ML19294C194
| ML19294C194 | |
| Person / Time | |
|---|---|
| Issue date: | 05/31/1978 |
| From: | Almeter F, Eisenhut D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19294C184 | List: |
| References | |
| REF-GTECI-A-04, REF-GTECI-SG, TASK-A-04, TASK-A-4, TASK-OR ACRS-SM-0151, ACRS-SM-151, NUDOCS 8003070301 | |
| Download: ML19294C194 (11) | |
Text
O Task A-4
}.;
f C'O'MBUSTION ENGINEERING STEAM GENERATOR Tubi INTEGRITY Lead NRR Organization:
Division of Operating Reactors (DOR)
Lead Supervisor:
Darrell G. Eisenhut, A/D for Systems and Projects, 00R Task Manager:
Frank M. Almeter, EB/ DOR Applicability:
Combustion Engineering Pressurized Water Reactors Projected Completion Date:
December 31, 1979 7
8003040 301
w Task A-4 Rev. No. 1 May 1978 1.
DESCRIPTION OF PROBLEM Pressu water reactor operating experience during the past five years shown that steam generator tube integrity can be degraded rh by co ion induced wastage, cracking, reduction in tube. diameter
~
(den
) and vibration induced fatigue cracks.
Since the steam generator tubes are an integrated part of the reactor coolant pres-sure boundary in the PWR system, the primary concern is the capabil-ity of degraded tubes to maintain their integrity during normal operation and under accident conditions (LOCA or a main steam line break) with adequate safety margins.
Palisades has been the only Combustion Engineering designed plant to experience tube degradation due to wastage and secondary side stress corrosion cracking with the use of a phosphate treatment for the secondary coolant.
Both types of degradation have been nominally arrested by conversion to AVT chemistry control.
However, tube degradation due to denting (but to a lesser degree than the Westing-house steam generators) occurred after the conversion to an AVT chemistry Recent inservice inspections at two sea coast facilities wi?.h CE designed steam generators, which used an AVT chemistry for the secondary coolant since initial startup, have shown that the prior use of phosphates is not a necessary precursor to cause denting in steam generatoi tubing.
Denting which leads to primary side stress corrosion cracking is the major problem at present and the principal focus of this technical activity.
However, as steam generator operating experience is accumulated and interpreted, it has become evident that condenser cooling water in-leakage resulting from the corrosion of condenser tubes can contaminate the secondary water of PWR steam generators and may be the principle source leaking to all types of steam generator tube degradation.
It has also become evident that the maintenance of secondary coolant water quality can-not be achieved if condenser in-leakage is allowed.
Because the condenser is an important component of the PWR secondary system, an approach must be developed to minimize condenser in-leakage to ensure adequate steam generator tube integrity.
2.
PLAN FOR PROBLEM RESOLUTION I
The problem will be resolved by reviewing the type and mechanism of tube degradation in operating reactors to evaluate the effects of tube structural integrity and failure probability under normal operation and accident conditions (LOCA, SSE and MSLB).
Assessment of the effects of degraded tubes on postulated accident conditions will be factored into the development of new criteria for tube plugging, acceptable levels of primary to secondary leakage, and ISI require-ments to ensure the safe operation of operating pressurized water A-4/1 m
fask A-4 Rev. No. 1 May 1978 reactors.
inimize tube degradation, priority areas where improvemen team generator design and criteria for the secondary care needed will be identified to develop licensing is).
coolant sys 54 positions the CP/0L review of new plants.
The specific activities directed at resolution of the denting problem in Combustion steam generators consist of the following issues and tasks:
A.
Generic Evaluation of ISI Results Review and evaluate the various eddy current inspection results; i.e., experience from operating reactors and evaluate these data as they relate to the generic determination of failure probabil-ity of degraded tubes.
In addition, evaluate the test programs and analytical studies to provide staff understanding suffi-cient to continue to provide justification of continued safe operation of operating reactors.
B.
Evaluation of Transients and Postulated Accidents Evaluation of failure consequences under postulated accident conditions (LOCA and MSLB) to determine the acceptable levels of primary to secondary leakage rates and the effect on ECCS The results will be used to define the acceptable performance.
number of tube failures that may be necessary as a licensing basis considering predicted fuel behavior and radiological dose during transients and postulated accident conditions.
Evaluation of Steam Generator Tube Structural Integrity C.
Evaluation of licensees' and CE's analysis of structural integ-rity of tubes under normal operating and accident conditions (LOCA, SSE and MSLB).
Information developed in this task will provide input for establishing a generic tube plugging criteria and recommendations for the revision of Regulatory Guide 1.121.
D.
