ML20205M918
| ML20205M918 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah, 05000000 |
| Issue date: | 03/17/1986 |
| From: | Asselstine J NRC COMMISSION (OCM) |
| To: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML20205M922 | List: |
| References | |
| FOIA-86-810 NUDOCS 8604160512 | |
| Download: ML20205M918 (1) | |
Text
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- [t sano UNITED STATES A'
'o NUCLEAR RESULATORY COMMISSION
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March 17, 1936 CFF;CE OF THE COMMISSIONER Victor Stello, Jr.
".EMORANDUM FOR:
Acting D ecutive Director for Operations James K. Asselstine FROM:
MAIN STEAM VALVE VAut.T TEf1PERATURE
SUBJECT:
During the March 11, 1986 Cemission meeting, TVA listed the main steam valve vault temperature as a najor unresolved issue at Sequoyah and I would indicated that its resolution may require some modifications.
appreciate receiving a surriary of the rature of that issue, an indication discussion of how and when the of whether the issue is generic, a brie #
issue was first identified and, to the extent that the issue is generic, it is appropriately resolved at other what we have done to ensure that plants.
Chairman Palladino cc:
Comissierer Peterts Comissioner Bernthat Comissioner Zech Rec'd Ott. EDO OPE Da'e _3, j q. g 9 OGC SECY I ' "' f Time -
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MEMORANDUM FOR:
Comissioner Asselstine FROM:
Victor Stello, Jr.
Acting Executive Director for Operations
SUBJECT:
MAIN STEAM VALVE VAULT TEMPERATURES AT SEQUOYAH In response to your memorandum of March 17, 1986, we are providing the background information that you requested as an enclosure to this package. This enclosure is an internal NRR staff report, about one year old. However, we believe it is an appropriate response to your request on the nature of the issue and provides a brief discussion on how and when the issue was first identified. The staff has not issued a generic letter on this matter; however, all licensees were advised that this issue needed to be addressed by "IE Information Notice 84-90:
Main Steam Line Break Effect on Environmental Qualification of Equipment," dated December 7, 1984.
Should a licensee of an operating unit, as a result of its investigation of the superheat issue, find itself at variance with 10 CFR 50.49, appropriate steps must be taken within the context of the rule. As noted in the enclosed report, only the Westinghouse PWR units are affected since CE and BW took superheat into consideration.
In October 1985, Westinghouse provided the revised Topical Report on Mass and Energy Releases to its customers. This report will serve as the basis of the new calculations that must be done to determine the impact superheat will have on equipment qualification.
As discussed in the report, the effects of superheat are greatly dependent upon plant design. There are major differences in the way main steam lines are routed in nuclear facilities and in the quality of the pipe in different areas l
of these plaats.
Postulated pipe breaks inside containment present less of a problem with the superheated steam for the reasons explained in the enclosed i
report. The facilities are reviewed on a case-by-case basis due to possible different approaches to resolution of this issue. For example, differences in the plant designs for the Diablo Canyon, Catawba, and Millstone 3 facilities allowed alternate approaches to resolution in each case.
TVA is actively pursuing this matter for Sequoyah and we expect a schedule from TVA in mid-April 1986 for the detailed analysis and resolution of the superheat t
issue.
It is our understanding that TVA will be utilizing an approach to resolution similar to that used at the Millstone 3 site.
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Contact:
Carl Stahle, NRR l
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Commissioner Asselstine In summary, because of the issuance of IE Infdraation Notice 84-90 and the Westinghouse Topical Report on this subject to the affected utilities, the staff proposes no additional generic action at this time. We expect the owners of the Westinghouse PWR operating units to respond appropriately and in a timely fashion on this issue in accordance with 10 CFR 50.49.
Upon submittal of such information, the staff will review each facility on a case-by-case basis with a particular emphasis on the superheat issue.
Victor Stello, Jr.
Acting Executive Director for Operations
Enclosure:
As stated cc:
Chairman Palladino Commissioner Roberts Conmissioner Bernthal Commissioner Zech OGC OPE SECY
MAIN STEAM LINE BREAK WITH RELEASE 0F SUPERHEATED STEAM REPORT I.
