ML20212L082
| ML20212L082 | |
| Person / Time | |
|---|---|
| Issue date: | 03/05/1987 |
| From: | Reyes J NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Mcknight J NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| FRN-52FR6334, RULE-PR-50 NUDOCS 8703100192 | |
| Download: ML20212L082 (31) | |
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i 05 MAR Y MEMORANDUM FOR:
James C. McKnight-
-Document Control. Branch Division of Information Support Services Office of Information Resources Management FROM:
Jose N. Reyes, Jr.
Reactor Systems Branch Division of Reactor Systems Safety Office of Nuclear Regulatory Research 1
SUBJECT:
TRANSMITTAL 0F DOCUMENTS RELATED TO THE PROPOSED REVISION TO 10 CFR 50.46 AND APPENDIX X (52 FR 6334) TO THE PUBLIC DOCUMENT ROOM The enclosures to this memorandum have been cited in the Notice of Proposed Rulemaking, " Emergency Core Cooling Systems; Revisions to Acceptance Criteria,"
(52 FR 6334) published March 3,1987, as being available in the'Public Document Room (PDR). Therefore I request that the enclosed documents be processed for
-the Document Control System and submitted to the PDR.
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Jose N. Reyes, Jjr Rpactor Systems Ltanch Division of Reactor Systems Safety Office of Nuclear Regulatory Research
Enclosures:
1.
Environmental Assessment and Draft Finding of No Significant Impact.
2.
Regulatory Analysis for the Revision of the ECCS Rule.
3.
SECY-83-472 cc: PDR, Advanced Copy
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i 070305 g2 2FR6334 m.
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ENVIRONMENTAL ASSESSMENT AND DRAFT FINDING OF NO SIGNIFICANT IMPACT ACCEPTANCE CRITERIA FOR EMERGENCY CORE COOLING SYSTEMS; AS PERTAINING TO THE NOTICE OF PROPOSED RULEMAKING (52 FR 6334; MARCH 3, 1987)
' NUCLEAR REGULATORY COMMISSION ~
[10 CFR PART 50]
Acceptance criteria for Emergency Core Cooling Systems; Environmental Assessment and Draft Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (The Consnission) is considering revisions to i 50.46 and Appendix K of 10 CFR Part 50 which specify requirements of emergency core cooling systems (ECCS) for light water reactors.
ENVIRONMENTAL ASSESSMENT IDENTIFICATION OF PROPOSED ACTION:
Section 50.46(a)(1) would be revised to eliminate the requirement to use the features of Appendix K when calculating ECCS performance during a loss-of coolant accident (LOCA). Section 50.46(a)(1)(1) of the amended rule would allow use of realistic analytical techniques and would require that the uncertainty of the calculation be evaluated and considered when comparing the results of the calculation with the temperature limits and other criteria of 950.46(b). Section 50.46(a)(1)(ii) would be added to allow continued use of the features of Appendix K as an alternative to the uncertainty evaluation required by the amended 9 50.46(a)(1)(1). Sections 50.46(a)(2) and 50.46(a)(3) would be revised to eliminate historical implementation sections and to specify requirements for reanalyses and reporting which are excluded from consideration in this environmental assessmen'. per i 51.22 of 10 CFR Part 51.
Appendix K of 10 CFR Part 50 would be revised to make minor technical changes to the acceptable features of the calculations.
NEED FOR PROPOSED ACTION:
The proposed revisions of 10 CFR Part 50 and Appendix K are required in order to pennit new knowledge of ECCS performance gained through research to be used in the calculetions of ECCS performance. The improved calculations 1
~ -
For the case of a small (i.e., 5%) increase in total power, there would be a correspondingly small. increase in fission product inventory, routine releases of-radioactivity and fuel use. - However, maximum allowed releases of-radioactivity during both accident situations and routine operation are-specified by technical specifications and other sections of 10 CFR Part 50 which are unchanged by this revision.
It is not expected that the small increase in total power that could result from this revision would result in difficulty in meeting the existing release limits. An increase in total power would increase the thermal discharge to the environment by an amount approximately proportional to the increase in power. -The discharge of heat to surface waters is regulated under the Clean Water Act by the U.S. EPA or-
' designated state agencies. NRC would defer to procedures under that Act to establish the acceptability of any increase in waste heat discharge.
It is not intended that NRC approval of increased power level affect in any way the responsibility of the licensee to comply with requirements of the Clean Water Act. These being the only potential environmental considerations, the NRC staff believes that site specific environmental impact assessments will not be of help to the decision process.
ALTERNATIVES TO THE PROPOSED ACTION:
The staff has considered a number of alternatives to revise 6 50.46 and Appendix K.
However, all the alternatives considered would allow similar increases in local or total power, with the exception of the alternative of making no changes. Since the environmental impact of the proposed revision is considered to be not significant and the revision is required to reduce unwar-l ranted restriction on the operation of some reactors, the Commission has l
decided to proceed with the proposed revision, l
l AGENCIES AND PERSONS CONSULTED:
The NRC staff consulted U.S. manufacturers of nuclear power plants to determine the maximum increase in local or total power that might result from application of the proposed rule revisions. The staff did not consult other agencies or persons.
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would allow relaxation of restrictions which are preventing optimal operation of_ some reactors and are not necessary to adequately protect the health and safety of the public.
ENVIRONMENTAL IMPACTS OF THE PROPOSED ACTION:
The proposed revisions would be likely to reduce the cladding temperatures that are calculated during a LOCA and allow an increase in the peak local power of the reactor, while still meeting the temperature limits and other criteria of i 50.46(b).
An increase in the allowed peak local power could be used in either or both of the following manners:
1.
The total maximum allowed power of the reactor would be unchanged, but plant efficiency would be improved by increased flexibility in the allowed power shape. More efficient fuel utilization, more flexibility in changing total power, and reduced derating of plants due to fuel limits might be possible.
2.
The total maximum allowed power of the reactor might be increased.
The expected maximum increase in total power for existing and currently planned reactors is approximately 5% based on practical limits of plant hardware.
Either of these actions would reouire an amendment to the plant license to change the technical specification limits and, therefore, would result in an environmental assessment specific to that particular plant and the specific amendment being considered. This environmental assessment is a generic evalua-tion considering the typical impact of the rule revision.
A change in the allowed peak local power, without an increase in total power, would produce no significant environmental impact. The total fission product inventory, routine releases of radioactive materials and themal i
releases to the environment would be essentially unchanged. Fuel cycle changes l
would be in the direction of improved use of fuel and should not significantly l
change the environmental impact of the fuel cycle unless major new fuel cycle methods (e.g., plutonium recycle) were adopted. Such changes are beyond the j
scope of this rule revision and are not considered.
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i DRAFT FINDING OF NO SIGNIFICANT IMPACT The Commission has determined not to prepare an Environmental Impact Statement for the proposed action. The foregoing environmental assessment of this action has concluded that the proposed action would not significantly effect the quality of the human environment.
