ML20010D730
| ML20010D730 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 08/14/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Davidson D CLEVELAND ELECTRIC ILLUMINATING CO. |
| References | |
| NUDOCS 8108310090 | |
| Download: ML20010D730 (16) | |
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Docket Files 7FC18f (16)
NRC PDR AUG 141931 Local PDR NSIC TERA Docket Hos. 50-440 TIC and 50-441 LB#2 File Attorney, OELD DEisenhut Mr. Dalwyn R. Davidson RPurple Vice President - Engineering RTedesco The Cleveland Electric Illuminating ASchwencer Company DHouston Post Office Box 5000 MService Cleveland, Ohio 44101 I&E (3)
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Dear !!r. Davidson:
JKudr M Subject; Request for Additional Information - Containment Systems In the performance of the Perry licensing review, the staff has identified concerns in regard to containment systems. The infomation that we require is identified in the enclosure.
We request that you provide the infomation not later than October 1, 1981.
If you require any clarification of this request, please contact M. D. Houston, Project Manager, (301) 492-8593.
Sincerely, 8I881'E18u0d bys Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
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AUG 141981 Mr. Dalyyn R. Dhvidson Vice President, Engineering The Cleveland Electric Illuminating Company P. O. Box 5000 Cleveland, Ohio 44101 cc: Gerald Charnoff, Esq.
Shaw, Pittman, Potts & Trowbridge 1800 M Street, N. W.
Washington, D. C.
20036 Donald H. Hauser, Esq.
Cleveland Electric Illuminating Company P. O. Box 5000 Cleveland, Ohio 44101 U. S. Nuclear Regulatory Commission Resident Inspr
's Office Parmly at Cen
<oad Perry, Ohio
- <081 Donald T. Ezzone, Esq.
Assistant Prosecuting Attorney 105 Main Street Lake County Administration Center Painesville, Ohio 44077 Tod J. Kenney 228 South College Apt. A Bowling Green, Ohio 43402 Daniel D. Wilt Wegman, Hesiler & Vanderberg 7301 Chippewa Road, Suite 102 Brecksville, Ohio 44141 Jeff Alexander 920 Wilimington Ave.
Dayton Ohio 45420 Terry Lodge, Esq.
915 Spitzer Building Toledo, Ohio 43604 r e
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480.0 CONTAINMENT SYSTEMS BRANCH REQUEST FOR ADDITIONAL INFORMATION N PERRY NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET N05. 50-440/441
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480.2 General Electric has provided, on the GESSAR-II docket, the latest loads criteria for both SRV and LOCA related pool dynamic loads. The staff is currently reviewing these load specifications.:.ad will pre-pare a technical evaluation report by early.1982.
In order for the staff to complete its review of th2 pool dynamic load definitions for year facility, we need the following:
a) A statement of your intention to completely utilize the load definitions presented in GESSAR-II as modified, if necessary,'
by the staff's evaluation report; or b) A detailed list of the exceptions to tie generic load definf-tion criteria (as modified by the rtaff's evaluation report) that will be used in the design of your facility.
480.3 Provide detailed plan and section drawings of the TIP station, equip-ment hatch, personnel hatch, and any other structure located within
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20 feet of the suppression pool surface whose width is greater than 20 inches. Show on these drawings your plan to extend these struc-tures into the suppression pool, thus el'iminating impact loads due l
to pool swell.
l 480.4 In the subcompartment pressure analyses of the reactor water cleanup rooms, that is, the heat exchanger room, filter demineralizer valve room, filter demineralizer room, and drain <alve nest room, blowout ENCLOSURE
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. panglsareshowntobeincorporatedintothedesignofeact$ room.
Identify the vent area used in the subcomp@tm'ent' analyses that were provided by the blowout panels. For these vent areas, discuss how t.ie flow area and flow resistance varies witt. time. Provide the ex-perimental data that support these assumptions or propose a testing program that will demonst ate.this capability.' Also, provide an analy-sis that shows there wil' be no missiles generated.
480.5 For each subcompartment analyzed, provide the following information:
a) Describe the nodalization sensitivity study performed to deter-mine the minimum number of volume nodes required to conserva-tively predict the maximum pressure load acting on the compart-ment. structure. The nodalization sensitivity study should in-clude consideration of spatial pressure variation; e.g., pres-sure variation circumferentially, axially, and radially within the compartment. Describe and justify the nodalization sensi-tivity study performed for the major component support; Walu-ation,':nere transient forces and moments acting on the compo-nents are of concern.
b) Discuss the manner in which movable obstructions to vent flow (such as insulation, ducting, plugs, and seals) were treated.
