ML19351A189
| ML19351A189 | |
| Person / Time | |
|---|---|
| Issue date: | 02/27/1981 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Ahearne J, Gilinsky V, Hendrie J NRC COMMISSION (OCM) |
| Shared Package | |
| ML19351A177 | List: |
| References | |
| REF-10CFR9.7, RULE-PR-50 NUDOCS 8106260217 | |
| Download: ML19351A189 (11) | |
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h NUCLEAR REGULATORY COMMISSION g
.y W ASHINGTON. D. C. 20555
.,3 D.%..... +#g E5 Ei issi MEMORANDUM FOR:
Chairman Ahearne Commissioner Gilinsky Commissioner Hendrie Commissioner Bradford FROM:
William J. Dircks Executive Director for Oprrations SUBJECY:-
PROPOSED RULEMAKING TO AMEND 10 CFR PART 50 CONCERNING ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) EVENTS (SECY-80-409)
As requested in _the February 18, 1981 memorandum from S. J. Chilk (M810212A),
the following revisions have been made to enclosures to SECY-80-409).
- 1. ' Enclosure "A" has been revised to add a paragraph to the proposed rule itself that excludes ATWS from consideration under Part 100.
A conforming change to Content of the Proposed Rule section of the FRN has also been made.
A discussion of the relationship between ATWS and Part 100 1
i is also provided (Enclosure "0").
2.
Enclosure "L" has been revised to include the pressures corresponding to the ASME Service Limits C and D in the limiting com; nents of the reactor coolant system.
The staff will be prepared to discuss these subjects at the next meeting with the Commi ssion.
751gned)T. A.P.chra l
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[ William J. Dircks l
Executive Director for Operations
Enclosures:
A - Revised pages to the Notice of J
Proposed Rulemaking L - Proposed ATWS, Acceptance Crite ion, Primary Pressure System (Revised) l 0 - Relationship between ATWS to Part 100 8106260 SIT cc:
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t A-14 toward assuring the~ integrity of the reactor coolant system and the reactor core following ATWS events..
e staff recognizes that failure to satisfy, these acceptance criteria dMs not necessary result in severe radiological consequences and has considered the additional safety margin in developing the proposed rule.
In formulating the proposed rule, the Commission consid-ered the need to compare for each plant the offsite doses that might result from ATWS events to the Part 100 guidelines.
Based on the conservative generic calculations, there is reasonable assurance that calculated offsite doses from ATWS will be within the Part 100 dose guidelines if the acceptance criteria of the proposed rule are met. Accordingly, the Comission has l
decided that applicants and licensees will not be required to calculate the potential offsite radiological doses resulting from an ATWS event, notwith-standing the requirements of 10 CFR 100.11. If only the guidelines of Part 100 for calculated offsite doses were specified, the flexibility of the designer i
would be increased, but the attainment of the safety objective would be more difficult to demonstrate.
If systems designs were specified, the flexibility of the designer would be reduced, and the demonstration that the safety objective were attained would be generic rather than for specified plants.
Prior attempts at such a generic demonstration.have been unsuccessful, as discussed above.
The level of safety, whether the mitigation of most or virtually all ATWS events, is specified through the criteria for acceptabel evaluation models. Since the i
parameters in the evaluation nodel are all uncertain to some dagree and some vary over the lifetime of the plant, the level of safety is determined to a large extent by the degree of conservatism in the parameters used in the eval-uation models,' which affect the conservatism of the calculated consequences of m -, ww -o
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. (3)
A description of the measures together with such proposed changes
~ in technical specifications or license amendments as may be neces-sary to ensure compliance with the criteria set forth in paragraphs
- *c)(1), and (c)(2) shall be ' submitted to.the Nuclear Regulatory Comission no later than July 1,1981 or prior to issuance of 1
an operating license, whichever is later.
(4)
Those measures required under paragraphs _ (c)(1) and (c)(2) of this. section shall be completed by July 1,1982 or prior to issuance of an operating license, whichever is later.
(d) Applicant or licensee is not required to calculate the potential offsite radiological' doses resulting from an anticipated transient without scram event notwithstanding the requirements of section'100.11 of Part 100.
(Sec. 161b and i, Pub. Law 83-703, 68 Stat. 948, Sec. 201, Pub. Law 93-438, 88 3
Stat.1242 (42 U.S.C. 2201(b), 5841).)
Dated at this day of, 1981_.
For the Nuclear Regulatory Commission.