Written comments or suggestions for consideration in connection with this Draft Finoing of No Significant Impact should be submitted to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention Dockcting and Service Branch. Copies of comments received may be examined in the Commission's Public Document Room at 1717 H Street NW., Washing-ton, DC 20555. The comment period expires (60 days following publication in the Federal Register). Comments received after that date will be considered if it is practical to do so, but assurance of consideration cannot be given except as to comments received on or before that date.
Dated at Rockville, Maryland this day of
, 1986.
For The Nuclear Regulatory Commission Eric 5. Beckjord, Director Office of Nuclear Regulatory Research 4
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REGULATORY ANALYSIS ACCEPTANCE CRITERIA FOR EMERGENCY (ORE COOLING SYSTEMS 3, AS PERTAINING TO THE NOTICE OF PROPOSED RULEMAKING (52 FR 6334; MARCH 3, 1987)
REGULATORY ANALYSIS FOR REVISION OF THE ECCS RULE 1.
STATEMENT OF PROBLEN
1.1 Background
1.2 Discussion of Proposed Rulemaking 2.
OBJECTIVES 3.
ALTERNATIVES 4.
CONSEQUENCES 4.1 Safety and Risk Effects 4.2 Costs and Benefits 5.
DECISION RATIONALE 6.
IMPLEMENTATION 1
o REGULATORY ANALYSIS l
FOR REVISION OF THE ECCS RULE 1.
STATEMENT OF THE PROBLEM
1.1 Background
Section 50.46 of 10 CFR Part 50 requires that calculations be performed to show that the emergency core cooling systems (ECCS) will adequately cool the reactor in the event of a loss-of-coolant accident (LOCA). Appendix K sets forth certain required and acceptable features that the evaluation models, used to perform these calculations, must contain. The r'esults of these calculations are used to determine the acceptability of the ECCS performance.
In many in-stances, these calculations result in technical specification limits on reactor operation (e.g., peak local power) in order to comply with the 2200*F cladding temperature limit and other limits of i 50.46. These limits restrict the total power output and optimal operation of some reactors in terms of efficient fuel utilization, maneuvering capability and surveillance requirements.
TheNRC, DOE (includingAECandERDA),U.S.nuclearindustryandforeign research on ECCS performance since the present ECCS rule was issued provides a technical understanding which shows that the existing ECCS rule restrictions are more stringent than required to protect the health and safety of the public.
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1.? Discussion of Proposed Rulemaking The proposed amendment would allow alternative methods to be used to demonstrate that the ECCS would protect the nuclear reactor core during a postulated design basis loss-of-coolant accident (LOCA). While continuing to allow the use of current Appendix X methods and requirements, the proposed rule would also allow the use of more recent information and knowledge currently available to demonstrate that the ECCS would perform its safety function during a LOCA. Procedural changes would also be made to relax requirements for certain reanalyses which do not contribute to safety.The proposed amendment would apply to all applicants for and holders of construction permits or operating licenses for light water reactors.
The likely effect of the proposed rule would be to reduce the peak cladding temperatures that are calculated to occur during a LOCA. This would allow some plants to increase allowed peak local power by increasing allowed peakino factors and/or increasing total power. This regulatory analysis discusses the effect of the rule change in terms of "value" (e.g.,
public benefits such as safety) and " impact" (e.g., consequences such as costs). The intent of the proposed rule is to reduce the prescriptiveness of the rule, which unnecessarily restricts applicants and licensees (negative impact), while continuing to ensure the health and safety of the public.
It appears that the proposed rule will result in a significant benefit for some plants. However, application of the proposed rule would be optional so each applicant or licensee can perfom its own analysis to determine if its use is advantageous. The value of the proposed rule may have some negative aspects since an increase in plant power may increase risk to the public by a very small amount. However, this may be offset by other positive benefits to safety.
Information used in this regulatory analysis was obtained from a number of sources:
3
(1) Previous studies sponsored by DOE (1,2)
(2) Fonnal responses from the major reactor vendors to a request by NRC forinput(3,4,5)
(3) Informal discussions with reactor vendor, utility, national ~labora-tory and NRC-staff.
In summary, this regulatory analysis will not be the same as a conventional value-impact analysis because the proposed rule would reduce the prescriptiveness of current ECCS regulations in order to remove unnecessary operating restrictions.
It is expected that the primary effect of the proposed rule will be an economic benefit both to industry and the public.
While there may be a very slight negative effect on risk, this effect may be offset for some plants. Under every circumstance, however, the proposed rule would continue to provide sufficient safety margin to absure the health and safety of the public.
2.
OBJECTIVES The objective of the proposed rule is to incorporate into the regulations the improved knowledge gained from recent research on ECCS performance, so as to remove unnecessary operating restrictions.
3.
ALTERNATIVES A number of alternative approaches have been considered by the staff and each approach was evaluated in terms of safety, impact on the industry, NRC and industry resources required, and possibility of challenge both during the rulemaking process and during application of the rule. The alternatives that were considered are:
A.
Retain the existing rule with its present conservatism (no change).
4
'B.
Modify the ECCS rule as stated in the advance notice of proposed
.rulemaking published in the Federal Register on December 6, 1978.
C.
Modify only.certain models contained in Appendix K, for which
.research investigations have been completed and a well documented
' data base exists. These changes have been selected in areas for-which new experimental data has shown that the existing models contain a larger degree of conservatism than justified by current data uncertainties or are obviously unrealistic.
D.
Eliminate the requirement to use Appendix K models and allow realis-I tic models to be used. Reduce the 2200*F and 17% oxidation limits of i 50.46 appropriately to ensure that sufficient conservatism exists to cover uncertainties in the realistic calculation.
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E.
Eliminate the requirement to use Appendix K models and allow realis-tic models with an evaluation of the uncertainty in the overall calculation added to the results, similar to that discussed in SECY-83-472, to be used. The 6 50.46 limits of 2200"F and 17%
oxidation would be unchanged.
These alternatives are discussed in Section 5 4.
CONSEQUENCES i
4.1 Safety and Risk Effects S
The value of the proposed rule must be evaluated in terms of the i
effect on safety. The proposed rule would probably result in increased local power within the reactor core and possibly increases in total power.
Power increases on the order of 5-10% will have an insignificant effect on risk. One effect of increased power would be to increase the fission product inventory. A five percent power increase would result in a five percent increase in fission products. Thus, five percent more fission products could
)
5
s i.c be released during core melt scenarios and potentially released to the environment during severe accidents.
The proposed rule would still require that fuel rod PCT remain below 2200*F. Because significant fuel damage will not occur until 2600*F, a 400'F safety margin will remain. However, reactors choosing to increase power by five to ten percent would be operating with less margin between the PCT and the 2200'F limit than previously. The increased risk represented by this decrease in margin and increase in fission product inventor within the uncertainties of PRA risk estimates.I9)y is negligible and falls In addition, other safety limits, such as DNB, and operational limits, such as turbine design, would limit the amount of margin reduction permitted under the revised rule.