INTRODUCTION 8
On December 7,1954, the NRC issued "IE Infomation Notice 84-90: Main Steam Line Break Effect en Environmental Qualification of Equipment".
In this notice, the NRC stated that the " staff considers steam superheating during steam generator tube bundle uncovery as a result of postulated main steam line breaks and subsequent release to compartments if omitted from plant analysis, to represent a potential deficiency in the equipment qualification required by 10 CFR 50.49."
In the notice it was also suggested that licensees review O
their main steam line break analysis with regard to this issue.
The NRC is currently completing the reviews of licensee programs for environmental qualification of electrical equipment and issuing the results of our review in Safety Evaluations. These evaluations are based on known temperature profiles both irdide and outside containment using mass and energy l
release data,celculated by previously acceptable codes and methods. To the i
extent these temperature profiles are modified by licensees as a result of the reviews suggested by the IE Information Notice, additional efforts may be required at that time to assure continued compliance with 10 CFR 50.49. This report presents the staff evaluation of efforts to date on the industry responses
- to the IE Information Notice.
principal response to date has come from the Westinghouse Deers Group. CE l
and B&W plants are, however, discussed in the Background section.
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2-II. BACKGROUND e
The environmental qualification programs for safety-relate'd electrical equipment must consider the time dependent temperature and pressure at the location of the equipment for the most severe design basis accident. For the
- MSLB event, the atmospheric temperature is calculated by the utilities for the applicable compartments using various vendor or AE thermal response codes which take into account the compartment volumes, heat sinks, and engineered safety features. The mass and energy released from the break is computed separately and supplied to the thennal response code as input. The mass and c
energy release from the break is typically provided by the reactor vendors and is a function of break size and location, power level, and steam generator type.
(A more precise and detailed discussion is provided in NUREG-0588
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- Interim Staff Position on Environmental Qualification of Safety Related ElectricalEquipment".)
i A methodology' for computing mass and energy releases for postulated MSLB accidents inside containments for Westinghouse g reactors was approved by the staff in August 1983. This methodology is described in Topical Reports WCAP-8821, and WCAP-8822. An earlier version of WCAP-8822 has been used on an interim basis for the licensing of most W design plants.
Interim use of l
WCAP-8822 (assuming no entrainment) was judged acceptable on the belief that the. computed mass and energy releases would result in conservative estimates of pressure and temperature effects.
In the course of our review of WCAP-8822, it was noted that the steam generator blowdown model dkd not account for the heat transfer from the uncovered portion of the steam s.-ve-,
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generator tube bundle to the escaping steam.
Final staff acceptance of WCAP-8822 was based, in part, on Westinghouse's inclusion of a medel to account for this heat transfer mechanism.
9 As a result of recent reanalyses by Westinghouse of their inass and energy release codes to account for steam superheating, as well as liquid entrainment, it was determined that certain break sizes would produce superheated blowdown and greater energy releases than had been previously calculated. The superheating of the steam occurs when the break area,is sufficiently large that feedwater flow to the steam generator cannot keep up with the steam loss through the break and the water level in the generator falls, thus uncovering the steam generator tube bundle. The steam produced in the generator is thus superheated by the uncovered portion of the tube bundle and is released through the break. Break sizes that do not lead to tube bundle uncovery do not produce appreciable superheat affects.
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r Because the magnitude of the,superheating effect and its duration following a i
postulated MSLB are plant specific and will vary as a function of several parameters (e.g. break size, containment heat removal system design, feedwater isolation time), Westinghouse notified each of the W plant owners of a possible unreviewed safety question concerning the temperature envelope used for the environmental qualification of equipment outside of containment. For equipment located inside Westinghouse large dry PWR plant containments, the impact of the superheat effect on the equipment qualification profiles was judged to be insignificant by W based on the results of scoping analyses performed for a large dry containment. However, a significant 1.npact on the equipment qualification profiles for equipment located inside ice condensers was predicted for the lower compartments.
