ML20151S856
| ML20151S856 | |
| Person / Time | |
|---|---|
| Issue date: | 01/27/1986 |
| From: | Ross D NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Schwink W Committee To Review Generic Requirements |
| Shared Package | |
| ML20151C834 | List: |
| References | |
| FOIA-87-714, REF-GTECI-A-43, REF-GTECI-ES, TASK-A-43, TASK-OR NUDOCS 8602100002 | |
| Download: ML20151S856 (54) | |
Text
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UtilTE D 87Af tl NUCLE AR REQULATORY COMMIS$10N
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January 27,1966 NOTE TD:
W. Schwink, CRGR FRON:
D. F. Ross Deputy Director, RES
SUBJECT:
GDC 4. CRGR MEETING #84 The following is sqy list of concerns on the GCD f4 package. Please distribute to CRGR menbers.
P. 2 of cover letter:
in consideration of Mojave experience, is the 1.failure of secondary piping "extremely likely" (however defined)?
2.
P. 3 of cover letter (and other places)...We should review the DBA containment tenperature and tressure as a function of break size, and the App. K PCT as a function of treak size (for various plants) to see in context what it means to be unconsistent.
3.
P. 5 of proposed rule: LOCA from sources other than pipe rupture do not place many requirements on safety system, I believe; we should get a reading from reactor systems staff.
4.
P. 5 of PR: The trgument that pipe whip restraints, if improperly installed, degrades safely is, in ey opinion, a weak argument.
If they are needed, then the argument should be that additional QA is merited. Many safety systems, if irproperly installed, could degrade safety. Should they also be removed?
j 5.
P. 6 of PR: Shouldn't there tie an explicit requirement to include aging effects?
6.
P. 7 of PR: 1 assume that a d.e. break, w/o restraints and jet deflectors, can lead to unmitigated core melt; (should verify or refute). This leads to the questions (also on p.7):
a.
10 per reactor year for what? core melt? Core melt w/early containment failure?
b.
If it is core melt, how does ?,h1s egmpare with latest thinking on dominant com. sequences?
Is 10" significant?
7.
P. 8 of PR: What does ELD think tbout the clause ' amendment of CP"? Seems like we have approved more significant changes than this (e.g. Vogtle containment) w/o amendment of CP.
(Wouldopportunityforhearingarise?)
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'8.
P.8, (Last line), safety argument: Senner or later the question (bottom line) will come up as to quantification (dCM) of effect of esbracing this rule. Let's so it now.
9.
P.9ofPR(andotherplaces): Clarify whether high eurgy piping in Category 11 structure may (following competent analysis) fall within the envelope of the P.R. (I believe such a demonstration to be entractable.)
10.
P. 11 of PR: Define; normal plus SSE loads. Some SARs have read *f" to mean RM5; others mean (absolute values.
11.
P. 13 of PR:
1s a revision in allowable outape times for leak detection equipment envisaged. Assuming new leak detection equipment is needed in places not heretofore regulated, what do e say about EQ, App. B Power Requirements ~,
Tech Specs, ete? Isn't more guidanse needed?
A i
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d D. F. Ross, Director Director Office of Nuclear Regulatory Research cc: GArlotto
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CSerpan JRichardson RBMinogue i
' RESPONSE TO R055 l
The' failed Mohave Generating Station Steam Line and its operating 1.
conditions are not typical at nuclear power plants.
In particular, the 1000'F temkrature is sirpificantly higher and makes the alloy steel material susceptible to creep rupture and creep fatigue; these conditions are not expected at nuclear power plants.
The material I
(1.25% Chromium, 0.5% molybdenum) is not used at nuclear facilities and is no longer used in new fo'ssil plants eitha.
Moreover, in approximate 1y' fifteen years of service, the Mohava Steam Line received no inservice inspection. Other conditions which distinguish the Mohave situation from conditions at nuclear power phnts are:
better welder qualification, leak detection requirements, higher quality NDE ouring construction, better design and analysis, better quality assurance and bettr2r control of pre and post weld heat I
treatments at nuclear power plants.
1 Extremely unlikely means 10-6 per reactor yen.
Page 2 of the Com-l sission paper states "extremely unlikely in certain piping".
"Cer-tain piping Ostems" means those that satisfy rigorou-acceptance criter?a.
2.
The inconsistency results because the postulated pipe rupture is eliminated for dynamic effects, but kept for other effects such as i
The review suggested in this list of 1
concerr.s is part of the staff's recomended 'long tenn evaluation" (mentioned in the Comission paper and FRN) which will lead to a f
replacement criteria eventually.
To undertake this review at this
f ',
s time in sufficient depth would delay reaping benefits for many NTOLS.
The suff has elected a step wise approach in which partial benefits are accrued as justification for individual actions are developed.
3.
Failures in bonnets of valves, blow out disks on tanks and heat
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exchangers and nanways on steam generators are examples where large LOCAs can result while piping integrity is maintained.
l 4.
The point is that pipe whip restrains are difficult to install prop-erly. At issue is proneness to malfunction.
Other safety systems can also be installed improperly, but achieving good installation is an easier task usually.
Pipe whip restraints, used only at nuclear power plants, require unattainable accuracy in maintaining tolerances j
and alignments, controlling gap sizes and assuring reliable snubber performance.
5.
Aging h expitettly treated in the fracture rechanics and corrosion evaluation. That is, pipe rupture ta modelled as a function of time; degradation with time is reported as crack size and population.
6.
In the vast majority of situations, an isolated double-ended pipe rupture without pipe whip restraints and jet irpingement barriers is not a serious threat to public health and safety.
Ever. in the few situations when special targets are involved (such as switchgear and instrumentation panels, for example),
redundant safety systems usually provide edequate assurances that fuel damage will not occur.
The figure of 10-6 per reactor year is for pipe rupture; the figure l
i 2
for core melt associsted with pipe rupture is less by an unquantified
~
- factor, I
l 1
7.
ELD (Joe Scinto) wrote the paragraph dealing with the amendment of the ConstructioE Pemit.
8.
For primary loop piping of PWRs, removing pipe whip restraints and jet impingement barriers as well as redesigning heavy component sup-ports has a negligible effect on 'c' ore melt frequency.
Pessimistic estimates indicate public risk may increase 0.003 man-ree.
Realistic estimates which take credit for improved inservice inspection, en-hanced fire protection and increased reliability have not been under-taken. Since the scope of affected piping is not determined, it is difficult to adequately calculate the 8epact of this rule on are melt frequency.
l 9.
The revised proposed rule has a statemer.t that "Piping systems, or portions thereof, located in non-seismic Category I structures must be shown to satisfy the extremely low mpture probability criterion to take advantage of leak-before-break tec:'cology." This will allow structural redesign as a ineans of eliminat'ng pipe whip restraints.
- 10. For piping, two or more dynamic loads are combined using SRSS. Stat-ic loads are combined with each other and with the resultant dynamic loats by using absolute sum.
This position, long enforced, is de-tailed in existing regulatory and industry documents. (See SRP 3.9.3 page 3.9.3-14 and NUREG-0484, "Methodology for Combining Dynamic Loads").
. t. *,
kl. At this time, no change in allowable outage times is envisioned.
When new leak detection is needed outside the containment, it should have equivalent leak detection requirements to those inside the containment, including factors relating to EQ, Appendix B and Power Requirements.
Whether or not Tech Specs are impacted is a policy decision.
Additional guidance will be developed later when policy decisions are made.
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, Note to Jim Snierek These are my comments and questions on the ODC 4 paper, l
Draf t Rule 1-p.4 It is probably okay to indicate that "containments' will still be covered by assumption of DESS Dig Drea6 LOCA even though the top of the page says "subcompartments and c.ompartments " are of f the hook.
I guess we can leave it to case by case rangling about how the compartment can take advantage of LSB but the ' containment" can't.
2-P.5 Middle of Page. What are the Loss of Coolant Accidents other than pipe breaks that place requirements on saf ety systeas??
I hope its not the SG Manhole cover story - bs:cause if the only thing that caused us to impose rapid pressuritation of containawnt and a very big ECCS were the 80 manholes, I'm sure the designs would quic.kly be fixed (We don't assume the PV head comes of f ).