Establish Tube Plugging Criteria Establish a generic tube plugging criteria that is consistent with the determined allowable leak rate, tube structural integ-These results will allow assess-rity and degradation rates.
ment of the adequacy of the requirements defined in Regulatory Guide 1.21.
A-4/2 h
Task A-4 Rev. No. 1 May 1978 E.
Secondary Coolant Chemistry Requirements
}
y.
" Evaluate the mechanism of tube degradation.
The results will be used to define the requirements for secondary coolant chemistry control including considerations for condenser in-leakage.
F.
Evaluation of ISI Methods Review the development of improved eddycurrent probes, coils and multi-f requency techniques to better quantify dents and growth of dents and increase sensitivity for detecting cracks j
in dented regions.
G.
Establish Criteria for Revision of Regulatory Guide 1.83 Integrate experience from inservice inspection results, the results from the evaluation of various ISI improvements and the plugging and secondary water chemistry requirements into criterion for possible revision of Regulatory Guide 1.83.
H.
Review Design Criteria for Plants Not Yet Licensed Review and evaluate design modifications proposed by applicants and CE to prevent denting in plants not yet licensed for operation.
3.
BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLE-TION OF TASK The safety issue addressed by this Task Action Plan is applicable to Pressurized Water Reactors with Combustion Engineering (CE) steam generators.
For CE PWRs currently licensed for operation, we have concluded that, pending completion of this TAP, continued operation does not constitute an undue risk to the health and safety of the public for the following reasons:
Primary to secondary leakage rate limits, and associated surveillance requirements, have been established to provide assurance that the occurrence of tube cracking during operation will be detected and appropriate corrective action will be taken such that no individual crack will become unstable under normal operating, transient or accident conditions.
A-4/3
Task A-4 Rev. No. 1 May 1978 Augmen Dservice inspection requirements and preventative
- L tube p ng criteria have been establishad to provide assurance
'y(
that,
great majority of degraded tute will be identified and removed from service oefore leakage develops.
Steam generator water chemistry control requirements are being considered to provide additional assurance that the potential for tube degradation during operation is minimized.
On a case-by-case basis, additional measures have been taken to (1) minimize contamination of the secondary coolant by in-leakage of condenser cooling water (e.g., condenser tubes with improved corrosion resistance have be installed) and (2) minimize buildup i'
in the steam generators of corrosion products generated in the secondary system (e.g., full flow condensate demineralizers have been installed).
Tube denting at tube / support plate intersections in CE designed stea.n generators has not been severe enough to result in through-wall cracks at dented locations.
However, if tube cracking were to occur in the dented region, it would be constrained by the support plates which would control crack stability and prevent tube failure (bursting) during postulated accidents.
Even if a LOCA or a MSLB were to occur during operation and some tubes were in a state of incipient failure, the radiological consequences of such an event would not be savere.
Continuous feedback from operating experience and the TAP efforts will be utilized to update interim criteria and requirements.
For plants experiencing severe degradation, the following additional interim bases were also considered:
The probability of the design basis accident during normal operation is small and the probability that the accident would occur during the short period of time between the leak detection and the plant shutdown is even smaller.
Even if an accident occurs when there are cracked tubes, the conservatively calculated consequences are still acceptably small until plant shutdown.
A small amount of leakage (e.g., less than the Technical Specification limit) can be tolerated during normal operation without exceeding the offsite dosage limits of 10 CFR Part 20.
A-4/4 h
Task A-4 Rev. No. 1 May 1978 The ie-mentioned rationale which constitutes the basis for
$-jk.
$ operation of licensed CE PWR facilities also support Tji con co ed licensing of new facilities.
Further, to the extent that is cticable, depending on the status of the design, fabrication and installation of the steam generators for facilities not yet licensed for operation, " state-of-the-art" design improvements and operating procedures which eliminate or at last minimize the poten-tial for steam generator tube degradation are required by the staff.
The following design and operational factors are considered by the staff in the conduct of its reviews:
Designs to provide improved circulation to eliminate low flow areos, and to facilitate sludge removal.
Designs to minimize flow induced vibration and cavitation.
Designs to provide increased flow around the tubes at the support plate.
I Selection of material for tube support plates with improved corrosion resistance.
Material se'..ction (chemistry), processing and heat treatment of minimize the susceptibility of tubes to stress corrosion cracking.
Secondary system water chemistry control.
Secondary side material selection (condensers, feedwater, heaters turbine discs and blades, elbows, etc.), and water cleanup system to minimize erosion and the resulting sludge and corrosion product buildup in the steam generators.
Designs to allow for installation of an ion exchanger (conden-sate demineralizer) in the secondary system to minimize feed-water contamination.
Condenser leakage detection systems.
