ML19269E782
| ML19269E782 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 05/15/1979 |
| From: | Hood D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7906290626 | |
| Download: ML19269E782 (30) | |
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- p MAY 151979 DOCKET NOS. 50-329 AND 50-330 APPLICANT: CONSUMERS POWER COMPANY FACILITY:
MIDLAND PLANT, UNITS 1 & 2
SUBJECT:
SUMMARY
OF APRIL 19-20, 1979 MEETINGS ON OPEN ITEMS REGARDING FSAR REVIEW On April 19 and 20, 1979, the NRC staff met in Bethesda, Maryland with representatives of Consumers Power Company, Bechtel Associates, and Babcock and Wilcox to discuss open items associated with the staff's review of the FSAR for Midland Plant, Units 1 and 2.
These open items are identified in the staff's letter of March 30, 1979 and served as the agenda for the meeting.
This is the second of two sets of meetings which were held to discuss the March 30, 1979 letter. The first set of meetings was held on April 10 and April 11, 1979.
Further meetings to discuss items of the remaining branches will be scheduled.
Meeting attendees 'or April 19, 1979 are listed in Enclosure 1, and for April 20,1979 in Enclosure 2.
The discussions in Enclosures 3 and 4 for this meeting summary corres-pond to the same-numbered items by technical review branches as specified in the March 30, 1979 letter.
'd e D. Hood, Project Manager Light Water Reactors Branch No. 4 Division of Project Management
Enclosures:
As Stated ecs w/ enclosures:
See next page 2143 148 7 906290(oRf ;
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Consumers Power Company i
CC:
l Michael I. Miller, Esq.
Mr. S. H. Howell Isham, Lincoln & Beale Vice President Suite 4200 Consumers Power Company l
One First National P1aza 212 West Michigan Avenue Chicago, Illinois 60670 Jackson, Michigan 49201 Judd L. Bacon, Esq.
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Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Mr. Paul A. Perry Secretary Consumers Power Company 212 W. Michigan Avenue Jackson, Michigan 49201 l
Myron M. Cherry, Esq.
One IBM Plaza j
r.hicago, Illinois 60611 Mary Sinclair 5711 Stmmerset Drive Midland, Michigan 48640 Frank J. Kelley, Esq.
Attorney General State of Michigan Environmental Protection Division 720 Law Building Lansing, Michigan 48913 Mr. Wendell Marshall l
Route 10 Midland, Michigan 48640 Grant J. Ferritt, Esq.
Thompson, Nielsen, Klaverkamp & James 4444 IDS Center 80 South Eighth Street Minneapolis, Minnesota 55402 Mr. Don van Farowe, Chief Division of Radiological Health Department of Public Ikalth P. O. Box 33035 l
Lansing, Michigan 48909 l
2143 149 e
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ENCLOSURE 1 ATTENDANCE LIST j
~%PRIL 19, 1979 NRC - STAFF CONSUMERS POWER COMPANY D. Hood J. Zabritski H. Daniels E. Olfier M. Bolotsky J. Webb H. Conrad A. Birkle M. Boyle R. Hermann R. Stevens BECHTEL B. D. Liaw J. Strossnider J. Kovach J. Shapaker M. A. Gerding D. Messing J. Clements B&W COMPANY C. E. Mahaney L. M. Lesniak J. Howard J. Shively S. Skaar J. Paster l
2143 150
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'4 ENCLOSURE 2 ATTENDANCE LIST APRIL 20, 1979 I
NRC - STAFF CONSUMERS POWER COMPANY D. Hood J. Zabritski B. Bosnak S. Hou R. Stephens BECHTEL l
D. Pickett l
C. Tinkler J. Clements i
J. Shapaker J. Lewis J. Dunne 1
J. Cartwright i
j B&W COMPANY t
j C. Mahaney D. Howell J. Howard J. Shively i
J. Anderton l
G. Vames J. Galford I
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2143 151 l
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ENCLOSURE 3 ITEMS DISCUSSED APRIL 19, 1979 Instrumentation and Controls Systems Branch The technical reviewers explained that the open items are based upon review through FSAR Revision 15 as indicated in the staff's letter of March 22, 1979.