I'm not sure they need this weak explanation now that the Commission paper p.3 (and the Dr f t rule more gently p.4-5) f esses up that this creates en inconsistency that the agency has to come to grips with (someday)
/
3-p.5 Bottom of page top of next page -- what does this statement imply?
It sounds l'ike its says we are going to do something different to give high confidence that the support system will ce reliable.
Do we mean that we e.re going to somthing different 7 If so what 7 If not either take this hyperbole out.
Alternatively, make it clear that we already do this.
4-p.6 3 don't see how under the rule the NRC could require ggjliiggg}, investigations of indirect failure mechanisms.
The rule calls f or a demonstration by analysis that the probability of piping system rupture is extremely low under This of course conditions consistent with tye design basis.
permits (t equir es) a case by case demonstration.
To the extent that indirect f ailure, mechanisms are part of the design basis the licensee would have to cover them under the rule - but to the wxtent they are not the staff wou)d have to challwnge its own rule in order to r e ctui r e addlilgoal 10Y25119511905 The criteria in the paper er staff positions on what the applicant should show to satisf y the rule -- if we mean to require mor e in certain casesi these cases should be reflec.ted in the criteria.
If this is only meant to say we might think of something to put on later I suggest we indicate that if so we will ollow the backfit rules.
/ 5 p.7 first f ull para. I agree with Sniereks comment
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.,1 6-p.7 1s "of the order of $0-6 per ret.ator year when all runture Incations are considered in the fluid mininu py11. ems" a clear statement in terms of "probability of fluid systes piping rupture" 7 7-p.7 Is the Commission stating that the implict t design
- goal of the deterministic review is 104 (somethings) per reactor year 7, 9-pt7 last sentence I don't understand.
It seems to suggest that our determinist c review presumq that things are engineerord to meet appl cable regulations.
If all we do is presume that they are enginesered to meet applicable regulaticns, we seems to be wasting a lots of resources in engineering reviews to see if things are in f act correctly designed and an a great deal of. inspection resources in looking to see thether they have in f act been engineered to seet applicable regulations l
9-p.9 first full paragraph make clear that the criteria that follow are stati gylggggg rather thn Comaiusion mandated standards (they would then be an authoritative Commission interpretation of i ts own rule that would require rulemaking or some other f orm of Commissm authoritsative action to change.
I to-p.9 the ref erence to 79-14 stuf f has to be fixed per discussion at CR9R meest! ng 11-p.11 is normal plus SSE the limiting 1 mad conditione f or all relevant cases 7 (LOCA and water hammer are to be excluded but are there no othur trt.nsient loading conditions that should be addressed ??)
12 p.11 next to last paragraph.
Is it clear what is expected by the statement...' final creck size is limited such that a double-ended pape break will not occur"7 Value f anact Starv j
13-p.v111 and 3 change ref erence to, rom the Exacutive Legal Director to the Of fice of the Executive Legal Director l
14-This was to be Icoked at f or wome better relationship with the rule being proposed 15-1 indicated my skepticism about the money ## but that was adequately discussed at the meeting.
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RESPONSE TO SCINTO 4
l 1.
We accept this coment.
It requires no changes to the rulemaking package.
2.
This is similar to the concern listed by Ross. Besides steam generator aanways, failure of valve bonnets and rupture disks on heat exchangers and tanks can be a source of loss of coolant which could impose design requirements on other aspects of facility design.
3.
Statement is deleted in response to Scinto connent.
4.
The word "additional" is deleted. 'This case-by-case situation results because plant layouts are so varied and the specific indirect mechanisms are unique for different piping systems.
In
- some instances, it e.ay be stated readily that indirect sources of pipe rupture are non-existent.
In other situation, detailed studies may be required to show that potential indirect mechanism of pipe rupture are sufficiently remote.
5.
This paragraph has been revised.
6.
Pipe rupture probabilities are calculated on a per location basis (more typically, per weld basis). The total pipe rupture probability is the sum of the probabilities of all the locations treated. The only meaningful way to discuss "probability of fluid systes piping rupture" is in terms of all rupture locations.
7.
The Connission is stating tgat the deterininistic design basis has an implicit design goal of 10 pipe ruptures per reactor year for fluid system piping.
No judgment is offered as to whether this design goal is achieved.
This then becomes a major issue, because it must in some fashion be demonstrated that the goal is attained.
8.
".... engineered on a deterministic basis" is revised to read,
.... correctly engineered on a deterministic basis (assuming verified fabrication and adequate inservice inspection)"
9.
Revised according to suggestion from Scinto.
- 10. Revised according to discussion of CRGR,
- 11. llorsal plus SSE is a convenient way of establishing margins. All plaht piping experiences earthquake loads when a seismic event occurs. For most piping, the SSE is the controlling design transient and is the only faulted condition dynamic transient.
- 12. The words ".... final crack size is limited such that a double-ended pipe break will not occur...." defines stable crack growth which was just mentioned in the text.
.t 2
' 13. Revised according to suggestion from Scinto.
- 14. Revised according to discussion of CRGR.
- 15. The numbers were based on a wide industry input and reflect the best that can be done at this tit.
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For:
The Comissioners From:
Victor Stello, Jr.
Acting Execut've Director for Operations PROPOSED BROAD SCOPE RULE TO MODIFY GENERAL DESIGN
Subject:
CRITERION 4 0F 10 CFR PART 50, To obtain Comission approval to publish a notice of
Purpose:
proposed rulemaking.
Category:
This paper covers a major policy issue. Resource estimates, Category I, preliminary.
This action expands the scope of affected piping in a Issue:
recent proposed modification to General Design Criterion 4 (GDC 4) which allowed exclusion of dynamic effects associated with postulated pipe ruptures in only primary coolant loops of pressurized water reactors (PWRs).
Consistent with the interim rule, dynamic effects are excluded, while non-mechanistic pipe rupture is still postulated as the design basis for emergency core cooling systems, containments, and environmental qu&lification, i
Operating plants, plants under construction, and future P ant designs are affected by this action.
l Sumary:
A limited scope amendment to GDC 4 was proposed to the Comission in SECY-85-108, dated March 26, 1985.
This amendment allowed use of "leak-before-break" technology only in the primary coolant loop piping of PWRs.
An extension of this limited application to all high energy piping systems in all nuclear power units that meet rigorous acceptance criteria is now being proposed in this
Contact:
J. A. O'Brien; PES 443-7854 K. R. Wichman, NRR 492-4679 W. M. Shields, OELD 492-8693 to
The Consnissioners 2
f paper. The amendment to GDC 4 now proposed would permit a potentially more extensive removal of protective devices i
such 'es pipe whip restraints and jet impingesent shields originally designed to mitigate the dynam' c effects of
~
postulated instantaneous pipe ruptures and other related changes.
Background:
The two-step approach to rulemaking was adopted because safety and economic benefits could irinediately be obtained by an amendment lir,ited to the primary loops of PWRs where sufficient technical evidence had already been developed, reviewed and accepted by the NRC staff, the ACRS and the CRGR.
Additionally, a number of applicants and licensees had requested exemptions within the purview of the limited scope rule.
While the interim rule was based on (but not limited to) the alternative resolution of US! A-2 previ-ously reviewed and endorsed by the ACRS and the CRGR, the broad scope rule required additional scrutiny by these two
- bodies, particularly with respect to the acceptance criteria which would be applied to piping other:than the primary loops of PWRs.
These acceptance criteria for applying leak-before-break analyses are annunciated by the NRC staff in NUREG-1061, Volume 3 November 1984.
The acceptance criteria were approved by the USNRC Piping Review Comittee, but were not formally reviewed at that time by the ACRS and the CRGR.
Discussion:
General Design Criterion 4 as applied in the context of the i
definition of "loss-of-coolant" accident has required the installation of protective devices in nuclear power plants to mitigate events which are now regarded as extremely unlikely in certain piping.
These protective devices impede inservice inspection and maintenance, reduce safety if improperly installed, and increase worker radiation exposures.
To overcome these difficulties, the addition of two sentences to GDC 4 is proposed which allows the use i
of analyses to demonstrate piping integrity in all high energy piping in all nuclear power units. As a minimum, a fracture mechanics evaluation including the effects of fatigue is undertaken.
Additional investigations of j
potential indirect failure mechanisms which could lead to pipe rupture may also be required on a case-by-case basis.
