ML20212G092

Forwards Documents Re Rev of 10CFR50,App K, ECCS Evaluation Models for Placement in PDR

ML20212G092
Person / Time
Issue date:	09/23/1999
From:	Joseph Donoghue NRC (Affiliation Not Assigned)
To:	NRC
References
NUDOCS 9909290081
Download: ML20212G092 (24)

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4 September 23, 1999 MEMORANDUM TO: Public Document Room FROM:

Joseph Donoghue, Reactor Systems Engineer /s/

Reactor Systems Branch Division of Systems Safety and Analysis

SUBJECT:

DOCUMENTS RELATED TO REVISION OF APPENDIX K, 10 CFR PART 50 Please place the attached documents in the public document room. The documents are referenced in the Federal Register notice for the proposed rule and should be filed together for inquires related to the rulemaking. The following information regarding the rulemaking is provided for your use:

Rule: 10 CFR Part 50, Appendix K, " Emergency Core Cooling System Evaluation Models"

' Rulemaking Identification Number (RIN): 3150- AG26 Contact Joe Donoghue at 415-1131 if you have any questions.

' Attachments:

1.

Regulatory Analysis 2.

Environmental Assessment 3.

Commission Decision on Rulemaking for Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled Nuclear Power Reactors, CLl-73-39, 6 AEC 1085

4. -

Generic Letter 88-16, " Removal of Cycle-Specific Parameter Limits from Technical Specifications," October 4,1988

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- Attachment 1

- Regulatory Analysis C

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I REGULATORY ANALYSIS REVISION OF 10 CFR PART 50, APPENDIX K Alternate Power Level Assumption for ECCS Evaluations 1.

STATEMENT OF THE PROBLEM Part 50, Appendix K, "ECCS Evaluation Models," contains a requirement that safety analyses used for evaluating the performance of the emergency core cooling system (ECCS) under loss-of-coolant accident (LOCA) conditions be conducted at 102 percent of the licensed power for the plant. The provision appears to have been intended to account for uncertainties attributable to instrumentation error. Licensees have proposed using instrumentation that would reduce the uncertainties associated with measurement of reactor power, thus allowing justification of a reduced margin between the licensed power level and the power level assumed for ECCS evaluations. Licensees could use a reduced ECCS analysis margin to facilitate small, cost-beneficialincreases to licensed power. If the uncertainties associated with power measurement instrumentation errors can be shown to be sufficiently small, then the current rule unnecessarily restricts operation. Therefore, the objective of this rulemaking is to allow the reduction of an unnecessarily burdensome regulatory requirement.

A.

B.gr*around A holder of an operating license (i.e., the licensee) for a light-water power reactor is required by regulations issued by the NRC to submit a safety analysis report that contains an evaluation of emergency core cooling system (ECCS) performance under accident conditions. In 6 50.46,

" Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," the Commission requires that ECCS performance under loss-of-coolant accident (LOCA) conditions be evaluated anc Pat the estimated performance satisfy certain criteria.

Licensees may conduct an analysis that " realistically describes the behavior of the reactor system during a LOCA" (often termed a "best-estimate analysis"), or they may develop a model that conforms with the requirements of Appendix K to 10 CFR Part 50. The majority of ECCS evaluations are based on Appendix K requirements. The opening sentence of Appendix K establishes the requirement to conduct ECCS analyses at a specified power level: "It shall be assumed that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level (to allow for such uncertainties as instrumentation error)." Licensees have proposed utilizing instrumentation that would reduce the uncertainties associated with measurement of reactor power, thus allowing justification of a reduced margin between the licensed power level and the power level assumed for ECCS evaluations. The proposed rule would revise this provision in Appendix K, thereby allowing licensees the option of using a value lower than 102 percent of the licensed power in their ECCS analyses.

Several licensees have expressed interest in using updated feedwater flow measurement technology (see Section IV, " Calorimetric Uncertainty and Feedwater Flow Measurement") as a basis for seeking exemptions from the Appendix K power level requirement and to implement power uprates. One licensee, Texas Utilities Electric Company (TUE), has obtained an exemptm from the Appendix K requirement for Comanche Peak Units 1 and 2 and is pursuing an increase in licensed power based on more accurate feedwater flow measurement

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capability. The prospect of additional exemption requests from other licensees provides the

' impetus for changing the rule.

If adopted, the proposed rule would give licensees the option to apply a reduced margin between the licensed power level and the assumed power level for ECCS evaluation, or they could maintain the current margin of 2-percent power. The proposed rule would provide licensees the opportunity to pursue voluntary power uprates without the need to reconsider ECCS evaluations, although the basis for the assumed power for ECCS analysis would change. Some licensees could benefit from the revision without increased licensed power through revisions to their ECCS evaluations at a lower assumed power level.

As presented in this regulatory analysis, the industry could realize a significant financial benefit by relaxation of this requirement. The intent of the rule, to assure margin to ECCS performance in the event of a LOCA, would still be honored and plant risk will not be significantly affected under the proposed rule.' However, the impact of raising the licensed power level for a plant must be evaluated on a plant-specific basis.

B.

Existino Reaulatory Framework Appendix K to 10 CFR Part 50 was written to define conservative analysis assumptions for ECCS performance evaluations during design-basis LOCAs. Large safety margins are provided by conservatively selecting the ECCS performance ciiteria as well as conservatively establishing ECCS calculational requirements. One conservative calculational requirement is to assume that the reactor is operating at 102-percent power when the LOCA occurs. The first section of Appendix K establishes the requirement to conduct ECCS analyses at a specified power level, along with other heat-source assumptions. As stated parenthetically in the rule itself, the power level requirement is impos'ed to account for uncertainties, including instrument error.

The 102-percent power requirement does not appear elsewhere in NRC regulations, but it has been widely applied in guidance dbcuments. The tables that follow list sections of the Standard Review Plan (SRP) (Reference 2) that contain the 102-percent power requirement.

The first table shows SRP sections that incorporate the 102-percent value, but that offer the possibility that a smaller value could be justified. The second table shows those SRP sections that give the 102-percent value without an altamative. The staff intends to review the affected SRP sections and will evaluate the impact of the proposed rule on those safety analyses.

Further, the staff is considering the need for specific guidance to help licensees appropriately account for power measurement uncertainty in safety analyses The only regulatory guide containing the 102-percent power requirement is Regulatory Guide 1 A9," Power Levels of Nuclear Power Plants"(Reference 3).

'NRC reviews of extended power uprates for two boiling water reactors (much greater than 1 percent increases) did not identify significant risk increases. The NRC staff has taken the position that risk evaluations are not expected to accompany applications for marginal licensed power increases (Reference 1).

m-_.-..._..-...._

. SRP Sections Containing the 102-percent Power Margin With an Option SRP Section Title 15.2.6 Loss of Non-emergency AC Power to the Station Auxiliaries 15.2.7 Loss of Normal Feedwater Flow 15.3.1-15.3.2 Loss of Forced Reactor Coolant Flow, including Trip of Pump and Flow

. Controller Malfunctions 15.3.3-15.3.4 Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break 15.4.3 Contml Rod Misoperation (System Malfunction or Operator Error) 15.5.1-15.5.2 Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction That increases Reactor Coolant inventory 15.6.1 Inadvertent Opening of a PWR Pressurizer Relief Valve or a BWR Relief Valve 15.6.5 Loss-of-Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary SRP Sections Specifying the 102-percent Power Requirement SRP Section Title 6.2.1.3 Mass and Energy Release Analysis for Postulated Loss-of Coolant Accidents 6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures 15.1.1-15.1.4 Decrease in Feedwater Temperature, increase in Feedwater Flow, increase in Steam Flow, and inadvertent Opening of a Steam Generator Relief or Safety Valve 15.2.1-15.2.5 Loss of Extemal Load, Turbine Trip, Loss of Condenser Vacuum, Closure of Main Steam isolation Valve (BWR), and Steam Pressure Regulatory Failure

- (Closed) 15.4.6 Chemical and Volume Control System Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant (PWR)

This proposed rule is not part of the proposed effort to revise Part 50 on a risk-informed basis, as described in SECY-98-300 (Reference 4). A risk-informed revision of Appendix K requirements, if undertaken, would involve a broad review of all ECCS analysis requirements and acceptance criteria.

II.

OBJECTIVE OF THE PROPOSED RUL.E The objective of this rulemaking is to remove an unnecessarily burdensome regulatory requirement. Appendix K was issued to ensure an adequate performance margin of the ECCS in the event a design-basis LOCA were to occur. The margin is provided by conservative features and requirements of the evaluation models and by the ECCS performance criteria.

The existing regulation does not require that the power measurement uncertainty be demonstrated, presupposing that the mandated margin is sufficient to account for uncertainties expected to be involved with measuring reactor power. By allowing a smaller margin for power measurement uncertainty, the proposed rule would not violate the underlying purpose of Appendix K.

A secondary objective is to avoid unnecessary exemption requests. The staff has previously

' sought rule changes to avoid the prospect of multiple exemption requests. In SECY-96-147 j

(Reference 5), the staff took steps to revise regulations that were associated with large

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-4 numbers of recurring exemption requests. In the cases addressed in SECY-96-147, the rules were being changed as a result of recurrent exemptions, which indicated an inadequacy in a regulation. In the case of this proposed change to Appendix K, the staff is anticipating j

recurrent exemptions and has determined that revising the rule at this early stage will be the 1

best course.

An economic benefit is a strong consideration for licensees. The economic benefit of an increase in licensed power can be considered in terms of replacement energy cost savings for utilities that no longer need to purchase the additional power generated as a result of a power uprate. Of course, plant-specific features and situations change the estimated benefit for any given plant either more or less favorably. Factors influencing the decision of a utility to upgrade a plant vary, and a plant-specific cost-benefit analysis would be required to determine whether a specific facility should pursue the uprate.

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Under the proposed rule, some licensees could realize savings without seeking a power

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urrate. By revising their ECCS analysis based on a lower assumed power level, licensees r uld gain margin that could lead to less stringent requirements for LOCA mitigation system v., ECCS) performance or in core thermal limits.

s Ill.

ALTERNATIVE APPROACHES

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Rulemakina Ootions The staff considered the following options:

No Rule Chance Instead of instituting a rule change, the regulation could be maintained in its current form and multiple exemptions to the existing regulation could be granted under 10 CFR 50.12.

A short term benefit to this approach would be that the NRC would avoid the costs of changing the rule and of implementing the revision. However, in the long term, this is not a satisfactory altamative from the standpoint of regulatory efficiency.

Each exemption request would need to be reviewed in accordance with the criteria of 10 CFR 50.12 in addition to reviewing its technical merits. The exemption request review would be handled as a separate regulatory step from the review of a power uprate request for each application, as is the case with the pending exemption request for Comanche Peak Units 1 and 2. Applying this process to a series of exemption requests would be an unnecessary expenditure of NRC and licensee resources, an expenditure not encountered under an amendment to a rule.

j Option 1 Maintain the provision requiring an analysis margin to account for uncertainty in power measurement, but remove the specification of the 2-percent value for the margin.

Licensees would then need to propose and justify the value used for their analysis.

This option is not preferred because it would not meet backfit criteria. Although it could provide relief to licensees that seek to reduce the margin, it would constitute a backfit on

. those licensees that would not wish a change from the currently required value but would nonetheless be required to justify a value. Because the proposed change is expected to have negligible risk impact, there is no basis for a compliance or adequate protection backfit for this option.

Option 2 Allow licensees the option to justify a smaller margin between licensed power and the assumed power level for ECCS analysis for their plant or to maintain the current margin now mandated.

This is the preferred option. Making this change to the rule would give licensees the opportunity to benefit from a reduced margin by demonstrating that power measurement uncertainty is sufficiently small. Licensees would if there is a sufficient benefit relative to the effort to justify the change in a license amendment request. Licensees could gain benefits from operation at higher power or relax ECCS-related technical specifications. In such cases, licensees would need to justify the reduced power measurement uncertainty as part of the license amendment request. Other licensees may elect to revise the ECCS analyses for their facility and seek benefits without increasing licensed power. Maintaining the current Appendix K requirements is not adverse to safety and should be permitted as an option.

Ootion 3 Eliminate the requirement for a margin between power level and assumed power.

This option is not preferred. The staff would need to investigate the feasibility of eliminating the requirement for an assumed power margin for analysis. Without a required analysis margin, licensees could seek benefits without addressing power measurement uncertainties. Justification for this option would involve demonstrating the acceptability of not accounting for any uncertainties behind the 2-percent power analysis margin. The technical effort involved in this option is probably not justifiable since a generic demonstration of the safety implications would be more costly than for option 2, and there is no safety benefit relative to option 2.

Ootion 4 Broadly revise Appendix K, addressing several conservative requirements.

The staff considered addressing several of the calculational requirements in Appendix K with the objective of reducing excessive conservatism. This would be a long-term effort, which, if pursued, would not avoid the exemption requests expected in the shorter term.

Further, given the existing option in 10 CFR 50.46 for licensees to apply best-estimate methodology to avoid Appendix K conservatism, and the substantial staff resource effort entai!ed in a broad Appendix K revision, the staff decided that this was not a preferred option.

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. IV. -

. EVALUATION OF VALUES AND IMPACTS Since the proposed rule does not in itself change any plant configurations or operating parameters, the staff evaluated likely benefits that licensees would seek to achieve from the i

revised rule. Those licensees electing to use the option afforded by the proposed rule to pursue licensed power level increases for their plants are likely to realize the la gest financial benefits as a result of the proposed rule. Therefore, the evaluation that follows emphasizes the costs and savings associated with a small (i.e., approximately 1 percent) increase in licensed power. However, as discussed in the Decision Rationale section, there are only slight differences between the costs and benefits associated with the options evaluated by the staff.

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Therefore, the main decision criteria became regulatory efficiency impacts of a large number of exemption requests that would be faced without a rulemaking and the desire to complete a timely rulemaking.

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. In conducting the evaluation, the staff followed the " Regulatory Analysis Guidelines of the U.S.

I Nuclear Regulatory Commission," NUREG/BR-0058, Revision 2 (Reference 6), including the use of a 7-percent discount rate to adjust values to 199g dollar values. First, benefits and costs are identified for the proposed rule, then the overall effect is evaluated for each of the rulemaking attematives considered. The values and impacts associated with rulemaking option 1 were not evaluated, because it was eliminated from staff consideration in view of backfit ramifications. Therefore, the evaluation that follows covers options 2 through 4

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compared to the no-rulemaking attemative. The staff considered the no-rule-change attemative as the base case. In the event the rule was not revised, numerous exemption requests are anticipated that would be similar to the exemption already approved for Comanche Peak.

Previously, the staff concluded that marginal power increases have little risk significance (see l

' Reference 1). Therefore, the staff considered value impact attributes related to health effects and property loss resulting from accidents to be unchanged by the proposed rule. Also, the financial benefits under each option evaluated are equivalent. As a result, the attributes contributing to the final selection of a rulemaking altemative are limited to regulatory efficiency implications. When data were readily available, the staff made quantitative approximations for the factors. However, the evaluation was eventually qualitative, since the benefit of regulatory efficiency maintained by avoiding large numbers of exemptions is difficult to quantify.

A. Values

1. Savings to Ucensees i

Licensees who want to use the option offered by the proposed rule could realize a significant economic benefit from an increase in licensed power. The benefit realized by a particular licensee will be influenced by a number of factors, including the market price of electricity, generating costs, and the mix of generating assets within the utility (i.e., types of units: nuclear, fossil, etc.). The staff estimated licensee savings under two sets of assumptions: replacement power cost savings and generation cost savings.

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_ a. Replacement Power Cost Savings On a purely replacement-power-cost-savings basis, the staff assumed that demand for electricity will increase such that any increase in generation by nuclear units will be purchased.

Naturally, the validity of this assumption could be affected by market factors and the particular

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situation of the utility considered. However, based on the average annualincrease in utility electnc production from 1990 to 1998 for all sources (about 1.7 percent - see Reference 7) and a generally greater annualincrease by nuclear units, use of added nuclear generating capacity of 1 percent appears to be a reasonable assumption. The licensee's benefit is considered on an average-plant basis using 1998 data from Reference 7. The retail price of electricity sold by electric utilities during 1998 averaged 6.74 cents per kilowatt-hour. Using the total amount of electricity produced in 1998 by nuclear generation,674 billion KWH (reflects an industry-wide capacity factor of 70 percent for 103 operating units) and assuming a typical power i

increase of 1 percent to be achievable from the proposed rule, the annual increase in electrical output for a single unit would be about 65.4 million KWH. Using these values, a unit could save about $4.4 million annually in replacement power costs, or $453 million for all operating units. However, increased power generation incurs some additional costs for the utility. The generating cost for nuclear power units during 1995 (Reference 8) was $19.23 per MWH (this value includes fuel, operation, and maintenance costs). For the average plant being considered, the increased generation would add about $1.7 million in annual costs (adjusted to 1999 value). Therefore, the not benefit for the average unit would be the difference between the replacement power savings and this additional generation cost, or $2.7 million. Over the average remaining lifetime of a U.S. nuclear power plant (about 17 years), the savings would be approximately $26 million (in 1999 dollars). The average lifetime does not account for expected license renewals.

b. Generation Cost Savings i

This estimate assumes that a utility would use the increase in nuclear generating capacity gained from a 1-percent power uprate to reduce the amount of power generated by units that are more costly to operate. No benefit from the sale of additional power is included in this scenario because the utility is assumed to sell the same overall amount of electricity after the nuclear unit power uprate. Comparison of power generating costs in Reference 8 shows that gas and oil-fueled units had higher generating costs than nuclear units, while coal-fueled units 1

had the lowest costs. It is reasonable to assume that a utility with units fueled by various means would use increased nuclear generating capacity to reduce more costly means of i

generation. Therefore, the staff assumed that a utility would apply the increased capacity of i

the average nuclear unit considered above to reduce the power generation by gas and oil-fueled units. The staff assumed that the reduction would be split evenly between the two

. types of units. Applying generation cost data from Reference 8, power generation costs from gas and oil-fueled units would decrease by about $2.7 million, which is offset by the increase

- in costs of power generated by the nuclear unit of $1.7 million, yielding a net savings for the utility of about $1 mWoon annually. Over the average lifetime of a U.S. nuclear power plant j

(17 years), the savings would amount to $9.8 million (discounted to 1999).

These scenarios represent a range of the benefits that licensees may expect if they choose to pursue the power uprate afforded by the proposed rule. A variety of factors could change the results for any particular utility, but the staff expects that for those licensees in a position to

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m pursue power uprate, the resuits would fall in the range between the two scenarios considered above, or between $1 million and $2.7 million annually.

The magnitude of the benefit from a license change not involving power uprate and the manner by which it would be applied are subject to plant-specific considerations. Licensees may decide to seek a change in technical specifications for ECCS systems based on revised analyses, rather than to increase liansed power. In other cases, licensees might pursue benefits by altering core performance characteristics based on the revised ECCS evaluation.

There is a wide range of possible scenarios and such savings would probably only add slightly to the industry savings realized from eventual power uprates. Therefore, the staff did not attempt to quantify the~ savings for plants that might make changes to their ECCS evaluations but would not seek power uprates.

2. Savings to NRC The monetary savings realized by the NRC through rulemaking are expected to be modest, in that they lie only in the difference between processing license amendments for power uprates or other license changes associated with the revised rule and processing exemption requests along with similar license amendment requests. The costs of processing amendment requests and exemptions are discussed later.

There is also a benefit from improved regulatory efficiency, because multiple exemption requests need not be considered under the proposed rule.

B. Imosets

1. Costs to Licensees Licensees electing to pursue the benefit offered by the proposed rule would incur costs of upgrading plant instrumentation that provide the basis for the improved accuracy in power measurement. There are also several costs incurred by those licensees seeking a benefit from the proposed rule. These include the resource investment to conduct analyses to support a license amendment request, whether it is a power uprate or other technical specification change, and costs associated with submitting the license amendment to the NRC. Finally, there are costs incurred to implement the changes to the plant to allow operation at higher Power.

For this evaluation the staff assumed that the acquisition and installation costs for an ultrasonic flowmeter or for other changes that licensees could make to improve the accuracy of thermal power measurement would be part of the overall power uprate cost. Costs of analyses to support a power uprate amendment request would be approximately $5 million, based on effort claimed by industry to support other power uprate requests (Reference 9). Some of these expenses could decrease as future applicants will realize efficiencies based on experience gained by earlier applicants. The staff also considered approximate values for both licensee and NRC costs that are available from NUREG/CR-4627 (Reference 10), which

- presents a cost estimate for a " complicated" technical specification change. For this assessment, the staff assumed that the analysis and submittal to justify a smaller assumed power margin incur costs equivalent to such a " complicated" amendment. Making adjustments i

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j for the period since 1988 when NUREG/C.R-4627 was published, the licensee's cost to justify a smaller assumed power margin could be about $75,000. Thus, using these estimates, each licensee would expend at least $75,000 to use a reduced analysis margin, and those licensees seeking the power uprate would incur costs of about $5 million.

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The staff estimates the licensee's cost of plant modifications to accommodate a small power i

uprate to be in the range of $5 million to $10 million, which accounts for hardware, procedural changes, and personnel training costs. This estimate is based on licensee power uprate cost

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estimates ranging from $150/KWe to $2250/KWe (Reference 11).2 The staff used the higher j

cost information in the analysis to ensure that licensee costs would not be underestimated.

2. Costs to NRC NRC realizes costs under any of the scenarios considered in this evaluation. The costs for 3

review and processing of license amendments or exemptions, as well as revisions to guidance i

documents and rulemaking costs themselves, are considered next.

NRC licensing action costs are based on dollar values, rather than on staff full-time-equivalent positions, given in NUREG/BR-0184 (Reference 12) for the expected NRC staff effort to implement new requirements and on a so-called complicated technical specification amendment review discussed eariier. NUREG/CR-4627 estimates that such a review would entail an NRC cost of $42,000, adjusted to present value. Assuming that the NRC cost to review the proposed assumed power margin reduction is comparable to that required for a power uprate amendment, the cost for each would be in the range of $42,000. Thus, the NRC would incur a cost of $42,000 for each proposed margin reduction, and an additional $42,000 to process each request for a power uprate.

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NRC costs to revise the rule and update review guidance were estimated by the staff. The rulemaking costs vary depending on the scope of the rule revision considered. Revision of associated guidance documents is estimated to be a one-time cost of about 0.4 FTE, or about

$54,000. To supplement the generic information discussed above, the staff also surveyed the NRC staff resources used for relatively recent licensing actions that might be representative of staff activity associated with the proposed rule, such as exemption requests and similar power uprate requests. This survey formed the basis of the staff's assumption that an Appendix K exemption request would require about 7 weeks of staff effort, valued at approximately

$21,000 (assuming $75 per hour for staff effort).

Savings might be realized as more exemption requests are approved and if generic submittals were made to address those facilities of similar design; however, the staff would need to i

ensure that plant specific features for certain facilities did not invalidate the generic assessment. Thus, some review would still be needed for each request.

l 2 The cost values from Reference 3 are in 1985 dollars. The total cost of $5 million to $10 million given here is a current value.

