ML20196F879

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Forwards Arguments Pro & Con for Retention of Specific Topics in SEP for Which Engineering Branch Has Responsibility
ML20196F879
Person / Time
Issue date: 08/03/1977
From: Shao L
Office of Nuclear Reactor Regulation
To: Ziemann D
Office of Nuclear Reactor Regulation
Shared Package
ML20196F442 List:
References
FOIA-87-854, TASK-03-02, TASK-03-03.A, TASK-03-03.B, TASK-03-03.C, TASK-03-06, TASK-03-07.A, TASK-03-07.B, TASK-03-07.C, TASK-03-07.D, TASK-03-08.D, TASK-03-10.C, TASK-03-11, TASK-05-12.A, TASK-3-10.C, TASK-3-11, TASK-3-2, TASK-3-3.A, TASK-3-3.B, TASK-3-3.C, TASK-3-6, TASK-3-7.A, TASK-3-7.B, TASK-3-7.C, TASK-3-7.D, TASK-3-8.D, TASK-5-12.A, TASK-RR NUDOCS 8803040235
Download: ML20196F879 (12)
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5G 3 1977 MEMORANDUM FOR: D. L. Ziemann, Group Leader, Systanatic Evaluation Program Review Group. 00R ,

i FRON: L. C. Shao, Chief. Engineering Branch, 00R

SUBJECT:

SYSTD4ATIC EVALUATION PROGRAM (SEP)

.l Enclosed are arguments pro and con for retention of specific

topics in the SEP for which the Engineering Branch has responsibility.

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L. C. Shao, Chief Engineering Branch Division of Operating Reactors

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III-2 WIND AND TORNADO LOADING l l

Pro l

1. Tornado missile protection is a governing factor in the design of l nuclear power plant structures. Many older steel containments and '

other safety related structures have not been designed for these missile loadings. It is important to maintain containment, core cooling and heat removal functions following a tornado to assure public safety.

2. The differential pressure loading associated with a tornado is not '

detrimental to containment structures which have already been designed for internal pressures, however, othey safety related structures important to maintain ECC and RHR capabilities have not been designed considering overpressure effects, especially the diesel generator building since the probability of loss of off-site power due to a tornado is quite high.

Con

1. Most structures are designed for 100 yr. recurrence interval wind velocities (80-100 mph winds), for which a 1/3 allowable increase is pennitted in building codes, but has not been pemitted in power plant design. Also, almost all power plants have been designed to resist seismic loadings which are more significant than wind loadings for the generally squat, massive power plant structures.
2. In addition, as stated in WASH 1400 the probability of a DBT at a plant in a given year is s5 X 10-6 while the probability of any tornado occurring at that plant in a given year if 5 X 10-4 For containments that are not missile protected, this probability is l combined with a low probability that the missile will penetrate the .

containment and damage a safety related structure.

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O O I!1<3.A EFFECTS OF HIGH WATER LEVEL ON STRUCT1JRES Pro Criteria governing the determination of maximum flood levels have become much more stringent over the past years. The high water level of operating plants should be reassessed to assure that all factors which .

l influence this level (waves, coastal flooding, dam failure, etc.) were assessed in the original design. If this high water level has increased, ,

it must be determined whether or not this increase will impair the con- l tainment ECC or RHR functions of the facility by affecting safety related l equipment functions and structural integrity.

i Con i

In general, plants are located well above the maximum flood level. In some instances flood protection has been required. A significant increase in the design water level would be required before the integrity of most structures would be in jeopardy.

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1 o o 1 III-3.B STRUCTURAL CONSEQUENCES W FAILURE OF UNDERDRAIN SYSTEMS Pro  :

1. Failure of underdrain system can affect a significant part of the plant, t
2. Even if structures do not fail, parts of structures may be subjected to flooding causing problems of ground water contamination, which is a public health hazard.
3. Rise in water table for significant length of time after failure of an underdrain system may lead to loss of structural integrity.
4. The extent to which underdrain systems are relied upon in operating plants is not known presently.
5. General Design Criterion #4 requires equipment compatibility to environmental conditions. Equipment not properly qualified to function under flooded condition would not meet the intent of GDC #4.

Con

1. Relatively few operating plants are expected to rely on underdrain systems.
2. High water level due to failure of underdrain system for relatively short duration may not pose a threat to the structural integrity.

I 3. Events requiring the services of the underdrain system are relatively infrequent. Therefore, the probability of the failure of the under-drain system, given the event, is low.

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III-3.C INSERVICE INSPECTION OF WATER CONTROL STRUCTURES Pro i

There is a high probability of dam failure (si every 1500 to 1800 dam I years) arising from lack of judgment, poor design, construction errors,

. and poor maintainence or inspection. There could be severe consequences for nuclear structures if there was a loss of water control structures, either from flooding or possible loss of the ultimate heat sink. There-fore, it is important that any flaws or iminent problems be identified and remedied at an early stage by stringent surveillance of water control structures.