The review of the Midland FSAR by this branch will resume in late July 1979 once prior workload comitments on other pro-jects are completed, and assuming sufficient information for an efficient review has been submitted on the Midland docket at that time.
1 & 5 - Unresolved - The staff stated that BTP-ICSB-23 is replaced by R. G.1.97 and that Midland is subject to the results of Generic Technical Activity A-34.
The staff reiterated its position that the post-accident monitoring recorders be qualified to operate after the SSE.
The applicant referenced a January 31, 1979 letter from R. DeYoung to J. Ward of AIF as the basis for its position that Position Cl of R. G.
1.97 is not applicable to Midland and also noted that no equipment on the market can be purchased which would meet Position C3.
The staff cautioned that the January 30, 1979 letter predated the TMI-2 accident and that this matter is being examined again in light of such recent developments.
The applicant stated it has formed its own task force to review the TMI-2 accident for impli:ation to Midland 1 & 2.
The Midland task force will include the adequacy of post-t accident monitoring instrumentation and R. G. 1.97.
l 2
- Unresolved - Additional information regarding the ultrasonic reverse flow monitors will be submitted in July or August 1979.
3
- The staff is reviewing the response in FSAR, Revision 18.
4
- Unresolved - The reply will be submitted in May 1979.
i 6
- Unresolved - One specific example for which two way comuni-cation devices have affected nuclear power plant operation is given in I&E Bulletin 76-03 dated 3/15/76.
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7
- Unresolved - The staff reiterated its position that the Midland design be revised to conform with BTP-ICSB #4.
l 8
- The staff's fonnal detailed drawing review will not begin until at least July 1979.
4 l
9
- Unresolved - The applicant will reply by FSAR Revision 20.
10
- Unresolved - This concern is based upon I&E Circular 78-19 and the term " bypass" is defined as used therein.
e 2143 152
. 11.
- Unresolved - Reply will be in FSAR Revision 20.
12.
- Unresolved - Reply will be in FSAR Revision 21.
l Power Systems Branch 1.
- Unresolved - The applicant clarified that the reliability calculation presented in FSAR Revision 19 is for the whole ESFAS, not the sequencer alone.
Further discussion of this item is needed.
2.
- Unresolved - The applicant will reply by FSAR Revision 20.
3.
- Not discussed.
4.
- Further information will be submitted by FSAR Revision 20.
- The staff is reviewing the applicant's response in FSAR 5.
Revision 19.
- Unresolved - The staff requested worst case examples 6.
involving the maximum use of jumpers and identification of other equipment involved.
7.
- Unresolved - The staff is reviewing the applicant's response and notes that Position No. 4 therein is not completed. An additional concern resulting from Arkansas Nuclear One, Unit 2 was described.
8a.
- The applicant intends to indicate compliance with the staff position by FSAR Revision 20 and to provide supporting analyses by FSAR Revision 21.
- Unresolved - The staff stated that one means of testing 8b.
fuses is by measuring resistance on a sampling basis.
Fuses which are installed are to be tested.
l 9.
- The applicant will clarify the FSAR to indicate that each diesel is provided with a 7-day fuel oil storage tank to l
meet the engineered safety feature loads.
The 3 1/2 day i
supply was based upon maximum rated diesel load.
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2143 153 1
. 1 Materials Branch (Metallurgy Section) 1.
- Unresolved - The staff requires mere specific information.
In each thickness of liner ar.d for each penetration for i
l which fracture toughness test data is required, the appli-cant was requested to provide the following:
the material and its specification, the applicable ASME code edition and addenda, the service metal temperature, the testing requirements, the acceptance criteria, and the ir.tual fracture toughness test data.
The staff noted that the information in Table 3.8-25 was insufficient in this regard.
Extra The staff needs info:mation on the hydrogen evaluation Item - rates inside contaiament after the design basis accident for the full range of pH conditions.
The staff requested data on pH as functions of time and temperature for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during which pH may be changing.
The staff referred the applicant to BNL-NUREG-24532 " Hydrogen Release Rates from Corrosion of Zinc and Aluminum," dated May 1978.
The applicant replied that it is participating in a study with Oak Ridge for the hydrazine spray additive, including effects on coatings, and sill submit a summary about June 1979.