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Samuel C. Chilk Secretary of'the Comission i
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PROPOSED ATWS RULE ACCEPTANCE CRITERION PRIMARY SYSTEM PRESSURE The specification of an allowable pressure following an ATWS event in the acceptance criteria of the proposed rule.should be based primarily on essurkg the integrity of the rentor coolant system.
However, the degree of assurance is' difficult to quantify since the probability of failure of the reactor coolant system as function of stress level is poorly defined.
--Therefore, theipecification of a pressure limit must be based almost entirely on-judgment. The ASME Boiler and Pressure Yessel Code Service Limits.are defined in terms of consequences and not in terms of probability of failure.
The level C Service Limits " permit large deformations in areas of structural-discontinuity," whereas the Level D Service Limits " permit gross general deformations with some consequent loss of dimensional sta-b il i ty... " The Level C criterion limits the primary membrane stress due to pressure and other mechanical loads to the yield strength of all mate-rial s..
In addition, for-ferritic materials, the primary membrane stress due to pressure only is limited to 90 percent of the yield strength. The Level D criterion ifmits the primary membrane stress to about 70 percent
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of the ultimate strength of materials. The internal pressures that would result in reaching the Level C limit would be approximately 20 percent to 30 percent greater than the design pressure and the pressure at the Level D limit would be another 20 percent to 30 percent greater than the pres-sure at the Level C limit.
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The ASME Code does specify overpressure limits for normal, upset, and emergency conditions while leaving to the designer the determination of i
the specific events to be included in each category. Overpressure for i
upset e oditions is limited to 10 percent above the Design Pressure.
In
- 1978 che Code was revised to include a specific r2quirement for over-pres
- Jre protection of ASME Class 1 components for emergency conditions.
The Jode now limits stresses in emergency conditions to those corresponding j
to the Service Limit C.
The Code does not provide rules for overpressure for faulted conditions.
None are provided because the Level D Service Limit was never intended to apply to situations where, as in ATWS etents, the major portion of the load results from pressure within the component.
Because of this, the Level D allowable stresses for materials such as those used for bolts were not developed and are not included in the Code.
In current NRC practice, the correlation between the conservatism of accept-ance criteria and the probability of an event is not uniform. For antici-
- pated trasients, the Level B limit is applied and overpressure is limited to approxi nately 10 percent greater than design.
However, the Level B l
Service Limit is specified as the acceptance criteria for accidents resulting from steam or feedwater piping failures. Other events of lower probability do not result in exceeding the design pressure of the reactor coolant system.
Although calculated pressures do not exceed the pressure corresponding to the Level B limit in PWR-rod-ejection or BWR-rod-drop i
accidents, the acceptance criterion is the Level C Service Limit.
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generator tubes are subjected to approximately the same differential pres -
sure during a steamline-break accident as during an ATWS event.
Integrity
.of the tubes following a steamline break has-been found acceptable based on extensive testing of tubes with defects.
Since tubes are subjected to stresses in the plastic range, use of the test data to demonstrate integ-rity of the tubes is equivalent to applying the Level D Service Limit to these components. The stresses in structural components are.also required to be calculated assuming the combined loads from a loss-of-coolant acci-dent and an earthquake. These stresses are limited to the Level D limit.
Although a directly applicable precedent is not available, the criteria applied to other components and structures provide some comparison. The acceptance criterion for fuel is that fuel damaga is limited to some clad perforations following anticipated transients.
Greater fuel damage, i.e.,
severe oxidation and ballooning, is permitted following a loss-of-coolant accident.
The acceptance criterion for the containment is more restric-tive.
While there is no containment pressure acceptance criterion for anticipated transients, the containment pressure following a loss-of-coolant accident is limited to the design pressure, which is analogous to the Level A Service Limit for the reactor coolant system.
The conservatism of an acceptance criterion is not only a function of the criterion itself, but also of the conservatism of the evaluation model used to calculate the parameter being considered. Thus, changes in the values of the parameters used in the evaluation model, such as the moderator tem-t perature coefficient, can affect the conservatism of the requirement as much or more than changes in the value of the acceptance criterion.
If adopted, the Level D Service Limit would not be the limiting factor l
for overprersure following ATWS events. The additional requirement to demonstrate the operability of components necessary for shutdown following an ATWS event would limit the pressure to below that which would result in l
Level D stresses:
In some cases, the demonstration of operability may be
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l a limiting factor even at the Level C limit. A specific demonstration of the capability of the safety and relief valves to close after discharging water at pressures corresponding to the Level C limit has not been required.