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. C. Health. Safety. and Environmental Effects j

in the Appendix K exemption recently approved for Comanche Peak, the instrumentation manufacturer (Caldon, Inc.) claims that a safety benefit will be achieved by using the i

instrument even during operation at a higher power level. The vendor quantified the benefit in terms of the probability sat the power level of the plant will exceed the licensed level at the initiation of the accident. Although the staff does not dispute the claim of a safety benefit, the oyerall safety impact of an increase in licensed power depends on a variety of plant-specific factors.

A slightly higher power level (i.e., about 1 percent) will result in a slight increase in decay heat load, but is not expected to affect the success criteria and required response time of ECCS i

equipment and the available operator response time following transients and accidents. In NUREG-1230 (Reference 13), the staff considered the risk impact of changes associated with the revised ECCS rules, including power increases, and determined that a power change of

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5 percent or less had little risk significance.

In Reference 1, the staff discussed its consideration of the risk impact from BWR extended power uprates, which are much greater than the marginal power change expected under the J

proposed rule, in these cases, the staff concludes that extended power uprates are expected to only slightly affect the risk profile of a plant. In Reference 1, the staff judged that marginal power uprates, of about 1 percent, were not expected to require an assessment of the risk impact on the plant. However, licensees requesting increased licensed power must demonstrate on a plant-specific basis that deterministic requirements are satisfied (e.g., those based on the general design criteria of Appendix A to Part 50).

D. Comparison of Altematives The operating reactor population used for this assessment was 103 units as of December

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1998. An assumption common to each option considered is that those licensees wanting to 1

pursue power uprate afforded by a rule revision would do so shortly after issuance of the final rule. Assuming the proposed rule is issued in final form during 2000, the average remaining plant lifetime is approximately 17 years, not accounting for expected license extensions, Not alllicensees are expected to seek a power uprate under the proposed rule. As described earlier, some would seek only to revise the ECCS analyses for their facility. For the purpose of this evaluation, the costs and benefits for these licensees are not considered, because a large range of options is involved and because the staff found that the proposed rule was justified by limiting the benefits to those plants seeking power uprate. For this evaluation, the staff assumed an approximately even split of the nuclear plant population between these two categories of 50 plants whose licensees sought a power uprate, and 53 plants whose licensees were not seeking a power uprate. If only 50 plant licensees pursue a marginal power uprate, they would share an annual benefit ranging from $50 million to $135 million, based on the two scenarios considered earlier.

e The table entitled, " Cost Estimates for Rulemaking Options," located at the end of this section, 1

summarizes the staff's cost estimates used in its comparison of the alternatives. For each attemative, the staff assumed that the costs applied to 50 plants, as indicated in the table.

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Note that the high estimate for licensee costs for power uprate is used in the table and that the NRC costs comprise salaries, benefits, and contract support.

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1. No-Rule-Change Altemative if the cunent requirement remains in place and no rule change is permitted, the staff expects that a significant number of licensees will pursue exemption requests, following the example of Comanche Peak. Licensees for at least 19 plants have expressed their interest to NRC in the staff review of the Caldon, Inc. ultrasonic system. It is not clear how many of these licensees, or if others, would eventually pursue exemption requests. The staff assumed that if licensees determined that the relaxation had financial benefits for them, then those licensees would seek the benefit whether or not the rule were revised. The licensee costs to support exemption and amendment requests were discussed earlier. The staff used values of $75,000 and $5 million, respectively. Also, costs for implementing the power increase total about $10 million. The typical NRC cost to review exemption requests were discussed earlier and are estimated to be about $21,000 per request. Added to this cost is the NRC cost to review the justification for i

the reduced power level margin of $42,000. The licensee and NRC costs associated with the power uprate would be the same as those considered for the 1-percent power increase assumed for options 2 and 3, about $42,000.-

2. Option 2 Under option 2, the change is not mandatory. Therefore, each licensee would first determine whether an investment to reduce the analysis margin is justified in light of the potential benefits. Ucensees opting to obtain a power uprate or other license amendment must conduct an analysis to justify a reduced assumed power margin, and then prepare license amendments to obtain a power uprate or technical specification change.

The costs for these activities were discussed earlier and are considered the same as in the no-rule-change case, although some savings may be expected because an exemption request is not involved. The NRC would incur a cost of $42,000 for each proposed margin reduction, and an additional $42,000 to process each request for a power uprate. The staff estimated that the rulemaking effort for opbon 2 would require 0.9 FTE or about $122,000. Once the current regulation is changed, any NRC SRP sections and regulatory guides that use the currently required value for assumed power margin would have to be revised to remain consistent with the regulations. These costs totaling $54,000 were discussed earlier. The staff estimated the

- total NRC cost for generic activity under this option to be $176,000.

3. Option 3 Onder this option, as in option 2, those licensees seeking a higher licensed power level, or other benefit, would incur costs. The costs would be associated with revising plant technical specifications and conducting those analyses necessary to amend the license to operate at a higher powerlevel. These costs are the same as those considered in option 2.

Under this option, the NRC assumes a much greater burden in that the rulemaking to eliminate a requirement, versus its modification, would be expected to entail a t,ignificant amount of technical and administrative effort compared to option 2. For instance, the NRC staff would probably use contractor assistance to help develop the technical basis for the revised rule. A protracted review of the revision would be expected and would entail significant staff costs.

NRC costs are estimated on the basis of the previous value for the staff review of a licensing-basis revision, or about $84,000 for each licensee submittal, and a one-time NRC cost of

$1.5 million assumed for the staff analysis of the generic issues involved and rulemaking to eliminate the requirement. This cost would be divided between staff effort and contractor services, as appropriate.

4. Option 4 Under this option, the staff would revise several parts of Appendix K, and some plants could then decide to seek higher licensed power !evels under the revision. Because a more far-reaching rule change would reduce conservatisms by more than just a change to the power measurement conservatism, a greater potential benefit should be expected. Thus, for this option, the staff assumed that plants might realize a 5-percent power uprate if Appendix K were broadly revised. The licensee costs involved with such a power uprate for a facility'could be expected to be somewhat more than costs assumed for the 1-percent change. The staff assumed that the costs to support and then implement such a change would roughly double, to about $10 million and $20 million per plant, respectively. The NRC rulem king and review costs for this option are more difficult to estimate, but an increase to about $5 million for a multi-year rulemaking effort requiring extensive technical support is reasonable. The NRC cost to review each more extensive amendment would also roughly double to about $100,000.

Thus, each licensee's cost would total about $27 million, and the NRC would incur costs of as high as $10 million for the overall effort involved. This option would also take much longer to implement than the others.

E. Backfit Considerations The NRC has determined that the backfit rule in 10 CFR 50.109 does not apply to this proposed regulation and that a backfit analysis is not required for this proposed regulation because the proposed rule does not involve any provisions that would impose backfits as defined in 10 CFR 50.109(a)(1). This proposed rule would amend the NRC's regulations by establishing an altemate requirement that licensees may voluntarily adopt.

F. Impacts on Other Proorams. Other Aconcies The only potential impact the staff foresees is that further changes to Appendix K could resu4 from the proposed risk-informed review of 10 CFR Part 50, discussed in SECY-98-300.

Cost Estimates for Rulemaking Options (1999 dollars) l NRC Costs Ucensee Costs (per plant)

Ucensee (per plant)

(generic)

NRC No. of Total Total OVERALL OptioM Rants

('

I* h MM Request Request Effect Process Process Rule &

opt on

,pt on)

Margin Power Power Margin Power Guide Change Uprste Uprate Change Uprate Changes No ' Rule 50

$75K8

$5M

$10M

$754M

$63K8

$42K

$5.3M

$759M Change 2

50

$75K

$5M

$10M

$754M

$42K

$42K

$176K

$4.4M

$758M 3

50 575K

$5M

$10M

$754M

$42K

$42K

$1.5M

$5.7M

$760M 4

50

$10M

$20M

$1.5B

$100K

$5M

$10M

$1.51B Notes:

1. Options 2 and 3 consider a 1-percent power uprate; option 4 involves a 5-percent power uprate.

Option 1 was not considered in the value-impact analysis.

2. Costs of preparing / reviewing the exemption request are included.

V.

DECISION RATIONALE The safety impact of options 2 and 3 is essentially equivalent to the ba.,eline, or no-rule-change attemative, because licensees for 50 plants are expected to submit exemption requests for the relief offered by the proposed rule, if it is not issued. The staff has previously oetermined that there is negligible risk impact from a marginal increase in licensed power, therefore, public health and safety and common uefense and security continue to be l

adequately protected. Therefore, the staff considered value impact attributes related to health l

effects and property loss resulting from accidents to be unchanged by the proposed rule.

Cost and benefit estimates are summarized in the table that follows. Differences in overall costs between options 2 and 3 and the no-rule-change attemative are small, and these values l

should be assumed equivalent. Also, the financial benefits under each option evaluated are l

equivalent. As a result, the attributes contributing to the final selection of a rulemaking l

attemative are limited to regulatory efficiency implications. When data were readily available, the staff made quantitative approximations for the factors. However, the evaluation was eventually qualitative, since the benefit of regulatory efficiency maintained by avoiding large numbers of exemptions is difficult to quantify.

The preferred rulemaking attemative is opbon 2. The no-rule-change attemative could not be eliminated on the sole basis of overall cost considerations. The staff then considered NRC precedent for revising rules to eliminate or avoid excessive numbers of exemption requests as a basis for narrowing the choices to options 2,3, and 4. Although a broad revision to Appendix K (option 4) could provide greater relief from ECCS analysis requirements (benefits are assumed to increase proportionally compared to the 1-percent power increase), such a change could not be completed in the short term. The NRC is currently prioritizing such a

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\\ revision along with other changes expected to be pursued to revise Part 50 on a risk-informed j

basis. Option 3 would take longer to implement than option 2 because of the more involved technicaljustificer >n that would be required, as discussed earlier, Also, the NRC costs are expected to be (imewhat greater for option 3. The anticipated benefits of the two remaining i

options are the came.

Decision Rationale Summary No. of Cost Benefit Option Plants Ucensees NRC Total Annual Ufetime No Rule Change 50 8754M

$5.3M

$759M

$50M-135M

$488M-51.3B 2

50 5754M

$4.3M 3758M

$50M-135M

$488M-51.3B l

3 50 5754M

$5.7M

$760M

$50M-135M

$488M-$1.3B 4

50

$1.5B

$10M 51.51B S250M-475M

$2.4B-6.6B I

i The industry has expressed its intention of submitting numerous requests for exemption from Appendix K to ease the assumed power level requirement. The exemption requests could be q

avoided or minim! zed by an expeditious rulemaking. In the interest of regulatory efficiency, the staff has proposed to revise the rule now, rather than proposing more involved action that will take much longer to implement. The simple revision contained in option 2, the proposed attemative, eliminates an unnecessary regulatory burden with little potential for adverse risk j

impact, and can be achieved relatively quickly.

VI.

IMPLEMENTATION The proposed rule will be issued for public comment. Following review of public comments and incorporation of any changes to the proposed rule, it will be issued in its final form and should be made effective 60 days following issuance.

Tentative Schedule:

Proposed Rule issued for Public Comment September 1999 End of Public Comment Period December 1999 Final Rule issued April 2000 Vll.

REFERENCES 1.

U.S. Nuclear Regulatory Commission, Letter from EDO to ACRS," Staff Response to ACRS Letter of July 24,1998 on General Electric Nuclear Energy Extended Power

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. Uprate Program and Monticello Nuclear Generating Plant Extended Power Level increase Request," September 14,1998.

2.

b.S. Nuclear Regulatory Commission, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, Washington, D.C.,

July 1981.

I 3.

- U.S. Nuclear Regulatory Commission, " Power Levels of Nuclear Power Plants,"

Regulatory Guide 1.49, Revision 1, Washington, D.C., December 1973.

4.

U.S. Nuclear Regulatory Commission, " Options for Risk-informed Revisions to 10 CFR Part 50 ' Domestic Licensing of Production and Utilization Facilities'," SECY-98-300, Washington, D.C., December 23,1998.

5.

U.S. Nuclear Regulatory Commission, " Planning for Pursuing Regulatory improvement in the Area of Exemptions Granted to Regulations," SECY-96-147, Washington, D.C.,

July 1,1996.

6.

U.S. Nuclear Regulatory Commission, " Regulatory Analysis Guidelines of the U.S.

Nuclear Regulatory Commission," NUREG/BR-0058, Revision 2, Washington, D.C.,

November 1995.

7.

U.S. Department of Energy, Energy Information Agency, " Electric Power Monthly,"

DOE /EIA-0226(99/04), Washington, D.C., Apri "

'evailable at www.ela. dos. gov).

8.

Utility Data Institute,"1995 Production Costs-Operating Steam-Electric Plants," UDI-201196, Washington, D.C., September 1996.

9.

" Plant Uprates Seen as Cheap Way to Meet Competitive Pressures," Nucleonics Week, Vol. 36, No.38, September 21,1995.

10.

U.S. Nuclear Regulatory Commission, " Generic Cost Estimates," NUREG/CR-4627, Rev. 2, Washington, D.C., February 1992.

11.

Westinghouse Corporation, Letter from E.P. Rahe, to Dr. D.F. Ross, NRC, "LOCA Margin Benefits," February 8,1985.

12.

U.S. Nuclear Regulatory Commission, " Regulatory Analysis Technical Evaluation Handbook," NUREG/BR-0184, Washington, D.C., January 1997, 13.

U.S. Nuclear Regulatory Commission, " Compendium of ECCS Research for Realistic LOCA Analysis," NUREG 1230, Washington, D.C., December 1988.

Environmental Assessment 1

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i ENVIRONMENTAL ASSESSMENT

]

REVISION OF 10 CFR PART 50, APPENDIX K This document examines the environmental impacts of its regulatory actions in accordance with 10 CFR Part 51, for a rulemaking addressing NRC's current emergency core cooling systems (ECCS) evaluation requirements for nuclear power reactors. NRC is proposing to modify these requirements, which are contained in Appendix K to 10 CFR Part

50. The proposed rule would provide a voluntary option for licensees to apply a reduced margin between the licensed power level and the assumed power level for ECCS evaluation.

The currently required analysis margin is 2 percent of licensed reactor power.

NRC's regulations for implementing Section 102(2) of the National Environmental Policy Act of 1969 (NEPA), as amended, are contained in Subpart A of 10 CFR Part 51.

These regulations require that an environmental impact statement or an environmental assessment be prepared for all licensing and regulatory actions that are not classified as

" categorical exclusions" in accordance with 10 CFR 51.22(c) and are not identified in j

10 CFR 51.22(d) as other actions not requiring environmental review.

This document presents the findings of NRC's environmental assessment of the proposed rule.

Identification of the Proposed Action A holder of an operating license (i.e., the licensee) for a light-water power reactor is required by regulations issued by the NRC to submit a safety analysis report that contains an evaluation of ECCS performance under accident conditions. Section 50.46," Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," requires that ECCS performance under loss-of-coolant accident (LOCA) conditions be evaluated and that the estimated performance satisfy certain criteria. Licensees may conduct an analysis that " realistically describes the behavior of the reactor system during a LOCA" (often termed a "best-astimate analysis"), or they may develop a model that conforms with the required and acceptable features of Appendix K to 10 CFR Part 50. Most ECCS evaluations are based on Appendix K requirements. The opening sentence of Appendix K establishes the requirement to conduct ECCS analyses at a specified power level: "It shall be assumed that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level (to allow for such uncertainties as instrumentation error)."

The proposed rule would give licensees the option to apply a reduced margin between the licensed power level and the assumed power level for ECCS evaluation. The current mr.rgin of 2 percent power may be maintained, if preferred. If licensees can show that the uncertainties associated with power measurement instrumentation errors are less than 2 percent, and a smaller margin can be justified, then the current rule unnecessarily restricts operation of some facilities by limiting their ability to operate at higher power levels, and.in other cases by imposing unnecessary requirements on ECCS performance.

Appendix K to 10 CFR Part 50 was written to define conservative analysis assumptions for ECCS performance evaluations during design-basis LOCAs. Large margins for important

\\

I safety parameters were provided by conservatively selecting the ECCS performance criteria as well as conservatively establishing ECCS calculational requirements. The staff has long recognized that Appendix K incorporated substantial conservatisms and previously had considered methods that would acceptably reduce safety margins. The conservatisms were necessary when the rule was written because of a lack of experimental evidence at that time.

When the NRC adopted changes to 10 CFP. 50.46 to allow *best-estimate" modeling, it concluded that experimental evidence gained since the original rule was implemented and analysis advances allowed the consideration of altamative approaches, in the proposed rule, the staff is extending the application of its understanding of ECCS evaluation conservatism to

' allow relaxation of one of several conservative analysis features.

The current analytical approach of assuming 102 percent of licensed power for ECCS evaluation is adequate to protect public health and safety; therefore, the NRC does not intend to backfit a change to the regulation on operating reactors. Because the proposed revision would not constitute a backfit, the bases for current ECCS evaluations must be preserved.

Therefore, the revision will retain the current requirement as an option for licensees.

Need for the Proposed Action The objective of this rulemaking is to allow the voluntary relaxation of an unnecessarily burdensome regulatory requirement. Appendix K was issued to ensure an adequate performance margin of the ECCS in the event a design-basis LOCA were to occur. The margin is provided by conservative features and requirements of the evaluation models and by the ECCS performance criteria. By allowing a smaller margin for power measurement wwrtainty, the proposed rule does not undermine the underlying purpose of Appendix K.

A secondary objective is to avoid unnecessary exemption requests. The staff has previously sought rule changes to avoid the prospect of multiple exemption requests. In the case of this proposed change to Appendix K, the staff is anticipating recurrent exemptions and has determined that revising the rule at this early stage will be the best course.

EnvironmentalImpacts of the Proposed Action The proposed rule would affect an analysis assumption for ECCS evaluation, not actual LOCA effects. Use of a reduced power margin alone cannot affect core damage frequency, the large early release frequency, or actual accident release consequences. The actual accident sequence and progression of a LOCA are not changed unless the licensee modifies its facility. However, the proposed rule may have indirect effects on the environment by allowing licensees to pursue changes to their facilities such as increases to licensed power.

The most obvious change a licensee might pursue under the proposed rule is to increase the licensed power of the facility without conducting ECCS evaluations at a higher power level. Licensees requesting higher licensed power levels are required to assess environmental effects of the change. However, the NRC expects only negligible effects on the environment from small power level changes, such as those that are likely to result from the proposed rule. The NRC previously considered the effects of smallincreases in licensed P

- =

b power level and concluded that such changes would present little change in risk. In NUREG-1230 (Reference 2), the staff considered the risk impact of changes associated with the revised ECCS rules allowing best-estimate analyses, including power increases, and determined that a power level inr:rease of 5 percent or less had little risk significance. This conclusion was, in part, based on the staff's estimate that a small power level increase would only slightly increase the fission product inventory. Also, the staff judged that a slightly higher power would not appreciably alter the potential for LOCAs or affect predicted accident progression.

The staff also considered the risk impact from boiling water reactor extended power uprates, which are much greater than the marginal power change expected under the proposed rule. In these cases, the staff concluded (Reference 3) that extended power uprates are expet.ted to only slightly affect the risk profile of a plant. The staff also stated that marginal power uprates, of about 1 percent, were not expected to require an assessment of plant risk.

An overall affect of a power uprate for a large number of plants is the possible increase in the amount of spent fuel generated by operating at higher power. For the purposes of this assessment, the staff assumed a linear relationship between power level and amount of fuel discharged, and a 1-percent power level increase for 50 plants. Using information on predicted fuel discharges contained in the " Generic Environmental Impact Statement tur i

License Renewal of Nuclear Plants" (Reference 4), the staff estimated that a marginal power increase for half the operating plants would amount to a total of approximately 70 additional discharge fuel bundles per year. This is less than the number of fuel bundles discharged during a typical reactor refueling for a plant. There is a potential cumulative effect associated l

with the anticipated annual increase in discharged fuel. However,it is not considered significant in light of the cumulative level of all fuel discharges during the lifetime of an operating facility.

Under the proposed rule, some licensees could realize savings without seeking power uprates. By revising their ECCS analysis based on a lower assumed power level, licensees could gain margin that could lead to a relaxation in requirements for LOCA mitigation system

.(i.e., ECCS) performance or in core operating parameters. Changes to technical specifications requirements for ECCS system performance will require license amendments and licensees will need to determine environmental impacts, in these cases !nvolving relatively small changes to ECCS analyses, the staff expects that no significant environmental impact would result.

The proposed action, as well at its indirect and cumulative effects, would not increase the probability or consequences of accidents; no changes are being made in the types of any effluents that may be released off site; and there is no significant increase in occupational or public radiation exposure. Therefore, there are no significant radiological environmental impacts associated with the proposed action. The proposed action does not involve non-radiological plant effluents and has no other environmentalimpact. Therefore, there are no significant non-radiological environmental impacts associated with the proposed action.

c

- ;- --- ; =,,3; _

= - - -

s a

-4 Alternatives to the Proposed Action As required by Section 102(2)(E) of the NEPA (42 U.S.C.A. 4332(2)(E)), the NRC has considered possible attematives to the proposed action. The staff considered the following rulemaking options: (1) maintain the provision requiring an analysis margin to account for uncertainty in power measurement but remove the specification of the 2-percent value for the margin and require licensees to assess power measurement uncertainty; (2) eliminate the requirement for a margin between power level and assumed power, disregarding power measurement uncertainty; and (3) broadly revise Appendix K, addressing several conservative parameters.

The attemative of retaining the existing assumed power requirement (i.e., no-action attemative) would essentially have the same environmentalimpact as rulemaking attematives 1 and 2 if licensees pursued exemptions from the current Appendix K requirement. Under the no-action altamative, licensees could also consider the more costly altamative of implementing a best-estimate ECCS evaluation under 6 50.46. However, fewer licensees are expected to take this course, because if there currently were sufficient benefit, they would have already done so. The potential power increase under a best-estimate evaluation is expected to be greater than the marginal power increase associated with the proposed rule. However, the fewer licensees that would use this option reduces the resulting overall environmental impact.

The staff assumed that the environmental impact for either scenario under the no-action altamative would be roughly equivalent.

The environmental effects for the first two altamatives would be roughly equivalent, because about the same number of licensees would seek benefits under any change that would allow a relaxation in the requirement. The main distinction between these altamatives is the course taken to revise the rule. But the end result is the same, in that a marginal power increase would be an indirect result. As discussed enriier, the staff considers marginal power increases to present little risk on a plant-specific basis and the overall effect of increased spent fuel generation is considered small.

The final rulemaking option, to broadly revise Appendix K requirements, could allow greater increases in licensed power for operating plants. However, since there is not a clear understanding of the magnitude of the changes that might result, the staff can only speculate that such a revision would lead to power uprates somewhat greater than those expected under the proposed rule change. The resulting power increases may be commensurate with those associated with previous changes considered by the staff, such as those discussed in NUREG-1230, which were not considered risk-significant.

Therefore, none of the altamatives considered by the staff is expected to significantly affect the environment.

Agencies and Persons Consulted The NRC developed the proposed rule and this environmental assessment. The proposed rule will be published in the FederalRegisterfor allinterested parties to review. All t

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P

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(

comments received within the stated time limit will be considered in developing the final rule.

The NRC is sending this environmental assessment to all State liaison officers for comment.

Finding of No Significant impact f

i On the basis of the environmental assessment, the Commission concludes that the proposed action will not have a significant effect on the human environment. Accordingly, the Commission has determined not to prepare an environmental impact statement for the proposed action.