Addi-The applicant will clarify the use of stainle:: steel tional ih the stm m gene rator.
Such use was said to be limited Item -
to the cladding.
Materials Branch (Materials Integrity Section) 1.
- Unresolved - The applicant will upgrade its inspection plan to the applicable code edition, once approved.
2 & 3 - Unresolved.
Extra - The discussion also included a review of the staff requests by letter of March 26, 1979 to which the applicant will i
i respond by a future amendment.
The staff also requested a summary of the fracture toughness data for each ferritic steel component of the reacter coolant pressure boundary.
i 2143 154 g
. Mechanical Engineering Branch 8.
- The applicant explained it can only comply with the " intent" of R. G. 1.121 because this guide was not written for the once-through steam generator design.
The applicant will clarify that the criteria of the guide will be meet, but that the analyses performed must be modified as appropriate to the once-through design. The staff noted that B&W has not provided a report on tube plugging.
Related The staff noted that this item (MEB, Item 8) was also item -
related to AAB Item 3 regarding turbine missile hazard, which is made less probable by frequent testing of the turbine stop valve; in Generic Technical Activity A-5 the staff notes a concern that frequent testing of the turbine stop valves may lead to degradation of the OTSG tubes as was observed at Oconee.
The staff also ex-pressed concern that less frequent testing may only defer tube damage to a later time. The B&W representative stated that more recent tests completed at Three Mile Island Unit 2, coupled with Oconee results, show that tube degradation will not occur if turbine stop valve testing is performed at reduced load and if the valve design provides for restroking of the governor valves to close at a slow rate.
The applicant has not com-pleted its review of this B&W finding. The generic resolution of this item, once determined, will be applied to Midland by the staff.
Containment Systems Branch 5.
- The applicant advised the staff that it is considering the deletion of the penetration pressuriz tion system, includiig the isolation valve seal water subsystem. The applicant noted that this system was provided in the design as a result of the staff's requirement expressed in a letter from fir. p. A.
Morris dated March 28, 1969.
l The applicant stated that Appendix J to 10 CFR 50 did not exist in 1969 and that, in its opinion, Appendix J l
negates the need for this system. The applicant further noted that the staff had permitted no credit for this l
system.
The staff stated it will further review the 1969 letter and advise the applicant further of its position.
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i 2143 155
ENCLOSURE 4 ITEMS DISCUSSED APRIL 20, 1979 Containment Systems B_ ranch 3.
- The staff noted that further information is needed regarding the nodalization sensitivity study as to the basis for which the applicant concludes the 28 nede scheme for the reactor cavity subcompartment is acceptable.
The FSAR states that the peak pressures converged in the 28 node model and that there was virtually no change in peak pressures in the 32 and 40 mode models. The applicant also indicated that the loads and moments acting on the vessel also converged and will address this convergence in the FSAR.
The staff questioned why the nodes above the shield plug (i.e., nodes 11, 12, 13, 14, 25, 26, 27 & 26 of FSAR Figure 6D-7) were neglected in determining the loads and moments, and expressed concern when changes are made in nodalization between peak pressure cases and the loads and moments.
It was noted that the loads and moments transients (time histories) vary between the various nodalization schemes and that it is difficult to deter-mine the worst case. The staff concluded that the worst case would be one with the peak magnitudes.
The appli-cant will submit a calculation of the loads and moments considering the region above the shield plug in order to justify its assumption that this region makes a neg-ligible contribution to loads and moments.
The reactor cavity model used in the applicant's analyses assumes that insulation blocks all flow paths to the in-core instrumentation tunnel.
The staff noted that this assumption prevents uplift forces on the vessel in the calculations and is not conservative for this purpose.
The staff requires that the applicant supplement its total asymmetric load analyses to determine uplift forces on the vessel without flow blockage by insulation.
The full cross-sectional area (l A) break in the primary cold leg piping submitted for the reactor cavity subcom-partment analysis occurs in the piping penetration.
The applicant's analysis assumes that the foam-glass insula-tion in the penetration is instantly displaced. The staff finds this assumption is very significant to the vessel calculations because it directs most of the blow-down to the steam generator subcompartment rather than into the reactor cavity.