Some extrapolation of test data at low pressures and engineering judgment have been used to justify that these valves will close following ATWS condi-tions. The industry now has underway a program to test the operability of safety and relief valves discharging water at design pressures. Operability would still require extrapolation from the test conditions. Operability at-pressures corresponding to the Level D limit would require even further extrapolation.
The pressures corresponding to the ASME Service Limit, Level C (or lower pressures if other limits are controlling) are provided ir Table L-1 for reactors designed by Babcock and Wilcox and Combustion Engineertag. The calculated pressures in plants-designed by Westinghouse are well within those corresponding to the Level C limit.
However, the Ifmiting components would be similar since Westinghouse plants have components ede by many of L-2 ob
l the same manufacturers.
Because the staff has not previously requested information.on the capa-111 ties relative to the Level D lief t, little information is available.
The pressures corresponding to Level D have been estimated for some of the components _ that are limiting at Level C.
This would give some indi,
cation of the allowable pressures corresponding to the Level D Ifmit.
There is no as'surance that other components will not reach the Level D i
limits at lower pressures.
The effect of the deformations that would occur at. pressures above those corresponding to the Level C. limit has also not been determined. - Non-j structural components such as bearings, rotors, valve internals and seals Evaluation of pressurizer top and bottom head are of particular concern.
deformation, pressurizer heater weld integrity, and the effects of moments produced by the deformation of safety and relief valve piping would also be required.
Deformation of the reactor vessel heads would result in moments being applied to the head pentratiot welds, a condition-that has not been evaluated and is now specifically prohibited in the ASME Code.
The effect of pressures above the Level C limit on closures, such as the-reactor vessel upper head and the steam generator manways has not been l
determined.
CE has reported that the vessel closure head would lift off the vessel at about 3750 psi. Whether this would result in vibrations and what effect such vibrations could have has not been evaluated. Lifting of the vessel head could also result in hydraulic loads on the core and vessel internals and thermodynamic effects on the core, neither of which have been analfzed.
For B&W plants, the pressures corresponding to the Level D Service Limit However, B&W reported that the are not available for any component.
stresses in the suction and discharge nozzles of the reactor coolant pumps manufactured by Bingham and installed in 177FA plants would exceed The maximum pressure capability of the reactor coolant yield at 4000 psi.
pump seals was reported to be 4700 psi.
CE reported (CENPD-263P) that 5700-6500 psi corresponds to the Level D The corresponding pressures limit for the reactor coolant pump casing.
for the shells of the pressurizer and the reactor pressure vessel were estimated,to be 5100 psi. All of t' a pressures were determined using These estimates are therefore a simplified elastic analysis technique.
misleading because the primary membrane stress (i.e., the average stressThe
.through the wall) would, in reality, greatly exceed tne yield stress.
pump casing stress is more than twice the yield stress at the pressure of The effects of the accompanying large inelastic deformations 5700 psi.
have not been evaluated.
The results of the preliminary fracture toughness analyses indicate that, for PWR vessels fabricated from materials with a relatively low " upper shelf" or high temperature toughness are more susceptible to radiation L
damage, brittle fracture of the vessel could be of concern at pressures L
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below those corresponding to Service Limit D, assuming the presence of an ASME Code reference flaw. At pressures below those corresponding to Service Limit C, brittle fracture would not be a concern.
In addition to technical arguments concerning the adequacy of the ATWS acceptance criteria, the staff has considered the amount of analysis by the industry and review by the staff necessary to demonstrate compli-ance with the criteria.- Beyond calculating the system pressure during postulated ATWS events, little or no analysis has been required to demon-strate that stresses are limited to the Service Limit C.
In most cases, the value of the already calculated normal operating or upset condition stress increased by the ratio of the calculated ATWS pressure to the design or upset condition pressure has been accepted. Adoption of the Service Limit D would require the licensees ar.d applicants to perform elastic-inelastic stress analyses of the reactor pressure boundary which are more complex than the elastic analyses needed to calculate Service Limit C stresses. The extensive deformations that would be calculated to occur.
at the Service Limit D would result in moments not present at the Service Limit C pressure levels, thus increasing the complexity of the calculations further.
The criterion of the proposed rule allows a demonstration of integrity and operability to be based on test data, sc that a few components, such as the i
steam generator tubes would not unnecessarily be the limiting factor, in determining the acceptable pressure. However, this alternative method would be limited to local areas in the reactor system.. Thus, acceptability for GE and Westinghouse designed reactor systems and most components of the systems designed by B&W and CE would be based on the Level C limit.
Exceptions would be permitted, but would require justification by the licensee.