1 I
t The Connissioners 3
Value-impacts resulting from this rule are greatest for future plants, where estimated costs can be reduced approximately $100 million per unit.
Of this sum, about
$30 million are direct costs and the balance stems from reduce'd financing costs and improved scheduling. Reduction in worker radiation exposures vary from plant to plant, but are in the range of 300 to 800 man-rem.
Public risk was l
not quantified, but is believed to decrease due to improved effectiveness of inservice inspection and enhanced safety.
The above quoted figures are based primarily on the elimination of pipe whip restraints and jet impingement barriers and do not treat other facility changes that could result from this rule.
For existing PWRs, considering primary coolant loops only, cost savings of $186 million and reductions of 34,000 man-rem are estimated for a population of 85 PWRs.
These figures did not include savings resulting from redesign of heavy component supports.
One licensee seeking to take advantage of the modification of 60C-4 is estimating a per plant cost savings of $20 million and reduced worker exposures of about 3000 man-rem associated solely with a redesign of reactor coolant pump supports.
For existing BWRs, considering only the recirculation loop piping, cost savings of $30 million and reductions of 8,600 man-rein are estimated for a population of 38 plants.
In existing PWR$ and BWRs, public risk is estimated to be insignificant 1y impacted, or if credit is taken for improved inservice inspection and enhanced safety, to be reduced by an unquantified amount.
The staff has not quantified situations other than those discussed above; however, 'it is believed that other high energy piping will also indicate favorable value-impacts.
This* rulemaking will introduce an inconsistency into the design basis for certain regulations by excluding only the dynamic effects of postulated pipe ruptures while still retaining non-mechanistic pipe rupture as the design basis for emergency core cooling
- systems, containments and environmental qualification. The staff recognizes the nted
i The Comissioners 4
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to address whether and to what extent fracture mechanics analysis techniques may be used to modify present requirements relating to other features of facility design.
However, this is a longer ters evaluation.
For the present, the proposed rule allows the removal of plant hardware which it is believed negatively affects plant performance, while not affecting emergency core cooling systems, containments, and environmental qualification.
The staff plans to begin development of a Regulatory Guide and Standard Review Plan section dealing with the proposed acceptance criteria associated with the Broad Scope Rule to modify GDC-4 after review and evaluation of solicited public coment have been completed.
Recommendation:
That the Comission:
1.
Approve publication of a notice of the proposed amendments in the Federal Register (Enclosure 1);
2.
Certify that the proposed rule, if promulgated, will not have a significant economic impact on a substan-tial number of small entities. This certification is necessary in order to satisfy the requirements of the Regulatory Flexibility Act 5 U.S.C. 605(b); and 3.
Note:
a.
The notice of proposed rulemaking in Enclosure I will be published in the Federal Register allowing 60 days for public coment.
b.
The Chief Counsel for Advocacy of the Small Business Administration will be infonned of the certifications and the reasons for it as required by the Regulatory Flexibility Act.
c.
The proposed rule does not contain a new or amended infonnation collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C.
3501 et seq.).
Existing requirements were approved by the Office of Management and Budget approval number 3150-0011.
TM Connissioners 5
d.
The Federal Register notice of proposed rule-making will be distributed to affected licensees and nonlicensees.
e.
A regulatory analysis (Enclosure 2) has been prepared for this rulemaking.
f.
The Federal Register notice of proposed rule-making has been reviewed by the ACRS and the CRGR.
g.
A backfit an'alysis (Enclosure 3) has been prepared in accordance with 10 CFR 50.109.
Victor Stello, Jr.
Acting Executive Director for Operations l
Enclosures:
1.
Federal Register Notice 2.
Regulatory Analysis 3.
Backfit Analysis 4.
Public Announcement 5.
Congressional Letters 6.
Environmental Assessment
[75M 01) i j
NUCLEAR REGUtATORY COP 9tl5510N 10 CFR Part 50 Modification of General Design Criterion 4 Requirements For Protection Against Dynamic Effects of Postulated Pipe Ruptures AGENCY:
Muclear Regulatory Comission, ACTION:
Proposed rule.
The Nuclear Regulatory Comission is proposing to expand the
SUMMARY
scope of a previous amendment to its regulations dealing with the protection of structures, systees and components important to safety A recent proposed against dynamic effects from postulated pipe ruptures.
rule (50 FR 27006, July 1,1985) was limited to the primary coolant loop piping of pressurized water reactors (PWRs), whereas this present action This would cover all high energy piping in all nuclear power plants.
expanded modification of General Design Criterion 4 (GDC 4) would allow denonstration of piping integrity by analyses to serve as a basis for excluding consideration of dynamic effects associated with pipe ruptures.
The amendment will not ir. pact other design requirements such as emergency core cooling system (ECCS) perforinance, containment
- design, and environmental qualification. However, a potentially more extensive removal of protective devices, such as pipe whip restraints and jet impingement shields, and other related design changes would be perinitted if rigorous acceptance criteria are satisfied.
)
[7690-01)
DATE:
Coment period expires (60 days after publication).
- Coments, rece'ived af ter this date will be considered if it is practical to do so, but assurance of consideration can only be given to coments received on or before this date.
ADDRESSES:
Send coments to:
The Secretary of the Camissis, U.S.
Nuclear Regulatory Comission, Washington, DC 20555, ATTN: Docketing and Service Branch.
Deliver coments to:
Room 1121, 1717 H Stnet, NW, Waf hington, DC between 8:15am and 5:00pm wekdays.
Copies of the regulatory analysis, documents referenced in this notice, and coments received may be examined at:
the NRC Public Document Room at 1717 H Street, W. Washington, DC.
/
FOR FURTHER INFORMATION CONTACT: John A.
O'Brien, Office of Nuclear DC Regulatory Research, U.S. Nuclear Regulatory Comission. Washington, 20555, telephone (301)443-7854
$UPPLEMENTARY INFORMATION:
Table of Contents
===1.
Background===
- 11. Scope of Rulemaking
!!I. Proposed Rule IV. Sumary of Acceptance Criteria y,
invitation to Corrent VI. Availability of Documents VII. Finding of No Significant Environmental Impact: Availability Vllt. Paperwork Reduction Act Statement 2
[769001)
IX. Regulatory Analysis X.
Backfit Rule XI. Regulatory Flexiblity Act Certification X11. List of Subjects In 10 CFR Part 50
~
BACKGROUND Background to this rulemaking can be found in the limited scope modification to GDC 4 published as a proposed rule in the Federal Register on July 1,1985(50FR27006). The Comission adopted a two-step approach to the modification because safety and economic benefits could be quickly realized without extensive and time consuming review and discussion if the scope were initially limited to the primary main loop piping of PWRs.
show # that the Substantial evidence had already been developed to "leak-before-break" concept was valid for primary main coolant loops of The Comission decided not to defer the limited application of PWRs.
leak-before-break technology while the detailed provisions of its Many near ters acceptance criteria were being reviewed and approved.
operating license (NTOL) nuclear power plant units and operating nuclea
/
power plant units had requested exemptions from the requirements of l
A broader application of j
end could benefit.from the limited scope rule.
i leak-before-break technology requires adoption of the proposed general acceptance criteria published in NUREG-1061 Volume 3, Chapter 5, Nove 1984.
i 3
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SCOPE OF RUttMAKlhG The direct dynamic effects of pipe rupture are missile generation, pipe whipping, pipe breek reaction forces, jet impingement forces, decompression l
f waves within the : ruptured pipe and pressurization in
- cavities, subcompartments and compartments.
To retain high safety margins, the application of leak-before-break technology to various piping systems should not decrease the capability of l
containments to perform their function of isolating the outside environment from pot 2ntial leaks, breaks, or malfunctions within the containment.
l Containments will continue to be designed to accommodate loss of coolant i
accidents resulting from breaks in the reactor coolant pressure boundary up to and including a break equivalent in size to the double ended ' upture of r
the largest pipe in the reactor coolant system.
As a consequence, l
t pressurization from such breaks of the volumes indicated below are still j
I included in the design basis even though other pressurizations due to pipe ruptures are excluded:
1 1.