Ice condensers are divided into
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4 upper and lower ccmpartments and all hava sprays in the upper compartment to remove heat, among other things. The lower compartments do not have sprays, except for D..C. Cook, and it is postulated that superheat releases in the region can result in higher temperature profiles.
The MSLB superheat concern appears to affect only Westinghouse PWR plants, since 1) B&W plants operate with steam generator superheat, which is accounted for in the mass and energy release calculations, and 2) CE mass and energy release codes currently model superheat if tube bundle uncovery is predicted to occur. Nevertheless, the staff has issued an infomation notice to the owners of all PWR operating plants, and applicants for licenses, informing them of the MSLB superheat concern.
The review of the program as it concerns each containment type and our evaluation of the justification for continued operation is presented below. A chronology of events related to the Nestinghouse Owners Group program is provided in Attachment 1.
III. Review of Program Resolution A.
Inside Ice Condenser Containments The Westinghouse methodology for computing mass and energy releases for postulated main steam line break accidents. Topical Report WCAP-8822, has been revised to account for possible superheating. For the ice condensers, the impact of superheated steam releases on lower compartment temperatures can be l
substantial since the lower compartment is not cooled by the containment i
sprays, except for D. C. Cook. Using the mass and energy release ~ data based on the methodology of WCAP-8822, the approved LOTIC-3 containment code i*
predicts compartment temperatures higher than equipment qualification temperatures. W has revised the LOTIC-3 code by adopting a new wall heat transfer model and by accounting for the energy removal capability of the ice condenser drains. These changes reduce the peak lower compartmey The revised temperatures to below equipment qualification temperatures.
LOTIC-3 code (Topical Report WCAP-8354, Supplement 3) is currently under staff review.
Based on the staff review of Catawba and Watts Bar, a number of concerns have been identified which require additional analysis. They primarily involve uncertainties in (1) the correlations used to predict the size of the droplets produced by the drain flow. (2) the simplifying assumptions used in determining the effect major equipment located in-the drain flow path would have on the droplet field, (3) the ice bed heat sink representation and
, corresponding ice-me.t rates, and "(4) the assumption that the lower l
compartment is well mixed. To provide an improved basis for the drain flow models and assumptions, Westinghouse has comitted to conduct a program of tests and analyses to specifically address each of the preceding concerns.
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This program'was initiated in mid-1984 and is scheduled to be completed in l
October-1985. The staff has concluded that interim operation of the ice condenser plants is acceptable pending completion of this work.
l For the above Topical Report review (WCAP 8822 and 8354, Sup. 3), the Catawba and Watts Bar plants are being used as the surrogate plant for supporting analyses. Other operating plants and plants under licensing review will be similarly considered following the conclusion of the Topical Report review.
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6 B.
Inside large Dry and Sub-atmospheric Containments On January 25, 1985 E met with the NRC staff to present the results of their latest sensitivity study on the effects of superheat releases following a MSLB inside large dry and subatmospheric containments. A study in 1942 using the MARVEL code with a simplified superheat model and a later study in 1984 using a modified version of the LOFTRAN code (incorporating a more detailed superheat model) has shown that the effect of superheat blowdown is not significant. Only one case studied showed an increase in temperature as a
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result of considering the superheat effect, and this temperature increase was considered by Westinghouse to be negligible. The W studies used a four loop reactor system with Model 51 steam generators as the limiting design. The
- four loop design has higher setpoints for spray and fan cooler actuation, and the model 51 steam generator has a longer tube design which leads to earlier l
tube bundle uncovery. W will forinally document the infomation provided at the January meeting in a supplement to WCAP-8822. This WCAP will provide the basis for concluding that the current mass and energy release calculations for large dry and subatmospheric containments (without superheat models) is sufficient for licensing, add therefore no reanalysis is needed. Staff review and if appropriate, approval Vf the supplement will complete the action on evaluating the impact of MSLB superheating inside containment for other than i
ice condenser plants. Further efforts will be conducted under the review of l
Topical Report WCAP-8822.