Also, the NRC is committed to following Executive Order 12898, " Federal Actions To

]

Address Environmental Justice in Minority Populations and Low-income Populations," dated February 11,1994. Since there are no significant offsite impacts on the public from this action, the NRC has determined that there are no disproportionately high and adverse impacts on minority and low-income parties. The NRC uses the following working definition of enviionmental justice: Environmentalfustice means the fair treatment and meaningful involvement of all people, regardless of race, ethnicity, culture, income, or educational level with respect to the development, implementation and enforcement of environmental laws, regulations, and policies.

References 1.

Code of Federal Regulations, Title 10, Chapter I, Parts 50 and 51.

2.

U.S. Nuclear Regulatory Commission, " Compendium of ECCS Research for Realistic LOCA Analysis," NUREG-1230, Washington, D.C., December 1988.

3.

U.S. Nuclear Regulatory Commission, Letter from EDO to ACRS, " Staff Response to ACRS Letter of July 24,1998 on General Electric Nuclear Energy Extended Power Uprate Program and Monticello Nuclear Generating Plant Extended Power Level increase Request," September 14,1998.

4.

U.S. Nuclear Regulatory Commission, " Generic Environmental irrpact Statement for

- License Renewal of Nuclear Plants," NUREG-1437, Volume 1, Washington, D.C.,

May 1996.

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Jde Mned UNITED STATES OF AMERICA CLI.73 39 has ATOMIC ENERGY COMMISSION

!(

om.., t2 a COMMISSIONERS:

no auth:rity

. Apart fr:m Diny Lee Ray, Chairman Clarence E. Larson lesi,we have William O. Doub proceeding, William E. Kriegsman 1 the basis of William A. Anders the report of ncies telating C inspectors in the Matter of at random in RULEMAKING HEARING Docket No. RM 501 he inspectors ACCEPTANCE CRITERI A FOR EMERGENCY

ction by the CORE COOLING SYSTEMS FOR LIGHT. WATER-I

. proceeding then appear, COOLED NUCLEAR POWER REACTORS December 28,1973 show cause The following opinion of the Commission, the concurring opinion of Commissioner Anders, and the inculy grants Appendix to the Commission's opinion are hereby issued this 28th day of December,1973 in Washington.

>urt,like the D. C.

By the Commi<sion l

Paul C. Bender

[

Secretary of the Commission s

Commission OPINION OF THE COMMISSION

1. INTRODUCTION The Atomic Energy Comminion herewith announces its decision in the rulemaking proceeding concerning acceptance criteria for emergency core cooling systems for light. water. cooled nuclear power reactora. The subject of emergency core cooling systems (ECCS) has become a focal point of attention for those concerned with the safety of nuclear power plants. As the manive record developed during this rulernsking shows, a wide spectrum of knowledgeable opinion exists concerning the adequacy o.f our current regulation on this subject-the Interim Acceptance Criteria-and with respect to the nature and scope of regulations which should be adopted at the present time. We have carefully, considered the entire record and the rnany points of viewit encompasses in reac!.ing the decision we a.nnounce today. We believe e

that our decision affords the required reasonable assurance of protection for the public health and safety with a subnantial margin in this introduction we briefly remw the history of this long proceeding and explain the principal reasons underlying the key elements of the decision.The introduction concludes with a brief discussion summarizing the technical context of the issues presented and outlining the changes introduced. The remaining sections set forth these changes and discuss the reasons for them in detsil.The Appendix to this decision set forth the amendments to 10CFR Part 50 which incorporate the rule announced herein,in the format in which those changes will be submitted to the rederclRegister.

On June 29, 1971, we published an immediately effective interim statement of policy establishing interim acceptance criteria for emergency core cooling systems for light water. cooled nuclear power reactors (36 F.R.12247). These criteria, which were adopted following a review by the Ceminission

(

Regulatory Staff and the Advisory Committee on Reactor Safegvards, provided the basis for our belief of 1085

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i Commission Decision CLI-73 39 l

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.us censonable assurance that such systems would be effective in the highly unlikely event of a toss of coolant

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accident (LOCA). The notice requested comments from interested persons and additiona!!y stated that we would consider holding a public rulemaking hearing on this interim policy staternent. Thereafter, on q

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November 30,1971, we announced our decision to hold a legislative type rulemaking hearing for the M

p purpose of aiding us in our determination as to whether or not the interim policy statement of June 29, q p.l 1971, should be retained as issued or criteria should be adopted in some other form. Expanded ground rules y p published on January 8.1972, established procedures, including broad rights of cross. examination, to y

guarantee development of a record that would be as complete as possible M

The Hearing Board consisted of Nathaniel H. Goodrich, Esq., presiding. Dr. Lawrence H. Quarles, and 9.$

Dr. John H. Buck. Participation in tiie rule making hearing was extensive.The primary participants included

.j; j the Commission Regulatory Staff, four reactor manufacturers, a consolidated group of electric utility 6 '[

companies, and the Consolidated National Intervenors (CNI), a group of about 60 organizations and individuals. In addition, three states, the Lloyd Harbor Study Group, and severalindividuals participated to '

$; o[

a lesser degree.The hesings lasted a total of 125 days and generated a record of more than 22,000 pages of 3

transcript and thousands of pages of written direct testimony and exhibits.We heard oral argument by the C

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@d D, "

seven principal participants on October 9,1973.

P Simultaneously with its participation in the hearings, the Regulatory Staff prepared both draft and final i

environmental impact statements in implementation of section 102(2)(c) of the National Environmental

!h,(

Policy Act. A separate phase of the hearings was scheduled to permit questioning of the Regulatory Staff r

F witnesses on the Final Environmental Statement, and of other participants concerning their comments on

%j.!

the Draft Environmental Statement. In its order of June 13,1973, the Hearing Board established a schedule for questioning which conditionally allotted approximately five and one half days for questioning by CN1

' y 4, during this phase, subject only to reasonable advance specification of the subject matter of such

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~l questioning. More than two additional days were allotted by the Board in its Supplemental Order of j

July 13,1973, for other participants to question CN1 witnesses on their environmental testimony.

Dissatisfied with these reasonable limitations, which were entirely consistent with our order of AS T

?YI December 12,1972, CN! withdrew its additional direct testimony and declined to participate in the N$

environmental phase of the hearings. See " Statement of CNI With Respect to Board Orders of June 13, July 10, and July 13,1973," docketed July 23,1973. This incident is especially disappointing to us, for it y

was largely at the request of constituent members of CNI that we experimented with the substantial use of adjudicatory type procedural features in this rulemaking. Moreover because of its broad based make up, td CN! might well have made a meaningful contribution to the environmental phase of the hearings. At oral I

argument before us, it was apparent that CN1 differed sharply from the other participants with reference to I

I cost benefit balancing. The absence of any development of its views on the record is therefore particularly disconcerting.

We followed the ongoing ECCS hearings with great interest. On numerous occasions during the course

. of those proceedings, questions were certified to us, and we issued severalinterlocutory orders dealing with, inter stia, discovery, subpoenas, treatment of proprietary information, and the permissible scope of I

questioning. We reaffirm all of these previous rulings, except insofar as we may specifically depart from them in this /iscussmaa.

During the course of these proceedings the Hearing Board issued nearly 100 written orders, often accompanied by leng~ thy opinions. Many more oral rulings are embodied in the transcript. CNI,in its conciuding statement, referred to its " continuing exception to each adverse ruling in the proceeding,"'

though it declined to specify where, within this massive record, each of the alleged adverse rulings" occuned. We have dealt with the major contentions presented by CNI, and by the other participants. We have not, nor do we believe we should have, scrutinized every page of the record for the possibility of minor procedural errors, which may well exist. While we might have differed with the board on occasional details, we conclude, on balance, that the Board generally exercised its discretion in an appropriate manner so as to develop a record-tested by abundant crossexamination-more than adequate for the formulation of the rule we announce today.

A major item of controversy was the method and manner by which the views of the Advisory Committee on Reactor Safeguards would be solicited. The Committee's views concerning the Interim

' Concluding statement-Safety Phase, of Participant Consotidated National Intervenors, March 15,19'l3. p. 2.5.

5 1086 J

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cf.c'oolant Acceptance Crit @ were contained in a letter of January 7,1972, which became part of the hearing record, ed that we in an order dr, January 26,1972, we denied the request of one participant to subpoena one or more cafter, en members of the ACRS in light of several considerations. Nevertheless, we did permit solicitation of an

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is for the expansion of the views of the Committee as a body through the use ofinterrogatories,a procedure which i

f June 29, was subsequently followed. We reaffirm our belief that this procedure struck an appropriate balance j

aund rules between the competing concerns, permitting a useful additional viewpoint to be added to the record, 3

nation, to without unduly inhibiting the deliberative process by which the Committee fulfilisits statutory purpose of advising the Commission. Following the certification of the evidentiary record to us,we again solicited the 1

arles, and views of the Committee. These were furnished by letter of September 10,1973,which was served upon all s included hearing participants who were requested to comment thereon at the oral argument before us on October 9,

utihty 1973.

mns and Another significant area of controversy during the proceedings related to the permissible scope of

pated to subject snatter to be included. In Chapter til ofits concluding statement,CN! takes strenuous objection to 3 pages cf rulings that certain subjects-5efense.in. depth, causes of a loss of, coolant accidedt, the acchnical basis for mi by the applying a single failure criterion, the matter of ECOS design and design changes, and finally the question of fuel densification-were beyond the scope of the hearing. For a variety of reasons we find CN1's arguments cnd final unpersuasive.

onmental There was an obvious practical necessity to place some reasonable li.nitations upon the subject matter stry Staff of the proceeding. Otherwise there would have been no way to compile a meaningful record and reach a j

decision within a reasonable time. As a practical matter, the proceeding could not have covered every

.ments en j

. schedule conceivable technical question arguably bearing upon the subject matter. This proposition has an important g by CNI corollary-4the rule announced can be no broader in scope than the record supporting it.Thus the rule af such deals with matiers developed on this record; matters outside the racord form no part of this decision. Where Order cf relevant, questions not explicitly addressed by the rule must be considered on a case.by. case basis in individual licensing proceedings. For example, we specifically require that the matter of fuel densification

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be taken into account in analysis but do not specify the method (see discussion on calculation of the initial i

te us.ne stored energy in the fuel, pp. Il001102 Mfrs). Matters such as pressure vessel integrity and steam ll June 13, generator tube failure are the subject of other Commission regulations which must be complied with in us,frit every case. A showing of compliance must be made on the record of each case. Moreover, our Rules of ii sr_l use cf Practice contain a mechanism whereby parties may show that such circumstances with respect to a j

make.up, particular reactor are such that a regulation would not serve its purposes.10 CFR 2.758. See also i

6. At cral Consolidated Edison Co. (Indian Point Unit No.2), memorandum and order dated October 26,1972, I

trence to dealir.g with the admissibility of evidence as to pressure vessel integrity.

-ticularly Three additional matters-the causes, probability and comequences of a 1.OCA were properly J

excluded from the " safety" phase of the hearing, which started with the assumption that the highly unlikely LOCA had occurred, for dat-. reason. That phase involved the question of the validity of es course i

ing with, Performance criteria for systems designed to mitigate the admittedly severe consequences of such a I

. cope of Postulated accident. With respect to the environmental phase of the hearings, however, these subjects art from assumed a far more prominent and relevant position. There the focus of attention was upon the balancing of costs-including potential adverse environmental effects-and benefits, the most notable of which is an ts, ofte]t increase in the margin of safety, CNI, which now argues that these subjects should have been treated in 1, b its greater depth, elected not so participate in the environmental phase, where discussion of such subjects eding,"8 would have been appropriate.

rulings" The matter of ECCS design and design changes is beyond the scope of the instant proceeding for yet ants. We another reason.Our General Design Criterlon number 35 (10 CFR,part 50, App. A, Criterion 35) requires bility cf prevision saf "a system to provide abundant emergency core cooling'." That criterion embraces the concepts casional of pyricrmance (reasonable assur'ance that the system will cool the core) an( retiab,ility (reasonable assurance that the individual components will work). Acceptance criteria for such systems-both our manner for the earlier Interim Acceptance Criteria and the new criteria announced 1oday-selste only to the performance aspect of this design criterion. In other words, ECCS acceptance criteria establish 'imits on design

\\dv ~v parameters (in terms of quantities such as time and temperature), which,if not exceeded, would provide I

assuran.cc that the intended function of cooling the core will be accomplished by the specific system provided in any given case. These ECCS criteria establish a uniform set of standards by which the 2.g.

combination of design and operating limits can be judged acceptable or not solely from the standpoint of l

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1087

~g calculated clad damage should an unlikely losvof coolant accident occur.The adequacy or superiority of a specific ECCS design in a!! respects would draw on more general considerations, and so is a subject peculiar to the specific licensing case. As noted, the remaining aspect of <riterion 35,eeliebsty, deah with the separate question of whether particular designs will work as intended and meet the limitations imposed by g,g; performance criteria. 9hisyestacc< remains proper for individual beensing cases and not for genenc cha plenuking.

p,g The rule we announce today is supported in every respect by the evidentiary record of the hearings, trar This is consistent with the procedures we spelled out in the SupplementalNotice of Hearing. 37 f.R. 288.

(January 8,1972).Those procedures also included the following proviso:

wh.

If reliance is placed on information which is not in the record, notice will be given of such information hea and an opportunity provided to comment thereon and to request an opportunity to respond thereto.

det Several requests to consider material furnished to the hearing board and to us subsequent to the conclusion lon of the safety phase cf the hearings are before us. The material referred to in these requests encompasses:

she (1) Appendix B to the Concluding Statement on Behalf of Babcock and Wilcox (February 22,1973);

(2)The information referred to in Attachment A of the February 22,1973, letter of Transmittal the inn accompanying the Concluding Statement of Position Submitted on Behalf of Combustion Engineering,Inc.t and (3)The October 26,1973, letter (with enclosures) from counsel for Babcock and Wilcox to each of the day Comrnissioners. In each instance, the submissions have been fumished to all other participants.

We have examired each of these extra record submissions but have placed no reliance on them in Prc reaching our decision. We agree with the Regulatory Staff that ample opportunity was afforded to all h/

participants in the proceeding to develop their case by way of sworn testimony subject to questioning at (re the time of the hearings. Accordingly, we deny the requests to utilize the mechanism of Rule 2 of the Bo i

Supplemental Notice of Hearing to elicit comments and responses with respect to these belated ap; -

int submissions.

5 in adopting this course, we are not blinding ourselves to new knowledge acquired as a result of ongoing Ac:

cu research. On the contrary, we believe that it is important that research programs-both analytical and experimental-continue, in order that we may increase the knowledge relevant to ECCS performance.The

'Per (3!

nuclear industry and the Commission have neveral such programs underway at the present time.

th<

We are aware that some information exists that may permit a more liberal allowance on maximum ap.

calculated clad temperature than our present criteria provide. This information is not unambiguous, F'

however, and there has been no adequate exploration ofit even if these extra record submissions were to be considered along with all of the record evidence concerning this subject. We have recently directed the 2*

Director of the Division of Reactor Safety Research to give priority attention to study to determine more H

Po exactly the temperature at which clad embrittlement ceases to be simply a function of oxidation.This is g

the one subject principally discussed in the three extra. record submissions referred to above. As new ll knowledge is acquired, the Comrmssbn will analyze it, and at an appropriate time consider the poulbility Im,

~

of amending the roe ws announce today. We do not, however, believe that the limited amount of comparatively recent knowledge now available, justifies delay in the issuance of a rule based upon extensive g,

examination of thisissue.

Proprietary Dats oc The Commisdon memorandum of June 6,1972, issued in this proceeding,made clear that the ultimate so O

rulemaking decision in this case would not necessarily result in permanent protection of data claimed to be ac!

k proprietary.We there secognized the " strong public interest in conducting a rulemaking proceeding which is th as open as possible to full public scrutiny," and explained that open consideration of the technicalissues w

was a motivating factor in the experimental use of a public rulemaking hearing. In that vein,we approved d["f procedures whereby all coumel-including those representing competitors of the participant offer,ing the data-were permitted to examine data claimed to be proprietary, and to attend and participate in he i

comera sessions involving such data. Subject to review by the board and by other participants,each counsel was allowed to submit issues involving the proprietary matter to his client for consultation or guidance.

P With respect to the decisional phase of the case,we explicitly stated:

'* ~ :

...we would underscore that our present holding is confined to treatment of proprietary information during the hearing phase of this proceeding Should such information form part of the basis for the y,

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1088 i

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ultimate rulemaking decision, the Commission will again-and in that context-address the question abject peculiar of that information's public disclosure.

Jeals with the Subsequently, Combustion Engineering (CE) voluntarily submitted data asserted to be proprietary with ns imposed by full knowledge that such data might be made public at a later time. Included within that submission is a it f:r generic chart appearing on page 65 of CE's Redirect and ' Rebuttal Testimony (Ex.1144). See page 1119 frr/ra. This item consists of data experimentally confirming the adequacy of cenain heat

,W hearings.

transfer correlations at low pressures. Each licenser must use one of these correlations in order to bring its

.37f.R.288.

reactor operations into conformity with the limitations here prescribed. The correlations are part of a rule

)

which will be used by the entire industry in strengthening the degree of rearonable assurance of public h information health and safety. The data conforming the correlations are thus part of **the basis" for the rule.

sd thereto.

Upon reexamining the matter of data claimed to be proprietary,in the context of promulgating the he conclusion decision itself, we conclude-in the circumstances of this case-that this aspect of the CE data should no encompasses:

longer be protected and should be disclosed in the public faterest. Accordingly, and unless good cause be ry 22,1973);

shown, this portion of the submitted data will be placed in the public document room 30 days following

( Transmittal the date of publication of the rule in the FederalRegister. Combustion Engineering and other participants ineering,Inc.;

may submit views as to the treatment of this item by filing appropriate papers with the Secretary within 15 D cach cf the days following the date of this decision.

For use in implementation of the rule, we have also approved two correlations claimed to be e on them in proprietary. First, we approve Westinghouse's transition boiling correlation (Exhibit 1152,section 25).See f:rded to all bifra. pp. I1091110,11161118. Second, we approve General Electric's Hench Levy CHF correlation utsti ning at (referred to in Ex.1001, p. 4 21; Ex.132, p. C.9; and Tr.14184 et acq.). See pp. I109,11131115,. infra Rule 2 cf the Both of these itams, previously approved for ae under the interim policy statement (36 Fed. Reg.12247),

1ese belated apply only to an individual corgany's evaluation models, and are more akin to matters involved in individual licensing proceedings, and thus must presently be considered to be distinct from the CE data.

.it f.oing Accordingly, our present inclination is to continue to protect these two items in a manner consistent with ca,

.and our approach in individual licerning cases (see 10 CFR 2.790)-subject, of course, to the outcome of our arrruu.The pending reexamination of policy and rules concerning data for which proprietary protection is requested (38 Fed. Res. 31543). At the same time, we are prepared to receive comments from all participants as to m maximum the public disclosure of these two items as well. Such comments may also be submitted by filing nambigurus, appropriate papers with the Secretary within 15 days following the date of publication of the rule in the c were t2 be FederalRegister.

directed the Our conclusion as to the CE item is limited to the partleular facts of this case.' Existing rules (10 CFR ermine more 2.790) sha!! continue to apply to individual proceedings. Similarly, the conclusion as to this item is without

. tion. This is prejudice to whatever determinations we may reach in the pending reexamination referred to above.

sve. As new it possibility Implementation Schedule amount of The matter ofimplementation of the rule ultimately adopted has generated strongly divergent views.

Indeed, at the oral argument on October 9,1973, the implementation schedule was the source of more controversy than any other issue. As appropriate in a rulemaking proceeding of this magnitude,the varied interests represented by the participants produced a broad range of views.

CNI, wh!!e arguing that the record supports no mle at a!!, apparently took the view that in any event the citimate cost. benefit balancing was irrelevant to implementation. The Regulatory Staff, recognizing that certain

' aimed ta be social and economic costs would flow from implementatfor ofits proposal, argued that those costs should ing which is nonetheless be borne in order to fashion a timely response. The majority of the industry participants, on mical issues the other hand, argued that the various costs were so hurdensome (and the benefits so minimal) as to ve approved offering the paM in h e Among she factors 3Meh combine to make this case unique are the followlag:

ach c:unsel (1)This is a rulemaking proceeding, involving Commission reliance on the CE data. claimed to be proprietary,in sr guidance."

promulgating safety rules applicable to a!! vendors and ticensees.

/

(2) This proceeding was conducted under its own rules and not under the rules of general appticabitity. Ses 10 CFR Part 2. Subpart G.

in m

(3) Prior to the voluntary submission of the item, a!! vendors sete placed on notice of the fact that proprietary data iasis, the forming a part of the basis for the ultimate decision might be pubticly disclosed.

1089 g

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"w warrant an implernentation schedule which would accord with the normal reactor fuel cycle-which means that for some reactors a rule would not be fully effective for as long as three years. Finally, one participant argued for permanent exemption of existing plants.

f.

We note that the calculations that must be made to conform to the rule will be time. consuming.They g

will require analysis well beyond that defined in the Interim Policy Statement. Additional phenomena must g

be taken into account in the evaluation models: clad deformation, clad bursting. expansion of the gas in the 4

gap between fuel and cladding, variable gap condunance (including effects of densification), clad oxidation a

(two sided oxidation, heat generation by oxidation), decay heat from actinides, flow redistribution. The sensirheity of calculated results to variations in noding and input parameten must be explored in repeated cortputer runs. Calculations will be required for a spectrum of pipe break sizes, including splits as we!! as

p; l

double <nded breaks. At least three values of the discharge coefficient must be used for each break q.

examined, spanning the range from 0.6 to 1.0.This large number of calculations wi!! now be needed for a!!

of the newer power reactors, as well as for other reacton cunently under licensing review. Only the reactor

) [

vendors will have the capability of performing the analyses for the plants they hase sponsored. Each vendor j '

1 will suddenly be inundated by needs to provide calculations for all the plants it has designed.

J We reject at the outset any schedule which does not take proper accacet of the steps needed for well founded implementation. It is evider.t that the vendon cannot develop the models and produce the necessary calculations in the four. month period proposed by the Staffin its Cone ading Statement.This is a

a

,1 l so even though prudent manufacturers may well have done advance calculations upon the assumption that the Staffs proposal would ultimately be the rule. Inasmuch as our rule is somewhat different from that 1

proposed by the Staff, the calculation process will have to begin anew. It is also unlikely that the AEC 4

I evaluation model can be completed and proof tested to be ready in a four month period.We do not wish to i

f l require a schedule for implementation that makes impossible demands on either the vendon or the 2

Regulatory Staff or that might lead to taking hasty steps on the basis of unconsidered analysis,perhaps 1i requiring retraction or sevision later. Rather, we wish to ensure that the calculations be thorough as well a timely.

A requirement of immediate compliance would be tantamount to an order shutting down or J[

substantially derating all reactors until requisite calculations were complete.We would not hesitate to take that action if circumstances so warranted. But the record shows,and we find, that the It terim Acceptance 7

E Criteria will provide reasonable assurance of protection for the public health and safety during the relatively 1

brief transitional period which will culminate it, mpliance with the new rule. In addition, there is not sufficient basis for presently accepting the view, es, 5 ed primarily by the utility and industry panicipants.

j which would defer implementation for as long as three years.