The containment pressure boundary, Channeling elements (vents, vent headers, and downconers) in BWR 2.
pressure suppression type containments, Corpartments necessary to the containment function, for example, PWR i
l 3.
ice condenser containment structures providing separation of enntainment volumes.
i 1
l 4
4 l
[7590-01) likewise, the design bases for emergency core cooling systems, and for environmental qualification still retain non-mechanistic pipe rupture.
However, further investigations may indicate a need to modify these requirements as well.
This proposed amendment to GDC 4 allows application of leak-before-break technolog[y to demonstrate by analyses the integrity of all high energy reactor piping in all nuclear power units that meet rigorous acceptance criteria.
There are two reasons for the decision to treat only dynamic effects in this rulemaking as opposed to other related requirements which could be interpreted to derive from postulated pipe ruptures, j First, loss-of-coolant accidents that place requirements on safety systems and structures include breaches of the fluid system pressure boundary other than those due to pipe rupture Second, studies completed to date by Lawrence Livennore National Laboratory under contract to the NRC indicate that adverse safety imr~ '" - "4n result from requiring protective devices to resist the dynamic effects assoc, rupture.
(See NUREG/CR-4263, "Reliability Analysis of Stiff Versus l
Flexible Piping, Final Project Report", May 19P,5).
The placement of pipe whip restraints degrades plant safety when thermal growth is inadvertently restricted, reduces the accecsibility for and effectiveness of inservice inspections, increases inservice inspection radiation dosages and adversely affects construction and maintenance economics.
4 5
l
[7590-01)
When leak-before-break is shown to be applicable to a fluid system piping, component supports may be redesigned excluding the effects of pipe rupture.
j The use of ASME code allowables in component support redesign is judged sufficient for preventing pipe rupture due to component support failure.
Nonetheless, redesign'ed component supports must have sufficient margins, a
consnensurate to the degree that leak before-break acceptance criteria are met, such that component support failure is a remote cause of pipe rupture.
PROPOSED RULE The proposed rule consists of two substitute ' sentences at the end of GDC 4 (replacing the sentence introduced by the limited scope rule) permitting the use of analyses to exclude dynamic effects of pipe ruptures in all high i
energy piping in all nuclear power units.
As a minimum, a deterministic fracture mechanic.s. evaluation including the effects of fatigue is undertaken.
Investigations of potential indirect failure mechanisms which could lead to pipe rupture may also be required on a case-by-case basis.
In order to demonstrate that the probability of fluid system piping rupture is extremely low, applicants and licensees may follow procedures and acceptance criteria developed by the ff.
Current procedures and acceptance criteria, are presented in Chapter 5 of Volume 3 NUREG-1061, j
l November 1984.
These acceptance criteria are sumarized in the next section.
The supporting safety analysis must demonstrate from the results of a l
fracture mechanics analysis that a substantial range of stable pipe crack 6
.c
[759001) sizes can exist for an extended period which provide detectable leaks, and that the fluid systems piping will not rupture under these conditions consistent with the design basis for the piping.
The language of the proposed rule specifies "conditions consistent with the design basis for the piping." The design basis for the piping means those conditions which were current when the FSAR was prepand, and may include 10 CFR Part 50 (especially the General, Design Appendix A to Part 50), applicable sections of the Standard Review Plan, Regulatory Guides and industry standards such as the ASME Boiler and Pressure Vessel Code.
The heading of GDC 4 is revised from "Environmental and missiJe design bases' to read ' Environmental and dynamic effects design bases" to clarify that this General Design Criterion covers other dynamic events than the effects of missiles.
The ters "extremely low" is used in this amendment to GDC 4 with reference to the probability of fluid system pipe rupture.
For reactor coolant loop piping, a representative value which would qualify as "extremely low" would be of the order of
- 10'0 per reactor year when all rupture locations are considered in the fluid system piping.
For other piping, representative values will be developed consistent with this definition as the need arises. The probability cf 10-6 per reactor year agrees with the implicit design goal of components and structures that are correctly engineered on-4 deterzinistic basis (assuming verified fabrication and adequate inservice inspection), and includes the probability of an initiating event occurring (such as an earthquake, abnonnal transient or an accident).
The 7
[759001) deterministic basis presumes that these components and structures are engin ered to meet the applicable regulations and NRC-endorsed industry 1
Codes.
l Modifications of the licensed plant design of operating plants may involve en unreviewed safety question under 10 CFR 50.59. Licensees of operating F
plants desiring to make modifications may submit a license amendment for NRC approval in accordance with revised General Design Criterion 4.
The i
license amenhent may also include provisions for an augmented laakage detection system or other license conditions developed during the rulemaking action.
Applicants for operating licenses seeking to modify design features to take I
advantage of this rule are required to reflect the revised desiga in an i
amendment to the pending FSAR.
If the design change modifies design criteria set forth in the PSAR, an amendment to the applicable construction j
permit may also be necessary. The amendment to the FSAR, and the applica-i tion for anenbent of the construction permit if necessary, may include i
j provisions for aupnented leakage detection or other design or operating j
features developed during the rulemaking action, i
j l
This rulecaking will introduce an inconsistency into the d:: sign basis by i
excluding only the dynamic effects of postulated pipe ruptures while still j
1 retaining non-mechanistic pipe rupture for emergency core cooling systems.
4 containments, and environmental qualification.
The Comission recognizes j
the need to address whether and to what extent fracture mechanics analysis t
techniques may be used to modify present requirements relating to other 8
ll i
i
[759001) j
)
features of facility design.
However, this is a longer tenn evaluation.
For the present, the proposed rule allows the removal of plant hardware which it is believed negatively affects plant performance and safety, while not affceting emergency core cooling
- systems, containments, and environmentalqualifidation.
Value-impacts resulting fran this rule are greatest for future plants, j
where estimated costs can be reduced approximately $100 million per unit.
Of this sum, about $30 million are direct costs and the balance stems from reduce'd financing costs and improved scheduling.
Reductica in worker radiation exposures vary from plant to plant, but are in the range of 300
]
to 800 man-rem.
Public risk was not quantified, but is believed to decrease due to improved effectiveness of inservice inspection and enhanced j
safety. The above quoted figures are based primarily on the elimination of pipe whip restraints and jet impingement barriers and do not treat other facility changes that could result from this rule.
For existing PWRs, considering primary coolant loops only, cost savings of
$186 million and reductions of 34,000 man-rem are estimated for a population of 85 PWRs.
These figures did not include savings resulting i
from redesign of heavy component supports.
One licensee seeking to take advantage of the modification of GDC-4 is estinating a per plant cost savings of $20 million and reduced worker exposures of about 3000 man-res associated solely with a redesign of reactor coolant pump supports.
[7590-01)
?.or existing BWRs, considering only recirculation loop piping, ccst savings of $30 million and reductions of 8.600 mn-rem are estimated for a population of 38 plants.
i 1
In existing PWRs and BWRs, public risk is estimated to be insignificantly impacted, or if credit is taken for improved inservice insp4ction and enhanced safety, to be reduced by an unquantified amount.
The Comission has not investigated otheF situations than those discussed abnve; however, it is believed that other high energy piping will likewise show favorable value-impacts.
l StMMRY Of ACCEPTANCE CRITERIA The Comission plans to begin development of a Regulatory Guide and a Standard Review Plan section dealing with the proposed acceptance criteria associated with the Broad Scope modification of GDC-S after review and evaluation of solicited public connent have been completed.
While NRC acceptance criteria are subject to further elaboration and revision as the results of ongoing studies become available, some details on existing staff guidance as reflected in Chapter 5 of Volume 3,
NUREG-1061 dated Nove:nber 1984, are given below, i
4 10
o 1,759001) i
'The leak-before-break approach should not be considered applicable to fluid system piping, or portions thereof, that operating experience has indicated is particularly susceptible to failure from the effects of corrosion (e.g.,
intergranular stress corrosion cracking), water Neuner or low and high cycle (i.e., thermal..mechar,1 cal) fatigue. To show that piping systems are not susceptible to failure from water hawner and corrosion, the extremely low probability criterion must be satisfied.
This can be recomplished through investigations of operating history and seasures to prevent or mitigate these phenomena.
