used for 231.33 (2) the rod differential growth and grid pressure drop and (4.2.1.2) l out-of-reactor flow tests and measurements which reported y confirm the calculations that show that grid position is i
Please show in greater t
well-maintained throughout life; detail how these calculations and experimen is sufficient to maintain grid position throughout life.
T'e response to 1st-round question 231.6 does not provide the l
requested information on dimensions, spring constants and h
231.34 experimental observations of the upper and lower pl (4.2.1.2) addition, show quantitatively that the resistance springs.
j flux, temperature, and stress l
withstand the worst postulated conditions, as asserted in the lat-round response.
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231.35 The response to 1st-round question 241.7 does not provide the (4.2.1.3) requested design bases fer Zirealey-4 irradiation growth.
Design "btses" are not synonymous with " values," as appears
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to be implied by the response. Please provide the design bases as. requested, and briefly outline the data which sup-port these bases.
231.36 The response to 1st-round q'uestion 241.8 requires clarifica-
.t on because of an apparent confusion of terminology. The i
(4.2.1.3) response appears to treat cladding strain and fuel rod de-flection as if they were synonymous. An intent of lat-round question 241.8, however, was to establish the displacement limit of B&W fuel rods from a rod bowing viewpoint. Such a i
l displacement limitation, when used in fuel design, should reflect a DNB correlation and power peaking factor calcula-tion. Provide the as-manufactured displacement limitation as well as the one imposed during operation. Discuss how one confirms that these limitations are not exceeded.
231.37 The response to 1st-round question 231.11 does not provide the (None) requested information on the currently used stiffness limita-tions on the spacer grid assembly and individual grid springs.
In addition to providing this requested information, please outline how the results of specific portions of the Mark C
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fuel assembly development program will be used to provide the information requested in 1st-round question 231.11.
231 38 The response to lat-round question 231.12 requires amplifica-(4.2.1.3) tion regarding the procedure for limiting the recommended power startup rate in the 0-205 power range. Please quantify this recommended limit in power startup rate and provide ex.
perimental quantitative. verification of the effect of reduction in power startup rate on defect propagation.
231.39 The response to 1st-round question 231.18 addresses the 15 strain (4.2.1.3) limit which is based on average cladding strain.- The R-2 re-actor power ramp tests, referred to in the response, were, however, performed on low exposure rods which were still ductile and, therefore, only demonstrated the ability of the rods to withstand pure mechanical loading. Describe any research pro-grams on analytical modeling development currently in progress or planned to evaluate the effects of local cladding strain due f
to pellet cracking on ridging, cumulative damage, and stress corrosion cracking.
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f 231.40 The response to tat-round question 231.21-indicated that in (4.2.3.2) experiments where irradiated A1 0 -B C was exposed to high-33 4 temperature high-pressure water, the B C reacted with the l
4 water to form H3B0. Thus, if the poison rod cladding were 3
perforated, the H B03 would be leached into the coolant.
3 Please discuss the potential safety implications of the re-activity insertion resulting from the loss of B-10 from the burnable poison rods'by this mechanism. Describe the re-activity anomaly that would result if all the B C were 4
removed from (a) one rod and (b) all the poison rods early in life.
Provide rate equations for the hydrolysis of B C 4
and rate of loss from perforated rods, and calculate these rates at (a) reactor coolant temperature and (b) local poison pellet temperature.
231.41 The response to 1st-round question 2.31.17 on fuel rod bowing (4.2.1.3) refers to examination measurements on the Oconee 1 Mark B (15x15) asse=blies which will be used as a basis for pre-
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dieting bowing in the Mark C (17x17) assemblies. Please discuss how the bowing data from 15x15 Mark B assemblies will used for 17x17 Mark C bowing predictions; i.e. how will 15x15 Mark B assembly data be related and applied to the 17x17 fuel?
Also provide the-following information:
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(1) Status of the 15x15 rod bowing data collection; (2) Schedule and scope of the 15x15 examination program; (3) Manner by which the 15x15 data and analysis will be reported to NRC, and approximate date for submittal of a topical report; (4 ) Plans for obtaining 17x17 fuel assembly bowing data; (5 ) Out-of-pile (if any) mechanical experiments which will provide input to a mechanistic bowinC model.