In view of the above, we conclude that issuance of Construction Permits and Operating Licenses, pending completion of this TAP, can continue with reasonable assurance that operation will not present an undue risk to the health and safety of the public.
A-4/5
Task A-4 Rev. No. 1 May 1978 4.
NRR TECHNICAL 0 IZATIONS INVOLVED
.,,W A.
Engineeri anch, Division of Operating Reactors, has the N
primary I responsibility for the overall review and evalu-ation of steam generator tube integrity in operating plants.
This includes operational experiences, tube failure mechanisms and potential repairs, plugging criteria, ISI requirements, tube failure probability studies, leakage rate limits, and secondary coolant system control.
This also includes the lead responsibility for determining the probability of LOCA and MSLB initiating events and the probability of tube failures during these events and responsibility for determining the number of tubes assumed to fail in LOCA and MSLB analyses.
Manpower Estimates:
0.1 man year FY 1977; 0.5 man year FY 1978; 0.5 man year FY 1979.
Environmental Evaluation Branch, Division of Operating Reactors, B.
has the lead responsibility for the review and evaluation of the offsite dosage related to the consequence cr probability of a Main Steam Line Break (MSLB) accident or a LOCA should EEB will also consult with EB such evaluation become necessary.
and provide support for the probabilistic evaluation of MSLB and LOCA initiating events and the probability of tube failures during these postulated events.
Manpower Estimate 0.1 man year FY 1977; 0.2 man year FY 1978; 0.2 man year FY 1979.
Reactor Safety Branch, Division of Opera;ing Reactors, has the C.
lead responsibility for the review and evaluation of:
(1) the ECCS performance re M ed to secondary tc primary leakage as a consequence of a LOCA, and (2) the effect of primary to second-ary leakage during a MSLB accident on ft.el failures should such evaluation prove necessary.
Manpower Estimates:
0.1 man year FY 1577; 0.13 man year FY 1978; 0.13 man year FY 1979.
Mechanical Engineering Cranch/Materiali Engineering Branch, D.
Division of Systems Safety, has respon.ibility in fsctoring all steam generator operating experierce into the review of new design / material concepts and new ystem component require-This will apply to PWR f acilit ies not yet licensed for ments.
operation.
A-4/6
Task A-4 Rev. No. 1 May 1978 1:
~ ctivities involved will include the review and evaluation UM'
'the applicant's and the NSSS's proposed improvements on the 2E' sign and/or operation of the steam generators; for items such
. as secondary coolant chemistry, design modifications to avoid denting, ISI requirements, recommendations for revision of Regulatory Guides, condenser design to avoid in-leakage and provisions for access opening and space in the containment to facilitate steam generator inspections.
i Manpower Estimates:
0.1 man year FY 1977; 0.5 man year FY 1978; 0.5 man year FY 1979.
E.
Analysis Branch, Division of Systems Safety, has the lead responsibility in developing analytical capabilities (computer codes, etc.) to evaluate the effects of steam generator tube rupture (s) concurrent with various reactor transients that include MSLB and LOCA accidents.
The purpose is to determine the equivalent number of tube failures that can be tolerated during transient events.
This information will then be factored into the overall program of determining an adequate sample plan for tube inspections.
Manpower Estimates:
0.2 man year FY 1978; 0.2 man year FY 1979.
F.
Reactor Systems Branch, Division of Systems Safety, has the responsibility of evaluating the design and performance of new associated auxiliary systems for CP/0L plants yet to be licensed, should any be required as the result of this tech-nical activity; e.g., full flow condensate demineralization, etc., for PWR secondary coolant.
Manpower Estimate:
0.15 man year FY 1979.
5.
TECHNICAL ASSI5TANCE A.
Contractor:
Brookhaven National Laboratory (BNL) - DOR, DSS Funds Required:
$98K FY 1977; $125K FY 1978; $225K FY 1979, This effort is funded as part of an overall program at DNL l
applicable to the three Category A Technical Activities (A-3, A-4, and A-5) related to PWR steam generators.
Funding values I
under DORSAT are not included.
This program is needed to obtain technical consultation and assistance to review information in areas of water chemistry and corros:on analysis, monitored jointly by EB/ DOR and HTEB/ DSS.
A-4/7
- s Task A-4 Rev. No. 1 May 1978 Stress and/or burst strength calculations are funded in part under D T contract on an as-needed basis.
This program will provide tance in accomplishing Tasks 2C, 2E, and 2G.
,5p$
- ?
B.
Contra Idaho National Engineering Laboratory (INEL) - DSS Funds Required:
$75K FY 1977; $100K FY 1978.
This effort is generic in nature and will be applicable to the three Category A Technical Activities (A-3, A-4, and A-5) related to PWR steam generators.