C. Outside Containments Those utilities with Westinghouse reactors need to detemine the temperature profiles resulting from a full spectrum of breaks in the main and branch stea-lines in order to detemine if the revised mass and energy releases result in higher temperatures than originally calculated in compartments odtside of containment. '(See additional requirements and criteria in SRP 3.6.1.) The t
revised analysis should assume the worst case single active failure concurrent with the pipe break. If the main steam lines in the valve compartments are designed to the break exclusion criteria of SRP 3.6.2 ("superpipe"), then only break sizes of one square foot and smaller need to be considered pnd a single active failure is not assumed.
If superheated steam is produced from these breaks, then the utility is required to verify that safety-related equipment is properly qualified to the new environmental conditions.
Those utilities th:t are seeking an operating license prior to November 30, 1985 must provide a justification for interim operation in the event there is equipment yet to be qualified at the time of licensing (10 CFR 50.49,1).
For plants currently being licensed, the staff is reviewing the revised temperature profiles for the affected compartments outside of containment.
The revised temperature profiles and justification for interim operation provided for the Catawba and Byron plants have been approved by the staff.
Pacific Gas & Electric (PG&E) has perfomed a specific calculation applicable
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only to Diablo Canyon, Unit 2 and ha's determined that insufficient superheated steam was generated to affect the qualification temperatures. The staff has approved PG&E's analysis suppcrting the previous temperature profiles.
Since June 6,1984, when Westinghouse first notified plant owners of a possible unreviewed safety issue, a number of meetings have been held and l
1etters exchanged discussing the Westinghouse Owners Group (WOG) program for resolution. The latest meeting was held on January 30, 1985. At that meeting, the WOG again described the program which would provide fomal documentation for mass and energy releases to each of the participating utilities. This program'is to be completed by August 1985. The NRC has
. projected that it would take 2-3 months following the WOG program for each licensee to obtain revised compartment analyses using the mass and energy release data. These analyses by the licensees, which should be complete about December 198b, would redefine the temperatura profiles in compartrents outside containment and would establish whether or not the existing equipment qualification for the area was appropriate. At that time plants with equipment qualification problems as a result of new temperature profiles would be expected to correct them in accordance with the provisions of 10 CFR 50.49.
The Westinghouse Owners Group Program for outside containment has been l
reviewed and found acceptable for those licensees referencing the program in l
1 response to the IE Infonnation Notice 84-90 issued December 7,1984.
IV. Environmental Qualification Concerns At this time we do not believe that any additional regulatory requirements are necessary to resolve the identified deffefencies.
' Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants " is specific on requirements and applicable provisions for operating plants and those being licensed.
In regards to equipment l
qualifications as a result of a MSLB with superheat releases, the following aspects, among other things, must be properly considered and addressed in order to demonstrate compliance with the requirements of 10 CFR 50.49.
In NUREG-0588, the staff has endorsed the suggesteJ values for margins l
identified in Section 6.3.1.5 of IEEE Std 323-1974 with the exception of
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radiation. These values include a +15'F margin on temperature and'the greater of +10% or 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> on the required operating time.
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.g Some equipment may be required to perform its safety function only within the first few minutes or hours into an accident. This equipment should remain functional in.the accident environment for a period of at least I hour in excess of the time assu.ned in the accident analysis unless a lesspr time margin can be justified. This justification must include, for each piece of equipment, (1) consideration of a spectrum of breaks, (2) the potential need.
for the equipment later in an event or during recovery operations. (3) a detennination that failure of the equipment after performance of its safety function will not be detrimental to plant safety or mislead the operator, and
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(4) a determination that the margin applied to the minimum operability time, when ccabined with the other test margins, will account for the uncertainties associated with the use of analytical techniques in the derivation of O-environmental parameters, the number of units tested, production tolerances, and test equipment inaccuracies..
If qualification is based on' thermal lag analysis, it must be shown that the surface temperature of the component, plus appropriate margin, does not exceed I
the qualification temperature.
If the calculated surface temperature plus margin exceeds the qualification temperature, the staff has required that (1)
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additional justification be provided to demonstrate that the equipment can l
maintain its required functional operability, or (2) requalification testing be perfonned with appropriate margins, or (3) qualified physical protection be provided to assure that the surface temperature plus margin will not exceed the actual qualification temperature.