Provide analytical and experimental justification that vent areas will not be partially or completely plugged by displaced objects. Discuss how insulation for piping and components was considered in determining volumes and vent areas.
3-c) Provide the projected area used to calculate these loads and
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identify the location of the area projectibns dn plan and sec-tion drawings in the select'ed coordinate system. This informa-tion should be presented in such a manner that confirmatory evaluations of the loads and moments can be made.
f d)
Provide the peak and transient loadings on the major components used to establish the adequacy of the supports design. This shold include the load forcing functions [e.g., f (t), f (t),
x fy(t)] and transient moments [e.g., M (t), M (t), M (t)] as re-X y
solved about a specific, identified coordinate system.
480.6 The provisions of Branch Technical Position (BTP) CSB 6-3 were not used in considering potential bypass leakage paths around the leakage collection and filtration system of the secondary containment. For example, the fact that lines penetrating the primary and s:condary containment have isolation valves does not preclude through-line leak-age. Therefore, identify all potential bypass leakage paths using the guidelines of Item 5 of BTP CSB 6-3.
Also, provide a realistic leakage rate for these potential bypass leakage paths and a discussion (includ-ing drawing) of provisions made to permit pre-operational and periodic leakage rate testing in a manner similar to the Type B or C tests of l
Appendix J to 10 CFR Part 50 for each bypass leakage path.
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480.7 Discuss the initial test program that will be carried out to verify the depressurization of the se:ondary containment to a negative pres-l sure of 0.4 inch water gauge. Provide the acceptance criteria for 1
N the drawdown time.
It is our position that'a [eriodic test program
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also be established to determine the operability of the system.
Also, provide a description of the pre-operational and periodic test program that will determine the secondary containment infiltration rate, including the acceptance criteria.
480.8 Heat loads from equipment inside the auxiliary building (e.g., the ECCS pumps) were not considered in the drawdown analysis for the secondary containment. Demonstrate that these heat loads will have no effect on the drawdown time in the secondary system or redo your analysis for the drawdown time considering all possible heat loads.
480.9 Provide the following additional information related to potential bypass leakage paths.
a) For each air or water seal, perform an analysis that will demon-strate that a sufficient inventory of the fluid is available to maintain the seal for 30 days, and describe the testing program and proposed er.tries for the Technical Specifications that will verify the assumptions used in the analysis. Provide the basis for the valve fluid leakage used in the analysis; and b) For each of these paths where water seals eliminate the potential for bypass leakage, provide a sketch to show the location of the water seal relative to the system isolation valves.
480.10 From'ihe discussion in Section 6.2.4.2.3 of the FSAR it is not clear whether or not debris screens are included in the design of the con-tainment purge system.
It is our position that Section B.I.g of BTP CSB 6-4 should be met. Guidance is provided below which, if followed, would represent,an acceptable debris screen design.
a) The debris screen should be seismic Category I and installed typically about one pipe diameter away from the inner side of the inboard isolation valve.
b) The piping between the debris screen and the valve should also be seismic Category I design.
c) The debris screen should be designed to withstand the LOCA differ-ential pressure; d) The debris screen onening typically should be about 2 inches by 1 3/16 inches.
State your intention to comply with our position and provide a descrip-tion of tne debris screen design.
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In Section 6.2.4 of the FSAR, an analysis for the containment vessel 480.11 and drywell purge system was provided assuming that the containment atmosphere was released through an 18 inch purge line. However, from the discussion, it is not clear whether or not the drywell purge sys-tem was assumed to be operational.
State whether or not this system was assumed to be operating for the analysis provided. Provide a dis-cussion of why this assumption is conservative.
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480.12 Section 9.4.6 of the FSAR states that the centainment atmosphere is continuously exhausted dur'ing normal operation. However, we believe that purging / venting should be miniinized during reactor operation be-cause the plant is inherently safer with closed purge valves than with open lines requiring valve action to provide containment isolation.
In fact, serious consideration should be given to a plant design such that purging / venting is not required during operation. Therefore, pro-vide a detailed discussion of the reasons why the Perry Nuclear Power Plant needs to purge, and an estimate of the number of hours per year that purging is expected through each particular valve.
480.13 From the discussion of the analysis of the containment vessel and dry-well purge system, it is not clear whether or not the containment at-mosphere was assumed to be released through only the 18 inch exhaust line or through both the supply and exhaust lines. Clarify this point and justify your response if the containment atmosphere was assumed to be released through only the 18 inch exhaust line.