The staff also noted that in-adequate discussion of the foam-glass insulaF a is given 2143 156
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. in FSAR Section 6.2.2.1.2.4, Tables 6.2-33 and-34. The applicant was asked to further describe in the FSAR the following characteristics of the foam-glass insulation: manufacturer, amount used, attachment methods, whether any sheet metal or other shroud surrounds the foam glass, and any other data relative to displacement or blocking assumptions. The applf-cant was also asked to address properties of the foam glass relative to the potential blocking of sump screens and com-patability of the foam glass material with the post-LOCA environment (including the potential of any of the insulation materials to form a slurry which could harm pump seals, foul heat exchangers, or insulate fuel). The staff requires that the applicant either (1) revise the analyses assuming the insulation remains in place, or (2) demonstrate analyti-cally or experimentally the instantaneous displacement of the insulation. The applicant noted the conservatism in the assumption of a full area (l A) break whereas physical restraints preclude in excess of a 0.244 break. The appli-cant will submit further analyses to justify the loads and moments provided.
The staff requested that the 40 node reactor cavity model of FSAR Figure 60-9 be expanded by including a horizontal plane through the nozzle centerline. The applicant will submit loads and moments using this modification to the model. The staff also questioned the basis for the verti-cal plane in the nozzle region between nodes 5-30, 2-29, 16-32 and 19-31; if these boundaries are intended to re-present the excore neutron detectors, then the staff notes that these vertical planes should be extended for the entire vessel length.
5.
- The staff stated its position with regard to the appli-cants consideration to delete the penetration pressurization system which was discussed the previous day. The staff does not support deletion of the system and finds that such deletion would require re-review of site acceptability issues.
The staff noted that this system provides post-accident consequence mitigating safety eouipment without which the staff advised in 1969 that the ' proposed site was unacceptable. The subsequent issuance of Appendix J does not change the need for increased protective measures at specific sites. The staff also notes that such deletion would represent a departure from the principal architectural and engineering criteria specified in PSAR Section 1.4.4.
2143 157
. Auxiliary Systems Branch 2.
- The staff stated that review of the information in the FSAR regarding the single failure proof hoisting mechanism for the cask handling crane is being reviewed in parallel with the generic review of the referenced topical report EDR-1.
The staff noted that Tablas 9.1-9 and 9.1-10 of the FSAR addresses certain positions of Regulatory Guide 1.104 as its affects the design of those portions of the auxiliary building crane that are not covered by the topical report.
However, the staff also noted that position C.l.C through and including C.l.F, and positions C.3d and C.S.a of Re-gulatory Guide 1.104 are not addressed in the FSAR and relate to portions of the crane not covered by the topical report. The staff requested that these positions be addressed in the FSAR and that justification for any deviations or alternatives to these positions be provided.
Mechanical Engineering Branch 5.
- The analysis of reactor internals for combined SSE and LOCA loads for the Midland plant, including asymmetric cavity pressurization, are based upon adjustments to a prior analysis performed for Davis-Besse 2 & 3.
A B & W representative explained in greater detail how these adjustments were made. Viewgraph slides used for this presentation are shown by Enclosure 5.
The Mid-land analyses will be submitted in March 1980.
6.
- Midland is the first B & W plant to perform dynamic, elastic-plastic analyses using break opening time histories for postulated guillotine breaks in the primary system.
The ANSYS computer program is used to determine the guillotine break opening area and time.
Pipe reaction forces are calculated using the CRAFT code.
The analytical method is based upon iterating between pipe reaction forces and the pipe time history displacements and areas until area convergence is reached.
Slides used for this B & W presentation are shown as Enclosure 6.
Operator Licensing Branch 1 & 2 - These items will be addressed by FSAR Revision 20.