The evidence has demonstrated that a number of irnprovements could be made in the models used to p

3 a

evaluate emergency core cooling systems, and these have been incorporated into the rule we announce fq

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toisy. These proceedings have reduced the number of unknowns, and thus reduced the degree to which c

unknowns must be bridged by conse vative safety assumptions. In short, the new rule provides a more e

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objective basis for safety analysis. Because of our confidence in the Interim Acceptance Criteria, we l' permitted them to remain in effect throughout our reevaluation. We also conclude that they will cor.tinu

[

3 to provide reasonable assurance for a transitional period. However, since we now have an impioved r 3

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supported by an evidentiary record, we choose to implement it at a rate which meets the following principles:

'$q 1.The schedule should allow for thorough development of the complex evaluation models necessary 1

f 2

to comply with the new rule;

?p' 3

2.The schedule should seek to effectuate the new rule's incrementalincrease in safety at the earliest 2

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practicable time; 3.lf possible, this gen! abould be accomplished without unwarranted disruption in the nation's

]

production of electric energy.

Guided by these principles, we adopt the following implementation schedule with respect to all 5

facilities for which operating licenses have previously been issued and for which operating licenses may i

during a period of one year from the date of this decision:

l.The rule shall become effective for the purpose of computis.; the time within which reiguired or permitted actions must be dorae 30 days following its publication in the FederaIRegister.

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2. Within six months fouowing said effective date, all licensm snau submit the requisite evaluation to J

participant

.the Director of Regulation for review. The evaluation shan be accompanied by such proposed changes in technical specifications or license amendments as may be necessary to bring resetor operation in ning. They conformity with the rule.

3. Any licensee may request an extension of the six month period for good cause. Any such request mena must a gas in the sha!! be submitted not leu than 45 days prior to the expiration of the six month period, and shall be 1 osidation accompanied by /: davits showing precisely why the evaluation is not complete and the minimum time uten. The believed necesse " omplete it. The Director of Regulation shaU cause notice of such a request to be a repetted published prc. ;,tly in the Fedenr1 Register;such notice shall provide for the submission of comments by

as wellas interested persons within a time period to be established by the Director of Regulation. If, upon reh break reviewing the foregoing submissions, the Director of Regulation concludes ' hat good cause has been ded frr all shown for an extension, he may extend the six-month period for the shortest avditional time which in his he reactor judgment will be necessary to enable the licensee to comply with paragrap 2 above. Requests for sch vendor extensions of the six month period, submitted under this paragraph, shall be rund upon by the Director of Regulation prior to the expiration of that period.

.eeded fer

4. Upon the submission of the matters specified in paragraph 2 above,(or under paragraph 3 above,if

- oduce the she six-month period is extended), the facility may continue or commence operation within the limits of rt. This is both the proposed technical specifications or license amendments submitted under the above procedure, Tisin that and aU technical specifications or license conditions previously imposed by the Commission.

ft:m that 5.Further restrictions on reactor operation will be imposed by the Director of Regulation if his the AEC review of the evaluations submitted under paragraphs 2 and 3 above so warrants, et wish to

6. Exemptions from the operating requirements of paragraph 4 above may be granted by the rs or the

• Commission for good cause. Requests for such exemption shall be submitted not leu than 45 days prior s,perhaps to the date upon which the plant would otherwise be required to operate in accordance with the as?-"as procedures of paragraph 4 above. Any such request shan be filed with the Secretary, who shall cause dlc notice ofits receipt to be published promptly in the FederalRegister; such notice shall provide for the submission of comments by interested persons within 14 days following FederalRegister publication.The

.er te 13 take Duector of Regulation shall submit his views as to any requested exemption within five days following xeptance expiration of the comment period.

relatively

7. Any request for an exemption submitted under paragraph 5 above must show,with appropriate tre 6 not affidavits and technical submissions, that it would be in the pubtle interest to allow the licensee a ticipants, specified additional period of tirne within which to alter the operation of the facDity in the manner required by paragsph 4 above. The request shall also include a discussion of the altematives available for s used is establishing compliance with the rule, mnounce in which Description of a LOCA s a more we have -

The Comminion noted. in the Interim Poucy Statement that:

continue Protection against a highly unlikely loss of coofant accident [LOCA] has long been an essential part of ved rule, the defense in depth concept used by the nuclear power industry and the AEC to assure the safety of aggowgng nuclear power plants. In this concept, the primary assurance of safety is accident prevention by correctly designing, constructing, and operating the reactor. Estensive and systematic quality assurance (cessary practices are required and apphed at every step to achieve this primary assurance of safety.

Nevertheless, deviations feom expected behavior are postulated to occur, and protective systems are r Carliest installed so take corrective action as required in such events. Notwithstanding all this, the occurrence of serious accidents is postuisted, in spite of the fact that they are highly unlikely, and engineered safety nation's features are installed to mitigate the consequences of these unlikely events.The less of coolant accident is such a postulated improbable accident; the emergency core cooling system [ECCS) is one of the engineered safety features installed to mitigate its consequences.

t to all ay issue The fo!!owing elementary description of a hypothetical LOCA is given to indicate the points of g

principal attention and to provide a background for a discussion of the changes in the regulations and their

[

conservatisnt It should be remembered that the calculations that are made of the effectiveness of the ECCS

(*

center on maintaining the integrity of the aircaloy cladding, since if it remains intact we can be sure that the uranium dioxide fuel pellets will be kept separate and coolable. To keep the zircaloy intact reautres j

controlling its maximum temperature and its oxidation.

\\

j iosi

Although the ECCS is intended to cope with a wide range of possible breaks in the primary system piping, most attention has been focused on the sudden severance of the large diameter cold les pipe for the pressurized water reactor (PWR) and of the recirculation jet. pump inlet pipe for the boiling water reactor (B%R). Postulated breaks of these lines define the onset of the hypothetical accident for the two kinds of reactor. Before the hypothetical accident, the reactor is arsumed to be operating at 102% of full power, thus including a rnargin for such factors as instrument error. The temperature of the zircaloy cladding at this tirne would be near that of the adjacent water,in the neighborhood of 600*F.The average temperature of the hottest UO: pellet would be well above 2000*F with a peak temperature in the center greater than 4000*F. The excess heat content of the UO: at.this average ternperature, called the " stored heat,"is approximately proportional to the power density and is affected by the therrnal resistance of the " gap" between the UOs pellets and the cladding. The stored heat is important because it would contribute significantly to the later temperature history of the cladding.

Very early in the LOCA the prompt fission heat would stop as the density of the water moderator decreases. At the same time, the cooling of the outside surface of the zircaloy cladding would diminish sharply because of t!.e altered hydrodynamic flow. Under these conditions, the temperature distribution 1

across the uranium dioxide and the zircaloy would tend to even out, dropping the peak temperature in the center of the U0, increasing the UO temperature near its surface and increasing the temperature of the 2

zircaloy, if there were no heat removal from the outer surface of the zircaloy, its temperature would quickly approach the averagt temperature of the UO. An appreciable amount of the stored heat would be removed by the rushLy water and steam as they escape from the reactor vessel, thus limiting the initial rise F

in temperature of the ccaloy.

After the aca.ident began the fuel pellets would continue to be heated by the decay of the fission A

products s.nd of the actinide elements (neptunium and plutonium) that were produced during reactor operatiori.da addition,if the zircaloycladding reached temperatures ofiboutl3800*F or above,its reaction

with steam to form zirconium dioxide would begin to add to the heat generation.These heat sources would cause the average imperature of the fuel rods to start to increase after the cooling effect of blowdown ceases, and the temperature of the zircaloy would now continue its increase, keeping pace with that of the j

fuel pellets. The temperature excursion would eventually be terminated as the ECCS begins to reflood the core. Both P%R's and B%R's have ECC systerns'in which water would reflcod the reactor. In B%R's the reflood would be provided by accumulation of mter from the low pressure injection system and the core spray system. Direct core spray is discussed below. To accomplish reflood in a reasonable time, the rate at which the ernergency cooling water would encroach on the core (the reflood rate)must be high enough to provide a heat transfer rate from the core that would be sufficient to counter the heat input rate from decay heat and from zircaloy oxidation. The Commission believes that the calculated reflood rate should

.?q have a substantial rnargin over the rate that isjust sufficient to turn the temperature excursion around in a 6

abort time.

' h As the cooling water reaches the hot core rrnch of it would be converted to steam, and it is this steam I

together with entrained water droplets that would provide the initial cooling of the hotter regions of the

-.Y core. For the reflood water to continue entering the core it must displace the cream,which would have to

{

escape from the reactor vessel and find its way into the containment atmosphere.In the pressurized water c

1

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reactors the steam would have to flow through the steam genentor and pump to escape through a cold leg

'~

break; the reduction of reflood rate by the relatively high resistance to flow of this path is called " steam (d.

binding". Steam binding would severely limit the rate of reflooding the core, reducing it from an intended 6 to !! inches per second to frasa 1.0 to 2.5 inches per second, depending on the reactor design.The rule we

.IK6 announce considers a5 the evidence in the record on this irnportant subject of steam binding and provides j

M an acceptable overa!! assurance of ECCS effectiveness. The inquiry, however, should not end there. Thus O'I the Commission urges the pressurized water reactor rnanufacturers to seek out design changes that would l

5 t overcome steam binding. This same point of view is reflected in the September 10,1973, letter of the

[Ir:

Advisory Committee on Reactor Safeguards.

j 1

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Bolling water reactors would not be subject to steam binding, because their system design provides a E

more direct path for the steam to escape, but the same requirement for rspid seflood would have to be met s

If excessive clad damage were to be avoided. BoUing uter reactors do have a core spray system that w'ould

( @.

I start about 30 seconds after occurrence of the break, but its cooling effect on the central rods of a feel bundle might be insufficient in itself to prevent exceeding the temperature limits we have art. The n,,:W i

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• 4 1092 va

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occurrence of reflooding within three.ninutes after a postulated break of the recirculation line would ry' erem pefcrthe terminate the excursion.

er reactor To recapitulate, after a large LOCA the principal cooling of the core would occur in two stages, first by a kinds of the uncontrolled escape of the pressurized water and steam during blow.down, and second by the reflood all power, of the core. The first would be effective in reducing the stored heat, albeit to an extent not yet fully adding at settled, and for which we require conservative estimates, and the second would overcome the afterheat generation rate of the core. Both would be important in determining the temperature history of the nperature ster than aircaloy cladding.

heat," is i

Without redesign and back. fitting, the only measures available to the operator in relation to limiting the the " gap design basis accident within the given design framework are to limit the power and the power density of the a

ontribute nactor. The power density can be manipulated somewhat independently of the total reactor power by adjustments of fuel enrichment and control rod action to provide more uniform power generation noderator throughout the core. The Commission notes that there has been a tendency to reduce the maximum diminish allowed peaking factor (ratio of the highest power density to the average throughout the core) to satisfy stribution ECCS criteria. These lower allowed peaking factors leave less margin above the normaloperating unge for tre in the maneuvering; thus greater care in reactor operation is required to ensure that these factors are not t:re of the exceeded.

are would would be Principal Changes From Interim Policy Statement inidal rise The interim Policy Statement includes: (1) general criteria for emergency core cooling systems applicable to all light. water power reactors (the Interim Acceptance Criteria, or IAC),(2) requirements for he fission analysis using a suitable evaluation model,(3) provisions for application to various classes of reactors by is mactor,

specified dates,(4) provision for variance under stated conditions, and (5) a listing of acceptable evaluation s maedon models. The new regulation has sections serving the same purpose as (1), (2), (3), and (4) above. No

5 ""ald complete listings of acceptable evaluation models accompany this decision. The required and acceptable h

features of evaluation models, however, will provide the basis for the Regulatory Staff to determine the hs

.ie acceptability of such models as may be furnished.

illood the The principal changes from the Interim Policy Statement are as follows.The old criterion number one, WsW specifying that the temperature of the tircaloy cladding should not exceed 2300*F,is replaced by two d the core criteria, lowering the allowed peak tircaloy temperature to 2200*F and providing a limit on the maximum W raw at allowed local oxidation. The other three criteria of the IAC are retained, with some modification of the

'"'"8h

wording. These three criteria limit the hydrogen generation from metal-water reactions, require rau km maintenance of a coolable core geornetry, and provide for long. term cooling of the quenched core.

su should The most important effect of the changes in the required features of the evaluation models is that i

round m a swelling and bursting of the cladding must now be tahen into consideration when they are calculated to occur, and that the maximum temperature and oxidation criteria must be applied to the region of clad this steam swelling or bursting when the maximum temperature and oxidation are calcuinted to occur there. Another sns of she important change is the requirement that,in the steady siste operation just before the accident, the thermal M have so conductance of the gap between the fuel pellets and the cladding should be calculated taking into consideration any increase in gap dirnensions resulting from such phenomena as fuel densification, and s edd leg should also consider the effects of the presence of fission gases. Men these effects are taken into ed "ateam comideration a higher stored energy may be calculated. Other changes in the evaluation models are mostly unded 6 in the direction of replacing previous brasd conservative assumptions with more detailed calculations where he mie =

new experimental information is available or where better calculational methods have been developed.

d provides The wording of the definition of a loss.of. coolant accident has been modified to conform to its

" ' D"'

longaccepted usage, limiting it to breaks in pipes. Justification for the exclusion of consideration of

- pressure vessel failures from the LOCA is extensively discussed throughout Volume 39 of the transcript

,g (Apr011,1972), and we have refened to it earlier (p.1087).

The new regulat'ons also require a more complete documentation of the evaluation models that are

"'* 0' ts he met q

f Conservatisms The Commission believes that the implementation of the new regulations will ensure an adequate i se...ne margin of performance of the ECCS should a design basis LOCA ever occur. This margin is provided by 1093 i

contribute to the conservative nature of the evaluations and the cri 8

highest estimated thermal resistance between the UOs(1) Stored Hea and the cladding provides a calculated stored heat over the extreme condition, it represents at least an assum not typical.

i (2)Blowdower. The calculation of the heat transfer during blowdown is made in a very

)

manner. There is evidence that more of the stored heat would be removed than calculated, I*,

0.0 not yet an accepted way of calculating the heat transfer most accurately.it is probable that thh be expected to carry over to a reduction in the calculated pea (3) Aste of Haer Genererlos. It is assumed that the beat generation rate from the decay g*

products is 20% greater than the proposed ANS standard.This represents an upper limit t E888" uncertainty. The assumption that the fission product levelis that resulting from operation at 102%

power for an inGnite time represents aa improbable situation, with a conservatism that h pro range of 5 to ISE The use of the Baker.Just equation for calculating the heat generation from t 8 Pred be del oxidation of zircaloy should also provide some conservatism, but the factor is uncertain.

h occ 1 to 2200*F and the stipulation that this criterion be applied to t includ provide a substantial degree of conservatism. They ensure that the core would suffer verylit the accident.

Suggestions L *e been made during the hearing for quite different types of criteria, bearing m directly on the

' ns available to reactor operators or to the mecharilsm by which the ECCS would claddir terminate the t.

rotere transient. For reactors that have already been constructed, the only limiu in this (4) sense available er,.no operator would be on the power of the reactor and on the distribution o amenal l within the core. This fact led to t.he suggestion by some of a criterion limiting the power density (5) fashion. (See, for example, Exhibits 1043 and 1044.)This proposal h rejected on the bash that core te to the ECCS is tenuous and arbitrary, and that the imposition of such a restriction might inhibit inn extends in reactor design. Simuarly, since the temperature transient would eventually be terminated b flooding, the suggestion has been made by some that the allowed power density be tied in a cons R.TEC way to the re!!ood rate (Exhibit 1113, pp.1418 and pp.174 to 17 5). Although this suggestion has III' considerable rnerit for application to the present design of ECCS for ITR's, the setting oflimits wou e m ed.

require calculation of the temperature history of the cladding by some evaluation model. For this reaso (2b this proposal was rejected in favor of modifying of the Interim Acceptance Criteria. In doing this, h El7 th the Commission has intentionally incorporated the effects of the conservative features indicated abo provide a suitable safety factor in the relationship between the capability of the ECCS and the p i

density in the reactor.

f"I i

In its Concluding Statement the Consolidated National Intervenors claim that there h an " inad

• • 3d*83 base on which to base predletions of the course of an accident" and that " uncertainties and errors...a E****

unknown", and in Chapter 5 they present a list of "information needs," for Commission guidance. The before a Commhslon realizes that the knowledge in regard to a number of facets of the analysis of a loss of coola accident is imprecise;it is partly for this reason that there la en on-going Water Reactor Research Pro i

ap g

The Commission is confident, however, that the criteria and evaluation models set forth here are more th sufficiently conservative to compensate for remaining uncertainties in the models or in the data.

Continuing research and development wf!! provide a more extensive data base for such items as heat transfer coefficients during blowdown and during spray and reflood cooling, oxidation rates for tirconium fission product decay heat, stean> coolant interaction, oseIIIstory reflood flows, fuel densification g

modeling and flow blockage. With the additional data it may become practical to assign a statistically meaningful measure of preelslon to the calculation. It is probable that, with a better data base, some Ther; relaxation can be made in some of the required features of the evaluation models. However, the intact to i Commhslon believes that any future relaxation of the regulations should retain a margin of safety abo array. Co.

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I U. ACCEPTANCE CRITERIA FOR EMERGENCY CORE COOLING ointunat SYSTEM EFFECTIVENESS ctor, and ared heat A.THE CRITERIA a margin (lpeak Cadding Tenpesture. The calculated maximum fuel element cladding temperature shall not a which is exceed 2200*F.

(2)Marinnen, Cadding Oxidation. The calculated total oxidation of the cladding sha!! nowhere exceed 1servative 0.17 times the total cladding thickness before oxidation. As used in this subparagraph total oxidation A tkn h means the total thickness of cladding metal that would be locally converted to oxide if a!! the oxygen vpresents absorbed by and reacted with the cladding locally were converted to stoicluornetrie zirconium dioxide.lf assonably cladding rupture is calculated to occur, the inside surfaces of the cladding shall be included in the g,

oxidation,beginning at the calculated time of rupture. Cladding thicknen before oxidation rneans the radial of fissi:n distance from inside to outside the cladding, after any calculated rupture or swelling has occurred but g,,,, gg kron significant oxidation. Where the calcuhted conditions of transient pressure and temperature lead to icf rated '

a prediction of cladding swelling, with or without cladding rupture, the unoxidized cladding thidness shall bly in the be defined as the cladding cross 4ectional area, taken at a horizontal plane at the elevation of the rupture,if it occurs, or at the elevation of the highest cladding temperature if no rupture is calculated to occur, the steam 8

divided by the average circumference at that elevation. For ruptured cladding the circumference does not a cladding include the rupture opening.

t fuel rod (3) Maximum #pdrogen Generation. The calculated total amount of hydrogen generated from the 3,

chernical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metalin the chdding cylinders surrounding the fuel, excluding the

-ing more cladding surrounding the plenum volume,were to react.

OS would (4)Coolable Geometry. Calculated changes in core geometry shall be such that the core remains

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amenable to coohns, (5) tong. Term Cooling. After any calculated successfulinitial operation of the ECCS, the calculated c.

t yin some core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the is relation extended period of time required by the long. lived radioactivity remaining in the core, nnovation i by core B. TECHNICAL DISCUSSION OF THE CRITERIA

,)

(1)Psak Chdding Temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200*F.

vould still (2)Marimum Cladding Oxidation. The calculated total oxidation of the cladding shall nowhere exceed e

his reason 0.17 times the total cladding thickness before oxidation. As used in this subparagraph total oxidation

. however' means the total mickness of chdding metal that would be locally converted to oxide if s!! the oxygen absorbed by and teacted with the cladding locally were converted to stoiciuo' metric tirconium dioxide.lf

. above to g pm, chdding rupture is calculated to occur, the inside surfaces of the cladding shall be included in the oxidation,beginning at the calculated time of rupture. Cladding thickness before oxidation means the radial nadequeu distance from inside to outside the cladding, after any calculated rupture or swc!!ing has occurred but before significant oxidation. Where the calculated conditions of transient pressure and temperature lead to g,,,,

ance. Tk a prediction of cladding twelling, with or without cladding rupture, the unoxidized cladding thickness shall ofcoolant I

be defined as the cladding cross 4ectional area, taken at a horizontal plane at the elevation of the rupture,if i Program.

it occurs, cr at the elevation of the highest cladding temperature if no rupture is calcuhted to occur, more than divided by the average circumference at that elevation. For ruptured cladding the circumference does not include the rupture openlag.

ns as heat

• irconium, Discussion of Peak Claddirns Temperature and Maximum Oxidation can, pump The purpose of these first two criteria is to ensure that the zircaloy claddingwould remain sufficiently tattticallY intact to retain the UO fuel pellets in their separate fuel rods and therefore remain in an casDy coolable m, some

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array, Conservative calculations indicate that during the postulated LOCA,the choding of m,any of the aw the rods would swell and burst locally with a long'itudinal split.The split cladding would remain in one piece if fe(

e it were not too heavDy oxidized, and would still restrain the UOs pellets.The possibility of destructive damage to the zircaloy cladding must be examined for times late in the course of the LOCA when oxidation N

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at high temperatures in the steam atmosphere could render the cladding brittle.The hmits specified in these criteria will assure that some doetility would remain in the zircaloy cladding as it goes through the guenching process, and therefore that the core would remain essentially intact,in a condition amenable to long-term cooling.

The steam oxidation process is described in Exhibit 1122, pp.21 o 2 7 (Scatena, of General Electric).

l Water rnolecules are absorbed on the surface and dissociate to hydrogen atoms and hydroxyl radicals.

Within the surface the hydroxyl radicals are (in several steps) converted into oxygen ions and hydrogen atoms. The hydrogen atons, wherever formed, combine into hydrogen molecubs at the surface and escape.

The oxygen ions diffuse into the surface, and are dissolved in the metal. When their concentration is high enough zirconium dioxide is formed.

The initial reaction rate is limited by the counter diffusion of gaseous water molecules to the surface j

and of hydrogen molecules away from the surface.This lim 3tation lasts for only a short time, until an oxide I

film is formed. Thereafter, at sufficiently high temperatures, the reaction rate is controlled by solid state diffusion processes: largely the diffusion of oxygen ions through the tirconium oxide and the metal. A gradient of oxygen concentration will exist from the surface to the center,with high concentrations on the i

outside surface and lesser concentrations in the interior.

At room temperature pure zirconium is in the alpha phase, with a close packed hexagonal crystal i

9 2:

structure. On heating above ll50*F,it is transformed to beta phase,with a body centered cubic structure.

The zirconium-oxygen phase diagram (Fig.2 4 of Exhibit 1122) is useful as a guide to the phases that will 9

be present under os.idizing conditions. At the temperatures at which the oxidation takes place (between 1400 and 1700*K in the figure), the outer surface M11likely be oxidized to stoichiometric ZrOs;inside this I

will be some alpha phase zirconium stabilized with a high concentration of dissolved oxygen;then a region

. of mixed alpha and beta phase; and in the interior, pure beta phase. When this is quickly cooled to room temperature three regions can be distinguished metallographically: the zirconium oxide, the unchanged J

alpha phase zirconium (called stabilized alpha phase), and a region of alpha phase zirconium that exis:<.d as 1

beta phase while it was hot (the so called prior beta phase). These three regions are shown in the photomicrograph of figure 2 5 of Exhibit 1122.

f i

): is well known that the oxide and stabilized alpha phases are brittle, and that what ductility and resistance to shattering is exhibited by the oxidized tircaloy is associated with the prior beta phase material.