The specification of an appropriate allowable pressure following 'an ATWS event has been an issue during the review of proposed requirements for ATWS events. The arguments discussed here have been addressed earlier and are presented in NUREG-0460 (Vol. I, Section 7.1.2, and Vol. 3, Appendix D, Section G.1).
The staff, the ACRS and the ASME Code committees have been l
involved in evaluating the issue. The consensus of these various groups is that the Service Limit C is the appropriate limit for ATWS events.
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PRESSURE CORRESPONDING 1
TO ASME CODE SERVICE LIMIT C Component Pressure,psii
.B&W 177FA Plant _
Pressurizer Heater Outer SheathReacto 32301 Reactor Pressure Yessel Head Seal 3460
.35002 Pressurizer Manway CoyerJ(Bingham)
Reactor Coolant Pump 2
I 3500 Pressurizer Shell & Surge Nozzle 3500 Reactor Pressure Yessel, Shell and Core 3750 3750 Flood Nozzle BaW 20ECA Plant Reactor Coolant Pump (Bingham)
Reactor Pressure Vessel Head Seal 3990 Pressurizer Shell and Spray Nozzle 3*a002
/
Reactor Pressure Yessel Shell and Lower 3750 3750 Head
- CE Plant Pressurizer Heater 3000 Reactor Coolant Pump Casing Pressurizer Shell 2900-3000 Reactor Pressure Yessel Shell I:
3500 3500 Reactor Vessel Head 37505 Notes:
1.
Failure Leakage S ction and discharge nozzlec 2.
t Diffuser vanes and volute wall 3.
4.
5.
Lift-off vessel L-5 I
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RELATIONSHIP BETWEEN ATWS AND PAP.T 100_
The relationship between Part 100 and the proposed addition to Part 50 that would specify requirements for ATWS is similar to the relationship between l
Part 100 and the other design criteria in Part 50.. As stated in Part 100, its purpose is "to describe criteria which guide the evaluation of the suit-i ability of proposed sites for stationary power and testing reactors subject Evaluation of a site is not only based on the characteristics te Part 50."
of the site such as population distribution, geology, seismology, hydrology and meteorology, but also on the characteristics of the facility to be built Part 100 specifies the criteria for the site characteristics while Part 50 and tne General Design Criteria in particular specify the on the site.
einimum requirements for the principal design criteria for the facility.
The principal design criteria establish the necessary design, fabrication.
construction, testing and performance requirements for structures, systems and components important to safety. Thus, the purpose of the cdteria in Part 50 is to provide assurance that the probabilities of accidents with significant releases of radioactivity are small enough so that there is These accidents are reasonable assurance such accidents will not occur.
primarily those that result in severe damage to the core or the contain-ment.
Under Part 100, the qu'antitati've analysis of the offsite consequences of one postulated accident is required as a key test of the suitability of th purposes of site analysis or postulated from consideration of possible acc facility and site.
dental events that would result in potential hazards not exceeded by thos from any accident considered credible."1 j
to determine compliance with Part 100 is termed a loss-of-coolant accident This accident is not bas?d on the same assumptions as the LOCA The acceptance criteria of section (LOCA).
required to be analyzed by Part 50.
50.46 are intended to assure that significant fuel melting will not occur The Part 100 LOCA is based on assumptions that although the containment is pressurized to a value calculated to result from a com-in a LOCA.
plete break in the largest pipe in the reactor coolant system, the fission product release from the core is of a magnitude generally as j
the containment or engineered safety features, including the containment melted fuel.
sprays and filters, to be designed to function under the conditions, other The Part 100 LOCA than the radiation field, that would accompany fuel melting.
f is based on the assumption that the containment and engineered safety features l
Thus, the Part 100 LOCA analysis is not an analysis of acc' dental events considered in the design bases of the plant, but an accident hypothe-function.
As such the Part 100 analysis is used sized for purposes of site analysis.
to assess whether the exclusion area and low population j heat removal system, and the effectiveness of the engineered safety features such as containment spray and filters are deceptable.
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10 CFR Part 100, Section 100.11, Footnote 1.
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Additional tests of the suitability of the faciliV at a site are made,by comparison of the calculated offsite doses resulting from a broad range of These accidents include failure of a main steamline or other accidents.
other small lines outside containment, failure of a steam generator tube, failures in the radioactivity waste handling and storage systems and spent These accidents are assumed to result in the fuel handling accidents.
release of the relatively small amount of radioactivity in the reactor coolant; the pellet-clad gap of a relatively few fuel rods or the radio-However, these accidents can result in the active waste handling systems.
release of the radioactivity by a more direct pathway to the environment Because some of these accidents are considered more pro-than in a LOCA.
bable than a major LOCA, the staff has required that for these accidents the calculated offsite doses be limited to a fraction cf the Part 100 This fraction is usually one-tenth but in some cases is one-The practical result of the aralyses of the pipe failures outside guidelines.
containment is to set the Technical Specification limits on the concentra-quarter.