There are also safety benefits derivable from alternative fuel management schemes that could be utilized if the proposed changes were implemented. An important safety benefit could be realized if restrictions on core power peak-ing were less stringent. High neutron leakages at the core outer boundary are used in PWRs to flatten the radial core power profile. This inefficient fuel management procedure is needed to maintain peaking within tight limits. In addition, the resultant high neutron fluence leakage can enhance vessel embrittlement resulting in pressurized thermal shock (PTS) concerns. The higher power peaking factors that would be allowed with the revised rule could provide greater fuel management flexibility when attempting to reduce neutron j
flux at the vessel. This can result in a corresponding reduction in risk from PTS.
l The reduced cladding temperatures that would be calculated under the pro-posed rule offers the possibility of other design and operational changes that could result from the lower calculated temperatures. ECCS equipment numbers, i
sizes or surveillance requirements might be reduced and still meet the ECCS designcriteria(ifnotrequiredtomeetotherlicensingrequirements). The diesel / generator start time duration could be increased from the current tech-nical specification limit of 10 seconds typical for BWRs to up to 70 j
seconds.I7) These two potential modifications could result in offsetting effects on risk. On one hand, elimination of an ECC system would tend to increase risk due to loss of redundancy (single failure criteria still l
6 i
required); however, increasing the time for diesel / generator start-up may result in improved operational flexibility and diesel reliability with concomitant decrease in risk. The improved diesel reliability could result from decreased stress, wear and possibly test frequency, during surveillance testing, as well as the effect of the relaxed design requirement (70 versus 10 secondstart-up).
In summary, the effect of the proposed rule on safety would have both potential positive and negative aspects. The potential for reduction of ECC systems in existing or new plants is present. However, several safety benefits may also be realized under the proposed rule. The net effect on risk is believed to be plant specific. However, a limited generic analysis of the effect on safety has been performed.I9) 4.2 Costs and Benefits LOCA considerations resulting from the present rule are restricting the optimum production of nuclear electric power in numerous ways. These restric-tions can be placed into the following three categories:
(1) Maximum plant operating power, (2) Operational flexibility and operational efficiency of the plant, and (3) Availability of manpower to work on other activities.
Maximum plant operating power at some nuclear facilities is limited by the present Appendix K licensing results. However, it can be very difficult to clearly separate these LOCA rule restrictions from other licensing issues and limitations. There are numerous limits that can restrict total plant power, as well as the ability to maneuver the power over a wide range. Typically, this limitisassociatedwitheitherpeakcladdingtemperature(PCT)calculatedto occur during LOCA transients, or as a result of departure from nucleate boiling l
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(DNB) restrictions. Additionally, there are limits to plant power because of NRC guidance on total allowable therwal power of 3800 MW and because of physi-cal hardware limits in existing reactors on the balance of the plant (turbine.
- condenser,~pumpsandsteamgenerators).
Regarding the second category of operational flexibility, some plants have very little LOCA margin. Such a limited margin necessitates additional core power surveillance to prevent peaking factor violations. This may also require special supplementary nuclear or safety analyses and restrictive fuel manage-ment schemes, resulting in inefficient fuel burnup and no extended burnup cycles. -
The third category concerns periodic reanalyses which are required by the current rule. If an error (i 20*F change in PCT for the limiting transient) is found in an accepted evaluation model, a new LOCA analysis must be performed imediately even if the error correction results in a decrease in PCT in the limiting transient. This was a major problem several years ago, and several cases have also occurred recently, with each reanalysis costing the licensee III about $150K and diverting both licensee and NRC staff from other, more productive activities. One NRC staff year is normally needed to perform a reanalysis review when required because of errors discovered or because of other moderate changes to existing analyses. New models and analyses would require about 3-4 staff years for review. Very often, this reanalysis contri-buted very little to plant safety. The proposed rule would ensure that a com-plate reanalysis will be required whenever changes to the evaluation model result in changes to the calculated peak clad temperature exceeding the original prediction by 250'F. If the new limiting transient calculation exceeds the criteria of 6 50.46(b), then imediate steps must be taken to achieve compliance.
If the criteria of i 50.46(b) are not exceeded, a reanalysis would be performed on a schedule proposed by the licensee and approved by the NRC.
The degree to which the proposed rule would benefit a particular plant depends on how limited the plant is by these LOCA restrictions. The BabcockandWilcox(B&W)andCombustionEngineering(CE)companieshave 8
informally indicated that they do not feel that the plants which they' design are limited by LOCA and, therefore, B&W and CE plants would not benefit from the first two categories. GeneralElectricCo.(GE)plantsdotendtobe limited in operation by LOCA restrictions and would benefit from relief from
.LOCA restrictions. However, this relief is already available for most GE plants through the recently approved SAFER evaluation model. An additional I4 relief due to a rule change would be of little further benefit Westinghouse (W) plants are the only plants which would appear to directly benefit from relaxation of LOCA limits. W plants represent the largest number of plants, however, with 47 plants operating and 10 additional plants being constructed. W indicates that most of these plants are limited by LOCA considerations.(3)
It can, therefore, be estimated that there are at least 47 nuclear plants on line that are limited by LOCA considerations either in total power and/or in flexibility of operation. Up to 10 additional plants may also eventually come on line which will be limited by LOCA considerations. Any rule change that produces a PCT decrease of 100'F can be translated into a total plant power increase of approximately 4% - 6% based on LOCA limit considerations. This number of a 45 - 6% increase in power represents a power increase which is within the capabilities of typical W plants based on existing hardware and is still well below other limits such as DNB limits. Calculations show that the proposed rule change would provide a reduction in PCT of more than 100*F.(11,12,13)
The economic impact of this increase in power can be viewed in terms of energy replacement cost savings.(14) Since Westinghouse plants would be most likely to implement these power upgrades, an analysis has been performed to determine the present values of the energy replacement cost savings which would be derived over the remaining life of each operating Westinghouse plant. The analysis was performed for the 47 plants currently in operation and used the following assumptions:
1.
Replacement energy cost penalties are assumed to be constant in real terms over the remaining useful life of the reactor. This means that 9
costs are not assumed to increase faster than the rate of general inflation.
2.
The comercial operating life of a reactor is assumed to be 30 years.
Thus the remaining useful life of a reactor equals 30 minus the number of years in operation prior to 1986.
3.
All costs will be expressed in 1984 constant dollars and discounted back to 1984 assuming a 10% real discount rate. The final cost estimate will represent a 1984 present worth value in 1984 dollars.
4.