Con Inservice surveillance of water control structures does not preclude failures under severe weather conditions.

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TOPIC: III-6 SEISMIC DESIGN CONSIDERATIONS Pro

1. Seismic vibratory ground motion can influence structures, systems, and equipment within a structure simultaneously.
2. Seismic input in terms of recurrence interval and SSE leval is unconservatively estimted for most operating plants. fx etor sitet with multiple units indicate that the seismic input for newer plants at the same site was increased significantly beyond that of the earlier units. For example, Davis Besse 1.15g. Davis Besse 2 & 3 .209 Ft. Calhoun 1.179 Ft. Calhoun 2 .25g, Indian Pt.1.19. Indian Point 2 & 3 .15g, Perry 1.15g, Perry 2 .20g, San Onofre 1.50g, San Onofre 2 & 3 .679 Reevaluation of seismic input for operating plants must J be undertaken before any assessment can be made of the seismic design adequacy of operating plants.
3. Many operating plants are without dynamic seismic design, and essential l structures and equipment in many instances are not classified as seismic Category I.

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4. Improper seismic design of structures, systems, and equipment may '

lead to potentially severe consequences.

5. In order to ensure the public safety, the integrity of the containment structure, the reactor vessel including the internals and the control rod drive system, operability of the ECCS and RHRS during a se'smic event must be demonstrated.  ;
6. It is difficult to quantify existing or inherent seismic conservatisms without detailed plant analyses, i
7. Vibratory testing of electrical and mechanical equipment was non-existent on some older plants.

Con

1. A significant seismic event at many sites is remote. '

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2. Massive nuclear power plant structures have inherent strength.
3. Current seismic design criteria are very conservative and are in excess of that which is necessary to the public safety, l

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III-7.A INSERVICE INSPECTION INCLUDING PRESTRESSED CONCRETE CONTAINMENTS WITH EITHER GROUTED OR UNGROUTED TENDONS III-7.C DELAMI)1ATION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES III-7.0 CONTAINMENT STRUCTURAL INTEGRITY TESTS These three topics can be covered under the ongoing generic program B-49. They should be retained in the topic list with an asterisk for the purpose of followup by SEP for specific plants.

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!!!-7.B DESIGN CODES SRP 3.8 ACCEPTANCE CRITERIA, LOAD COMBINATIONS, AND REACTOR CAVITY DESIGN Pro

1. Older plants have not been designed to current codes with existing design techniques and may not incorporate all the current load combinations. .
2. Realistic safety margins and acceptance criteria need to be developed so that the adequacy of the older operating plants to withstand the i new load combinations can be demonstrated. This would eliminate the -

need for modifications that would otherwise be required.

Con

1. There may be adequate margin in the old criteria to offset current refined analysis techniques and load combinations.
2. Some load combinations would be reviewed under seismic and tornado loading effects.

. . O O TOPIC:  !!!-8.0 IRRADIATION DAMAGE, USE OF SENSITIZED STAINLESS STEEL AND FATIGUE RESISTANCE Pro

1. High neutron irradiation levels that will be experienced by stainless steel internals may result in significantly degraded material strength and toughness properties and may produce dimension change.

Because of this degradation and because of the now known unconservatisms in the design fatigue curves for stainless steel contained in the ASME Code, the safety margins for internals are uncertain.

Con

1. While the safety margins for reactor vessel internals are uncertain and may decrease during service, the consequence of internals cracking is minimized since devices are employed in many plants to limit core drop to a few inches, should failure of the support plate occur. Some older operating plants, however, may not lave the limiting devices and core drop, should the internals fail, may be one or two feet. ~
2. Experience has indicated that sensitized stainless steel can under certain stress, environment and flow conditions produce significant cracking in stainless steel. While cracking of internals has been observed on some BWR's and it is obviously important to determine if significant cracking exists in reactor vessel internals, the chemistry control and flow conditions generally present in the reactor vessel would tend to minimize the potential for cracking in stainless steel. j l

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O O TOPIC: III-10.C PUMP FLYWHEEL INTEGRITY pro

1. Because the ASME Code proceedures do not apply specifically to reactor coolant pump flywheels and because flaw induced fracture was possibly not included as a design criterion for flywheels in older operating plants, the fabrication methods, preservice inspection techniques and consequently our knowledge of the ability of these flywheels to resist fracture is uncertain.
2. Because the reactor coolant pump flywheel is not considered part of the reactor coolant pressure boundary it is not included under the inservice inspection rules of Section XI of the ASME Code.

Additionally older operating plants built prior to the issue of R.G.1.14 had no requirement for flywheel ISI and until a review is conducted the status of flywheel ISI at these plants is uncertain.

3. Because of the uncertainty of the flywheel initial design criteria, fabrication methods and preservice intpection techniques and because no means are provided to mitigate the consequence of flywheel failure should it occur, ISI is necessary and the only available means of protection for older operating plants.