O For this reason it has been the custom to correlate the strength remaining in the oxiud metal with either the sum of the thicknesses of the oxide and the stabilized alpha material (called XI,t (), or, conversely.

the fraction of the original thickness remaining as prior beta phase (called Fw). For example,Hobson and R;:.enhouse (Exhibit 509) preferentially use XI to correlate their data on ductility and hardness, while ph.

Combustion Engneering (see for example Exhibit 1144) has used Fw.

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The situation is complicated by the fact that not all of the prior beta phase is equally strong or ductile.

'y since these properties depend on the amount of dissolved oxygen. This fact has been suspected for some time and the basis for using Xi or Fw, aside from the case of measurement, has been the suumption that the oxygen content in the prior beta phase is closely related to the thickness of the oxide and stabilized i

alpha layers. For example, figure 7 of Exhibit 509 gives a correlation between the hardness of the prior beta j

Jt phase and XI. Westinghouse,in their concluding statement (page A 5), indicated a belief that a different

'Q exposure parameter, related to the amount of oxygen in the prior beta phase,would be more appropriate.

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The matter is discussed at Trameript pages 20,935-41 and 21,629 30; and in Exhibit 1133, pp.27 and

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29 33.

As a result of the diffusion process, the concentration of oxyBen in the prior beta phase will not be uniform within a sample, but will depend upon the depth below the surface.(See, for example, figure 2 3 u

of Scatens, Exhibit 1122, or figure A 3 of the Westinghouse direct testimony, Exhibit 1078.) From the phase diagram, given by both Scatena and Westinghouse,it is obvious that it is possible for the beta phase

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zirconium to take on a higher oxygen content at 2600*F than at 2000*F. Furthermore,since the diffusion rate depends exponentially upon temperature, ene might expect a greater incursion of oxygen into the beta phase for a given thickness of oxide and stabilized alpha phase at higher temperatures. Westinghouse has

"'5 calculated three cases (page A.10 of Exhibit 1078), and although the results are not unequivocal, the comparison of the 2200*F and the 2000*F cases indicates that it is possible to have significantly more oxygen in the beta phase at the higher temperature for about the same value of XI.

. h.i In discussino th: Av. compression tests of Hobson, the Regulatory Staff in their supplementary

[;g,gg testimony (Exhmit 1113, page 1814) stated that "... the 2400*F specimens seem to be more brittle than d M.

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tnd Fw values would indicate, compared io the lower temperature specimens. The basic inference Hobs pf feels should be drawn from this observation is that embrittlement is not just a monotonic func or Xi penetration, but is also related to the exposure temperature. The most likely explanation for this

,g,,,

behavior is increased oxygen in the beta phase, due to increased oxygen solubility and to m diffusion kineties."

,f Others have also observed that the resistance to rupture depends upon the temperature et which g,,y gatidation occurs as well as the extent of oxidation. For example, Combustion Engineering in Exhibit page 2 2, asated with respect to their compressive load tests,"The limiting value of VDt of 0.012 c g.,p,

,gp' meuure of the extent of oxidation)is based upon experimental tests at 2300 to 2500*F. Oxidation tests at 2100*F demonstrated that the limiting value of 0.012 was algnificantly increased so that the 0.012 cm increasingly conservative with decreasing temperature below 2300*F." Babcock and Wilcox, altho

,,7,

att nsly disagreeing with much of the Oak Ridge interpretation of their results, apparently confirmed the oxide major thesis that &mbrittlement is a function of both the extent of oxidation and the, oxidizing

,gy, 891MPerature.1n their concluding staternent, page 239,Ithey stated,"However, the tests confirmed, under tal. A similar experimental conditions, the implication from the Oak Ridge tests that a conelation between i,

,g ductility and remaining beta fraction does not adequately characterize cladding ductility above a threshold

,y,g temperature ranging from 2200 to 2400*Ffumace temperature " (B&W has stated a belief, however, that the actual temperatures of the zircaloy specimens were at least 100*F higher than the furnace temperatures.)

To recapitulate, measures of zircoloy oxidation, whether by percent, XI, or Fw, are largely or who neen determined from the brittle layers of zirconium oxide or stabilized alpha phase, while the ductility an e Ws sitength of oxidized zirconium depend upon the condition and the thickness of the prior beta phase.

yion According to Westinghouse *s sample calculations (Exhibit 1078, page A 10) the proportion of the total oxygen content that is in the prior beta phase is of the order of 4 to 10%.hs a criterion based solely on d

she extent of total oxidadon h not enough, andsome additional criterion is needed to assure that t

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' beta phase is not too brittle.The speelficatio'n of a maximum temperature of 2200*F adequately.The data cited in Exhibit 1113 would not support a choice of a less conservative limit.

There h relatively good agreement among the industrial participants as to what the limit on total cnd oxidation should be. Babcock and Wilcox and General Electric have suggested a limit of 17% of the crial.

aircordum being oxidized, while Westinghouse proposed 16%. Although Westinghouse has preferred the uher limit to be expressed as percent total equivalent oxidation, they equated their 16% to a ratio of brit

,,j,y thickness (Xi) to original thickness of 0,47. Combustion Engineering's recommendation of a minimu tnd of 0.65 is also equivalent to a maximum XI ratlo of 0.47, so that allof the reactor venders are in essential site agreement. The Utility Group also recommended the 17% oxidation limit, but said that if one wants to be

.igge, more conservative and avoid brittle behavior even when the clad were to be cooled to room temperature,a i

I 129 limit would be reasonable,7he Regulatory Staff in their concluding statement compared various onw that measures of oxidation (page 90) and concluded that a 17% tof al oxidation limit h satisfactory,if calculated by the Baker-Just equation. The Consolidated National Intervenors in their direct testimony (Exhibit seed 1041) Indicated satisfaction with the Ritsenhouse criteria, and as argued by the Regulatory Staff.

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that the 17% oxidation limit h within the Rittenhouse criteria. Thus a remarkable uniformity of opin rent seems to exht with regard to the 17% oxidation limit, tie.

None of the resetor manufacturers agree 6 with the Stafra proposeo stipulation of a 2200*F maximum and calculated temperature (Cor,cluding Statements and Responses to the Staff Concluding Statement).

Westinghouse proposed a maximum calculated temperature limit of at least 2700*F; Combustion

. he Engineering and the Utility Group agreed on 2500*F as the peak allowable calculated temperature on the

.3 basis that much of the data on oxidation and its effects stops at 2500*F. Babcock and Wilcox suggested she -

more conservative 2400*F as the peak calculated temperature to be allowed, presumably because l

, 3e "significant eutectic reaction and an excessive metalto-water reaction rate would be precluded below w

240(fF,"(conch 6g Statement, p.242.) General Electric argued strongly that the limit should not be eta reduced Io 2J' O*ff;that 2700*F is really all right as far as embrittiement is concerned, but that the Interim has the Acceptance Criterion value of 2300*F should be retained. In addition to being consistent with their ceg expressed desire not to change any of the criteria, the GE recommendation of retaining the 2300*Flimit is j

intended to ensure that the core never "gets into regions where the meta!-water reaction becomes a serious,

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concern."(Initis! Closing Statement, Vol. 2, p. M-49.)

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The bases for the higher temperature recommendations of the reactor manufacturers are: (1) c cuhtlons of the stresses imposed on fuel rods during the LOCA and particularly during quenchi

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(2) measurements of the stresses and strains that oxidized zircaloy tubes can withstand, and (3) q oxid.

tests of oxidized zircaloy tut es to deterrnine their resistance to shattering. The calculations are imp are c to both items (2) and (3) because they tend to show that the greatest stresses that the rods would fract encounter are these of the thermal shock during quench.The thermal shock streu calculations are usefulin Anot s

indicating the general order of magnitude, or limit, of these strenes. Taken as a whole, howver, than not provide a confide' t basis for an securate number. Combustion Engineering calculates a peak thermal n

these stress of 24,600 lbs/sq in: General Electric-38,000 lbs/sq in.; Babcock and Wilcox-23,000 lbs/sq in.: a core Westinghouse-36,000 lbs/sq in. in Exh.1078, later corrected to 3500 lbs/sq in, in Exh. I151. Some of t Wilc< i difference between various values may be due to different assumptions about the hett transfer coeffic has i during the replo quench. (Trans. 21,619 20) local Other stresses on the fuel rods that have been calculated include those from circum 1

lead temperature variation, internal gas pressure, rod-rod interaction, assembly restraint, and cross flow, a sitt f Westinghouse's rod to rod interaction seems to be the same as GE's circumferential temperature v snore and aside from the thermal shock stresses, these seem to be the largest calculated (1000 to 2000 ps!)

is some lack of certainty as to just what nature of stresses would be encountered during the LOCA. As an hypo example it is stated by Westinghouse, Exhibit 1151, page 16 2: "At the inception of the LOCA,a pressure part.

T wave passes through the system and imposes a dynamic loading on the fuel assembly. However, at this tim caleu l the fuel rod ductility has not yet been reduced by oxidation and hence this blowdown load is not of F

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interest to this analysis" It seems possible that such a load might distort the ductile fuel rods and leave went them in a state that would aggravate the later rod-to rod interaction. The only point in mentioning this the o possibility is to draw attention to the fact that it may not be possible to anticipate and calculate all of the oxids stresses to which fuel rods would be subjected in a LOCA. Although we believe the calculations of thermal cessa-shock stresses are worthwhDe and informative, we agree with the regulatory staff that they are not

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sufficiently well defined to depend on for regulatory purposes, chest.

Since the principal stresses calculated for the fuel rods were those of thermal shock during quench.it isthat natural that in the testimony considerable reliance has been placed on the demonstrated ability of the i

j cladd exidised clad to withstand a rapid quench. General Electric in Exhibit !!22, page 3 3, showed a plot of I

El data indicating that no zircaloy tubes shattered on quenching if they were oxidized to less than 17% orif that i !

their oxidation ternperature was less than 2700*F. Additional data and plots of others did not change this Critti conclusion. However, many of the oxidation tests of zircaloy have involved quick quenching all the way

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from the peak temperature at which oxidation took place. Scatena pointed out (Exhibit 1122, page 2 4 Pe8ee inters;

" the time spent in the alpha + beta transition region car: whance embrittlement."Scatena attributed the

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additional embrittlement to precipitation of zirconium oxide along grain boundaries, although the Regulatory Staff in their concluding statement (page 83) indicated that a more likely explanation is the amen j D!

precipitation of alpha zirconium in the grain boundaries whDe the bulk of the materialis stal in the beta fusei!

phase. This explanation is fortified by Hobson's observation of pronounced incursions of alpha phase into gener !

the beta phase for specimens exposed at 2400 and 2500*F. (Exhibit 1126, page 11). One bit of exida experimental evidence on the effect of slow cooling is given by the ANL experiment number 3, as discussed 4

into ;

by Scatena, page 3 7. This sample remained at a temperature between 1500 and 1600*F for several ECCS minutes, and failed upon final quenching, although the amount of oxidation was less than that at whlen exten faDure was expected. The transition temperature from beta to alpha is in the range of 1500 to 2000*F.and calda judging from the PWR FLECHT tests (e.g., Exhibit 150, page C.15), the cooling rate experienced through Exh.

this temperature range during a LOCA might be quite slowas compared to the fan uench typical of most s

Ti laboratory tests. (General Electric stated that the quench from peak temperature in BWR would be fast, but sugge-the curves dispisyed in evidence do not provide quantitative data in this regard.) The uncertainty criter. ;

introduced by the effect of cooling rate in the temperature region above !$00*F casts some doubt upon the f

coola' applicability of some of the quench tests that have been carried out. Nevertheless we find the quench Consi results encouraging in that they provide assurance that the 2200*Flimit is conservative.

situat Our selection of the 2200*F limit results primarDy from our belief that retention of ductility in the i

funda.

zircaloy is the best guarantee ofits remaining intact during the hypothetical LOCA.The stress calculations, objee-the measurements of strength and flexibility of oxidized rods, and the thermal shock tests all are reassuring, as a c but their use for IIcensing purposes would involve an assumption of knowledge cf the detaued process This :,

taking place in the core during a LOCA that we do not beDeve isjustified.

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l All of the reactor manufacturers except Combustion' Engineering objected strongly to the application of

enetung, oxidation and maximum temperature criteria to the hottest spot of the cladding,especially to regions that equench are calculated to have swollen and burst open. It was argued that this represents only an extremely small iportant fraction of the reactor core and that even if this small amount became fractured it would do no harm.

s vould Another contention is that the criteria are so conservative that even if these hot spots were oxidized more usefulin than the criteria allow they would remain intact. it is aho pointed out that if any damage were to occur to they da these hot spots it would happen as they were being quenched at about 1000*F after the great bulk of the thermal core had aheady been quenched and was at a temperature in the range of 300*F. As put by Babcock and Wilcox in their concluding staternent, page 255, it is difficult to conceive of conditions in which the core in.; and te of the has been substantially cooled, the transient completely anested for the bulk of the core, and yet find a cfficient local fragmentation of a negligibly small portion of the core causing a significant thermal effect.**

These are valued argurnents, supporting the thesis that damage to a small part of the core would not firential lead to more extensive damage. However, they must be recognised as opinions as to what would happen in ess flos a situation that has never yet occurred. Others are not so sure that a local faDure would not be propagated ariati:n, more widely throughout the core. (CNI, Exhibit 1041, p. $.67.) In view of the lack of experience in this

~

1).There hypothetical sf astion, we think it prudent to apply our criteria to all of the core and not to exempt any part.

A. As an There remains the question of the extent to which oxidation on the inside of burst cladding should be pressure calcuhted. The two principal pertinent experiments are the in-pue experirnents FRF.! and FRF.2. FRF.1 his time,

s not cf went to only 1800*F, but the oxidation on the inside of burst c.ladding was approximately the same as on snd leave the outside in the region of the burst area. FRF.2 went to a higher temperature and had significantly less oxidation of the inside than the outside. However, this experirrent was steam limited, as indicated by ning this all of the cessation of steam effluent during the time when the peak temperaures occurred.Thus there is no basis for f thermal assuming anything but equal access of steam to the inside of the burst tubes, as stated in the criterion.

(3) Maximum Hydrogen Generation. The calcuhted total amount of hydrogen generated from the are not chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metalin the cladding cylinders surrounding the fuel, excluding the s

cladding sunounding the plenum volume, were to react.

)

.ne a plot af Discussiort The object of this criterion is to ensure that hydrogen would not be generated in amounts I?? orif that could lead to explosive concentrations. The criterion is essentially the same as Interim Acceptance aanpe th5 Criterion #2, but is more explicit in detaDing how much of the aircaloy is to be used as the base for the one I

the way percent calculation. This criterion is'non. controversial. However,its purpose was misunderstood in some pape 2 4) intervenor and vendor analyses. It has nothing to do with the need to retain strength of cladding.

auted the (4)Coolable Geometry. Calculated changes in core geometry shall be such that the core remains

. ugh the amenable to cooling.

Discussion. If there were no emergency core cooling after a LOCA, the core would probably eventually i.e a the the beta fuse together into a large mass with insufficient external surface area to allow the fission product heat generated within it to be transfened away. it.termediate steps in arriving at such a state might be the hase mto

,e bit of oxidation and melting of the tirealoy cladding, allowing the uranium dioxide fuel pellets to fall together Jisassed into a heap that would be difficult to cool.Other difficulties have been envisaged, some the result of an

.it several ECCS that was not fuDy effective. Those most often mentioned are the bulging of the cladding to the at which extent of closing off the coolant passages and the possibility of zircaloy becoming embrittled by steam u'F.and oxidation and shattering during quench, allowing the fuel pellets to fallinto a heap.(Exhibit 1041, see.7; J through Exh.1007B).

41 of most Thinking largely of the latter difficulty, Combustion Engineering in their Concluding Statement i fast, but suggested that, in view of the restrictions on cladding oxidation placed by criteria 1 and 2, a specific criterion on coolable geometry is no longer needed. $1milarly, Babcock and Wucox omitted a criterion on

.<citamty upon the coolable core geometry from their proposed criteria in appendix A of their Concluding Ststement.

.c quench Considering all of the required features of the evaluation models, we are inclined to agree that, for any situation that we have been able to anticipate, this criterion should be superfluous. However,in view of the

.:1) m the fundamental and hhtorien! Importance of maintaining core coolability, we retain this criterion as a basic objective, in a more general form than it appeared in the Interim Acceptance Criteria. it is not controve L ulations.

as a criterion, although the extent of flow blockage resulting from clad swe!!ing is a matter of controversy.

g.

This subject is dheussed in the section on Required and Acceptable Features of the Evaluation hfodels, r

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3 (5)Long Term Cooling. After any calculated successfulinitial operation et the ECCS, the calculated h

core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long lived radioactivity remaining in the core.

Discussion. Although most of the attention of the ECCS hearings has been focussed on the events of k{r the first few minutes after a postulated major coolingline break,up to the time that the cladding would be

,y cooled to a temperature of 300*F or less, the long term maintenance of cooling would be equally p.

Important. The intent of this criterion is self evident and it is non controversial.

III. REQUIRED AND ACCEPTABLE FEATURES OF THE EVALUATION MODELS

[,

An evaluation model is the calculation frarnework for evaluating the behavior of the rer*or system

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during a postulated leas 4f coolant accident (LOCA). It includes one or more computer programs and all ether information necessary for application of the calculationt.1 framework to a specific LOCA,such as 7

c Ti, Q

mathematical models used, assumptiore included in the programs. procedures for treating the program input and output information, specification of those portions of analysis not included in computer

,,T, programs, values of pararneters, and all other information necessary to specify the calculational procedure, pc Loss-of coolant accidents (LOCA's) are hypothetical accidents that would result ' om the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double ended

[g rupture of the largest pipe in the reactor coolant system.

pf m A. SOURCES OF HEAT DURING THE LOCA For the heat sources listed in paragraphs I to 4 below it shall be assumed that the reactor has been j

operating continuously at a power level at least 1.02 times the licensed power level (to allow for such uncertainties as instrumentation error), with the maximum peaking factor allowed by the technical specifications. A range of pour distribution shapes and peaking factors representing power distributions

,i that may occur over the core lifetime shall be studied and the one selected should be that which results in

?

the most severe calculated consequences, for the spectrum of postulated breaks and single failures analyzed.

3 I

1. The /nitialStored Energy in the fuel. The steady. state temperature distribution and stored energy in 7

the fuel before the hypothetical accident shall be calculated for the bum up that yields the highest alculated cladding temperature (or, optionally, the highest calculated stored energy.) To accomplish this.

the therrnal conductivity of the UOs hall be evaluated as a function of burn-up and ternperature, taking I

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i into consideration differences in initial density, and the thermal conductance of the gap between the UO:

and the cladding sha!! be evaluated as a function of the burn-up, taking into consideration fuel densification and expansion, the composition and pressure of the gases within the fuel rod, the initial cold gap dimension

y with its tolerances, and cladding creep.

2. Arsion Heat. Fission heat shall be calculated using reactivity and reactor kinetics. Shutdown 4

reactivities resulting from temperatures and volds shall be given their minimum plausible values including 94

' 8 allowance for uncertainties, for the range of power distribution shapes and peaking factors indicated to be ik studied above. Rod trip and insertion may be assumed if they are calculated to occur.

l 3.Decey of Actinides. The heat from the radioactive decay of actinides, including neptunium and

+

plutonium generated during operation, as well as isotopes of uranium, shall be calculated in accordance

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with fuel cycle calculations and known radioactive properties.ne actinide decay heat chosen shall be that appropriate for the time in the fuel cycle that yields the highest calculated fuel temperature during the U

LOCA.

4. fission Product Decay. De heat generation rates from radioactive decay of fission products shall be s

assumed to be equal to 1.2 times the values for infinite operating tirne in the ANS Standard (Proposed

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American Nuclear Society Standard " Decay Energy Release Rates Following Shutdown of Uranium-I Fueled Therrnal Reactors" Approved by Subcommittee ANS 5, ANS Standards Committee

• October 1971). The fraction of the locally generated gamma energy that is deposited in the fuel (including the s

cladding) may be different from 1.0; the value used shall be justified by a suitable calculation.

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5. Metal-Water Reaction Rare. The rate of energy relene, hydrogen generation, and cladding

C** dated oxidation from the metal / water reaction shall be calculated using the BakerJust Equation (Baker, L,Just,

.wd for W L C.," Studies of Metal Water Reactions at High Temperatures, III. Experirnental and Theoretical Studies of the Zirconlum-Water Reaction," ANL.6548, page 7, May 1962). The reaction shad be asumed not to I

be steam limited. For rods whose cladding is calculated to rupture during the LOCA, the inside of the

'"8 cladding shall also be assumed to react after t!:e rupture. The calculation of the reaction rate on the inside I "' '

of the cladding shall aho follow the BakerJust equation, starting at the time when the cladding is calculated to rupture, and extending around the cladding inner circumference and axially no less than 1J

[

inches each way from the location of the rupture, with the reaction assumed not to be steam limited.

13 del.5

6. Reactor Internair Heer Dor sfer. Heat transfer from piping, vessel walls, and non. fuel internal hardware shall be taken into account.

i

4. tor system

. 7. hessurized Water Reactor himary toSecondary Heat Thmsfer. Heat transferred between primary rams and all and secondary systerrs through heat exchangers (steam generators) ahaD be taken into account. (Not i

s'A, such as applicable to Boiling Water Reactors.)

l 1

tw program

.n ownputet DISCUSSION

'((

f

1. 7he Initial Stored Energy in the fuel During reactor operation just before the accident is assumed to happen the temperature of the UO would depend strongly on the radial position in the fuel pellet.The y,p, pes m temperature would be highest at the center of the rod and much lower at its outer surface next to the vbie.cnded zirealoy cladding. The heat generated anywhere within the UO: rnust be conducted outward through the remaining UOs, and all of heat generated within the UO must pass through a poorly conducting region at the boundary between the UO: and the zircaloy, ca!!ed the pp. The poorer the heat conductivity,either within the UO: or in the gap, the higher would be the average temperature of the UO.The stored energy is defined to be the energy that would be released by the UO if its temperature were reduced to that of

• % heen the zircaloy cladding. Just before the postulated accident the average of the radial temperature distdbution i

veh.

of the 00 in the hottest region would be near the melting temperature of zircaloy, and since the heat

.wal -

transfer inte between the zircaloy and the water would drop abruptly within a fraction of a second aftei the si Autams initiation of the accident, the stored energy would be an important heat source which could cause a sharp t: ersuits m rise in the temperature of the cladding during the blow down period.The increase in zircaloy temperature

% anahird resulting from the redistribution of stored energy tends to contribute to the calculated peak temperature of J entre m the aircaloy..

+e hitheSt Although the importance of stored energy was recognized at the time the Interim Policy Statement was

Th5h hn-no general rule was fonnulated. Rather, the rnethods for arriving at the initial fuel temperature were

~.*e 18Lm8 specifle in the descriptions of the individual evaluation models. A considerable arnount of evidence

(@h (

y regarding tored energy has been introduced in these hearings, indicating a wide diversity of approach.