The purpose of this program is to determine the effect of steam generator tube plugging on the predicted peak clad temperatures following a postulated LOCA.
The primary activity is to produce a reliable computer code to aid the evaluation of the effects of tube plugging on the ECCS performance.
An addition to the pro-gram will be needed to consider steam generator tube failures concurrent with MSLB or a LOCA.
This program will provide assistance in accomplishing Tasks 28 and 20.
C.
Contractor:
Sandia Laboratories - 00R Funds Required:
550K FY 1977; $100K FY 1978; $150K FY 1979.
This work is of generic natura, and will be applicable to all PWR steam generators.
The purpose of this program is to perform a statistical analysis of steam generator tube failures in operating reactors in order to establish the bases for the sampling plan for inservice inspection.
This is a new program to augment staff effort in steam generator safety reviews and will assist in addressino Tasks 2A, 2F, and 2G.
6.
ASSISTANCE REQUIREMENTS FROM OTHER NRC OFFICES A.
Office of Nuclear Regulatory Research, Division of Reactor Safety Research, Metallurgy and Materials Branch and Proba-bilistic Analysis Branch.
RES has funded, at the request of NRR, a major confirmatory The experimental program at Pacific Northwest Laboratory.
activity of this program consists of a series of tests to verify the burst and cyclic strengths of degraded steam generator tut es and the leakage rate data.
This program is managed by Metallurgy and Materials Branch, (Task 2C).
A-4/8
Task A-4 Rev. No. 1 May 1978 RES has funded, at the request of NRR, a program to address the
)
. tors which determine Inconel 600 susceptibility to stress 9
rosion cracking in primary water.
Metallurgical condition, hemistry, temperature, stress and environment will be considered, (Task 2E).
The Probabilistic Analysis Branch funded the program to assist EEB in probaoilistic analyses, (Task 28).
i B.
Office of Standards Development, Division of Engineering Stand-ards, Structures and Components Standards Branch.
OSD has funded a confirmatory research program at Battelle Columbus Laboratory to evaluate eddy current methods for inspecting steam generator tubes as a subcontract to Brookhaven National Laboratory, (Part of Task 2F).
C.
Office of the Executive Director for Operations, Applied
{
Statistics Branch.
I Provide assistance to EB/00R for statistical assessment of steam generator tube integrity, (Part of Tasks 2A, 2F, and 2G).
D.
ACRS This task is closely related to one of the generic items iden-tified by the ACRS and, accordingly, will be coordinated with the committee as the task progresses.
7.
INTERACTIONS WITH OUTSIDE ORGANIZATIONS A.
Licensee (s) of Combustinn Engineering Nuclear Facilities At present all CE plants experiencing tube denting will be monitored to evaluate the progress of denting.
Each licensee will submit an analysis of the consequences of tube denting on tube integrity and demonstrate that adequate safety margins exist for continued safe operation.
B.
Combustion Engineering The primary interactions with CE has been and continues to be related to their investigation program for the resolution of the tube denting problem at CE designed plants, and its generic implications, such as the licensing bases or justifications for continued operation of CE plants with known tube degrada-tions.
For interim periods of operation until the cause of tube A-4/9
Task A-4 Rev. No. 1 May 1978 c'enti '
tidentified and corrective measures (s) implemented, h
this raction will be needed to ensure that CE develops F
capab.
ties for the evaluation of ECCS performance for postu-lated accidents concurrent with tube failures, should such a licensing basis become necessary.
In conjunction with licensees, CE will be requested to submit a test program and corrective action plan for Maine Yankee and Millstone Unit 2 and an analysis of the structural integrity of degraded tubes under normal oper-ating and accident conditions (LOCA, SSE and MSLB).
In addition, CE will be requested to keep NRC informed of steam generator design changes and modifications in secondary water treatment systems to alleviate tube degradation in future CE plants.
This information will be incorporated into all Tasks of the program.
C.
Interactions with other organizations such as the Electric Power Research Institute (EPRI) and the "ad hoc" organization of PWR owners may also be required because of the mutual interests in the safe operation of steam generators in general and, in par-ticular, the various problems associated with the operation of steam generators.
Current programs sponsored by EPRI include the CE model boiler studies and the round robin program for ISI techniques.
8.
POTENTIAL PROBLEMS It should be anticipated that required feedback from related programs funded by outside organizations may delay the timely completion of certain subtasks.
However, it is hoped that effective participation of NRC representatives at "ad hoc" organizational meetings will improve mutual interests in NRC goals.
Any delays in submittals required by licensees and NSSS vendors would certainly delay the review and evaluation of tasks defined in the program.
Timely input is req' sired from all technical organizations involved.
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