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231.42 The treatment of the seismic and LOCA analyses for the Mark C (None)
(17x17) fuel assembly is inadequate. An in-depth safety analysis of.the seismic and LOCA response of,the Mark C (17x17) fuel assembly has been requested (letter, Ross to Schwencer, July 25, 1974) and a commitment to submittal of a topical report in early 1976 (at least one year prior to the filing of the first FSAR incorporating the Mark C fuel assembly) was made by B&W (letter, Mallay to Schwencer, September 3, 1974). Our evaluation of the B&W seismic and LOCA analyses for the Mark C assembly cannot be completed until the requested report has been received.
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232.0 Reactge, Physics 232.26 The response to Question 232.17 is inadequate. Flesse (15.1.2) identify the'205 FA plant for which the analysis was performed.
s 232.27 The response to Question 232.11 (as presented in the response to Question 212.71) implies that power shapes i
with "large" negative offsets were used in the deriva '
tion of the pcwer range scram reactivity curve. Please confirn and indicate the range of negative offsets con-sidered. In particular was consideration given to a scram while in the recovery from a load following j
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311.0 ACCIDENT ANALYSIS l
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(1) Your respcnse to 211.13 indicates that you have conducted radiological accident analyses to determine the seismic design classification of the deborating i'!c/'
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N demineralizer. The details of this analysis should be provided including justification of the assumptions used.
Incl,ude an explanation of the assumptions associated with the 100 minute time period prior to
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I (2) Prc'/ide similar information pertaining to the reactor
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coolact degasifie'r (211.14).
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(3) Summarite all similar analyses, if any, used to J
justify :omponent design classification as non-seismic Category I.
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(4) The respogse tg 211.13 indicates a limiting x/Q of 7.0 x 10- s/m. We do not agree that the x/Q is
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likely to be conservative for anticipated -Babcock-205 1
sites. The staff h s doc mented. in WASH 1361, that i
ki the x/Q of 7.0 (10- ) s/m is a limiting value for.
f only one half of the large number of sites examined.
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Our position is that you must provide explicit identification cf the limiting 'x/Q as an interface l
requirement on the BOP supplier.
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l-400.0 PROJECT MANAGEMENT 400.3 At a meeting with the staff on August 6,1976, your Assadsent 2 response to request 211.2 was discussed and certain connaitments were made by B&y regarding additional information concerning the staff position and request for information embodied in 211.2, as well as 1
several other numbered requests.
A sununary of the August 6,1976 meeting is enclosed for reference.
Your adequate response to these issues within the time allotted 4
for Round 2 responses is required to complete the BSAR-205 review effort within the review schedule presented in our letter of April 22, 1976 to Mr. K. Suhrke.
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400.3 Docket Ho: STN 50-561 VENDOR: Babcock & Wilcox Company (B&W)
SUMMARY
OF MEETING WITH NRC STAFF On August 9, 1976, members of the Nuclear Regulatory Commission (NRC) staff met with representatives of BitW to discuss B&W's July 19, 1976 responses to several of the staff Round 1 questions issued to B&W on May 14, 1976. Questions discussed included numbers 012.27, 30, 31, 32, 36 and 211.2. The staff requested the meeting to develop a more complete understanding of, and to discuss observed deficiencies in, the B&W responses. An attendance list is enclosed.
The discussion on request number 211.2 generally included a description '
by staff members of the evolution of the issue embodied in the request, and a joint B&W/ staff discussion of what should constitute an adequate response to the request. B&W concluded that while they felt they could provide at least some of what the staff was looking for in the imediate future, detailed commitments regarding their complete responsa would need more work and could not be scheduled at this time.
The staff stated that "RC pump' auxiliaries that are an incegral part of each RC pump as:,en,My" is intended to include auxiliary components on both the pump and the driving motor, which together comprise the RC pump
" assembly." B&W statcd that they would provide the clarification requested regarding Quality Group classification and Safety classifica-tion of auxiliary components and offer further justification of their contention that their designs are in conformance with Regulatory Guides 1.20 and 1.29.
B&W also stated that they will provide additional specific design criteria for RC pump (and motor) auxiliaries in the BSAR document to include structural design criteria, materials and quality class data.