"" *e i Ur (Exhibit 1113, p.1010 & ff). Furthermore, a new factor affecting gap conductance and therefore stored

'

• 5d - 8 ' * *

(0 \\ - energy nas oeen orougnt io agnt (fuel densification: see transcript 15,242 3: 15,296.) We therefore believe

• (d'

'a

• that it is proper to require a more urdform approach to the calculation of stored energy, and to require that

• ' J' *

• the evaluation of gap conductance and stored energy be made on a case by. case basis.

^ Nd*8 A surranary of several reviews of the thermal conductivity of uranium dioxide is given in Exhibit 1113

. auJ i.+ N pagea 10 23 to 10 29. The views of the Regulatory Staff are given on page 10 29 of that document, and the i

Commission believen that these views are reasonable and can serve as a reference for judging the adequacy of the values of thermal conductivity used by licensees.

N'* * '"d

'" *' 38 "d The steady state pp conductance depends on the thickness of the gap (ranging from 0.000 to 0.012

1. W **'

inches) and the thermal conductivity of the gas filling the gap.lf the gap is closed its thermal conductance

• 8 *"" 8 "

• depends upon the pressme of the contact between the UO: and the zircaloy. Thus any change in fuel density, whether by thermal expamion, fusi dersification, or swlling w!!! change the gap conductance. The

#'u N pp dirnensions wDI also change if the external pressure causes the tircaloy to creep inward. As fission gas is A

' #* '* d released the composition of the gas in the gap changes from pure helium, which has a high thermal 1.

conductivity, to a mixture with xenon and krypton, which have low thermal conductivities. In the past,6 Babcock and Wilcox and Westinghouse have been calculating the gap conductance as a function of bum-up.

Babcock & Wilcox (transcript p. 5536 and Concluding Statement, page 70) included all'of the above features except fuel densification and cladding creep. This latter neglect was conservative, resulting in lower gap conductivities and yielding the highest stored energy at the end of fuellife. Westinghouse has provided 1101 1

\\

been much higher at the end of fuellife,similar calculations, but inc r gap conductances have I

advent of the fuel densification problem, new fue r

urthermore, with the stored energy on a case by. case basis.

d neceuary to evaluate in their initial closing statement (Vol.II, K 1 through K I 8

(3 ory Staff (pp. 60-70) and their argurnent depends upon the fact that the Axed va a

t1 ap conductance. Much of suitability. Their argument also depends on the fa Pi re t 1001 affirmed its x

seems to be directed toward PWR's, and that BWR's were seldom mentioned scuulon in the record much was learned during the course of the hearings, both from new expe ca ommhslon notes that e

and from further study ofinformation that snay have been previously availab to culational results th expressed in 1101. The Commission also notes tha result the Commission t an on those encased in zircaloy and that the uranium dioxide is subject to the fissioning cc ue e with uranium dioxide variatiom much the same as in the pressurized water reactors. There is therefore fac process and temperature bu the basic physical processes are similar in the two reactor types and th C every reason to believe that exempt boiling water reactors from the provislom of this section The importomm e

P-LOCA has been well established, as has its dependence upon the stead anes of stored energy in the ample evidence that the initial gap conductance of fuel rods in boiling waterp co Ass (Exhibit i113, pages 1011).

reactors can vary widely, the

2. Fission Heat. This represents no change from previous practice

3. Decay o/ Actinides. The decay of actinides (isotopes of U Np or Pu) p Bak contribute a significant amount of heat, as much as 10% of that of the fissio by i

ro uced in the fuel can by the Regulatory Sta:T would have required that the rnaximum decay heat fr con e rule proposed objected to this in their respomes to the Regulatory S b l rept om actinides be usedin the De and Babcock & Wilcox d

physicaDy irnpossible to have the maximum actinide decay heat simultane and j on the grounds that it is energy and the maximum peaking factor. The rule adopted by the Commission r

,I e maximum stored 75 7 i and is in accord with the ptinciple ofimproving the realism of the calculation ecognizes this situation,

4. Fission hoduct Decay. The major source of continuing heat producti zirer {

them by neutron absorption. The Regulatory Staff reviewe narm '

on after reactor shut.down is abou ro uced from pp. 3 24 to 3 26, and concluded that the tentative ANS standard gave thI{

i shou l

available information. The ANS standard h derived from the review and c i

resentation of the largel l the region of most interest to the ECCS (up to 1000 seconds after sh t d re which,for Westi experimental work done since the Shure review was pub conel a aso comidered committee recommended uncertainty limits of +20%-40% The Staff concluded th starva

, page 22 3). The ANS uncertainty factor of tempi Statement they were providing a small margin above what th Ridge e Interim Policy Although no new experimental work was presented during the hearings new accun actor to be.

I Intervenors (Exhibit 1152, pp. 2.2 2.6). England's work evider computer calculations an ph j e onsolidated National Us l capture in the fission product chains. As originaDy prese puter calcuhtion and theBa of neutron abose the ANS prescription, particularly for high neutron fluxes and fuel b arge deviations 0"T'P' !

For e>

Exhibit !!!3, p. 22 5L However a series of errors in both input data and the calcu ee, for example, Exhibi l onal program were Dl

• K. Shure, Tiuion Product Decay Energy"in WAFD BT.24, Dec.196 t, pp.1 17.

3 equath International Conference on the Peaceful Uses of Atomic Energ

~

Geneva 1958.

e Second (J.N.

of180' o.

, pp.49 54, United Nations.

6.1 Primar,.

1102

,,k..._..--.-

found both by England (Exh.1178, p. 7) and in the course of a review by Shure (Exhib aave markedly reduced the deviations found by England's approach. With the corrections ma re, with the deviations found by the England approach from the ANS standard we nowhere greater tha 13 evaluate pnerally snuch less. (Ill3, at 2215). In addition, there is the possibuity that the selection ofin (fission product yicids and decay energies) may not have been the best (1113,22 8 and 22 9).

60 70)and While England's approach is a valuable contribution,it is only one piece of work out of a rec:rd to the record;furthermore it presents no new experimental determinations. On the basis of the re

e. Much of proceedings, however, one is led to believe that the ANS standard curve may be about 5)Elow in e approved j

nglon of principal interest, namely, zero to fivs minutes after shut.down. England's revised values flirmed its within the previously expressed limits of uncertainty, and to the extent of the credence given th the record calculations, they tend to nanow those lirrits of uncertainty. At present it appe : to us that the notes that for of the ANS decay heat formula fairly represents the uncertainty and does not provide any inal results that uncertainty. It is still conservative.

VOL.

{

ommission There is some margin provided, houver,in the prescriptio requiring tha the reactor shall have a on those considered to have operated continuously at 1.02 times rate power, with the maximum m dioxide factor, for an infinite length of tinv.. The exact amount of rgin is uncertain, and it will vary with time, moerature but it is probably in the range of 5 to 15% (Exhibit 1137, pp.113 to 5:and Staff Concluding S g

aljeve that W.Il4 reason to Considering all of the above, the Commission believes that the prescription of A.; + 20% for the rgy in the fission product decay heat is reasonable and should be continued.

id there is

5. Metal-Water Reaction Rare. A rate equation is needed for the calculation of hydrogen y widely, the extent of cladding oxidation, and the heat seneration from oxidation of the zircaloy by steam Baker.Just equation has ao far been used in evaluation models for this purpose.This equation w y

by Baker and Just from their measurement of the rate of oxidation at the melting point of zirco fuel can prf

.I cordunction with lemmon's and Bostrom's data at lower temperatures. The equation is a straig representation of a plet of the logarithm of the reaction rate vs. the reciprocal of the absolute tempe A

ne slope of this line is the activation energy, and depends in an important way on the single point o A Wiscox and Just at the inetting point of zirconium.

that it is The Baker Jusi equation has been criticized extensively, principally on three bases. (Ex m stored 75 78 & 137140; Exh. 258, p. 40; Exh.1122, p. 211.) One is that the data point at the melting ituation.

, zirconium (3365'F) is quite unrelated to the phenomenon of oxidation at the temperatures ofint namely,1800*F to 2200*F. The second is that more recent data yield lower oxidation rates,

-dowa 6 about 2300*F and above. The third basis for criticism is the theoretical one that a single ac cd from i

should not be expected for different crystal forrns at different temperatures.

it1001.

Recent proposals for new rate equations (with the exception of Klepfer's, Exhibit 258, page i of the largely upon the data of Hobson and Rittenhouse, Exhibit 509, An example is the derivation ih. for Westin5 ouse in Exhibit 1078, pp. 75 78 and pp.137140.

h i

However, Babcock and Wilcox in their

{

(rom a concluding statement (page 241) suggested that the Oak Ridge experiments may have suffered from s isidered starvation, and also suggested that the temperatures may have been higher than assumed from the furnac is ANS temperature rnessurements. These two observations, which may be jusefied, cast some doubt on the Oak eas an Ridge results and possibly others of sirrn*e origin. The required measurements are diffleult to make Policy accurately and there is usually a wide spread in the results, especia!!y from different investigators. This a

j l

rAdent, for example, from figure 34, page 137, of Exhibit 1078, where values of the oxidation thickness stations e plotted against a measure of the time-ternperature history, ational Until new data are obtained and present doubts are resolved we be!Ieve it best to continue the use o in and the Baker.)ust equation. It apparently gives about the correct results at 2000*F, and although it estron over predicts the oxidation at 2200*F, thb over-prediction does not exceed the range of the dats available.

utions

. For example, the Baker Just squation fairly represents the oxidstion data depleted in fire 1, pag i

ample.

Exhibst 258, op to about 2372* F, well above our allowed maximum temperature.

, were.

here is evident need for new and better experimental data to resolve this issue and to provide a ra i

equation with a more representative activation energy, if we were to establish such a new equatlod at thi t(

time, we would choose one that provides about the same calculated oxidation over the temperatur

• *g of 1800 to 2200*F as the Baker.Just equation, so that its continued use makes little practical differenc

6. Reactor Intemalt Hear 7)anifer. A substantial amount of heat is stored in the mets! parts of the g

pimary circuit of a reactor system. Thl hest would be transferred rather slowly to the coolant and would t

1103 i

is I

_ _.. ~ _. _ _ _ _ _ _ _...

~.

1' the Staffs concluding statement objected to its broade c n their response to signi requirement should be applied to all evaluation models in the interest of havin ss on rules that this LOC phenomena comprised as is reasonable.

c in the important and does not represent any change from present p

• j l e

e as const estirr.

B. SWELLING AND RUPWRE OF Tile CLADDING AND FUEL Elect ROD THERMAL PARAMETERS whic!

• The Each evaluation model shall include a provision for predicting cladding s regio.

consideration of the axial temperature distribution of the cladding and the di rom P.21 the inside and outside of the chdding, both as functions of time To be acceptable I

anyotherapplicablevariables.

n with :

calcuhtlons shall be based on applicable data in such a way that the deg e swelling and rupture taken supture are oot underestimated. The degree of swelling and rupture shall be ta block calcuhtions of gap conductance, cladding oxidation and embrittlement, and hydrog conductance and other thermal parameters u functio Conel accou comn dependent variables. The gap conductance shall be varied in accordance with core.'

ons and St DISCUSSION. In the postulated LOCA the reactor system pressure would in mc aoon fall below the pressure of the helium and fission gases within the fuel rod snade pressure would exert an expansive force on the cladding. At the same time, as the M

dropped, the temperature of the cladding would increase rapidly, decreasing inad cladding. At some time during the LOCA the yield strength of the zircsloy mig e

sempe i tensile stresses exerted by the differential pressure, and the cladding would then s the di For example, Babcock and Wilcox, using the evaluation model of the Interim P aps burst.

radiaB 3

seconds after the maximum size cold leg break. (Exhi

.spacin I

adjace the - differential pressure would be about 200 psi and the cladding temperature Fr Westinghouse, in a simihr calculation, conservatively estimated that 25% of the

~

u II" t' (Exhibit 1078, pp.D48 and D49.) Combustion Engineerin urst (ExhR P.D4 in both cases, as judged from the blockage, they were c oc age th' #8 oftlw 1

every fuel assernbly. The major difference between the pressurized and unpressur ry usualh s

rnig6.1 he expected. (Exhibit i14A, sec. 5, using material from temper II'* I provide a longer period of assured effective cooling of t From i 8

and & l temperature of the cladding is not so great. Furthermore, the pressure within the fuel ro b,

Ib' so that ballooning of the cladding would not be expected to occur during the blowdo e ow, experie

p. 2 24.) General Electric offesed one calculstion for a 1967 product line BWR re8Pect,

'1 temperature was 2105*F. (Exhibis 1148, sec.F.) Using some of the assumptions mad 8"CI"' {

hottest bundle would perforate. CNI, using the probably x

t ApPent It 35 "

communication between the hot spot and the fission gas plenum at the top of the fuel rod and &

22% of all the fuel rods in the whole reactor would rupture. They said that this cornp where a 8.2-1, reference is made to a calculation for a Bolling W as den.

D expected to rupture by the tine the ECCS core sprays came on,with 75% of the pins inches.

ultimately.

riumbes and channel blocking in the BWR FLECHT test Is.2. The c obsem

'" ""Y 1104

-a

=-.:

VQ';.

'E LOCA, even for the " hot" bundle. (Exhibit 1148, p.P j,

09 g g, was aperoximately 2250*F and 39 out of the 49, or 80%, vi the heater rods perforated.

pp.53 54.)

g, From the above it is obvious that, when the course of the LOCA is calculated esthnated to occur, in abundance. Three of the reactor m l

Electric, and Westinghouse) have objected to applying the ECCS criteria to which swehg is calcu!sted (Concluding Statements). On the other hand, Comb "The extent of oxidation thus limited"(by CE's version of the criteria)"for the w t

region assures that the remaining cladding will remain intact in a LOCA."(CE Conc

p. 213.) The Consolidated National Intervenors in their Concluding Statement were Wu' W taken into consideration, but were concerned that there me t8""

ceW 88"'

I Concluding Statement, recomm:nded "that no changes

- #88"' #

e.

• e"8"' * '

account for effeus of clad thinning and inside oxidation near the locations of fuel ro commenting that "it seems most appropriate to continue to focus on the other more tha

'*** b M cote."

• d'* '"

Some of the objections to inctusion of the effects of cladding awelling and perfora in more detail below, but the Commission sees no merit in ignoring these consequ rnade with an officially approved evaluation model

• '8 #

i in which aircaloy tubes were pressurized and heated at sp M

the degree of smelling. Many of the tests have been made a e of the an radially, Other tests (multirod terts) have been made with the airconjura tubes arra pand

• )

,specing simuer to that which they would have in a reactor, in this kind of test, interfere

• N adjacent rods can limit their expansion.

From these tests, curves can be constructed relating the internal pressure to the the tubes bunt. Such curves have been published by ORNL (Exhibit 1007b fig 1) Gen (Exhibit 137, fig.9), Combustion Engineering (Exhibit 1066, fig.2.4) and Westinghous c

p. D.64. The amount of swelling that the cladding would experience before ruptu the temperature at which it would begin to deform,in dependence on the ductility of th usually plotted as a function of the internal pressure, whic flow blockage as a function of rod internal pressure (e.g From this type of information, together with detaDed information on the internal pre l

and their temperature history, the degree of expected clad swelling for each reactor can The data from the rod bcrst tests show a great deal of scatter, particularly in ;he

. c, experienced. (See, for example, fig. 2.5 of Exhibit 1066).The greatest controveny, howeve

. ese respect to interpretations and predictions of the resulting blockage to coolant flow,(See s.,

concluding staternents:

Appendix B: CNI Ch. 5, sec. A;ECCS Utility Group,p. 50; Co i

u a.

. e a.

It is expected that variat_ ions in cladding thickness, fuel pellet properties, gap j

=.,

and the texture of the zircaloy because ofits anisotropic character would go into t

= aw where along the length of a fuel rod the perforation would occur. (Transcript 11,51518).

.* 4 The swoDen and perforated region is expected to be about !$ to 3 inches long, and io wt as determined by the above variables over a length of relatively uniform temperature of f

• = = =

inches. (Trans.12,701; Exhibit 1066, fig. 2.7; Exhibit 1144, p. 5.2.) Thus it is not e asaw number of adjacent rods would have their max! mum swelling in the same plane. T observed to date in any rnuftirod experiment containing 16 channels or more has bk on any horizontal plane. (Trans. pp. 9166 7).

1105 L

I

.-y_.

r 4

the core. wide now area reduction in the plane of grea i

. (','

, p.212 namely,that i

flow channel reductions, over perhaps a 4 x 4 array of fuel rods would n not exceed 60% and that local

. studies th-

. calculations of blowdown and ECCS heat transfer considered elsewher As shown by o;t above estimated fuel rod swelling and rupture would hot rend (6)The se s

ow area, ss on concludes that reaches th.

i fuel pellets and the cladding (the gap conductance e uncoolable.

this rule,i nce between the the stagnar Interim Policy Staternent has the capability of calcalating the change of ths proved under the The C4 during the course of the postulated accident and the effects of this change eametry of the cladding believes it thermal radiation terre in the calculation of the gap conductance a term th ey include a i

conservatis:

temperatures and for large gap dimensions (Concluding Statement of at becomes important at high from the e temperatures the effect of prior aveuing would be to decrease the gap co dthe Regulato fact that th

.there were an appreciable amount of fission product gas mixed w n uctance drastically,especiallyif one (seen fo Exhibit i113, pp 10 21.) A decrease in gap conductance during the blow General

'cladens from the fuel peUets, seducing the amount of stored energy remo ed nsulate the stated that keepias the temperature of the claddinglow ducirs this time.(Exhibits 1151 p ur ng blowdown, and evalusuon r v

p.103.) For pressurized water reactors " uring refill and especially during reflo

,sec.10; 1137 (and as of '

o peUets to the cladding would become important for the rods with the swolle n rom the fuel Statement, y the rate of heat removal from the cladding. The greater stored onessy th a y exceeding

• GE strongl.

would tend is cause the peak temperature of the cladding to exceed that obtained b penetradon -

radiation term and so do not properly calculate the peak tempe y present evaluation purpassa." ((

o ave the that the sequ '

For boiling water reactors, the change in gap conductance from clad swelli larply en th awelling of the cladding would be expected to have a would have little PWR's.

owever,the Go eral i and radiation. It is true that GE aported a calculation leading to their conclusion spray cooling Concludlag 5 transfer is little affected by clad bulging (Exhibit 132, p. D-55) but that cal l i radiative heat g

and that prop the doubtful asn'unption of a uniform circumferential temperature for each (g

cu at on may have contained proposed by a BWR FLECHT experiment were ched as evidence for th "in a causerva b

have theespot e r.

temperature measurements were made at the positions of maximum bulging W b li ng, although no effects of sict assessments need to be made of these effects,

. e e eve that additional done using the In addition to the primary heat transfer effects of taking lato conalderation G As sealed a by the steam. Higher temperatures would lead to increa and rupture of a BWR,as sisi.

the effect mist increase in temperature, and the opening in the cladding would allow caldation o u e to s further make sum the increasing the calculated temperature.

n the inside, again ngard to saking would not have a significant effect on cladding ternperatur rods, we have a clad swelung temperatserihr In IDCA evaluatione (p. Ig3). The Regulatory Staff challenyd the assertion of insi not be considered the highest essr thinning or for radiation from the fuel to the cladding, and ad gnificance in their into considerath.

unt for clad Westindmus peak temperature can be algnificant.

a a SI'Fincrease in to believe that }

With regard to the related subject of transient gap conductance calculations Ba a

the transiset wc number of reasons why they contend that the problem of transleht gap condu Staff). However, cox save a calculation and not a significant problem in any physical sense."(Concludin localized damage e

Response to the Concluding Statement of the Regulatory Staff,pp.48 5g )These reas require that the a 134 193; initial stored energy is overestimated. (2)The peak powsr iensity that is a are:(1)That the be taken into see aDows. This would be espeelany important locause, if tr ssumed is higher than can be in the Coast technical issues. !

on during blowdown. (4)B&W interpreted the Staff statemmt on page 1711 of Exhibit !

Section A CNI kr rupture oscDistory flow during reflood would greatly improve reflood heat transfer,in our dis to mean that ajt would appser heat transfer we urge caution in making this interpretation, ut least for the presen in soma LOCA ci on reflood there are differen r ng to B&W and that thesisa 110g

\\

-c

' at I

studies the.:alculated peak temperature is insensitive to variations in gap conductance. However, as pointed d tha local out above, their calculation did not include radiative heat transfer between the UOs and the cladding.

s shown by (6)The requirement that no heat sha!! be considered lost from the fuel rods during refill until the water

, flow area.

reaches the bottom of the core was said to be extremely conservative. Although this is not a requirement of icludes that this rule, it makes little difference because there would be very little heat 1:ansferred during this period to

able, the stagnant superheated steam that would be pesent.(Exhibit 1001, at 3 36),

netween the The Commission agrees that items (1). (2) and (6) provide moderste degrees of comervatism, and d under the believes it probable that item (3) provides a substantial degree of conservatism. These are welcome he cledding conservatisms;it is intended that thre should be a margin to allow for extremes in statistical fluctuations y include a from the expected behavior of the systens, components, flows, cc' ling mechanisms, and materials. The tnt at high fact that the calculation is intentionally conservative should not,in the opinion of the Commission, keep Y).Atlow one from following the consequences of that calculation to its logical conclusion.

25Pecially if General Electric, in opposing any change in the Interim Polley Statement for Boiling Water Reactors, rans. 3536; stated that the " preponderance of the evidence in the record provides no basis fer changing the BWR msulate the evaluation model." With regard to clad bulging, they said "...the overwhelming weight of the evidence down,and (and all of the reliable evidence)it:dicates the flow blockage is not a concern for the BWR."(Closing

.10;1137, Statement, p.0 4.) With regud to taking into consideration clad thinning and oxidation inside burst rods,

)m the fuel GE strongly opposes the... suggestions...that this highly localized wall thinning and inside oxygen cxceeding penetration should form the basis of any aspect of the embrittlement assessments used for licensing er cladding purposes." (Closing Statement, p.M 22). On pages K.35 to K.40 of their Closing Statement they argued evaluation that the requirement for calculating changes in gap conductance during the LOCA can not apply to BWR's, at have the largely on the basis that the calculations showing the possible magnitude of the effects were made for PWR's.

. t have little General Electric further commented on " Clad Swelling and Rupture" in their Response to the ewever, the Concluding Statement of the Staff Vol.2, pp.5176. Some differences between our wording of the rule ta hg and that proposed by the Regulatory Staff are in response to some of the comments on page 51.The rule sa,

at proposed by the Staff was criticized by GE as vague and unworkable. Phrases such as " applicable data" and con wned "in a conservative way" were characterized as imprecise. The Commission believes that the reactor vendors ely on the have the capability to interpret the existing data in formulating adequate evaluation models that include the if the Zr 2 effects of clad swelling and rupture. For imtance, calculating the incidence of cladding rupture could be though no done using the methodology described in Exhibit 1144, pp. S.1 and 5 2.

additional As stated above, we believe that the effect of calculated clad swehg on heat transfer may be small for a BWR, as claimed by CE, although the evidence of this in the record is not very extensive. However small rupture cf the effect might be, we believe that ignoring it would be nonconservative, and so it should be included to ie claddinf make sure the evaluation modelis as comprehensive as the present state of understanding allows.With oa f rther regard to taking into consideration the calculated thinning of the cladding and internal oxidation of burst

4. attain rods, we have already ruled in connection with our discussian of the first two criteris ' hat the maximum temperature limit and the oxidation limit should be applied to the section of the cladding that would reach

'd swelling the highest temperstwe and be most oxidized. Therefore the estimated burst of cladding must be taken

.wsidered into consideration.'

x b their Westinghouse objected to the inclusion of clad swelling and rupture on the basis that there is no reason it for clad to believe that highly localized and limited rod cracking or severance during quench or at any other time in meresse in the trnsient would impair cose coolability. (Comments on the Concluding Statement of the Regulatory Staff). However, since there is insufficient evidence of a substantial nature to the effect that that such "cos pave a localized damage would prove harmless, the Commission must make the conservative choice.We therefore fact of the require that the most highly damaged places calculated to be found along the length c,f the fuel rods sha!!