Con

1. As an alternative to dncluding pump flywheel ISI in the Systematic Review Program, an attempt can be made by NRC to have flywheel ISI included in Section XI of the ASME Code, even though it is not now considered part of the RCPB, This action would ensure periodic ISI and review of pump flywheel integrity under 10 CFR 50.55a(g).

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l TOPIC:  !!!-11 COMPONENT iTi SRITY Comment: Present on-going , grams such as A-2; Asynmetric LOCA Loads, .

and B-24; Seismic I, alification of Electrical and Mechanical l Components do not include within their scope such concerns es design codes, SRP 3.9 and 3.10 acceptance criteria, load combinations and the structural 1.itegrity and functional oper-ability of all Class 1, 2 and 3 components *ad the structural integrity of their supports.

Pro

1. Mechanical and electrical equipcent and their supports of older plants have not been designed to current codes with existing design  ;

1 or testing techniques and may not incorporate all o'f the current load combinations.

2. Safety margins must be established and acceptance criteria needs to be developed so that the adequacy of older plant components and their supports, to withstand the latest load combinations, can be determined. Testing criteria need to be re-evaluated.  ;

Con

l. There ray be adequate margin in the old criteria to offset current  !

refined analysis and testing techniques and load combinations. l l

2. Testing techniques could be reviewed under B-24. 1 i

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i TOPIC: V-12.A WATER PURITY OF BOILING WATER REACTOR PRit%RY COOLANT I

Need for Review:

l. Maintenance of high water quality is important during operations, for water is of ten transferred from one part of the plant to another during outages, and this water can easily find its way to the hotwell l and from there to the reactors.
2. Possible leaks in the condense tubes can result in circulating water entering the hotwell and introducing impurities into the primary coolant. When the circulating water in the condensers is sea water, the consequences of very small tube leaks can be quite significant i.e., high chloride levels in the primary coolant can lead to stress corrosion cracking of BWR piping systems resulting radioactive release, and therefore special precautions are required in sea water cooled plants.
3. The hotwell water is deaerated to reduce the oxygen content of the water, and then passes through full flow condensate demineralizers in order to obtain the purest possible feedwater. The condensate demineralizers remove insoluble corrosion products by filtrations and disolved ionic materials, i.e. , metals and salts, by ion exchange. They afford limited protection from condenser tube leaks until corrective action is initiated. The feedwater passes through the feedwater heaters and enters the reactor at about 3850F. Since there is no "blowdown" (as there is in a fossil reactor or a PWR),

the reactor water concentrates all of the soluble and insoluble materials that enter via the feedwater. The reactor water cleanup system removes soluble corrosion and fission products from the reactor water, but it has only limited ability to remove insoluble corrosion products. Thus extremely pure feedwater is a necessity. ,,

4. Tighter limits on quality control and better defined provisions in l the event of eMineralizer breakthrough will contribute to safer 1 operations q 2he resctor and should provide early warning of the

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potential of a chloride intrusion event, such as ocurred at Millstone 1. 1 l

Potential Disadvantages of Review:

1. With regard to the impact on operating plants during the review stage, the principal impact will be the requirement for tighter limits on conductivity and chloride content, and new requirements for pH control.
2. For operating plants, any tightened requirements for conductivity and chlorides and new requirements for pH measurement may dictate the installation of additional conducitivity meters.

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AUG 0 71977 [

i:D;0PMDUM FOR: b'illian P. Gam;11. Assistant Director for Site Technology, OSE FROM: Darrell G. Eisenhut. Assistant Director for Operational Technology, D0R SUNECT: SEISHIC REVIEW - OPERAT!!!G PL?.NTS As you are aware the Systentic Evaluation Program (SEP) Group has developed a list of topics for consideration in the review cf 9perating reactors. Seimic considerations is one of the inst significant topics on the list and will probably be one of the nost difficult is reconcile. ,

The support of Site Technology will certainly be needed in coming to '

grips with this issue. The purpose of this memorandum is to request that you further consider criteria for acceptance of the older operating plants with regard to seismic events.

The broad scope of the sr.isnic review should inclucie an evaluation of the edequacy of design and operation and docunent how the plants compare l

with present ifcensing requirerents. It also should include a basis for acceptable departure from these requirements and for any backfitting required.

If the Cornission approves the progran as expe:ted, the eight oldest plants will be reviewed first and these reviews could begin as early as, Never.ter 1977. The SEP Group will keep your apprised of its progress I and schedule and will work closely with your staff. As indicatd in an earlier discussion wi'h Carl Stepp, the SEP Group would like to n.<: w th nenbers of your staff to discuss the review progran and to get a bette}r l

understanding of the seismic review that will be required for old operating

!. plants, if pu have any questions, please c.all Dennis Zienann, on j extension 27330. He will be contactino y:,u in the near future in hopes of establishing a meeting date during tne week of August 8,1977. l

,0riginal signed by

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!' D. G. Eisenhut, Assistant Director for Operational Technology 1 Division of Operating Reactors 9

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