2143 158 i
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Enclosure S MIDt.AND HYDR 0-DYNAMIC LOADS I f i f DAVIS BESSE DAVIS BESSE CORE BOUNCE SYSTEM ANALYSIS DYNAMICS LOADS i f 1 r
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VERTICAL HORIZONTAL SEISMIC LOAD LOAD LOAD RATIOS RATIOS RATIO >
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DYNAMIC SIDE LOADS ON CSA n
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INCLUDES LATEST ANALYSIS METHODS e
ASME CODE, SUBSECTION NG ANALYSIS e
STRESS CATEGORIZATION e
ALLOWABLE STRESS LIMITS o
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F MIDLANDS STRESS ANALYSIS EACH COMP 0NENT WILL BE ANALYZED EACH LOAD CAN BE RATI0ED DIRECTLY SEISMIC LOADS WILL BE RATI0ED AND ADDED DIFFERENCES FROM BASELINE WILL BE ADDRESSED RESULT WILL BE CODE SAFETY FACTORS FOR EACH COMPONENT 2143 161
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,. g =7p- ~.: Comparison of break opening areas HAND CALCULATIONS LOCATION CALCULATED BY ANSYS BASED ON GEOMETRY e HOT LEG RV N0ZZLE O. 39 A-0.388A e COLD LEG STEAM GENERATOR O.15A 0.144A N0ZZLE e COLD LEG PUMP SUCTION 0.53A 0 4A N0ZZLE EXAWLE HAND CALCULATION ASSUMPTIONS ANSYS CONSIDERATIONS (FOR PUMP SUCTION GUILLOTINE) 1. PIPE HINGED AT OTSG N0ZZLE. 1. PIPE STRESS IS MONITORED THROUGHOUT FOR YIELDINi. 2. MOTION LIMITED BY GAPS. -2. RESTRAINT COMPRESSION IS CONSIDERED. 3. NO CONSIDERATION OF PIPE 3. EXACT GEOMETRY CONSIDERED. YIELDING. N 4. NO RESTRAINT COMPRESSION 4. MOTION IN ALL DIRECTIONS CONSIDERED. A CONSIDERED. u 5. ONLY PLANAR MOTION G CONSIDERED. .6. NO PUMP MOTION CONSIDERED.
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l FSAR 3.9.3.1 FSAR 3.6.3.1 h0R!!AL PIPE PIPL BREAK CRili kl A LOADS Y BREAK LOCATIONS FSAR TABLES 3.6-20, 3.0-21 AND TYPES f FSAR 3.6.2.2 JET BARRIER & PIPE WHIP 3.6.2.3 RESTRAINTS B&W TOPICAL REPORT BAW-10132P FSAR 3.6.3 FSAR G.2 FSAR 6.2 f PIPE REACTION BRfAK OPENING I4 ASS & ENERGY SUBCOMP A'illif N T ARIA / RELEASE (lA & 24) PR ESSilRE S \\ LOADS + 2 / FSAR 3.6.3 JET TARGET IDENilFICATION I I FSAR V
- 6. 2 B&W TOPICAL REPORT C dItfu rif.h l JET TARGET P
RE BAW-10132P AREA FSAR 3.6.2.2 3.6.2.3 f SUPPORT & RESTRAINT TARGET B&W TOPICAL REPrRT --pa Littli LOADS -op BAW-10132P JET LOADS (E.G. SN UD BE P.S ) fSAR 3.6.3 SY S I E t4 VIA RISPONSE TO LOCA LOAD NRC QULSil0N 110.30 1 ISSUE PREllMINARY HARDWARE SPECIFICATION FIX I W / CERflFY ACCI.PTANCE NO RL,'aNS( YES IN STRESS REPORTS STOP REFINr RESOLUTION -2 'CCEP.'hCE ANALYSIS 9 i ). ACCeiiRAff Of 151 j I
MEETING
SUMMARY
DISTRIBUTION H. Denton Docket File /NRrPDR MAY 151979 Local PDR TIC NRR Reading Branch File #4 E. Case D. Bunch R. Boyd D. Ross D. Vassallo W. Gammill J. Stolz R. Baer
- 0. Parr S. Varga C. Heltemes L. Crocker D. Crutchfield F. Williams R. Mattson R. DeYoung Project Manager Darl Hood Attorney, ELD M. Service IE (3)
ACRS (16) R. Denise L. Rubenstein NRC
Participants:
H. Daniels M. Bolotsky H. Conrad M. Boyle R. Hermann R. Stevens B. D. Liaw J. Strosnider J. Shapaker R. Bosnak S. Hou R. Stephens D. Pickett C. Tinkler 2143 177 .}}