3 " I"3; be taken into consideration.

6 That the in the Comolidated National Intervenors' Concluding Statement, ChapterV discusses some of the san can he.

technical issues. Sections A on Flow Blockage and C on Gap Conductance are pertinent to this section.In

.shlanon Section A CNI implied that the state of knowiedge of flow blockage is defielent, and they stated further:

• I '"P8N "It would appear that swe!!ing can and likely will yleid an unacceptable degree of coolant channel closure a that In some LOCA circumstances." With regard to the state of knowledge, the Commission recognizes that d

there are differences of opinion as to how flow blockage should be calculated from the basic swelling data, and that there is a high degree of scatter in the swehg data. Neither of these situations seems unnatural;in 1107

.-...-_.;.m...

~

p w..

fact they make it possible to understand better the degrees of conservatism in the different calculations.

Additional multirod tests will help to refine the calculations, but if proper regard is shown for the divergent views, the present situation aDows sufficiently conservative estimates. As to the possibility of "an unacceptable degree of channel closure,"the Commission has seen no hard evidence that this is at alllikely, and believes that the risk of such a circumstance occurring is acceptably small.

C. BLOWDOWN PfiENOMENA

1. Break Characteristics and Flow

a. In analyses of hypothetical loss.cf. coolant accidents, a spectrum of possible pipe breaks shall be considered. This spectrum shall include instantaneous double. ended breaks ranging in cross sectional area up to and inclur"ng that of the largest pipe in the primary coolant system. The analysis shall aho include t..

the effects of longitudinal splits in the largest pipes, with the split area equal to the cross. sectional area of f

thepipe.

b. Discharge Nodel. For aD times after the discharging fluid has been calculated to be two phase in

~<

composition, the discharge rate shall be calculated by use of the Moody model(F.J. Moody," Maximum

,b4 Flow Rate of a Single Component, Two. Phase Mixture," Joumal of Hear hansfer. Transactions of the American Society of Nechanical Er.gineers, 87, No.1. February,1965).The calculation shall be conducted with at least three values of a discharge coefficient applied to the postt. lated break area, these values 5

spanning the range from 0.6 to 1.0. If the results indicate that the maximum c1t i emperature for the t

t l

~

hypothetical accident is to be found at an even lower value of the discharge coefficient, the range of I

d2scharge coefficients shall be extended until the maximum clad temperature calculated by this variation l

has been achieved.

3 c.End of BloWown. (Applies Only to Pressurized Water Reactors.) For postuhted cold leg breaks,all emergency cooling water injected into the inlet lines or the reactor vessel during the bypass period shallin he the calculations he subtracted from the reactor vessel calculated inventory. This may be executed in the calculation during the bypass period, or as an altemative the amount of emergency core cooling water calcuhted to be injected during the bypass period may be subtracted later in the calcuhtion from the water p:,

remaining in the inlet lines, downcomer, and reactor vessel lower plenum after the bypass period. This i

bypassing shall end in the calculation at a time designated as the "end of bypass,"after which the expulsion 3 [ gg or entrainment mechanisms responsible for the bypassing are calcuhted not to be effective. The Ga end.of. bypass definition used in the calculation shaU be justified by a, suitable combination of analysis and experimental data. Acceptable methods for defining "end of bypass" include, but are not limited to, the 7

I L

{ 'i following: 1. Prediction of the blowdown calcuhtion of downward flow in the downcomer for the i')

F remainder of the blowdown period; 2. prediction of a threshold for droplet entrainment in the upward J

1 'f I

velocity, usinglocal fluid conditions and a conservative critical Weber number.

):b l..

{'

d.Noding Near the Break and the ECCS In/ccrion Polnis. The noding in the vicinity of and including Ij the broken or split sections of pipe and the points of ECCS injection shall be chosen to permit a reliable j,i i

analysis of the thermodynamic history in these regions during blowdown.

i M 3Y

2. Frictional Pressure Drops g i V1 The frictional losses in p! pes and other components including the reactor core shall be calculated using j

i

'l modeh that include realistic variation of friction factor with Reynolds number, and realistic two phase d

'0 friction multipliers that have been adequately verified by comparison with experimental data, or models j

that prove at least equally conservative with respect to maximum clad temperature calculated during the

]

hypothetical accident. The modified Basoczy correlation (Baroczy, C.J., "A Systematic Correlation for Two-Phase Pressure Drop," Chent Enging. hog. Symp. Series, No. 64,Vol. 62,1965) or a combination of

?

7 the Thom correlation (Thom, J.R.S., " Prediction of Pressure Drop During Forced Circulation, Boiling of Water," Int. J. of # car d Mass Transfer, 7, 709 724, 1964) for pressures equal to or greater than 250 psia

^

f i

and the Martinelli Nelson correlation (Martinelli, R.C., Nelson, D. B. " Prediction of Pressure Drop During d

Forced Circulation Boiling of Water," Transactions ofASME,695 702,1948) for pressures lower than 250 psia is acceptable as a basis for calculating realistic two-phase friction multipliers.(See Combined Discussion of Frictional Pressure Drops and Momentum Equation below.)

e

\\

j j

1108 j

3. Momentum Equation e dwergent "The following effects shall be taken into account in the conservation of momentum equation:

89 of "an 1.?tmporal change of momentum, 2. momentum convection, 3. area change momentum flux, 4.mo.

.i all t'ely.

rrmtum change due to compressibility, 5. pressure loss resulting from wall friction,6. pressure loss resulting from area change, and 7. gravitational acceleration. Any omission of one or anore of these terms under stated circumstances shall be justified by comparative analyses or by experimental data.

4. Critical Heat Flux

a. Correlations developed from appropriate steady state and transient. state experimental data are acceptable for une in predicting the critical heat flux (CHF) during IJDCA transients. The computer es shall be programs in which these conelations are used shall contain suitable checks to assure that the plyric31

.s, mal ans parameters are within the range of pa:ameters specified for use of the conelations by their respective saa velude authors.

J

.nni an s of

b. Steady. state CHF conelations acceptable for use in LOCA transients include, but are not limited to, the following:

• • ,haec in C) W J. L. S. Tong. " Prediction of Departure from Nucleate Boiling for an Axially Non uniform Nasimum Hest Mux Distribution," Journal ofNuclear Energy, Vol. 21,241248,1967.

d5 l 8k (2) Bd W.2. J. S. Gellentedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, L. J. Stanek," Correlation of

.'*Juned Critical Heat Flux in a Bundle Cooled by Pressurized Water," N> Phase MowandHeat Transferin Rod w ulues Bundles, ASME,NewYork,1969.

.9" I*' *

(3)Hench Levy. J. M. Healzer, J. E. Hench. E. Janssen, S. Levy " Design Basis for Critical Heat Flux

• v sance of Condition in Boiling Water Reacton," APED 5186, GE Company Private report, July 1966.

'8 *2'*"

(4)Macbeth R.V. Macbeth,"Ar; Appraisal of Forced Convection Burnout Data," Proceedings of the instsrute ofMechanicalEngineen, t9651966.

./

all (5) Bamett. P. G. Bamett,"A Conelatbn of Burnout Data for Uniformly Hested Annu11 and its Uses for Predicting Burnout in Uniformly Heated Rod Bundles," AEEW.R 463,1966.

r (6)# usher. E. D. Hughes, "A Conelation of Rod Bundle Critical Heat Flux for Water in the Pressure "8 "

.'e ?tv C11ef Range 150 to 725 psia,"IN.1412, Idaho Nuclear Corporation, July 1970.

ein! N c.Conelations of appropriate transunt CHF data may be accepted for use in LOCA transient analyses if a ng.

comparisons between the data and the conelations are provided to demonstrate that the correlations

,..,,, The Predict vanes of CHF which allow for uncertainty in the experimental data throughout the range of w and pararneters for which the conelations are to be used. Where appropriate, the comparisons shall une

..,,, s. the statistical uncertainty analysis of the data to demonstrate the conservatism of the transient correlation.

w int

d. Transient CHF conelations acceptable for use in LOCA transients include, but are not limited to, the i,.t. sed fouowing (1)CE transient CNF. B. C. Slifer, J. E. Hench," Loss.of. Coolant Accident and Emergency Core

. s sing Cooling Models for General Electric Bolling Water Reacton," NEDO.10329, General Electric Company, w

saMe Equation C.32, April 1971.

e./m ter CHF is rest predicted at an axial fuel rod location during blowdown, the calculation shall not f

mee mucMte boDag heat tramfer conelations at that location subsequently during the blowdown even if the calcula:24 Jocal fluid and surface conditicas would apparendy justify the reestablishment of nucleate esdsted seg bolling. Heat transfer assumptions characteristic of return to nucleate bo!!ing (rewetting) shall be permitted tue9 ase when justified by the calculated local fluid and sqtface conditions during the reflood portion of a LOCA.

h a

% 88

5. Post.CHF Heat Tresfer Correlations

-j' a.Conclations of heat transfer from the fuel cladding to the surrounding fluid in the post.CHF regimes of transition and film bolling shall be compared to applicable steady. state and transient. state data using statistical correlation and uncertainty analyses. Such comparison shall demonstrate that the correlations g

predict values of heat transfer coefficient equal to or less than the mean value of the applicable i

experimental heat transfer data throughout the range of parameters for which the correlatioris are to be used. The comparisons shall quantify the relation of the correlations to the statistical uncertainty of the e

,y ~

spplicable data.

'Y 1109 g

1n I

,a

u..

b. The Groeneveld flow film bolling conelation (equation 5.7 of D.C. Groeneveld,"An Investigation of Heat Transfer in the Uquid Deficient Regirne," AECL.3281, revised December 1969), the Dougall.

e' Rohsenow flow film boiling correlation (R. S. Dougall and W. M. Rohsenow," Film Boiling or. the Inside of 1

Venical Tubes with Upwsrd Flow of the Fluid at Low Qualities," MIT Report Number 9079 26, 1:

Cambridge, Massachusetts, September 1963), and the Westinghouse conelation of steady. state t: mition bolling ("Propdetary Redirect / Rebuttal Testimony of Westing! ouse Electric Corporation," U.S.A.E.C.

t.

Docket RM 50 I, page 25-1, Gnober 26,1972) are acceptable for use in the post.CHF boiling regimes.In d

addition the trutsition boiling correlation of McDonough, Milich, and King (J. B. McDonough, W. Milich, c

E.C. King, " Parti.a! Film Bolling with Water at 2000 psig in a Round Vertical Tube," MSA Research Corp.,

C Technical Report 62 (NP-6976), (1958)) is suitable for use between nucleate and film boiling. Use of all 4

these correlations shall be restricted as follows:

(1) the Groeneveld conelation shall not be used in the region near its low. pressure singularity, d

(2) the first term (nucleate) of the Westinghouse conelation and the entire McDonough,Milich,and e

King correlation shall not be used during the blowdown after the temperature difference between the clad 11 and the saturated fluid first exceeds 300*F.

ei (3) truition bolling heat transfer shall not be reapplied for the remainder of the LOCA blowdown, ti even if the clad superheat returns below 300*F. except for the reflood portion of the LOCA when justified by the celculated local fluid and surface conditions.

m a'

6. Pump Modeling The characteristics of rotating primary system pumps (axial flow, turbine, or centrifugal) shall be derived from a dynamic model that includes momentum transfer imween the fluid and the rotating

{

membLt, with variable pump speed as a function of time. The pump model resistance used for analysis should be justified. The pump model for the two phase region shall be verified by appliesble two-phase pump performance data. For BWR's after saturation is calculated at the pump suction, the pump head may be assumed to vary knearly with quality, going to uro for one percent quality at the pump suction, so long f

as the analysis shows that core flow stops before the quality at pump suction reaches one percent.

E

7. Core Flow Distribution During Blowdown i

(Applies only to pressurized water reactors.)

di s.The flow rate through the hot region of the core during blowdown shall be calculated as a function of es time. For the purpose of these calculations the hot region chosen shall not be greater than the size of one Ti fuel assembly. Calculations of everage flow and flow in the hot region shall take into account cross flow M

between regions and any flow blockage calculated to occur during blowdown as a result of cladding swelling se i or rupture. The calculated flow shall be smoothed to eliminate any calculated rapid oscilhtions (period less A' j ac than 0.1 asconds).

l

b. A method shall be specified for determining the enthalpy to be used as input data to the hot channel as i

hestup analysk from quantitles calculated in the blowdown analysis, consistent with the flow distribution th j calculations.

ac DISCUSSION

1. Break Characteristics and Flow

[

s. The Interim Acceptance Criteria and the models that have been accepted for their application require us considering the consequences of postulated rupture of primary system pipes. The requirements differ th 3

somewhat from one accepted model to another.In general, analysis must be made of the effects of rupture of pipes whost areas range up to that of the largest pipe in the primary system. Double. ended rupture must fo be considered, widch snea is that it must be assumed that the pipe is severed instantaneously by a (B

L circumferential break, and the two parted sections undergo mutuallateral displacement so that each can M.

discharge primary system water without interference from the other.

For pressurized water reactor models, a requirement was also established for analysis assuminginstead ds of double-ended breaks, longitudinal splits whose areas ranged from a factor of 0.6 to 1.0 times that of the Te p

maximum. No such requirement was placed on analysis applied to boiling water reactors.

T i.

g 1110

't 4

L

m.:

- -~ - - ; y.

..... m v ^..... -. -

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For the approved AEC Model, Babcock snd Wilcox Model,and Combustion Engineering 3g

]'

p i, end of the break spectrum to be analytd % defined as 0.5 ft. For the Westinghouse Model, no j

8 lower limit was set, but exceptions to the Westinghouse Model were stated only for pipes de d.

2 Spectrum of pipe breaks to be considered was not an issue during the hearing. The original in

'j'If.5.A.E.C.

888 tim 08Y Presented by Aerojet indicated little difference to be found between the effects double ended break of the lasyst pipe for a 3 loop PWR, though Ybarrondo stated in questio

,eynws.in en theoretical grounds that a split should lead to lower rates of fluid dacharge (Transcript y Milich.

i

,'y c,p,,

Other information in the hearing record also confirms the esuntial equivalence of effects o!

l double. ended breaks.

g g,gg The rule proposed by the Staffin its Concluding Statement of Position has rectified the existen differences in assumed break sizes and types for models to be accepted, but has done ao a hand clarity of direction. We believe that the rule should be explicit in its requirement that analysis be

[ww clad the effect of aptits as well as double. ended guillotine breaks. We agree with the Staffin eli reference to a murdmum in the size of break to be analyzed, We have worded our statem to reflect these points.

mdown,

b. Ascharge Nodel The blowdown period can be divided into two parts, from the standpoi InrA den models to describe the discharge of primary coolant. During the first, the fluid leaving the postu would be liquid c'r almost entirely liquid. The second would begin when the fluid is definitely The record of the hearing considers at length an analytical rnethod based on use of a correlati; Moody, incorporated in all accepted calculational models for treating the rate of discharge l

rais shall be coolant during the second interval. Moody's correlation was derived assuming the two phases

'se rot 88*8 discharge nosale to be in thermodynamic equilibrium. This assumption leads to a relations vs analysis maximum rate of fluid discharge.

e swofhase The record shows that predictions based on the use of the Moody model have been extensivel f head msY c'n e k*8 against blowdown experiments in several facDitiet. These include blowdowns of the Containment Sy Experiment at Hanford, the LOFT Semiscale vessel at NRTS, and simple pipe blowdowns by E

}

England. Some comparisons have also been made with the results of other experirnents in vendors.

The Regulatory Staff, ANC, and the four vendors all agreed that calculated blowdown rates based on direct use of the Moody correlation overestimate the observed rate of discharge for any given a o,e e;f especiaBy for the larger breaks. ANC testirnony and answers during questioning of the Staff (Ross, we of nne Transcript, pp.18001806) and ANC personnel during the hearing attributed the difference possibly to now Moody's use of equDibrium thermodynamics. The testimony by Roy (Babcock and Wilcox) expres

., i.eism,.

same view. The lack of thermal equilibrium would be expected to have a greater effect when cold sesi,id icis accumulator water is irdected in the nozzle near the hypothetical cold leg break. Ybarroedo stated that

~ ANC calculationa can now 8: discharge rates without use of a critical flow model such as Moody's

.,e channel associated use of a discharge coefficient (see below). But the Stafrs Supplementary Testimony attributed i itm6iuu,e this to ANC's incorporation of the momentum flux into the computational model used, rather than a non-equilibrium calculation.

Because the calculation overestimates the discharge rate and underestimates the time unta and of blowdown, ANC r.nd the vendora have aD tried modifications of their calculations based on use of a ac> called discharge coefficient. This coefficient is applied directly to the break area for which the salt.dation is made; thus the calculated blowdown assuming a break area A and a discharge coefficient D would be made assuming a mythical area AD.It is found that if a fixed value of the discharge coefficient la

..n wquise used,it is chosen best to be below unity, especiaDy for the larger breaks. This reflects the need to reduce nos Jeffe the calculated discharge rate to agree with observatloa.

.irvrirse The initial Staff testimony and the Staff Supplementary Testimony reported a best value of about 0.6 l

-ave must for the discharge coeflicknt for large LOCA breaks; this value was supported by Roy's testimony vih Ig a (Babcock & WDcox), and by Zane (Transcript p.10791), Cermak (Westinghouse. Transcript p.15134), a

• end ses Woose (Westinghouse, Transcript p.15504).

There was also widespread speement that a variable discharge coefficient provides a better fit to the mewatead data than a constant one (Brockett, Transcript p. 7484; Moore, Transcr!pt p.15161; Staff Supplementary Testimony; Babcock and WDcox Redirect and Rebuttal Testimony; Bingham, Transcript p.21143; Roy, y e

Transcript p.21144).

1111 I

There was some difference of opinion as to the effect of the value of the discharge coefficient on the calculated clad temperature. The Babcock and Wilcox Direct Testimony report a reduction in maximum calcuhted clad temperature when the discharge coefficient was taken as 0.6 instead of 1.0. Colmar's Testirnony stated that a reduced discharte coefficient could increase the maximum clad temperature through delaying the onset of ECCS, and that a variable discharge coefficient can lead to a higher maximum g

calculated clad temperature than a fixed value of unity. Ybanondo (Transcript p. 6362) reported the result of an ANC calcuhtlon with a discharge coefficient ialtially 2.0 and hter in the blowdown 0.6, where the a*

fint peak in the clad temperature exceeded by about 100* the value obtained with a fixed discharge h

coefficient of 1.0. Roy (Transcript p.12944) reported lowered maximum clad temperatures when a q.

variable discharge coefficient was used by Babcock and Wilcox in place of the value 1.0.lanra(Transcript

-h pp.1406814071) reported a reduced maximum clad temperature calculated by General Electric when a M-discharge coefficient less than unity was used. Moore (Transcript E.26) reported a similar result from proprietary Westinghouse analyses.

The Direct Testimony of the Consolidated National Intervenors argued that the use of the Moody model leads to in orrect res&lts becau e the liquid discharged would be nrtastable (supercooled) and so r

would emerge at a rate about 1.7 times that predicted by Moody. This would lead to higher transient containment pressures and to faster core dryout which would cause maxirnum calculated clad temperatures to be higher. It was stated that this tendency would be found especiaUy if the break were near the vessel.

Data from experiments by Barton and by Fauske were cited in support of ChTs view. During questioning.

Dawson (Transcript pp. 1834818349) added that CN1 also has some supporting data based on cryogenic fluids. The CN1 position was discussed at length during questioning reported on pp.182!r018653 of the 3

Transcript.

Ij addressed the points made by CN1, and disagreed with CN1 both in detail and in general. General Electric The Redirect and Rebuttal Testimonies of Babcock and Wilcox and of General Electric generally 4

testimony claimed that CNI misinterpreted the Fauske data and ignored data of Barton that do not support their case. GE reported that Edwards' blowdowns, Bogarty data, and other data show metastability to have decayed in less than I millisecond which is less than the transport time of fluid from the nozzle entry to the hypothetical break. The Babcock and Wilcox testimony pointed out that the Fauske data were obtained with a % inch diameter orifice near the bottom of a vessel 10 ft. high,and the liquid at the bottom of the

?

vessel was subcooled because of the liquid head. For this and other stated reasons Babcock and Wilcox chimed that the Fauske data used to support the CN1 position were not relevant.

j The Staff took the position in its concluding Statement that the Moody modelis we!! supported, apart fem the possible need to use a discharge coefficient, and that this model should be used for quality exceeding 0.02.

i

~

We agree with the Staff position as to correctness of use of the model based on critical flow,since the i

length of time available during the blowdown far exceeds the amount needed for nucleation and build-up of Iwo phase discharge. Furthermore, the evidence is strong that use of the Moody conelation does not 3

underestimate observed experimental discharge rates, as would be the case if discharge were really t

metastable, but in fact it definitely overestimates the discharge rates.

'.g We require calculations with at least three values of the discharge coefficient ranging from 0.6 to 1.0.

V.-

Use of a discharge coefficient less than unity is equivalent to assuming a smaUer break, and the regulrement

'k to calcuhte the consequences of a spectrum of breaks whose sizes range up to the area of the largest y

primary coolant pipe makes the calculations with discharge coefficients less than unity seem unnecessary.

f However, the practical effect cf the redundancy seerrs nef gible. The need to calculate for smaller breaks li j

takes care of any concern raised in connection with Colmar's view as to the greater severity of longer 8

discharge thnes.

We note finally the point raised in the ANC direct testimony, to the effect that fits of calcuhted I-T blowdown transients to experirnents were based primarily on comparisons of pressures. Measurements of fluid density or of rnass discharge rate were not reliable enough to be used in tests of analytical methods.

~~

1.ack of thermal equDibrium in the as-yet tobe discharged fluid can lead to incorrectly estimated fluid inventory even when the pressure is correctly calculated. This is not likely to be very much of a problem 0

near the end of blowdown, however, where rates of change in pressure and flow rate have cased o'ffin 35 rapidity. At this point the primary system inventory is mostly vapor, and therefore the aver Ee mass discharge rate over the entire transient wlU turn out to have been predicted adequately if the pressure time curve has been predicted properly.

j

(

1112 q

L

-n q>'5 eggive.asssaetigh es#E+ esp-regWuStPM9Pyeg49 -

it eri the c.End of Blowdown. De event that was most influential in raising questions about the adequacy of aximum emergency core cooling was the publicity given to the loss of injected coolant during blowdown in A, n's Semi 4cale test 845 at Idaho, and one of the major features of the interim Policy Statement was the P"stu" requ rement that all emergency core coohnt injected during blowdown should be considered lost. Although j

• • I""*

there is still objection to comparing PWR's with the Semi. scale tests (Combustion Engineering Response to the Reguhtory Staffs concluding Statement, p. 23), there appears to be no major objection to the rule.

he result

. here the The rule moderates the arbitrariness of the " accumulator bypass" section of the Interim Policy lischarge Statement in a conservatin way, requiring that the calculation of the end of bypass be supported by when a I

suitable experimental data. Although the Reguhtory Staffin their concluding statement reject for the time nnscript being the Wallis correlation proposed by. Westinghouse and the " Waterfall" concept suggested by when a Combustion Engineering, these methods of calcuhting the extent of cooling water bypass can clearly be alt fnm considered under the rule when adequate calculational detail and experimental verification are available.