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V. Justification for Continued Operation of Operating Ice Condenser Plants and Plants Having large Dry Containments e
i Although the main steam line break (MSLB) accident is an important design i
basis accident in determining several design features of the plant, the accident sequences associated with MSLB generally are not regarded as major contributors to public risk; by itself, a MSLB will not uncover the reactor As long as either main or auxiliary feedwater is available, core core.
cooling is achievable. For the particular concern here, i.e., superheated conditions causing equipment failure, the equipment which fails must in turn cause failure of auxiliary feedwater, main feedwater and alternate sources of coolant to all steam generators in order for a core melt to be of concern.
The events thus fonn a sequence of probabilities which, when properly considered with other decay heat removal strategies such as feed-and-bleed, make the MSLB with superheat releases.less of a concern as an accident.
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le The letters d,ated August 2, 3 and 8,1984 from the licensees for Sequoyah, Cook, and McGuira respectively concern the high energy line break superheat effect inside containrent. The letters provide justification for continued operation for the operating ice condenser plants based on the results of containment analyses perfonned using a revised containment respor.se code.
The July 26 and August 20, 1984 letters from the Westinghouse (wner Group (WOG) concerning the MSLB superheat effect outside containment provide justification for continued operation (JCO) based on 1) preliminary analysis provided by Westinghouse and 2) a review of equipment located in pffected areas. Each of these justifications is discussed below.
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, The Sequoyah containment analysis, based on traditional licensing assumptions, was compared by TVA to a revised analysis performed by Westinghouse using updated mass and energy release data for Catawba and a modified version of the LOTIC-3 containment code which accounted for heat transfer to ice, condenser drain flow. TVA found that, in the revised analysis, the lower compartment and dead ended compartment temperatures were less than or equal to the original qualification temperature. For the valve room outside containment.
TVA detemined that sufficient time exists for safety related equipment to perform its safety function and that sufficient instrumentation exists for the identification and mitigation of the MSLB accident. The TVA assessment was based on the use of Catawba (inside containment) mass and energy release data with consideration of line losses between the steam generators and valve vault.
McGuire, like Sequoyah, has an ice condenser containment without sprays in the lower compartment. For McGuire, W perfomed an analysis similar to that for Se quoyah. The results show peak containment temperatures below that originally predicted for the MSLB accident. For MSLB accidents outside containment,' Duke's analysis has shown that affected equipment and instruments will operate before the superheated steam produces temperatures above the qualification profile and
- hat sufficient equipment and instruments are available to control or mitigate the MSLB accident.
In this region valve operator heater circuits, which might be affected and cause reopening of some valves, have been disconnected.
For D. C. Cook, the lower compartment inside containment has sprays and the superheat issue is not expected to be a concern.
Foroutsideconkainment,the licensee evaluated all of the potentially affected equipment and
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instrumentation needed to perform a safety function, assuming their function must be complete within 1.5 minutes, which is the time for the onset of superheat steam release to reach the valve vault. The licensee concluded that t
the equipment would function before the allotted time or other equipment or instrumentation exists to control or mitigate the MSLB.
For outside containment, the WOG has developed several generic arguments related to MSLB accidents for use by member utilities in justifying continued operation of their plant. The generic arguments rely upon (1) the results of
}{ scoping studies which indicate isolation valve activation occurs before superheated steam release begins, and 2) the fact that many plants have open areas surrounding the steam lines which lessens the effect. The WOG also presented additional arguments for consideration including the mechanistic treatment of high energy line breaks and their probability of occurrence, the thermal lag of equipment exposed to superheated steam, the unlikely single failure of a main steam isolation valve (MSIV), and the small probability that
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a break will occur between the containment wall and MSIV.
We have considered all elements of the generic JCOs and find that operating plants which can justify their applicability may continue operation for the interim period until plant specific analyses, if required, can be completed.
For the operating ice condenser plants the staff's topical report review (discussedabove)hasprogressedtothepointwherecontinuedoperationis justified 'for the interim period pending final resolution and plant specific application of the revised Westinghouse containment code.