480.14 In Table 6.2.32, the primary and secondary mo,de of actuation for the containment isolation valves are given. The information provided does I
not state whether or not power-operated isolation valves are automat-ically operated upon receipt of a containment isolation signal as the primary mode of actuation or if these valves can be remote manually operated from the main control room as the secondary mode. Provide this information.
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N 480.15 Figure 6.2.60 shows the arrangement of the v,arious, isolation valves listed in Table 6.2.32.
Many of the lines penetrating the contain-ment have test lines between.the isolation valves. Provide justifi-cation why these test lines should not be treated as branch lines and incladed intthe containment isolation valve tables and tested in accordance with Appendix J.
480.16 Table 6.2.32 lists the positions of the containment isolation valves for post-accident conditions. Many of the valves that are required to be open or closed during an accident fail in the "as is" posit.fon upon loss of power.
It is our position that all containment isolation valves fail in the position of greater safety in the event of power failere to the valve operator during an accident. Therefore, justi-fication should be provided to demonstrate compliance with the above position or the appropriate plant modifications should be made.
480.17 Some systems have lines that are sealed from the containment atmos-phere because their lines terminate below the water level of the sup-pression pool. Therefore, these systems are not vented and drained for the Type A containment leak rate test. However, to be considered a sealed system, the piping between the suppression pool and isolation valves should meet the following requirements:
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- 1) The piping is protected against missile and pipe whip; 2)
The piping is designated seismic Category I; and
- 3) The piping is classited Safety Class 2.
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N State whether or not the piping between the su,ppression pool and isola -
tion valves meet the above requirements for the penetration mentioned abov e.
Also, specify the fluid that is used to pressurize the valves to perform the Type C test.
f 480.18 Closed systems outside containment having a post accident function be-come extensions of the containment boundary following a LOCA. Certain of these systems may also be identified as one of the redundant con-tainment isolation barriers. Since these systems may circulate con-taminated water or the containment atmosphere, system components which may leak are relied on to provide containment integrity. Therefore, discuss your plans for specifying a leakage limit for each system that becomes an extension of the containment boundary following a LOCA, and leak testing the systems either hydrostatically or pneumatically.
Also, discuss how the leakage will be included in the radiological as-sessment of the site.
480.19 Discuss and schematically show the design provisions that will permit the personnel airlock door seals and the entire airlock door seals to be tested. Discuss the design capability of the door seals to be leak tested at a pressure of Pa; i.e., the calculated peak containment internal pressure.
If it will be necessary to exert a force on the doors to prevent them from being unseated during leak testing, describe the provisions for doing this and discuss whether or not the mechanism can be operated from within the airlock. Also, discuss how the force on the door will be monitored.
N 480.20 Provide a complete drawing for each closed 3ystem outside containment for which credit is claimed 'as an ~ isolation barrier. Show all piping connecting to the closed system up^ to a second isolation barrier.
Identify all lines connected to the closed systems that leave the secondary containment.
480.21 Figure 6.2.60 shows the isolation valves and the test connections.
However, from the figures it is not clear what test connection is going to be used to test which valve. Provide this information. For any valve that is being tested for leakage in which the pressure.is*
applied in the opposite direction as that when the valve would be re-quired to perform its safety function, provide justification that the results from the tests for the pressure applied in a different direc-tion will provide equivalent or more conservative results.
480.22 With regard to system venting and draining for the Type A containment leak rate test provide the following information:
1)
Itemize each system penetrating containment and discuss the vent-ing and draining provisions for each system.
Systems that are not designed to remain intact following a LOCA should have the isolation valves exposed to the containment atmosphere to permit the test differential pressure to be applied across them; i.e.,
the system should be vented and drained both upstream and down-stream of the isolation valves. For each system penetrating con-tainment that is rat vented and drained, provide justification,
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N 2)
Identify any gas filled lines that will, pot.be. vented for the.
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Type A test and provide' justification for not doing so.
429.23 List the systems which penetrate the containment and are not vented and drained for-the Type A containment leak rate test.
Those systems that are not vented and drained for the Type A test must meet the following requirements:
- 1) The system is protected against missiles and pipe whip;
- 2) The system is designated seismic Category I;
- 3) The system is classified Safety Class 2;
- 4) The system pressure is greater than the containment pressure at all times during the course of the accident;
- 5) The system will remain full of water for 30 days; and
- 6) Botn items 4 and 5 will be maintained when a single active failure is assumed in the system.
State whether or not these systems meet the above requirements.