The avera e cost to upgrade equipment in order to increase power is
$150/kWe.5)
Table A. presents the average daily energy replacement cost for each Westinghouse plant, the yearly energy replacement cost savings resulting from a five percent power increase and the present value (1984 dollar) of these cost savings over the remaining life of each plant. The total present value of the energy replacement cost savings for a five percent power increase in all 47 Westinghouse plants is estimated to be $3.2 Billion dollars. As part of a sensitivity analysis, a 5% real discount rate applied to the same plants resulted in a total present value of energy replacement cost savings of 4.9 Billion dollars. Neither of these values include the cost to upgrade plant equipment which is small compared to the present value for most plants.
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TABLE A.
PRESENT VALUE OF ENERGY REPLACEMENT COST SAVINGS DUE TO A 5% POWER INCREASE (Millions of 1984 Dollars)
- Average Daily
- Yearly Energy Present Value Energy Replacement Replacement Cost of Cost Savings Cost Savings Over Plant Life Beaver Valley 1
.272 4.96 42.8 Beaver Valley 2
.272 4.96 46.8 D.C. Cook 1
.289 5.27 44.1 D.C. Cook 2
.300 5.48 47.9 Comanche Peak 1
.710 12.96 120.4 Comanche Peak 2
.710 12.96 122.2 Salem 1
.580 10.59 87.7 Salem 2
.595 10.86 97.4 Braidwood 1
.383 6.99 65.9 Bryon 1
.556 10.15 94.3 Bryon 2
.556 10.15 95.7 Callaway 1
.353 6.44 59.8 Kewaunee 1
.091 1.66 13.6 Point Beach 1
.116 2.12 15.9 Point Beach 2
.116 2.12 15.3 Zion 1
.524 9.56 76.7 Zion 2
.524 9.56 77.5 Prairie Island 1 097 1.77 14.4 Prairie Island 2
.097 1.77 14.7 Ginna 1
.306 5.58 41.1 Haddam Neck 1
.427 7.79 53.1 Indian Point 2
.581 10.60 86.9 Indian Point 3
.650 11.90 101.3 Includes plant specific capacity factors and power.(10,14)
- Yearly savings associated with a 5% power upgrade.
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TABLEA.(cont.)
PRESENT VALUE OF ENERGY REPLACEMENT COSTS DUE TO A 5% POWER INCREASE (Millions of 1984 Dollars)
- Average Daily
- Yearly Energy Present Value Energy Replacement Replacement Cost of Cost Savings -
Cost Savings Over Plant Life Millstone 3
.856 15.62 147.3 Seabrook 1
.856 15.62 147.3 Yankee Rowe 1
.128 2.34 17.8 Farley 1
.353 6.44 55.9 Farley 2
.359 6.55 58.8 Harris 1
.360 6.57 62.0 McGuire 1
.515 9.40 85.4 McGuire 2
.515 9.40 86.7 North Anna 1
.377 6.88 60.1 North Anna 2
.382 6.97 61.7 H.B. Robinson 2
.296 5.40 49.1 Sequoyah 1
.306 5.58 50.1 Sequoyah 2
.306 5.58 50.7 Surry 1
.312 5.69 52.3 Surry 2
.312 5.69 52.5 Turkey Point 3
.430 7.85 72.1 Turkey Point 4
.430 7.85 72.4 V.C. Summer 1
.388 7.08 64.8 Watts Bar 1
.309 5.64 52.4 Wolf Creek 1
.426 7.77 72.2 Diablo Canyon 1
.851 15.5 144.0 Diablo Canyon 2
.836 15.3 142.1 San Onofre 1
.297 5.42 36.1 Trojan 1
.305 5.57 47.4 Includes plant specific capacity factors and power.(10,14)
- Yearly savings associated with a 5% power upgrade.
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Table B. provides the present value cost savings for several different increments of power increase and includes the average cost of upgrading plant equipment. This table assumes a 10% real discount rate.
TABLE B.
T CHANGES IN TOTAL PRESENT VALUE COST SAVINGS DUE TO
_ '[
A PERMANENT POWER INCREASE (Billions of 1984 Dollars) 1 Power
- Total Present Value Plant Upgrade Increase Cost Savings Cost
- Net Cost 1.0 %
-0.64 0.06
-0.58 2.0 %.
-1.28 0.13
-1.15 3.0 %
-1.92 0.19
-1.73 4.0 %
-2.56 0.25
-2.31 5.0 %
-3.20 0.32
-2.88 j
6.0 %
-3.84 0.38
-3.46
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7.0 %
-4.48 0.45
-4.03 i
8.0 %
-5.12 0.51
-4.61 9.0 %
-5.76 0.58
-5.18 10.0 %
-6.40 0.64
-5.76
- Negative values indicate a cost savings.
These potential cost savings represent hypothetical maximum savings which may not be realized.
It is difficult to estimate how many plants would take advantage of such a rule revision and upgrade power. Factors influencing the decision of an individual utility to upgrade a plant would vary and depend upon the need for additional capacity and other means of obtaining additional capacity, other limits such as environmental factors (themal pollution),
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'po,tensiallocal opposition!toplanted!ficationandplantspecific cost-benefit analyses.
The reduced LOCA restriction would result in fu'rther savings due to more
' ' efficient plant; operation. i nespective of wh6ther or not the total power of the plant was upgraded. These improvsentslinclude improvef fuel utilization 4
j, and improved maneuvering capabilities. - Core managuent and advanced fuel management cur.cepts are c licated subjects and LOCA limits are only one of many factors to consider.
Thus obtaining precise estimates of potential savings is. difficult. However, savings n! 3 to 6 million dollars per plant per year would not> be unreasonable.N. Even if a utility did net increase power or change fuel management, simpler generic reload calculatiors pessible with less l
restricifve. LDCA limits would save $Js0,000 per plant per year.II)
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i 1-4.3 Potential' Impacts 1
In contrast to the econmic and safet.y conse_qu.ences described in section l
4.2, there are some detrinwnta,1 Impacts that mi6ht result from the proposed i
rule. Some believe that a%ule change might destabilize the present licensing l
process, which, while,,perhaps overly conservativer,. is well known and
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predictable.. However,'others believe that the li'cansing process will not be I
fully stable untti the, rule is revised: Any disruption of the licensing processmaybecifcum$ntedby"grandfathering";thatwillgiveutilitiesthe optionofhioptingthenewanalENsmethodsorcontinuingwiththeoldAppendix Kproceduhs'.
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Perhaps the single most sihaificant concerr' is thh rotential for reopening l
ECC rulemaking. These original learings lasted for a year and consumed the j
timeandtalt.ntsofmanyengineersandattorneys. The manpower and resources thatwouldbtneededtoconductnewhearingswobdcausemajorproblemsforthe NPO and an industry in which these items are already in short supply. The
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'vendorsfeelthatnew'hearingsarepnacceptableandprefernottomodifythe
- >le rule if new hearings are required.1he potential for a public hearing will be evaluated etter public conronts are received.
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DECISION RATIONALE The various alternatives of Section 3 were considered and evaluated before finally settling on the reccmmended regulatory action. The considerations follow:
Alt. A:
. Retain the existing ECCS rule with its present conservatism (nochange).