The leak-before-break approach should not be considered applicable if there is a high probability of degradation or failure of the piping from tectreet causes such as fires, missiles, and damage from equipment failurys (e.g.,
cranes), and failures of systems or components in close proximity to the pipe. Piping systems, or portions thereof located in non Seismic Category I structures may take advantage of leak-before-break technology when it can be shown that these structures, either as originally des 10ned or through redesign, can resist SSE loads with acceptable margin.
]
j l
The leak-befor portions thereof, in which analyses are perfomed using the as-built configuration (as opposed to the design configuration).
The as-built configurstion description should be sufficiently accurate and reliable such that analyses yield credible results.
1 I
The leak before-break approach is limited to piping systerns wb,tre the material h not susceptible to cleavage-type fracture over the full range of systems operating temperatures.
[7590-01)
,Leaka ge detection systems should be suf ficiently reliable, redundant, diverse and sensitive so that a margin no less than 10 on detection of unidentified leakage from throughwall flaws exists. Also, residual welding stresses and cold springing stresses on a case-by-case basis may require l
special remedies.
i
\\
In performing the fracture mechanics evaluations, the following procedure meets NRC acceptance criteria.
Specify the type and magnitude of the loads applied (forces,' bending and torsional moments), their source (s) and method of combination.
Identify the location (s) at which the highest stresses coincident with I
poorest material properties occur for base materials, welttnents, and safe.
ends.
Identify the types of materials and materials specifications used for base metal, weldments and safe ends, and provide the material properties including apppropriate toughness and tensile data, long-term effects such as thermal aging and other limitations.
Postulate a flaw at the location (s) where the highest stresses coincident with poorest material properties occur for base materials, weldments and safe ends and that would be permitted by the acceptance criteria of Section 1
XI of tW ASME Boiler & Pressure Vessel Code. Demonstrate by fatigte crack growth enalysis that the crack will not grow significantly during service.
12 1
[7590-01)
P,ostulate a throughwall flaw at the location (s) above.
Th* size of the flaw should be large enough so that the leakage is assured of detection with margin using the installed leak detection capability when the pipes are subjected to normal operating loads.
If auxiliary leak detection systems are relied on; they should be described.
Assume that a safe shutdown earthquake (SSE)
- curs prior to detection of the leak to demonstrate that the postrhted leakage flaw is stable under normal operating plus SSE loads for a long period of time; that is, crack growth, if any, is minimal during an earthquake.
Determine flaw size margin by comparing the selected leakage size flaw to critical size crack. Using normal plus SSE loads, demonstrate that there is a margin of at least 2 between the leakage size flaw and the' critical size crad to account for the 'Jncertainties inherent in the analyses and leak detection upability.
Determine margin in terms of applied loads by a crack stability analysis.
Demonstrate that the leakage-size cracks will net exprience unstable crack growth even if larger loads (at least Etimes the normal plus SSE loads) are applied.
Demonstrate that crack growth is stable and the final crack size is limited such that a double-ended pipe break will not occur.
The piping materials toughness (J-R curves) and tensile (stress-strain curves) properties should be determined at temperatures near the upper range of normal plant operation.
The test data should demonstrate ductile behavior at these temperatures.
13
[7590001)
The J-R curves should be obtained using specimens whose thickness is equal or greater than that of the pipe wall. The specimen should be large enough to provide crack extensions up to an amount consisten. dith J/T condition determined by analysis. Because practical specimen size limitations exist, the ability to obtal the desired amount of experimental crack extension may be restricted. In this case, extrapolation techniques may be used.
The stress-strain curves should be obtained over the range from the
~
proportional limit to maximum load.
The materials tests should be conducted using archival material for the pipe being evaluated.
If archival material is not available, tests should be conducted using specimens from three heats oi material having the same
/
material specification. Test material should include base and weld metals.
At least two stress-strain curves and two J-resistance curves should be developed for each of a minimum of three heats of materials having the same material specifications and tiermal and fabrication histories as the in-service piping material.
If the data are being developed from an archival heat of material, a ninimum of three stress-strain curves and three J-resistance, curves frce that one heat of waterial is sufficient.
The tests should be conducted at temperatures.near the upper range of normal plant operation.
Tests should also be conducted at a lower teinperature, which may represent a plant candition where pipe break would present' safety concerns similar to normal operation.
These tests are intended only to determine if there is any significant dependence of 1
14
[7590-01) toughness on temperature over the temperature range of interest.
One J-R curve and one stress-strain curve for one base metal and weld metal are considered adequate to determine temperature dependence.
There are certain ljaitations that currently preclude generic use of limit-load analyses to evaluate leak-before-break conditions for eliminating pipe restraints.
However, limit-load analysis can be used to demonstrate acceptable leak-before-break margins, provided the limit moment is greater than the applied (normal operaiton plus safe shutdown earthquake (SSE)) moment at any location in the pipe run by a factor of ai, least three.
INVITATION TO C0fEENT Coment is invited on the following topics:
l 1.
Value-impacts associated with this expanded modification to GDC 4, with particular reference to experience with the use of pipe whip restraints and jet impingement shields near nuclear reactor piping.
(The value-impact analysis prepared by Lawrence Livennore National Laboratory is available for inspection and copying for a fee in the NRC Public Document Room,1717 H Str2et NW, Washington, D.C.)
2.
The scope of piping which could or should be affected, supported by technical justifications.
15
[7590-01) 3.
.The decision to limit impacts of this modification of GDC 4 to only dynamic effects associated with pipe rupture.
4.
The acceptance criteria which the Comission proposes to use to evaluate whether leak-before-break technology is appitcable to specific situations.
Acceptable allowables for pipe-connected component supports which 5.
would provide adequate asstirance that component support failure would not be a source of the pipe ruptur'e loads being eliminated from the
. design basis.
AVAILABILITY OF DOCUMENTS Copies of NUREG-1061, Volume 3, may be purchased by calling (202) 1.
275-2060 or (202) 275-2171 nr by writing to the Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, D.
C.,
20013-7082, or purchased from the National Technical Infonnation Service, Department of Comrnerce, 5285 Port Royal Road, Springfield, VA 22161.
2.
Copies of NUREG/CR-4263, may be purchased by cailing (202) 275-2060 or (202)275-2171 or by writing to the Superintendent of Documents, U.S.
Government Printing Office, Post Office Box 37082, Washington, D.C.,
20013-7082, or purchased from the Nationni Technical Infonnation Service, Depannent of Conmrce 5285 Port Royal Road, Springfield, VA 22161.
16
l l
,' a
[7590-01)
FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY l
The Comission has determined under the National Environment Policy Act of 1969, as amended, and the Concission's regulations in Subpart A of 10 CFR Part 51, that this rdle, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. Although the removal of certain plant hardware could result, this will not alter the environcental impact of the licensed activities as set out in the Final Environmental Impact Statement for each' facility.
The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room,1717 H Street, NW, Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from John A. O'Brien, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Comission, j
Washington, DC 20555, telephone (301) 443-7854.
PAPERWORK REDUCTION ACT STATEMENT This proposed rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing information collection requirements under 10 CFR Part 50 were approved by the Office of Management and Budget approval number 3150-0011.
17
-,-,n
-.,-,,-,,_n--,---
[7590-01)
REGULATORY ANALYSIS The Comission has prepared a draft regulatory analysis on this proposed regulation.
The analysis examines the. costs and benefits of the alternatives considered by the Comission. The draft analysis is available for inspection in the NRC Public Docum nt Room, 1717 H Street W.
Washington, DC. Single copies of the analysis may be obtained from John A.
O'Brien, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington, DC 20555, telephone (301) 443-7854. The Comission requests public coment on the draft regulatory analysis. Coments on the draft analysis may be submitted to the NRC as indicated under the ADDRESSES heading.
BACKFIT RULE l
A backfit analysis (see 10 CFR 50.109) has been prepared and is available for inspection in the NRC Public Document Room, 1717 H Street W, Washington, D.C.
Single copies of the backfit analysis may be obtained from John A. O'Brien, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission Washington, DC 20555, telephone (301) 443-7854.
REGULATORY FLEX 1BILITY ACT CERTIFICATION i
As required by the Regulatory Flexibility Act of 1980, (5 U.S.C. 605(b)),
the Comission certifies that this rule, if adopted, will not have a 18
T
[7590-01]
,significant economic impact on a substantial number of small entities.