480.24 Discuss the capability available to detect leakage and to take ap-propriate action in those lines needed for safe shutdown of the plant, or which are part of the engineered' safety features, and that have a remote-manual valve or check valve inside containment or a remote-manual valve outside containment.
480.25 For those lines that are required to be open following an accident which have either two or more isolation valves outside containment
N and no isolation valves inside containment,,or a single remote-man-ual isolation valve outside containment.in a closed system and no isolation valves inside containment, show that:
- 1) The remote-manual isolation valve nearest the containment and the piping between the containment and the valve is enclosed in a leak-tight or controlled leakage housing; or
- 2) The design of the piping up to and including the first remote-manual isolation valve conforms to the provisions of SRP Sec-tion 3.6.2.
In either case the design of the valve and/or the piping compartment should provide the capability to detect leakage from the valve shaft and/or bonnet seals and to terminate the leakage.
480.26
'n'here more than one bellows is utilized on a penetration, provide as-surance that each bellows will be subjected to Type B testing.
480.27 Table 6.2-40 identifies those containment isolation valves that will not be Type C tested. Therefore, justify that they do not constitute potential containment atmosphere leak paths following a LOCA. In this regard, a water seal may be shown to exist that will preclude contain-ment atmosphere leakage.
If this approach is taken, discuss how a water seal can be established and maintained using safety grade pipes and com-ponents, and considering single failure of active components. System drawings showing the i-outing and elevation of piping should be used to show the existence of a water seal.
N When operation of a system is need'ed to m'aiItain a'watte seal in khe system, the ECCS for example, show that the system will keep its water seal for a sufficient period of time if the system is removed from op-eration.
t 480.28 Identify (1) the location of the hydrogen sample points in the drywell and suppression chamber; and (2) the location of CGCS suction and dis-charge points, with respect to local structures and equipment.
480.29 The statement is made in Section 6.2.5.3.a of the FSAR that there will be no hydrogen generation resulting from spray water contact with any aluminum or zinc components because the containment spray water con-tains no boron. The staff has determined that hydrogen release from aluminum and zinc corrosion following a postulated loss-of-coolant ac-cident should be considered in the analysis of hydrogen production and accumulation within the containment. Therefore, justify your position that corrosion of aluminum and zine will not occur, or provide the fol-l lowing:
(a) the -
os on rate as a function of temperature for all materials in the.s.. =.nment that could become a source of hydrogen due to corrosion; (b) describe how the corrosion rates for the various materials were es-tabitsbed.
In so doing, identify the experimental data used as a basis (and provide references) and discuss the conservatism in the I
applicability of the data in view of the environmental conditions that are expected following a LOCA;
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A (c) Graphically show the hydrogen concentradon~inside the containment
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as a function of time, with no recombiners operating, with one re-
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combiner operating (minimum engineered safety features), and with both recombiners operating; and 1
(d) Graphically show the contribution of each source of hydrogen as a function of time.
In either case, provide the mass and surface area of aluminum, alumi-num paint, zinc paint, galvanized steel and other corrodible material
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in both the drywell and the wetwell for our confirmatory analysis'.
480.30 The accident at Three Mile Island, Unit 2 involved a large amount of metal-water reaction in the core with resulting hydrogen generation well in excess of the amounts considered in 10 CFR Section 50.44 of the Comission's regulations. During the past year the staff has been studying the potential of excess hydrogen generation, the effects such concentrations of hydrogen would have on the various types of plants, and the effectiveness of various mitigation systems in protect-ing the plant against such situations. The results of our studies to date are presented in the SECY 80-107 series'of documents.
In these re-ports, we recomend that all BWR Mark I and II containment plants be inerted and that owners of all other plants be. required to provide a proposed design (or designs) to mitigate the consequences of large amounts of hydrogen in containment. The associated proposed interim I
rule was published in The Federal Register on October 2,1980.
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N Subsequent to the issuance of SECY-80-107, a~subst4ntial amount of ad '
ditional work has been performed on this issue with emphasis on ice condensers. With respect to-ice condensers, and specifically Sequoyah, the Commission has decided that the matter of hydrogen control for de-gradedcoreaccidentsinplantswithsmallcontiinmentsneedstobere-solved in in the near term, i.e., the resolution should not be deferred to rulemaking.
In SECY 80-107, the staff showed that Mark III containments are similar to ice condenser containments in regard to their ability to accommodata large amounts of metal-water reaction without jeopardizing containment.
integrity.
We, therefore, request Cleveland Electric Illuminating Company to provide a description of its program to improve the hydrogen control capability at the Perry Nuclear Power Plant, Units 1 and 2.
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