PRO:
a.
The current well-established and stable licensing process would be retained, f
b.
No staff resources would be required for rulemaking.
CON:
a.
Many plants would continue to be unnecessarily re-stricted in operation by the current rule.
b.
Many licensees would continue to seek relief from restrictions through requests for exemptions or by using the approach discussed in SECY-83-472. Both these approaches are interim measures and should not substitute for revising the rule to make it consistentwithcurredtknowledgeandpractice.
Alt. B:
Modify the ECCS rule as stated in the advanced notice of proposed rulemaking published in the Federal Register on December 6, 1978; 43 FR 57157.
,P,RO : Consistent with previously-stated plans and would allow minor changes to be quickly implemented.
CON: b.
Substantial changes would be delayed until a later phase.
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b.
Does not resolve comments received on the advanced notice recomending more substantial changes.
Alt. C:
Modify certain models contained in Appendix K, for which research investigations have been completed and a well-documented data base exists. These changes would be selected in areas for which new experimental data has shown that exist-ing models contain a larger degree of conservatism than justi-fled by current data uncertainties or are obviously unrealistic.
PRO:
a.
Plants would no longer be limited in operation by current ECCS rule restrictions.
b.
This would be in agreement with Comission policy (NUREG-0885)toincorporatetheresultsof research into the licensing process.
CON:
a.
The revised ECCS rule might not contain sufficient and quantified conservatism to account for calculational uncertainty. Additional analyses would be required to demonstrate that sufficient conservatism remained in the calculation.
b.
The ECCS rule would have to be changed in the futare to make use of research results or other information which may become available.
Alt. D:
Eliminate the requirement to use Appendix K models and allow realistic models to be used. Reduce the 2200*F and 17%
oxidation limits of 9 50.46 appropriately to ensure that sufficient conservatism exists to cover uncertainties in the realistic calculation.
15
+.,
PRO:
a.
Maximum use of completed research could be made in licensing to relax unnecessary operating restrictions. This would be in agreement with Commission policy (NUREG-0885) to incorporate the results of research into the licensing process.
b.
Licensing models would provide more realistic cal-culations to allow more accurate determination of the effect of equipment changes or failures and operating procedures.
CON:
a.
The conservatism used to account for uncertainties would be fixed through the revised i 50.46 limits and could not be varied to account for more accurate calculations of uncertainty which may be available in the future (i.e., little incentive for furtherimprovement).
b.
Additional staff resources would be required to establish fixed conservatisms applicable to all plant types.
c.
The introduction of a less prescriptive rule would provide a greater opportunity to challenge licensing amendments.
Alt. E:
Eliminate the requirement to use Appendix K models and allow realistic models combined with an evaluation of the uncertainty in the overall calculation included, similar to that discussed in SECY-83-472. The 9 50.46 limits of 2200*F and 17% oxidation would be unchanged.
17
PRO:
a.
Maximum use of completed and future research could be made in licensing to relax unnecessary operating restrictions. This would be in agreement with Commissionpolicy-(NUREG-0885)toincorporatethe results of research into the licensing process.
b.
Licensing models would provide more realistic cal-culations to allow more accurate determination of the effect of equipment changes or failures and operating procedures.
c.
The uncertainty evaluation would quantify the con-servatism in the calculations which could change as the accuracy of the calculations improved.
d.
The industry and NRC staff are already investing effort to follow this approach.
CON: The introduction of a less prescriptive rule would provide a greater opportunity to challenge licensing amendments.
In all alternatives considered, the current Appendix K would remain avail-able for those applicants or licensees not desiring to use a revised
[
eva?uation model.
l The staff believes that Alternatives A and B, which would provide little or no change in the ECCS rule, are unacceptable. The ECCS rule should be changed because:
l (1) A data 5ase now exists that supports relaxation of the ECCS rule.
(2) A revised ECCS rule would remove unnecessary operating restrictions on plants.
I le
~
~ ~ __.
(3) Almost all U.S. research on LOCAs has been completed. The remaining portions of the MIST, Semiscale and 2D/3D programs are expected to provide valuable information for assessment of models, but should not affect the proposed rule.
(4) Nuclear reactor vendors are currently working on future plant designs which would be influenced by the revised ECCS rule.
4 The staff has also considered Alternative C which would modify certain models in Appendix K, for which research investigations are completed.
The revisions considered under Alternative C include:
a.
Reanalysis requirements, b.
Post-CHF heat transfer, c.
Return to nucleate boiling, 1
d.
Refill and reflood transfer (steam cooling below reflood rate of one inchpersecond),
e.
Fission product-decay, f.
Metal-water reaction, and g.
Discharge model.
Based on recent supporting analyses performed by vendors and national laboratories, the staff has determined that if Appendix K were to be revised according to Alternative C, it is possible that the remaining overall conservatism in the evaluation models would be on the same order or less than the uncertainty of the calculation.(15) This would be unacceptable since one could no longer assume that Appendix X contained sufficient conservatism to account for the total uncertainty in the i
19
0; calculation. Thus, use of Alternative C without supporting uncertainty analysis was dropped from consideration.
The staff also considered revising Appendix K in a manner similar to Alternative C. but requiring an additional uncertainty analysis to ensure that.the evaluation model contained sufficient conservatism. This' option would. require two calculations, a realistic calculation with uncertainty analysis and an evaluation model calculation. This option was also dropped from consideration since the licensee would be required to perform two calculations, of which one, the evaluation model, would provide little benefit to safety.
Based on the current understanding of ECCS performance, the approach of a prescriptive Appendix K is no longer believed to be appropriate. A realistic calculation, taking into account the overall uncertainty in the analyses, is believed to be the correct approach to ensure the safety of the public without unnecessarily restricting applicants and licensees.
Thus, the ECCS rule should be revised accordingly and the requirement to use Appendix K eliminated. Therefore, Alternatives D and E were considered, both of which use realistic calcultions. The difference between the alternatives is in the treatment of uncertainties.
I Alternative D would reduce the i 50.46 limits of 2200*F and 17% cladding oxidation to cover uncertainties in the calculation and uncertainties in the point at which substantial core damage would occur. Alternaive E would require an uncertainty factor to be added to the best estimate calculation. Alternative D was not selected because (1) the i 50.46 limits of 2200*F and 17% cladding oxidation are believed to already be appropriate and conservative limits below which substantial core damage will not occur (16) (2) the conservatism used to account for uncertainties in the calculation would be fixed and could not be varied to account for i
more accurate calculations, and (3) further staff analyses would be required to support establishing these limits.