The anal-tion of radioactivity, primarily iodine in the reactor coolant.
ysis of the fuel handling and radioactive waste system accide.ts can result in the addition of filters or limits on the quantity of activity stored.
The requirements in both Part 50 and 100 have the purpose of limiting the amount of radioactivity that could be released to the environment from possible accidents and one set of requirements supplements the other.
However, the requirements of Part 50 are directed toward assuring that core damage and the amount of radioactivity available for release isAlthough limited and that containment integrity is assured.
requirements could be applied to accomplish these same objectives, the analyses of doses and comparisons to the Part 100 guidelines have been used to specify the performance level of the containment, that is, the Only for non-LOCA leak rate including %ssotion by sprays or filters.
type accidents have the Part 100 criteria been used to limit the quantity of activity that might be available for release from the containment.
Thus, the Part 100 requirements have primarily resulted in controlling the pathways for leakage of radioactive materials.
If the ATWS mitigating systems that would be required under the proposed rule limit system pressure and core power and provide adequate cooling, some fuel damage is still possible but would be limited only to the Because the reactor coolant system pres-rupture of the fuel cladding.sure is very high during an ATWS event, fissi the fuel pellet clad gap could leak through steam generator tube cracks and thence to the environment through the steam generator and the vei. dors, offsite doses following ATMS events safety valves.
The staff has required conservative to exceed the Part 100 guidelines.
calculations for other accident doses but realistic dose calculations would be consistent with the realistic system calculations in the pro-Realistic calculations result in doses that are only a posed rule.
small fraction of the guidelines.
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1 If the calculated offsite doses from ATWS events in each facility were, required to be compared with the Part 100 guidelines, in the worst case this could require additional features to control leakage from the con-tai nment.
Such features would not reduce the probability of severe core or containment damage resulting from an ATWS. However, the risk from ATWS events is negligible if the acceptance criteria of the proposed rule are met and severe damage to the core and containment is prevented because the probability of fuel damage would be low and any consequent dose would be small. Therefore the cost of such additional leakage control features is not justi fied.
Since the leakage rates from the containment, steam generators or residual ;ieat removal system and the effectiveness of sprays and filters are determined by the analyses of. doses from other accidents, plant-specific
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analyses of doses from ATWS events would not be required to determine any system performance requirement.
On the basis of the above, and the conser-vative generic calculations reported in Vol. 2 of NUREG-0460, the staff concludes that no significant safety benefit would be achieved by requiring the submittal and staff review of plant-specific, Part 100-type dose calcula-tions for ATWS events given that the acceptance criteria of the proposed rule are shown to be met.
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' TRANSMITTAL TO:
$ ~ Document Control Desk, 016 Phillips ADVANCED COPY-TO:
O The Public Document Room
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-g DATE:
June 17, 1981 p.
Attached are the PDR copies of'a Commission. meeting Q
transcript /s/ and related meeting document /s/.
Gey 6
are being forwarded for entry on the Daily Accession List and placement in the Public Document Room.
No other: distribution is requested or required.
Existing DCS identification numbers are listed on the individual g
. documents wherever possible.
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Transcript of:
Discussion of Anticipated Transients Without SCRAM, June 16, 1981.
(1 copy) a.
Copy of vugraphs presented at above meeting:
Why_an Alternative ATWS Rule?
(1 copy)
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b.
Memo from J. Hendrie to the Commissioners dated
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June 9,.1981,
Subject:
ATWS.
(1 copy) c.
Memo from W.
Dircks to the Commissioners dated d
Feb. - 27.91, -Subj r - Environmental Impact Assessment 4
For Proposed Rulemaking to Amend 10 CFA Part 50 y
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' ~ - ~ ~(ATifS) ~ Events.
(1 copy)
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Concerning Anticipated Transients Without Scram q
B d.
Memo from W. Dircks to the Commissioners dated Feb. 27. 81, Subj:
Proposed Rulemaking to Amend 10 CFR Part 50 Concerning Anticipated Transients Without Scram (ATWS) Events ( SECY-80-409).
(1 copy) ja brown Office of the Secretary D
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