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VI Conclusion The IE Information Notice 84-90 issued on December 7,1984 has been responded to by the Westinghouse Owners Group on behalf of specific participants. We have reviewed the WDG program and schedule and find them reasonable for the work to be done. We have also reviewed the generic justifications for continued operation provided by the WOG and agree that the program participants rey use those applicable to their plant and configuration.
Furthermore, the provisions of 10 CFR 50.49 are sufficient for resolution of environmental qualification if new temperature profiles are found necessary.
The Topical Report Program should resolve the problems with inside containment for Westinghouse plants. Any action necessary as a result of those reviews 1
will be handled separately as appropriate.
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ATTACHMENT I Chronology of Events - Westinghouse Owners Group (WOG) Program 1977 NRC issued inquiry regarding heat transfer from uncovered steam generator tube bundles, as part of WCAP-8821/8822 review December 1979-NRC issued NUREG-0588, " Interim Staff PosMtion on Environmental Qualification of Safety-Related Electrical Equipment".
Appendix' A lists WCAP 8822 as under review but acceptable for mass and energy release calculations provided entrainment is not assumed.
February 1982-Westinghouse reported results of sensitivity analysis to NRC which indicates that containment temperature response may exceed environmental qualification temperature profiles for ice condenser lower compartments as a result of MsLB.superheat.
February 1983-Catawba SER issued which required a refined M5LB analysis to account for lower compartment temperature increases.
June 6,1984-Westinghouse notified owners of possible unreviewed safety question concerning temperature profiles for compartments outside containment. Westinghouse was continuing efforts for ice condensers lower compartments reported as having been addressed and detemined to be negligible.
July 20, 1984-Meeting between NRC, ice condenser owners and Westinghouse.
It was agreed that owners of operating Tce condenser plants will provide JCO's, and plants under licensing review will be handled on a plant-specific basis.
Westinghouse has notified the W Owners Group's (WOG) Regulatory Response Groups (RRG) of potential problem outside Westinghouse plants - (See meeting suninary July 30,1984).
July 26, 1984 - Letter from WOG RRG on activities since W June 6:
Issued Nuclear Notepad to Owners on July 19.
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- On July 23, began defining a generic program for owners consideration.
- Concern confined to outside containment.
August 2, 3, 8, 1984 - JCO's provided by Sequoyah, Cook and McGuire l
on ice condenser lower compartment and outside containment problems.
August 2, 1984 - Meeting between WOG and NRC. WOG proposing a two phase program to provide generic JC0 (Phase I) and have Westinghouse provide " typical" mass and energy release rate data for MSLB both inside and outside containment based on the category of plant, e.g.,
nunber of loops (Phase II). See meeting summary September 6,1984.
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August 20, 1984 - Letter from WOG in response to the August 2 meeting, fonnalizing the program for outside containment. Regarding this matter inside containment the letter states: Westinghouse has '
advised RRG that this issue inside containment has been addressed in the NRC review of a topical report and believes it to be resolved.
Any further review of high energy line break inside of containment is very plant specific and generic discussion is not pdhsible".
October'24, 1984 - WOG letter providing status of efforts.
Phase I: generic JCO's provided to owners on 9/11/84.
Phase II: being revised.
November 19, 1984 - Telecon update of WOG efforts. Three groups of plant owners' identified; Group 1 - have no problem due to configuration of compartments outside containment, equipment location, etc. Group 2 - can provide justification based on pipe break mechanics and other arguments, Group 3 - will pursue W assistance to resolve issue. New Phase II to be discussed at December WOG meeting, January 85 is earliest date to meet with NRC on status.
l January 30, 1985 - WOG Meeting with NRC.
Utility participants identified.
Phase II mass and energy release calculations and final documentation to be complete by August 1985.
NRC ~ expects revised temperature profiles to be in place at each utility be end of 1985; equipment qualification, if necessary, to,. follow.
L February 25,1985 - WOG 1etter listing participants, restarting generic JCO's, and confiming schedule from January 30, 1985 meeting.
l March 21,1985 - WOG 1etter including TVA as a participant.
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