It was decided that Alternative E be adopted. This alternative would require that the licensee show that the criteria of 6 50.46 are met using 20
.,.-_____..,_.._..._,.,m_
,.,_....-.,,c.-
a realistic calculation combined with an evaluation of the uncertainty of-the overall calculation. This uncertainty evaluation, combined with-the additional conservatism in the 2200*F peak clad temperature and the 17% cladding oxidation criteria, should ensure a negligible risk to the public. This approach to licensing is consistent with the interim method discussed in SECY-83-472 except that the additional Appendix K calcula-tion, which contributes >1ittle to safety, would not be required. As discussed in SECY-83-472, the combined conservatisms in Appendix K methods resulted in a calculated peak cladding temperature within 100*F to 200*F of the 2200*F limit. Extensive research has shown that the Appendix K models provide excess conservatisms in the calculated peak cladding temperature and thereby afford greater margin than deemed necessary.
Realistic calculation methods indicate that the peak cladding temperature during a LOCA ranges from 1400 F to 1700*F. The margin of excess i
conservatism in the current Appendix K therefore ranges-from 400 F to 600*F. Thus, increases in the operating linear heat generation rate can beobtained,whilecontinuingtomeetthecriteriaof50.46(b). Further, Appendix K would remain available (with minor modifications) as an alter-native. Therefore, licensees and applicants who do not need nor desire relief from current operating restrictions would have no new requirements and could continue to meet existing Appendix K requirements. The burden of performing new calculations would only be placed on those applicants and licensees who elect to gain relief from LOCA restrictions.
The proposed rule would also provide relief from the reanalysis requirements which do not contribute substantially to safety, as well as 3
allow use of research data that has been obtained since the current rule was written. The modification would allow applicants and licensees relief from unnecessary operational restrictions resulting from loss-of-coolant accident (LOCA) analyses ard still result in an adequate level of conservatisim in the ECCS analyses. The net effect would be to allow increased operational flexibility in the form of increased linear heat generation rates and more optimum fuel utilization while retaining the conservative margins set forth in 50.46(b).
l l
21
6.
IMPLEMENTATION 6.1 Schedule No implementation problems are now anticipated. Each applicant or licensee may, at its discretion, continue to utilize the existing Appendix K criteria for development of the ECCS evaluation model. With regard to the reporting of significant changes to the accepted evaluation models, the proposed rule provides for the establishment of a mutually agreed upon schedule for completing required actions.
6.2 Relationship to the Existing or Proposed Requirements In view of the fact that an integrated schedule is to be used for prescribing necessary actions, it is not expected that actions resulting from other requirements will be seriously affected. Although a backfit analysis is not required by 10 CFR 50.109 because the proposed rule does not require applicants or licensees to make a change but only offers additional options, the factors in 10 CFR 50.109(c) have been analyzed. A backfit analysis, developed primarily from this regulatory analysis has been included in the Notice of Proposed Rulemaking.
22
1..
' REFERENCES 1.
" Revision of Loss-of-Coolant Accident (LOCA) Rule Licensing Requirements -
A Survey of Opinion within the Nuclear Industry," NUS-4221, April 1983.
2.
P. Wei, et al, " Boiling Water Reactor Uranium Utilization Improvement Potential," GEAP-24965 June 1980.
3.
Ltr fm Hochreiter to Beckner, "Results of Westinghouse Calculations Using a Modified Version of Appendix K," April 23, 1984.
4.
Ltr fm Quirk to Beckner, " Impact of Proposed Appendix K Rule Changes on General Electric SAFER /GESTER ECCS Results," July 26,1984(Proprietary).
5.
Ltr Rahe to Ross, "LOCA Margin Benefits," February 8, 1985.
6.
"Probabilistic Risk Assessment (PRA) Reference Document," NUREG-1050, September 1984.
7.
"Effect of Diesel Start Time on BWR/G Peak Cladding Temperature," NSAC-96, January 1986.
8.
" Emergency Core Cooling System Analysis Methods," SECY-83-472, November 13, 1983.
9.
" Compendium of ECCS Research for Realistic LOCA Analysis," Draft NUREG-1230, Chapter IX (to be published).
10.
J. C. VanKuiken, et al, " Replacement Emergency Costs for Nuclear Electricity Generating Units in the United States," NUREG/CR-4012, October 1984.
- 11. Ltr Hochreiter to Beckner, "Results of Westinghouse Calculations Using a Modified Version of Appendix K," April 24, 1984.
23
- 12. "Assessement of Proposed Changes to Appendix K on LOCA Limits," B&W Report 77-1150444-00, April 1984.
- 13. Ltr Scherer to Beckner, "NRC Proposed Changes to 10 CFR 50, Appendix K,"
LD-84-021, May 25, 1984.
14.
S.W. Heaberlin,et al., "A Handbook for Value-Impact Assessment,"
NUREG/CR-3568, Pacific Northwest Laboratory, December 1983.
- 15. Ltr Quirk to Beckner, " Impact of Proposed Appendix K Rule Changes on General Electric SAFER /GESTR ECCS Results," July 26, 1984 (Proprietary).
- 16. Van Houten, " Fuel Rod Failure as a Consequence of Departure from Nucleate Boiling or Dryout," NUREG-0562, June 1979.
1 24
INFORMATION REPORT j
i November 17, 1983 SECY-83-472 For:
The Comissioners From:
William J. Dircks Executive Director for Operations EMERGENCY CORE COOLING SYSTEM ANALYSIS METHODS
Subject:
To inform the Comission of revised analysis methods the
Purpose:
staff proposes to find acceptable for conformance to Appendix K to 10CFR50.
Backoround:
ECCS research conducted over the past 10 years has shed considerable light on the safety margin provided by Appendix K.
Results from major RES-sponsored experimental programs such as LOFT, Semiscale and TLTA (Two-Loop Test Apparatus) have provided an abundance of data to verify the advanced, realistic, ECCS analysis computer codes which have been developed.
Specifically, LOFT tests L2-2 and L2-3 have.been used to demonstrate the ability of advanced computer codes such as RELAP5 and TRAC-BD1 to predict the realistic behavior of a large cold leg break loss of coolant accident in a PWR.
Similarly, eight Two-Loop-Test-Apparatus (TLTA) tests simulating both 7x7 and 8x8 fuel assembly designs in BWR/4 and BWR/6's have been used to demonstrate the ability of both TRAC-BD1 (BWR version) and GE's SAFER code to predict the realistic behavior of large break LOCA's in BWRs.
The Office of Nuclear Regulatory Research is currently preparing a proposed revision to the Appendix K rule.
This proposed revision will account for new information, and will propose changes to certain ECCS model features presently required by Appendix K.
However, this rule CZTACT:
B. W. Sheron, RSB/DSI/NRR X27460 upp.
TheCommissinnprs,( -
revision process is potentially lengthy, and, if a hearing is required, it could take well over a year to implement. Until this rule revision is implemented, the staff. proposes to accept, at the licensees' or applicants' option, revised analysis methods for demonstrating conformance to Appendix K to 10CFR50 for large break LOCAs.
Discussion:
10CFR50.46 specifies the acceptan'ce criteria for Emergency Core Cooling System (ECCS) perfomance.