This rule affects only the licensing and operation of nuclear power plants.
The companies that own these plants do not fall within the scope of the definitions of "small entities" set forth in.the Regulatory Flexibility Act or the Small Business' Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121.
LIST OF SU8JECTS IN 10 CFR PART 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations Nuclear power plants and reactors.
Penalty, Radiation protection. Reactor siting criteria, Reporting and recordkeeping requirements.
l For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing to adopt the following amendments to 10 CFR Part 50, 1
PA'RT 50 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FA 1.
The authority citation for Part 50 continues to read as follows:
I AUTHORITY:
Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, I
937, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2133, 2134, 2201, 2232, 2233, 2236, 2239, 2282);
)
secs.
1 19 3
,,,e
},' ; "
[7590-01) l 201, 202, 206, 88 Stat.1242,1244,1746, as amended (42 U.S.C. 5841, 5842, 5846), unless otherwise noted.
Section 50.7 also issued under Pub. L.95-601, sec.10, 92 Stat. 2951 (42 U.S.C. 5851).
Se'ctions 50.57(d), 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat. 2071, 2073 (42 U.S.C. 2132, 2239).
Section 50.78 als'o issued under sec.122, 68 Stat. 939 (42 U.S.C. 2152).
Sections j
t 50.80-50.81 also issued under sec.184, 68 Stat. 954, as amended (42 U.S.C.
2234).
Sections 50.100 - 50.102 also issued under sec. 186, 68 Stat. 955 j
(42U.S.C.2236).
For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C.
2273), il 50.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54, and;50.80(a) j are issued under sec.161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b));
il 50.10(b) and (c) and 50.54 are issued under sec. 1611, 68 Stat. 949, as amended (42 U.S.C. 2201(1)); and il 50.55(e), 50.59(b), 50.70, 50.71, J
50.72, 50.73, and 50.78 are issued under sec. 161o, 68 Stat. 950, as
)
amended (42 U.S.C. 2201(o)).
2.
In Appendix A General Design Criterion 4 is revised to read as fcllows:
APPENDIX A - GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS 20
~
..1:,
[7590-01)
~ ~ '
- CRITERIA
!. Overall Requirements I
Environmental and dynamic effects design bases.
Criterion 4 Structures, systems, and cocponents important to safety shall be designed to accorinodate the effects of and to be compatible with the environmental conditions associated with nonnal operation, maintenance, testing, and postulated accidents, in'cluding loss-of-coolant accidents. These structures, systems, and components shall be cppropriately protected against dynamic effects, including the effects of missilet, pipe whipping, and discharging fluids, that may result ~ from equipment failures and from eyents and conditions outside the nuclear power unit.
However, dynamic effects associa*ed with postulated pipe ruptures in nuclear power units may be excluded fror the design basis when analyses demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
These analyses must include, as a minimum, a detenninistic fracture mechanics evaluation of the piping.
Dated at Washington, D.C. this day of 1986.
I For the Nuclear Regulatory Comission.
Samuel J, Chilk, Secretary of the Comission.
21
s I
USI A-43, CONTAINMENT Ef15RGENCY SUMP PERFORMANCE PROPOSED RESOLUTION PREPARED FOR:
COMMITTEE TO REVIEW GENERIC REQUIREMENTS l
T. P. SPEIS, DIRECTOR A. W. SERKlZ, TASK MANAGER DIVISION OF SAFETY TECHNOLOGY OFFICE OF NUCLEAR REACTOR REGULATION SEPTEMBER 9, 1985
1 BACKGROUND 1)
LAST MEETING ON A-43 WAS HELD WITH CRGR ON 7/11/84.
2)
.SINCE THEN:
A)
CRGR QUES 110NS & COMMENTS HAVE BEEN CONSIDERED AND ADDRESSED.
B)
IHE REGULATORY ANALYSIS HAS BEEN REVISED UTILIZING RESULTS OBTAINED FROM NORE RECENT PRA AND SEVERE ACCIDENT RELATED STUDIES AS WELL AS CONTAINMENT
+
PERFORMANCE STUDIES.
l 3)
THE NET EFFECT HAS BEEN A RELAXATION OF PREVIOUSLY PROPOSED REQUIREMENTS.
- s 2
REASSESSMENT OF PLANTS 1)
AS NOTED BEFORE, ALL PLANTS CANNOT BE TREATED AS A HOMOGENEOUS GRO'P DUE TO DIFFERENCES IN PLANT DESIGN AND J
CONTAINMENT CAPABILITY.
WE HAVE REFINED THE REGULATORY ANALYSIS FROM THE VIEW POINT OF DIFFERENT CONTAINMENT i
CAPABILITIES.
2)
CORE MELT FREQUENCIES, AVERTED RISK ESTIMATES AND VALUE/
IMPACT RATIOS WERE DEVELOPED FOR THE FOLLOWING CONTAINMENT DESIGNS:
A)
PWR LARGE DRYS t
B)
PWR SUBATMOSPHERIC l
j C)
ICE CONDENSERS D)
MARK IS & IIS E)
MARK lilS
3 REVISED IMPLEMENTATION LANGUAGE DUE TO CONCERNS RAISED REGARDING IMPLEMENTATION LANGUAGE IN THE PROPOSED REVISIONS TO RG 1,82 AND SRP SECTION 6,2.2, THE FOLLOWING LANGUAGE IS PROPOSED AS A SUBSTITUTE:
"!S APPLICABLE TO:
1)
CONSTRUCTION PERMIT APPLICATIONS AND PRELIMINARY DESIGN APPROVALS (PDAS) TdAT ARE DOCKETED AFTER SIX (6) MONTHS FOLLOWING ISSUANCE OF REGULATORY GUIDE 1.82, REVISION 1, 2)
APPLICATIONS FOR FINAL DESIGN APPROVAL (FDAS),
AGGEPTABLE FOR FORWARD REFERENCING THAT HAVE NOT t
RECEIVED APPROVAL AT SIX (6) MONTHS FOLLOWING ISSUANCE OF REGULATORY GUIDE 1.82, REVISION 1, j
OUR INTENT IS TO MAKE CLEAR THAT RESOLUTION OF THIS USI WILL NOT IMPACT PLANTS ALREADY UNDER CONSTRUCTION,
i PARAMETERS USED IN REVISED A l43 ANALYSIS PARAMETERS USED IN SUMP BLOCKAGE PROBABILITY 2
DEBRIS SCREEN AREA 50 To 200 FT FA oW RECIRCULATION a en RATE 6000 TO 10,000 GPM BREAK DESTRUCTION ZONE 3 To 7 L/D NPSH AVAILABLE 1 To 5 FT OF WATER OPERATOR RECOVERY CONDITIONAL PROBABILITY USED TO CALCULATE CORE MELT PROBABILITY 0.5 PL ANTS EVALUATED IN CALCULATION OF 0FFSITE RISKS PWR LARGE DRY CONTAINMENTS WITH SGFCS l
PWR LARGE DRY CONTAINMENTS WITHOUT SGFCS PWR SUBATMOSPHERIC CONTAINMENTS PWR DRY OR SUBATMOSPHERIC CONTAINMENTS WITH SPRAY RECOVERY PWR ICE CONDENSERS BWR MARK I AND 11 CONTAINMENTS
/
BWR MARK 111 CONTAIMMENTS BWR MARK IS AND llS WITH CONTROLLED VENTING PARAMETERS USED IN COST ANALYSIS 6
LIMITED PLANT DESIGN CHANGE 0.f4 x 10
$/ PLANT 6
INSULATION REPLACEMENT 1.5 x 10
$/ PLANT
.~-
t c
5 FINDINGS 1)
THE ESTIMATED CORE MELT FREQUENCY, CMF, FROM LOCA PLUS SUMP BLOCKAGE EVEN FOR PLANTS WITH SIGNIFICANT BLOCKAGE VULNERABILITY IS LESS THAN 3 x 10-5/Rx YR.
2)
SUMP BLOCKAGE PROBABILITIES ARE BASED ON PIPE BREAK ESTIMATES DERIVED IN 1977, FOR WHICH THE DATA BASE INCLUDED PIPING FAILURES OF ALL TYPES KNOWN AT THAT TIME INCLUDING MATERIALS NOT USED IN NUCLEAR PLANT PIPING.