B&W's July 19,1976. response to request number 211.2 was generally intended to demonstrate to the staff that the RC pump will operate for
.at least 30 minutes after a loss of component cooling water without impairing the pump coastdown capability. Additional automatic pump trip l
signals (nonsafety grade) were proposed to be included in the bearing temperature alarm system. Staff reviewers pointed out that the B&W response was deficient in that:
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(1) The response did not demonstrate that the component cooling water system design would be required to meet the requirements 3
of condition 2 of 211.2 regarding single failure protection j
and instrumentation and control provisions (safety grade detection and control room alanc),
(2) The response did'not provide the requested analytical justi-fication of successful RC pump operation for 30 minutes following loss of component cooling water, and (3) The response did not address the RC pump test that request number 211.2 stated would be required in the event that B&W elected to justify 30 minute pump operation.
The staff brought out during discussion that the full-scale, operating condition test of an RC pump assembly was required in part to verify the vendor's ability to analytically predict temperature rise in the Once confidence in the vendor's predictive analyses complex system.
and pump performance at maximum temperature is achieved, it should not be necessary to test a given RC pump assembly over the entire range of operating temperatures, and it may not be necessary to proto-l type test pump assemblies of similar design and construction but from different manufacturers.
012.27, 29 and 32 led to a clarification Discussion of request numbers of interface requirements for B&W designed equipment to be mounted by the balance-of-plant (B0P) designer on or in safety-related structures.
B&W was requested to provide preliminary design information in BSAR-205 describing static loads and the locations of these forces on the equip-ment items. Where applicable, electrical service and other B0P re-quirements should also.be specified in the interface data.
Regarding request number 012.30, B&W agreed to specify in BSAR-205, the
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numerical values of decay heat load resulting from the amounts of spent fuel described in the request for information. The response to date is inadequate in that the heat load values were not given as requested, although a commitment was made to submit the data directly to the B0P designer at an unspecified future time.
9 Discussion of 012.31 resulted in a B&W comitment to provide, in a future BSAR-205 amendment, a specified minimum water depth over the spent fuel array, as requested by the staff.
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In a discussion of 012.36, B&W pointed out that their July 19 used in, and uncertainty factors added to the equ
, 1976 standard ANS 5.1 to calculate the decay heat rates following reactor shutdown.
i They reiterated that their response did describe a method that gave results identical to the staff position APCSB BTP 9 2 staff reviewers stated that specific details, including the governi
. The staff to verify that the B&W calculational method method specified in the Acceptance Criteria of the staff's Standard Review Plan, Section 9.2.5.
B&W agreed to provide complete details i the BSAR-205 document, including a copy of the version of the p n
standard (ANS 5.1) that was used.
roposed BSAR-205 pages no later than the date schedul to Round 2 positions, which is October 18, 1976.
Thomas H. Cox, Project Manager Light Water Reactors Branch No. 3 Division of Project Management
Enclosure:
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Attendance List Babcock & Wilcox Company cc:
ATTN:
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Mr. Kenneth E. Suhrke Mr. A. H. Monteith I
Manager, Licensing Ohio Edison Company Nuclear Po.ser Generation 47 North Main Street P. O. Box 1260 Akron, Ohio 44308 Lynchburg, Virginia 24505 Mr. W. E. Kessler Washington Public Power Supply System Commonwealth Associates, Inc.
ATTN:
Mr. J. J. Stein 209 East Washington Managing Director Jackson, Michigan.49201 300 George 'ashington Way Robert J. Xafin, Esc.
Richland, Washington 99352 n
s w York 12801 Mr. Robert Borsum Bethesda Representative Babcock & Wilcox B. G. Shultz, Project Engineer Nuclear Power Generation Division Stone & Webster Engineering Corp.
Suite 5515, 7735 Old Georgetown P. O. Box 2325 Road Boston, Massachusetts 02107 Bethesda, Maryland 20014 j
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AUG 2 41976 f\\
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ENCLOSURE j
i ATTENDANCE LIST MEETING OF AUGUST 6,1976 NRC AND B&W i
NRC:
B&W:
T. H. Cox (DPM)
E. Swanson R. Kirkwood (P.S.).
G. Anderson j
L. Riani (APCSB)
J. R. Hamilton A. R. Ungaro (APCSB)
G. J. Brazill V. T. Leung (APCSB)
J. G. Newton D. L. Tibbitts (DPM)
J. J. Happell i
G. Mazetis (RSB)
- 0. D. Parr (DPM).
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