Included is the requirement that analysis models used to calculate the themal-hydraulic perfomance of the ECCS conform to the requirements specified in Appendix K to 10CFR50. Appendix K to 10CFR50 includes such requirements as the use of the 1971 proposed ANS Standard on decay heat increased by twenty percent, use of the Moody break flow model, and the assumption of the worst single active failure.
In addition, Appendix K identifies other analysis models and correlations, primarily for the large break LOCA, which while not required, are stated as acceptable for use. These required and acceptable features of evaluation models have been used to provide a substantial level of conservatism in large break LOCA ECCS performance -
analyses.
There are two other sources of conservatism in ECCS analyses. One source is imposed by the NRC staff and results from the NRC staff review of ECCS evaluation models.
In many instances, specific models and correlations used in evaluation models (but not specified in Appendix K) may have large uncertainties. To account for these uncertainties, the staff required that the models and correlations conservatively bound the phenomena or the experimental data. A second source of conservatism is imposed by the industry and is derived from the industry's approach to analysis.
Many models, correlations, and initial conditions used in ECCS I
evaluation models were conservatively specified by the industry during their development.
In many instances, the industry chose to conservatively model certain phenomena rather than spend the engineering and financial resources to develop and justify a more realistic but less conservative model.
The combined effect of conservatisms in ECCS perfomance analyses has resulted in calculated large break LOCA peak fuel cladding temperatures (PCTs) approaching (usually within 100*F to 200*F) the 10CFR50.46 cladding l
temperature limit of 2200"F. Originally, this limit restricted the operation of only a few plants.
- However, in more recent years, more and more plants are operationally restricted by the 10CFR50.46 PCT limit combined with the Appendix K requirements.
One reason is f
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- 1he' Commission.ers -
that proposed fuel operating limits optimized for maximum burnup and fuel management considerations (also low neutron leakage) result in higher calculated PCTs.
Another reason is that more and more plants are plugging steam generator (SG) tubes, and this increases the calculated PCT.
In addition, correction of earlier coding errors has sometimes resulted in higher.
calculated PCTs. The cumulative conservatisms from the various sources previously mentioned have caused calculated large break LOCA PCTs to be close to the 2200*F limit of 10CFR50.46. Subseouent changes to ECCS models to account for new information such as SG tube.
plugging can result in the calculated PCTs exceeding the 2200*F limit.
In order to justify continued operation of their plant in such situations, licensees have expended considerable engineering resources proposing and making modifications to their analysis models to offset the PCT increases and bring the calculated PCTs back below the 2200*F limit. This in turn consumes considerable staff resources during the review and approval process to
. assure compliance with the regulations.
For the most part, these efforts have resulted in a negligible impact on actual plant safety, since the outcome of such exercises has usually not involved any signi.ficant change in operational flexibility or any operational limit.
Recent analyses indicate that the most probable ~ PCT,that would be experienced during the limiting large LOCA would be 1000*F to 1200*F for both BWRs and PWRs.
These results have been obtained from advanced computer codes (TRAC and RELAPS), developed independently at two separate national laboratories.
Industry calculations with realistic LOCA computer codes reach the same conclusion; namely that there is approximately a 1000*F to 1200*F margin between the PCT expected during the limiting large break LOCA and the 10CFR50.46 limit of 2200*F. These analytical estimates are now well verified for both classes of reactors by th.e LOFT and TLTA experiments.
"The limiting large LOCA is defined by the combination of break size, location and worst singie failure which results in the highest calculated PCT.
For some plants the limiting break size may be less than the area of a full double ended break of the largest pipe in the
+
However, even for discharge coefficients as small as 0.4, the resultant ecuivalent break area for a PWR cold leg break or BWR recirculation line break falls well within the range of those sizes considered in WASH-1400 to be large breaks.
e
The Commissioriers.'. -
Based on the above, the staff has concluded the following:
- 1. The safety margin in peak cladding temperature provided by current evaluation models to assure compliance with 10CFR50.46 limits is approximately 1000*F to 1200*F for the large break LOCA.
- 2. -This margin is more than. adequate to assure successful ECCS performance in the event of a LOCA.
- 3. This margin can be reduced without adverse effect on plant safety.
- 4. Acceptable reduction in this margin may be warranted to avoid unnecessary restrictions in operation as a result of excessive conservatism imposed in ECCS evaluations.
Proposed Approach:
The staff has established an approach that would permit a reduced margin in large break LOCA ECCS analyses without changing the required features of Appendix K. Such reduced margin can come from careful consideration of alternatives for treatina
~
the acceptable (but not required) features of Appendix K, the conservatisms that have been imposed as a result of NRC reviews, and the conservatisms that were otherwise contained in industry analyses.
In the proposed method, the licensee or licensees would employ "best estimate" models to calculate the PCT, both at the realistic, or most probable (50 percent probability) level and at the more.,,
conservative 95 percent probability level.
When calculating the PCT at the 95 percent probability
. level, other uncertainties, such as the precision with which the code can calculate actual behavior, input or plant parameter uncertainties (such as power level, initial temperatures and pressures) and nuclear parameters not otherwise considered would be accounted for.
Acceptable methodologies for calculating the 95
- This approach is currently only proposed for the large break LOCA analysis.
Many of the features in Appendix K do not apply to Small breaks and the acceptability of this approach for small break LOCAs has not been thoroughly examined at this time.
- This approach could be used on a generic basis by groups of owners, l
such that many similar plants could utilize a single evaluation.
j
"** Appendix A to this paper provides a discussion of the basis for l
I selecting the 95% probability. L'e interpret the 95 percent prcbability level of the peak cladding temperature to mean that in 95 out of 100 LOCAs of a particular size, the peak cladding temperature would be below this value.
-B-percent probability level would have to b9 fonnulstud by the industry and approved by the staff.
It would be expected that the industry would take.into account systematic as well as random errors when establishing probability estimates. Once an acceptable methodology was established, the realistic PCT at the 95 percent probability level would be calculated.
The licensee or licensees would then perform a conventional ECCS-analysis, except-they would be permitted to'use their realistic model augment'ed only'with the required features of Appendix K.. This would constitute a new evaluation model which conforms to Appendix K.
The PCT calculated using this model would then be compared to the realistic PCT at the 95 percent probability level.
If the evaluation model PCT exceeded the PCT at the 95 percent probability level and remained below the 2200*F limit, we would find the evaluation model analysis acceptable.
If the evaluation model' PCT were calculated to be less than the realistic PCT 95 percent probability level, this would indicate that the overall evaluation model, including the required Appendix K conservative features, was not pro-viding sufficient margin in excess of.the estimated statistical uncertainty. Therefore, we would require additional margin be included in the evaluation model such that the calculated Appendix K temperature was in excess of the 95 percent probability level.*
The proposed approach would allow a reduction in the excess conservatism of ECCS analyses with the remaining' conservatism being quantifiable at or above the 95 percent probability level of the realistic PCT. We conclude this level is sufficiently conser.vative based on:
The now extensive experimental data base supporting the LOCA analysis methods 0
The probability of a large break LOCA combined with the probability of a core melt resulting from a LOCA.**
i This approach does not relax any regulations and, in fact, was always available to the staff since the model changes we propose to allow are not prohibited by i-Appendix K.