3)
THE LOCA PROBABILITY (OR INITI ATING EVENT) IS THEREFORE UVERESTIMATED WHEN COMPARED TO CURRENT ESTIMATES OF PIPE BREAK AND BREAK TYPE, AS PREDICTED BY CURRENT FRACTURE-MECHANICS ANALYSES.
t l
4)
THE CALCULATED WIDE RANGE FOR SUMP BLOCKAGE PROBABILITY, t
CORE MELT PROBABILITY, OFFSITE CONSEQUENCES, AND VALUE-lMPACT RATIOS REFLECT THE RANGE OF PLANT DESIGN FEATURES AND COST ESTIMATES.
5)
EVEN WITHOUT CONSIDERATION OF CONTAINMENT OVERPRESSURE CAPABILITIES VALUE/ IMPACT RATIOS ARE MARGINAL.
6)
BASED ON THE ABOVE, WE DO NOT BELIEVE THAT A BACKFIT REQUIREMENT IS JUSTIFIED.
l
j 6
PROPOSED A-43 ACTIONS (1)
ISSUE THE STAFF'S TECHNICAL FINDINGS (NUREG-0897, REV. IA)
FOR USE AS TECHNICAL INFORMATION SOURCE.
(2)
ISSUE RG 1.82, REV. 1 AND SRP SECTION 6.2.2, REY. 4 WHICH HAVE BEEN REVISED TO REFLECT THE STAFF'S TECHNICAL FINDINGS.
THE 50% BLOCKAGE CRITERIA HAS BEEN REMOVED FROM R.G.1.82 AND A METHODOLOGY FOR ESTIMATING BLOCKAGE HAS BEEN PROVIDED.
THIS REVISED REGULATORY GUIDANCE WOULD APPLY ONLY TO NEW CP APPLICATIONS AND STANDARD PLANT DESIGNS.
(3)
ISSUE A GENERIC LETTER, FOR INFORMATION ONLY, TO ALL HOLDERS OF AN OPERATING LICENSE OR CONSTRUCTION PERMIT OUTLINING SAFETY CONCERNS REGARDING POTENTIAL DEBRIS BLOCKAGE AND ATTENDANT POTENTI AL FOR RECIRCULATION FAILURE.
7 GENERIC LETTER CONTENTS
{
1)
INFORMS LICENSEES 8 APPLICANTS OF GENERIC SAFETY CONCERNS RELATED TO LOCA GENERATED DEBRIS BLOCKAGE 8 LOSS OF NPSH.
2)
POINTS OUT THAT SAFETY CONCERN IS PLANT DESIGN AND INSULATION TYPE DEPENDENT.
3)
POINTS OUT THAT 50% BLOCKAGE CRITERIA UTILIZED PREVIOUSLY VIA R.G. 1.82, REV. O IS UNSUBSTANTIATED AND SHOULD NO LONGER BE USED.
PROVIDES REFERENCES FOR GUIDANCE FOR A MECHANISTIC EVALUATION OF DEBRIS BLOCKAGE.
4)
ADVISES LICENSEES AND APPLICANTS TO REVIEW SUMP DESIGNS AND SELECTION OF INSULATION, PARTICULARLY AT TIME OF t
INSULATION CHANGE OUT.
l 5)
HIGHLIGHTS PLANT DESIGN AND OPERATIONAL CONSIDERATIONS WHICH ARE IMPORTANT (E.G., SMALL DEBRIS SCREEN AREAS - LESS THAN 100 FTJ HIGH ECCS RECIRCULATION FLOWS - GREATER THAN 8000 GPM; LOW NPSH MARGIN - LESS THAN 2 FT. OF WATER.)
8 PWR LARGE DRY CONTAINMENTS (W/SGFCS, 57 PLANTS) 1)
CALCULATED RANGE OF CORE MELT FREQUENCY = 1.5 TO 25 x 10-6/Rx-YR.
2)
THESE CONTAINMENTS ARE DESIGNED TO ABSORB ALL ACCIDENT LOADS, HAVE LARGE VOLUMES AND HIGH DESIGN OVERPRESSURE CAPACITY (I.E.
110 PSIG).
3)
SGFCS HAVE THE CAPABILITY TO INDEPENDENTLY REJECT DECAY HEAT LOADS, THEREBY MAINTAINING PRESSURE LEVELS WITHIN t
DESIGN LIMITS EVEN IF CONTAINMENT SPRAY SYSTEMS WERE LOST DUE TO SUMP BLOCKAGE.
4)
IF CONTAINMENT INTEGRITY IS NOT BREACHED, FISSION PRODUCTS ARE RETAINED AND THERE ARE NEGLIGIBLE OFFSITE CONSEQUENCES.
5)
THUS BACKFIT FOR THIS CLASS OF PLANTS IS NOT SUPPORTABLE.
1 9
PWR LARGE DRY CONTAINMENTS (W/0 SGFCS, 14 PLANTS) 1)
EST'D. CORE MELT FREQUENCY DUE TO SUMP BLOCKAGE IS 1.5 TO 25 x 10-6/Rx YR.
WE BELIEVE THAT THE DESIGN OF MOST PLANTS IS SUCH THAT THE BEST ESTIMATE CORE MELT FREQUENCY IS AT THE LOWER END OF THIS RANGE.
2)
CALC'D AVERTED RISK, ASSUMING CONTAINMENT FAILURE, IS20-300 PERSON-REM /RX WHICH IS A LOW-TO-MODERATE LEVEL.
IF CORRECTIVE ACTION WERE TAKEN TO RESUME CONTAINMENT SPRAYS, THIS EST'D. RISK WOULD BE REDUCED BY A FACTOR OF 10.
t 3)
CALC'D. VALUE-lMPACT RATIOS ARE:
r 50 TO 800 PERSON REM /$M ($0.4M/Rx COST)10-200 PERSON REM /$M ($1.5M/Rx COST) 4)
THUS BACKFIT FOR THIS CLASS OF PLANTS IS NOT SUPPORTABLE.
10 PWR SUBATMOSPHERIC CONTAINMENTS (17 PLANTS) 15 THE EST'D. CORE MELT FREQUENCIES, OFFSITE CONSEQUENCES AND VALUE-!MPACT RATIOS FOR THIS CLASS OF PLANTS ARE THE SAME AS FOR PWR "DRYS" W/0 SGFCS.
2)
THESE PLANTS ALSO HAVE LARGE VOLUMES AND HIGH INHERENT 4
OVERPRESSURE CAPACITY.
l 3)
CONTAINMENT PRESSURE 8 TEMP. TRANSIENT DUE TO LOSS OF SPRAYS WOULD BE GRADUAL (SIMILAR TO "DRYS" W/0 SGFCS).
t 4)
THUS BACKFIT FOR THIS CLASS OF PLANT IS NOT SUPPORTABLE.
11 PRW ICE CONDENSER PLANTS (10 PLANTS) 1)
THE ESTIMATED CORE MELT FREQUENCY DUE TO SUMP BLOCKAGE IS 0.5 TO 4,5 x 10-6/Rx-YR.
(LESS THAN FOR LARGE DRYS) 2)
EXCLUSIVE USE OF REFLECTIVE METALLIC INSULATION (RMI)
ON THE PRIMARY SYSTEM PIPING 8 COMPONENTS.
UNLIKE FOR FIBROUS INSULATION DEBRIS, 100% FLOW RESTRICTION IS UNLIKELY FOR REFLECTIVE METALLIC INSULATION, 3)
LOW SUMP APPROACH VELOCITIES WHICH INHIBIT OR REDUCE DEBRIS TRANSPORT, 4)
RELATIVELY LARGE DEBRIS SCREENS AND NPSH MARGINS FOR MOST OF THESE PLANTS, l
5)
ICE CONDENSER PLANTS (ICP) ARE MOST SUSCEPTIBLE TO OVERPRESSURE FAILURE & DO NOT HAVE SCFCS, 6)
THE CALCULATED AVERTED RISK IS60-560 PERSON-REM /Rx, 7)
CALCULATED VALUE/ IMPACT RATIOS ARE:
160 TO 1400 PERSON-REM /$M ($0,4M/Rx COST) 25 TO 380 PERSON-REM /$M ($1,5M/Rx COST) m
i 12 PWR ICE CONDENSER PLANTS
)
(CONT'D.)
8)
THUS BACKFIT FOR THIS CLASS OF PLANT IS MARGINAL ON VALUE/ IMPACT BASIS AND IS NOT SUPPORTABLE BASED ON THE i
ABOVE RESULTS AND THE KNOWN VALUES OF THE SIGNIFICANT DESIGN PARAMETERS FOR THIS CLASS OF PLANT.
b 1
13 MARK I & 11 CONTAINMENTS (23 MARK IS, 10 MARK llS) 1)
ESTIMATED CORE MELT FREQUENCY BASED ON NRC AND INDUSTRY PRAS IS 2 TO 10 x 10-6/Rx-YR.
2)
EST'D. AVERTED RISK IS 250-1250 PERSON REM /Rx.
3)
CALC'D. VALUE/ IMPACT RATIOS ARE:
630 TO 3100 PERSON-REM /$M ($0.4M/Rx CusT) 170 TO 830 PERSON-REM /$M ($1.5M/Rx COST) 4)
THE HIGHER EST'D. CONDITIONAL OFFSITE CONSEQUENCE ASSOCIATED WITH CORE MELT IS THE PRINCIPAL CONTRIBUTOR To t
THE EST'D. CONSEQUENCES FOR THIS CLASS OF PLANTS..
5)
ON THE OTHER HAND:
A)
MARK l & IIS HAVE HIGH CONTAINMENT DESIGN PRESSURES (1.E., 45-60 PSIG).
B)
BWR'S APPEAR TO HAVE HIGHER AVAILABLE NPSH, THEREFORE CAN TOLERATE HIGHER DEGREE OF BLOCKAGE (i.E.,
llMERICK-1 CALCS).
-~
r
l 14 MARK 1 8 11 CONTAINMENTS l
(CONT'D)
)
C)
MARK l PLANTS HAVE A VARIETY OF ALTERNATE WATER SOURCES (1.E., CWST) WHICH COULD BE UTILIZED TO RECOVER SPRAYS TO MINIMIZE CONTAINMENT OVER-PRESSURIZATION.
D)
CONTROLLED VENTING (IN PROCESS OF BEING IMPLEMENTED)
WILL MlHIMIZE CONTAINMENT OVERPRESSURIZATION & LIMIT RELEASE OF FISSION PRODUCTS.
6)
ON THE BASIS OF CONSERVATIVE ESTIMATES OF LARGE LOCA PROBABILITY AND CONSERVATIVE ASSUMPTIONS REGARDING DEBRIS TRANSPORT AND STRAINER BLOCKAGEJ AND RECOGNIZED MITIGATING t
ACTIONS WE BELIEVE THAT THE CALCULATIONS HAVE OVERSTATED THE OFFSITE RISKS AND THAT THEREFORE NO BACKFIT ACTIONS t
ARE REQUIRED.
l
15 MARK Ill CONTAINMENTS l
(8 PLANTC) 1)
THE ESTIMATED CORE MELT FREQUENCY IS 2 TO 10 x 10-6/Rx-YR.
2)
PWR BLOCKAGE CONSIDERATIONS FOR THESE PLANTS ARE SIMILAR TO MK IS 8 IIS.
THE WETWELL/DRYWELL STRUCTURAL DESIGN WOULD LIKELY IMPEDE DEBRIS TRANSPORT DUE TO THE INTERCEEDING WEIR WORK.
3)
THE ESTIMATED AVERTED RELEASE IS 25 TO 125 PERSON REM /Rx.
E 4)
THE VALUE IMPACT RATIOS ARE:
60 TO 300 PERSON-REM /$M (s $0.4M/Rx COST) r 17 TO 80 PERSON-REM /$M (a$1.5M/Rx COST l
5)
MARK lilS PLAN TO IMPLEMENT CONTROLLED VENTING PROCEDURES.
6)
THUS BACKFIT FOR THIS CLASS OF PLANT IS NOT SUPPORTABLE.
16 Table 7, Overview of Ice Condenser Plant Design and Operational Features RHR Debris NPSH Approach Flow Screen Margin Insulation Velocity Plant 12Em}
(So Ft)
(Ft Water)
Used (Ft/Sec)
Catawba 1&2 6000 135 7.9 (RHR)
Reflective
.10 Metallic DC Cook 1&2 6000 90 21.9 (RHR)
Reflective
.15 Metallic McGuire 1&2 6800 120 6.9 (CSS)
Reflective
.13 16.9 (RHR)
Metallic Sequoyah 1&2 9500 43 2.8 (RHR)
Reflective
.49
. Metallic Watts Bar 1&2 8000 43 11.5 (RHR)
Reflective
.41 8000 1642 Metallic
.11 1
- Trash rack area; this structure would intercept large size insulation debris before transport to the sump debris screen structure could occur.
i
.e f
/
NUREG-0869, Revision 1 E-15
17 1
Table 5.1, Sumary of Calculated Values and Impacts Associated with Various Plant-Containment Designs for Resolution of USI A-43 Estimated Core i..lculated CalculatedVge/
Melt Probability (1)
Risk Averted, AR Impact Ratio Type Containment (1/Rx-yr)
.(person-rem /Rx)
(person-ren/$JM
-6 PWR Dry w/o 1.5 to 25 x 10 19 to 310 48 to 780 12 to 210 SGFCs and Subatmospheric
-6 PWR Dry 1.5 to 25 x 10 w/SGFCs M IC'
-6 Condenser (2) 0.5 to 4.5 x 10 60 to 560 160 to 1400 25 to 380
-6 PWR Dry w/o 1.5 to 25 x 10 2 tn 31 5 to 33 1 to 21 SGFCs and Subatmospheric w/ Spray Recovery
-6 Mark I and 11 2 to 10 x 10 250 to 1250 630 to 3100 170 to 830
-6 Mark III 2 x 10 25 to 125 60 to 310 17 to 80
-6 Mark I and 11 2 to 10 25 to 125 60 to 310 5-17 to 83 Y
w/ Venting and Spray Recovery (1)The estimated core melt frequency is based on the conditional consequences discussed in Appendix E, the sump blockage frequency estimates discussed in Appendix 0, and the assumption that 50% of the time that blockage occurs leads to loss of NPSH and core melt follows.
This assignment of a conditional core melt probability of 0.5 is felt to be realistic from the viewpoint of potential detection of flow degradation, potential operator followup action to correct this situation and sump design (2) variability.A separate estimate of sump blockage probability was made for the i condenser plants (see Appendix E) which takes into account their specific (3)designfeatures.The value/ impact ratios have been calculated for an estimated cost of
$0.4M/Rx (this assumes retrofit costs would be minimal) and for an estimated cost of $1.5M/Rx (this cost assumes replacement of troublesome insulation (s)); see Appendix G for a discussion of estimated costs.
NUREG-0869, Revision IB "Pris v
T&ble 8, Overview of Consequences Associated with Sumo Blockage Est* mated Estimated Assumed
~
F Conditionel Blockage Core Melt Estimated 1
Consequences Probal_,11ty Conditional Risk Averted **
h Type Containment (person-ram)
(1/Rx-yr)
Probability *
(AR, person-ren/Rx)
=>
-6 3 to 50 x 10 0.5 PWR D y w/SGFCS 6
1 to 9 x 10-6 PWR Ice ConA nser 5.x 10 0.5 40 to 560 PWR Dry w/o SGFCs 5
-6 and Subatmospheric 3 x 10 3 to 50 x 10 0.5 19 to 313 4
-6 PG Dry w/o 3GFCs 5 x 10 3 to 50 x 10 0.5 2 to 31 and Subatmospheric, m
h w/ Spray Recovery 6
-6 Mark I and II 5 x 10 4 to 20 x 10 0.5 250 to 1250 6
-6 Mark III 5 x 10 4 to 20 x 10 0.5 25 to 125
-6 Mark I and II 5 x 10 4
20 x 10 0.5 25 to 175 w/ venting or spray recovery TL-
"The assumption is made that 50% of the time that blockage occurs, core melt would occur.
assignment of a conditional core melt probability is realistic in view of potential operator dection and mitigatiag actions which could be taken.
- An outstanding n actor life span of 25 years has been assumed.
5 l
s