It also allows for productive use of research results and encourages design improvements.
Appendix B to this paper describes an approach proposed by GE to assure that the Appendix K PCT is always in excess of the 95 percent probability.
"* See Appendix A I
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The Commissitners;
-b-Expected Impact Present estimates of the difference between the 95 percent probability level PCT and the realistically calculated nominal PCT during a LOCA are on the order of 200*F to 400*F.
Adding this uncertainty to a' realistic PCT of approximately 1000*F to 1200*F would result in realistic predictions of PCTs at the 95 percent probability level of about 1200*F to 1600*F.
We also anticipate that realistic analysis codes combined with'the required Appendix K features would predict PCTs on the order of 1400'F to 1700*F using current linear heat generation rates.
This indicates that a maximum of 500*F to 800'F of margin would be available for the industry to use in the form of increased linear heat generation rates, optimized fuel utilization, etc.
Because this approach has not yet been implemented (nor do we intend to make it a requirement), we do not yet have a definite assessment of how the industry will ultimately use the additional margin being made available. However, we do not anticipate that the industry will treat. Appendix K as a " speed limit" and use all of the available margin in the form of increased power levels or increased linear heat generation rates to move the calculated peak cladding temperature back up to the 2200*F limit.
This conclusion is based on a number of reasons:
o A prime objective of the industry is to
{
reduce the resources expended on ECCS model revisions to offset the penalizing effect of new information that occasionally arises.
Licensees and vendors have told us that they will use the margin between their calculated PCT and the 2200*F limit to reduce these resource expenditures.
o One vendor, General Electric, has indicated it would use the margin primarily in obtaining extended fuel cycles, higher burnups, and in general, more efficient fuel management.
o For many plants, if not most, other current fuel performance limits (e.g., departure from nucleate boiling ratio) will restrict the degree to which this margin can be used, Although operating limits are expected to o
increase, overall plant power level increases are not likely to be proposed since most plants are limited in their power output by the installed turbine / generator
- capacity, o PWRs could use the marcin to imolement neutron flux reduction proarams to mininize
ThG CommissiCners,
pressurized thernal shock potential while maintainino the same power levels.
The staff will closely monitor the implementation of this 6pproach by the industry. We will keep the ACRS periodically advised of our experiences regarding implementation, and will report back to the Commission if necessary.
Implementation:
The approach described in this paper is not a requirement.
Present ECCS evaluation models that' comply with Appendix K would still be acceptable for performing ECCS analyses to demonstrate compliance with IDCFR50.46 criteria.
Use of this approach by the industry would be voluntary.
We have had discussions with the General Electric Company and we understand this approach would be beneficial to GE-designed reactors.
GE indicated it would pursue this methodology with the staff.
Westinghouse has also recently approached the staff to explore methods for obtaining relief from Appendix K restrictions. Therefore, we anticipate a positive industry response to this approach.
Concurrence:
The ACRS has been briefed on this proposal and has voiced no objection. The Office of the Executive Legal Director has no legal objection to the use of this approach.
/
Willi m J. Dircks Executive Director for Operations
Enclosures:
1.
Appendix A - Basis for 4
95 Percent Probability i
2.
Appendix B - Proposed GE Model i
APPENDIX A BASIS FOR 95 PERCENT PROBABILITY The establishment of an acceptable probability level is a difficult task.
Because it is statistically based, it is not considered a fixed, deterministic criterion, but ultimately involves, quantifying an acceptable level of risk.
The problem is initially simplified if, instead of establishing an acceptable level of risk, an acceptable probability of core melt is established.
However, since there is no I
universally acceptable probability of core melt, a more reasonable approach is to select a probability level and from it infer that the probability of core melt is acceptably low.
The probability level selected for use with this methodology was 95 percent, or 1.645 (one-side limit) times the standard deviation (1.645e ).
Ninety-five percent was selected for a number of reasons.
Primary was its historical significance in regulatory matters involving themal-hydraulic performance. Many parameters, most notably the departure from nucleate boiling ratio (DNBR) were proposed by the industry and accepted by NRC to be conservatively established at the 95 percent probability level.
Secondary was the fact that in a similar approach previously presented by GE, the 95 percent probability level was proposed and defended.
With a probability of a large pipe rupture of approximately 10-4 per reactor year (as reported in WASH-1400), and assuming the acceptable probability of exceeding the 2200*F limit in 10CFR50.46 is a maximum of 0.05 (corresponding to the 95 percent probability level), the probability of large pipe rupture resulting in a PCT exceeding 2200 F would be
approximately 5x10-6*
It should also be noted that peak cladding temperature analyses are for the hot pins only, which comprise only a few percent of the total number of fuel pins. Therefore, even if the hot pins exceeded 2200*F and fragmented upon quenching, there is a good chance that the resulting damaged core wduld be coolable.
In fact, the TMI-2 core, which apparently lost more than 35% of its fuel cladding integrity, remained coolable.
If it is assumed that 1 out of 10 such damaged cores utlimately resulted in gross core melting, the overall probability of core melt would still be approximately 5x10-7.
This probability is well below many other significant contributors to core melt risk such as loss of all AC power or small break LOCAs with no high pressure injection available.
It would not, therefore, represent a dominant contributor to risk.
- Note that most plants will have Appendix K PCTs below 2200 F, so the probability of exceeding 2200*F for most plants will actually be lower than 5x10-0.
-,.,.._m
APPENDIX B PROPOSED GE MODEL The model proposed by GE was designed to' satisfy' two basic requirements.
.These are:
1.
Satisfy Appendix K Specifications _for ECCS Evaluation Models 2.
Show that the Appendix K licensing analysis results in PCTs that bound the 95 percent probability level of the realistic PCT.
To accomplish this, GE proposes to calculate the Appendix K PCT as follows:
I
) Realistic * ^UU (PCT) Apcendix K (ADDER)2 = [(PCT) Realistic Model + Appendix K - ( CT) Realistic Model Specifications (hPCT):
+
lant variable uncertainties PCT
= (PCT) parameter i perturbed ~(
) nominal from nominal As can be seen, if the plant variable uncertainties are set to zero, the Appendix K calculated PCT reduces to that calculated with the realistic model augmented with only the required features of Appendix K. l l
Therefore, the comb'ination of terms in the ADDER equation should result in PCT that is in excess of a PCT calculated using a realistic model augmented with the required Appendix K features.
We believe the basic methodology presented by GE'is acceptable and meets Appendix K.
The plant variables selected, as well as their associated 95 percent probability level, have not yet been reviewed in detail.
9 e.