The primary importance of calculating the correct curve of system pressure versus time for a PWR is Moody found in calcuhtMn of the time at which blowdown ends and effective emergency coolant injection is

) and so assumed to start. Several dermitions of the end of blowdown have been proposed, and each has some

)

ransient dif6culties. The Staff Direct Testimony stated that end of blowdown is assumed to be the first time when matuns fluid flow from the break ceases. Parks (B&W, Transcript pp.1252612527) stated that this may not be the o nesul, best dermition, because zero break flow may never be reached for some breaks such as hot leg breaks and

"'8885"8 very small breaks. But he added (Transcript p.12527) that B&W calculations show a time of zero break

'90FalC flow for all cold leg breaks but the smallest ones. Rosztoczy (Transcript p.13230) reported that no CE l

3 of the calculation has shown the core recovered before zero break flow occurs,and so zero break flow is adequate to define the ead of blowdown.

y "mIlY The B&W Redirect and Rebuttal Testimony discussed calcuhtions of accumulator water entrainment I

Electric which considered the end of blowdown to be the time at which entrainment ended. Roy stated however sr that B&W does not at this time recommend such a definition (Transcript pp. 21164 21165).

~0 Definition of"end of blowdown" is replaced in the Staffs Concluding Statement by "end of bypass,"

716 d which means assumed end of accumuhtor water bypass.This is the time at which calculations supported by

?

ibtained experiments indicate the ECC bypass or entrainment mechanisms to be no longer effective.We believe that a of the it will be possible to provide calculations satisfying these requirements,and this procedure for defining the i17ilcox start of effective emergency coohnt injection is superior io the one used formerly. The proposed procedure on page 188 of the Regulatory Staffs Concluding Statement is correct and precise,and we have elected to d, apart adopt it verbatim.

%**\\itY d.Noding Near the Break and the ECC$' injection Points. A number of witnesses in the course of the hearing pointed out that inco rect noding near the assumed break or the point of ECCSinjection can lead ince the to non-conservative conclusions. Excessiwly hrge nodes near ECCS injection could lead to unrealistic Idep er calculated cooling of the water in the downcomer and/or the lower plenum, and more rapid reduction of

.oes not the driving force during blowdown than would really be expected. The noding detailin the vicinity of the e really break as discussed on page 40 of the Staffs Concluding Statement is directed to avoiding this problem.

l The advantages of correct thermodynamic description of the discharge podes have been discussed

' 80 M

earlier, itement

? larpst 2 and 3. Discussion of Frictional Pressure Drops and Momentum Equation

-cessarv.

r bNs (a)/dentification of the Terms. The momentum equation describes the Newtonian behavior of the f lonpr coohnt as it is subjected to such forces as fluid pressure and gravitation. Two terms in the rnomentum equation received special attention during the hearing. One of these is the momentum flux.De other is the Iculat:d frictional pressure drop.

ients of These two terms have assumed extra importance because they can become raore than usually significant 3

iethods.

when the coolant is two. phase.

j ed fluid (b) Momentum Flux. The momentum flux represents the spatial convection of momentum.In tegions l l problem of the system where accelerations of the coolant are occurring,and where heat transfer changes the, ratio of df liquid water to steam, the momentum flux differs from zero. Thus the mcmentum flux assumes finite

. p(

values at phecs where area changes in flow paths occur (form losses)and where heat sources and sinks are are time k

1113

found. It is also finite if phase changes are the result of system pressure changes, such as accompany factor is 4

(

depressurization.

that the ! '

he Staffs supplementary testimony pointed out that rione of the vendor computational models import n includes the momentum flux as it appears la the momentum equation. The form of RELAP 3 approved for The\\

use with the interim criteria did not incorporate the momentum flux, either. The models used by lumped t !

Combustion Engineering and by Babcock and Wilcox did include the area change part of the momentum Thec flux. The Westinghouse model, the General Electric Model, and the ANC model contained empirical form correlatic,

losses or abrupt losses to account for the area change and the heightened turbulence at such locations as the (NEDOli core entrance and exit. Different methods were sometimes used to modify form factors for two phase Martinelli conditions, and to change magnitudes of pressure losses when flow reversal occurred.

Evide:

The questioning of witnesses led to a rather uniform set of statements to the effect that the momentum the choic flux is not very important to analysis of the transient. Bingham said that the B&W rnodel does not include red stribu the density change component of momentum flux,but calculations that had been done accounting for this calculatio component of the momentum flux had not changed the results significantly (Transcript p.12621).

(Transcrif Rosztocry reported insensitivity of the peak clad temperature as to whether calculations are made with and accounts I without the momentum aux term. Cermak said that the Westinghouse model used the momentum flux in These 1967, but the term was removed because it led to computationalinstabilities and had only a small effect on blowdowr

' the result (Transcript 15592). Rockwell said that the acceleration terms in the momentum flux amount system pr on',' to 4 percent of the bundle pressure drop during the early part of a BWR transient,and the effect blowdowr decreases as the transient proceeds (Transcript p. 21420). He also said that the momentum Hux is it.cluded empirical in GE's empiricalloss coefficients (Transcript pp.21421,21542 3).

factor cor:

The Staffs Supplementary Testimony expressed the view that vendors' codes should be modified to The ti include all momentum aux terms. Thh view was supported by consultants at ANC and at ORNL.The calculatini Concludmg Statement of the Staff in its proposed rule took the same position, but aho indicated that Testirnon) specific terms can be left out if the omission isjustified by comparative analysis or experimental data.

Staffs Su; l The B&W Concluding Statement took issue with the Staffs position, generally on the basis that the

chawls,

[

B&W CRAFT code can already predict the course of a blowdown so as to achieve a conservative assessment accordmal g

of system response to a hypothetical LOCA,and aho that the record does not support any need toinclude De S' all terms of the mornentum flux in the model. These views were also stated in B&%"s Response to the requiring t.

Concluding Statement of the Regulatory Staff.

basis for.

The Concluding Staternent of GE and also GE's Responsive Closing Statement, Respon&ng to correlatior l Concluding Statement of Position of the Regulatory Staff, also gave similar reasons why the GE computer Thom con i model need not be modified to include the momentum flux. We note that the question of the need to friction. T i include the momentum Oux in the momentum equation wasin fact argued during the hearing. Views both psi) are ac pro and con were expressed, though it appears that the arguments against the need for momentum Dux can. The I were a little more quantitative than those favoring the need, adequately ;

On the other hand, the evidence in all respects was not overwhelming. Since the momentum equation GE arg does contain the momentum flux term,it seems that the burden of proof must rest on those who wish to made its c leave this term out. The contentions in the record do not add up to such proof, The Staffs proposed course is fair and is supported by the record. Momentum flux is seen to be a We hm the GE Co reality, and in any stated circumstance its eomponents must all be included in the analysis or any omission the flow li:.

must be justified. We have adopted the Staffs careful wording for this reason.

need to ca The record contains only a few references directly or indirectly as to including momentum nux in However, t ;

calculating the effect of any core blockage that might occur during blowdown, it should be clear that the given later {

requirements as to use of the momentum flux apply to all LOCA blowdown Dow calculations.

reactors.

j (c) 7hrhase Nultipliers. The practice in estirnating fluid friction for a two. phase fluid is normally to Therec calculate the friction coefficient for a single. phase fluid and then to multiply this value by a factor that time. The corrects for the quality. This is the wo phase friction factor, for BWR's j The vendors' practices differ in calculating the single phase friction coefficients, as is seen on page 4 5 defect.We j of the Staffs Supplementary Testimony. Westinghouse and GE blowdown codes do not include single BWR accid j phase friction factoes with explielt dependence on the Reynolds number. "Ihe B&W and CE models do appropriate include such a dependence, as does the GE transient CHF model. Parks said that B&%"s calculation of results at 1. :

Reynolds number dependence is not extended into the laminar flow regime, because no laminar now is two-phase ;

encountered during blowdown (Transcript, p.12766 7). Rosztoczy stated that the CE single phase friction choice pros 1114

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i

T7 7_..

.. _..,.,.._j.._.._.

factor h applicable to turbulent flow and not laminar flow, but only turbulent now is expected. He said u ompany that the Reynolds number dependence used has only a mild effect on the transient,and it is not one of the cal modch important factors.(Transcript p.13285 6).

j The Westinghouse Redirect and Rebuttal testimony stated that friction factors and form factors are ppreved fo, h und by lumped together before application of the two-phase correction correlation.

The correlations used to correct for the two-phase effects also differ. Westinghouse uses the Arrnand comemum correhtlon (Westinghouse Redirect and Rebuttal Testimony). GE uses the Martinelli Nelson correlation wriest form

',,ons,s th, (NEDO 10329). B&W and CE use the Thom correlation at pressures above 250 psi and the Martinelij Nehon below.

r two.phaw Evidence was provided that the results of calculatiom of LOCA consequences are not very sensitive to the choice of two-phase friction factor (but see our reservation below relative to the effect on flow momentum redistribution.) Routoczy also discussed such conclusions (Transcript pp. D.12,13). Hench said that in GE not indude calculations the entire friction in the core only accounts for about 10% of the core pressure drop tmp for this.

p, g 3;i p.

(Transcript p.14443 4). Moore said that in Westinghouse calculations the entire friction in the core accounts for as rnuch as 45% of the core pressure drop during blowdown (Transcript p.15591).

Je with and These comments as to adequacy are only relevant to the one. dimensional calculations with the tum nus in blowdown code and the heat balance code. Here the core pressure drop is only a small part of the full

,g,rr,,. on system pressure drop, and so large changes in the core pressure drop can be made without affecting the M amount blowdown signincantly, it may be added that changes of this kind can be compensated by changing the J she effect empirical discharge coefncient, so that matching exp:rrimental blowdowns with a given two-phase friction

. n anduded factor correlation and the discharge coefficient as a free variable proves nothing.

)

The record makes it clear that proper treatment of the two. phase friction is more irnportant for m&fied so calculating now redistribution in the core. This point was made by the Staff in its Supplementary

. Mt.. The Testimony, and by Morgan (Transcript p.12678 9). It was also made in the Oak Ridge comments on the Jaaied slut Staff's Supplemet m i. stimony. Because the quality will vary significantly from the hotter to the colder j,

channeh, the fluid wmity corresponding to the common pressure drop plenum to plenum will also vary he-accorkngly.

n

.sent he Staff in its Supplementary Testimony and in its Concluding Statement revealed an intention of J io mdude requiring use of r+:listic wo phase friction conelations. Review of the adequacy, realism,and experimental

,in,toshe-basis for comp +1 possibilities has led the Staff to the conclusions that either the modified Baroczy correlation should be used, or the MartinellbNelson correlation should be used below 250 psi, and the

eJ ng to Thom correlation :i es 250 pai. These choices are expected to provide realistic estimates of the awo phase i gomputer friction. The Staff states that correlations that overpredict the friction (e.g. Martinelli Nelson above 250 "a need i,,

t iens both '

psi) are not obviously conservative in all applications,and so realism should be the guiding principle in this e m m nus case. The Staff points out (Supplernentary Testimony) that the Armand correlation has not been adequately tested by comparison with the two phase multiplier versus quality data used in other cases.

GE argues in its Concluding Statement and in its Responsive Closing Statement that the Staff has not iequati.m

.. enh so made its case that the Martinelli. Nelson correlation inay not be used throughout for BWR calculations.In the GE Concluding Statement it is said that use of Martinelli. Nelson provides good fits to blowdown data.

l

,en.. be a We hav -oted abow the inadequacy of this argument. GE stater that the use of channelwalls to confine the ih insts possible flow redistribution in a BWR;thus an argument for realistic correlations based on a

.. muw.m need to calculate flow redistribution accurately would not apply to a BWR. We do not altogether agree.

f 4m tb m However, this point was not well explored during the hearing, and in fact the discussion of now reversal es, wt the given later does propose that flow' redistribution be considered only in connection with pressurized water reactors.

4r un> in The record sustains the Staffs position relative to umeceptability of the Armand correlation at this

.t..i Wi time. The Staffs position relative to acceptability of the Martinelli. Nelson correlation throughout the range for BWR's is not clearly based on the record,in arriving at a rule based on the record we must rote this e ;upe 4a defect.We conclude that unacceptability of the Martinelli. Nelson two. phase friction correlation for use in

..Je enpe.

BWR accident analysis cannot be part of a rule based on the record of this hearing. However,it would be eJeh Ja appropriate to have a rule that required use of realistic two phase multipliers or other multipliers that led to

.4 results at least equally conservative. We have therefore rewritten the Staffs proposed requirement as to two phase friction multipliers to permit GE to use the Martinelli. Nelson correlation throughout if this n

• eni... n,m choice proves et conservative as the realistic correlations. Any use of Martinelli Nelson this way would have

'6

• 1115

_ 7

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to be supported by sufficient parallel calculations using an acceptable realistic correlation to show that the latter choice will not lead to higher clad temperature during the hypothetual accident.

4. Discussion of Critical Heat Flux (CHF)

The critical heat Dux would be important during blowdown because it marks the point at which a transition would be made from the more ef0cient heat transfer from fuel to clad produced by nucleate i

boibng. to a less efficient heat transfer regin.e. The change of heat transfer mode is called departure from v

nucleate boiling (DNB).

h The values of the critical heat flux used in practice have usually been determined in experiments with y

slow transients, such as a slow rate of increase of heat Oux, or small step wise increases in heat flux. The j

transients during a postulated loss of-coolant accident from a large pipe rupture would be much more rapid.

For a PWR subjected to an assumed instantaneous rupture of a cold leg pipe, flow reversalin the core is N

calculated to occur within about 0.1 seconds of the break. At approximately this time,for any point in the core, How stagnation would exist. Just prior to flow reversal the conditions in the coolant would be such the DNB would have taken place if the transient were slow.The computational models accepted according to the Interim Policy Statement have assurr.ed that DNB would take place at this time,and afterwards the heat would be transferred through stable film boibng or a transition boihng which is interme& ate between nucleste boiling and stable film boiling.

For a BWR subjected Io an assumed instantaneous break of a recirculation pipe, conditions appropriate to DNB would occur much later, from 5 to 10 seconds into the transient, and could accompany reduced core flow as the jet pump noules are uncovered.

E CHF is predicted in interim Acceptance Criteria models through use of a number of correlations.Most are based on experirnents conducted under quasi steady state conditions and these are called steady state correlations. GE has used a correlation based on transient data, that leads to a brief delay in onset of DNB.

The hearing explored the use of steady state heat transfer correlations for DNB under transient constions. The evidence was uniformly that under rapid transients DNB if anything would be delayed and so actual heat removal would be better inan calculati:d essuming steady state correlations.

Combustion Engineering's Redirect and Rebuttal Testimony cited numerous experiments showing this, though there is some confusion from grouping together results of experiments with different kinds of transients (now decrease. pressure decrease). But this and other evidence considered during the hearinglead s

us to agree with the position stated in the Staffs Supplementary Testimony, that the steady state correlations can be used to predict the CHF realistically in slower, quasisteady state blowdowns and to predict the CHF conservatively in fast blowdown transients.

z Also in its Supplementary Testimony, the Staff suggested that the steady state CHF correlations should t

be modified in use to reflect statistical uncertainty. The suggestion contained a techni:al flaw, but the

.,;j intem was clear. in its Concluding Statement, the Staff withdrew this suggestion and appears content now Io accept realisrnin this part of the calculation in place of conservatism.

The PWR vendors have all used steady state CHF correlations. GE has used its own correlas.lon based on transient data. GE has reported sensitivity studies that show the results to depend only makly on the use 3

of a transient model.

S in its concluding staternent the Staff proposed to accept use of several steady state CHF correlations t

where comparisons with data show that the correlations predici values of the CHF that allow for

{

uncertainty in the experimental data throughout the range of parameters for which the correlations are to be used. Acceptance of the CE correlation us also proposed on this basis.

We believe that the record upholds the conservatism of this approach.

Westinghouse has proposed to add the condition that DNB should be assumed to occur in reality only if conditions appropriate to DNB persas for a period greater than 50 ms. Although the record includes reports of laborat rk studies showing that DNB does not occur during rapid flow reversal, we believe that this observation is not weR enough substant sted under the conditions that would prevail during a reactor i

blowdows so be accepted in LOCA computational models. We agree with the Staff's omission of the Westinghouse assumption from its approved evaluation models.

On the same bar.s. we agree with the omission of the similar proposal by GE, to the effect that DNB be f

neglected if conditions appropriate to it persist less than 3 ms.

[

1116 f

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~ * * * * ' ~ ~

1

(

e f'~j We have again accepted the caretul wording used by the Regulatory Staffin the t

I to estabbsh the requirements as to calculation of critical heat flux.

ent

5. Discussion of Post CHF Heat Transfer Correlations:

(DNB) is vital to estimation of the course of a hypothet wa s

.smi calcuhted to occur within about a tenth of a second after a postulated instamaneo s

, we history of the chd during blowdown and the possibility

,,, cd would also determine the effectiveness of removal of heat from the De energy in the fuel at the time refill of the plenum by ECCS fluid starts.

p3 The lack of flow reversal in the design basis accident to a BWR leads to a very

,ea regards DNB and post CHF heat transfer. DNB would not occur until about 5 to 10

, on blowdown, when the jet pump intakes were uncovered by the blowdown. T subsequent to this would then be a principal determinant of heat removal from th l

+4 later in the blowdoan. By this time, however, the heat generation rate from fis e **

would have decayco further, e

/nierim policy Statemerrt NodcIs forPWR *st The interim Policy Statement'has req transfer coefficients for PWR's calculated using stable film boiling correlations afte "3

been defined. Westinghouse has been perrnitted use of a proprietary transition bo ons have meant to bridge the region of time between nucleate boiling and stable aim boiling,and and ANC haw similarly used the transition boiling correlation of McDonough, Milic same purpose. The transition period affected by this bridging typically occupies a small frac T'

after DNB, and its use instead ofimmediate use of a coefficient appropriate to sta M

effect on the maximum calculated clad temperature. The use of a transition boiling

• * ~ ~

mathematical oscillations in the results of computer code calculation

/

1 The practice as regards correlations for film boiling has varied from case to case.

Westinghouse practice has been to continue use of the 9

the quality is above zero. The first part of this correlation is a nucleate boiling term whic after the first brief interval, and the remainder of the correlation leads heat trans mportant by the Groeneveld correlation. The THETA code used for hot channelanalysis by A Groeneveld correlation, as does the RELAP blowdown code used by ANC. The S codes used by Combustion Engineering incorporate the Dougall Rohsenow film bo (T s/ Ten,4 $ suggested by McEbget o correct for cases of h u

correlation was derived from both steady state and transient data.All The validity of these practices followed according to the Interim Policy Staternent w length during the hearing, and a number of points bearing on the acceptability of were considered. These are discussed below.

luterim ?olicy Statemerrt Modelfor BWR's: GE has been required to use the Groene after DNB.as long as the quality of the fluid is between acto and unity.The extens i

through the period oflower plenum thshing is discussed at greater length below l

Blowdoun # art Trarrtferi The transition from nucleate boiling to stable film boiling w heat transfer from clad to fluid by about two ordets of magnitude. Evidence was pre heat transfer would not depressurization would cause the coefficient to remain near i

This evidence was mostly provided by B&W, who have carried out an extensive res Alliance, Ohio facility on blowdown heat transfer. The B&W reports indicate that heat t dominated by bulk nucleation until the channel is almost empty of liquid. These expe performed with a fluid velocity several times that appropriate to a LOCA.and with heat those that would exist at appropriate times during a LOCA. They were also performed w ee.

j

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long, about half the length used in water reactors. Some questions were therefore raised by ANC and the 3

Staff as to the direct appbcability of th: results to analysis for a PWR.

Transient Hear Tmntfer: Some criticism was expressed as to the use of steady state conelations during the fast transients analyzed for a large 1.DCA.These views were stated by bwson (Transcript,pp. 5766 7),

Ybarrondo (Transcript, pp. 6069, 6272,10282,10890,10906 7), and Brockett (pp.7480,7588). The 4

tenor of the criticism was that evidence was not conclusive that steady state correlations overpredicted the transient coefficients or predicted them accurately.

,1 Considerable evidence was provided nonetheless to the effect that during depressurization the use of steady state correlations for stable film boiling was a conservative course. The strongest evidence was the gi B&W work csted abow. Additional work in the U.S. was referenced by B&W as indicating that transient heat transfer during blowdown will be better than predicted using the stesdy state correhtions; this i.

included work at Columbia University and at the Bettis Atomic Power 1.aboratory. Other experiments referenced were performed at MAN in Germany and in the Soviet Union. In our view the evidence is near overwhelming that the use of steady state conelations for stable film boiling after CHF will provide a conservative estimate af heat transfer during blowdown.

t Rewetring and #psteresis.like Effects: It was stated by the Staff that Westinghouse used its transient g

heat transfer correlatiom throughout calcuhtions of blowdown for safety analysis of the Point Beach 3

reactor. This extended use of a transition boiling correktion was questioned by the Staff, because it implies that rewetting occurs as soon as the fluid conditions again becerne appropriate for nucleate boiling after

,l e

Q DNB has once occurred. This is equivalent to assuming instantaneous rewetting after the superheat has

,l fallen below the bidenfrost value.

b The issue is posed in a somewhat different way in that both GE and Westinghouse have proposed the y -

use of switching criteria for Critical Heat Flux. Westinghouse has proposed that if conditiom providing DNB persist for less than 50 ms,it should be assumed that nucleate boiling continues unimpaired.This gl period of interest for pressurized water reactors would be that attached to flow reversal almost instantly

,(

after the break. The Westinghouse position would then be essentially that the full transition bothng y

,k conelation with its nucleate boiling term should be used throughout most of the blowdown phase of the

i transient. In this form, the Westinghouse argument is that rewet does not need to be considered because y%

there has not been enough time for dryout.

.g '

GE's proposed switching criterion would lead to continued me of nucleate boiling heat transfer t

A coefficients if the conditions appropriate to DNB persist for less than 3 ms. The practical result wod! be ff use of a heat transfer coefficient calculated from the GE nucleate boiling correlation during essentially the 1;

j ;y entire period oflower plenum flashing of a blowdown of a BWR.

3;

" l We note the inconsistency of vendor positions that would rely on hysteresis like effects as the basis for M h switching criteria such as those above,as compared to other positions calling for instantaneous rewet.We

,l T 3 note also the inconsisteney of Staff positions to the contrary in both cases.The point remains that there is Mi swt adequate understanding of either rewet after CHF or of hysteresis.like effects during flow reversalin

,l 5j fast transient. The Staffs position has been to approve use of only stable film boiling once DNB has been calculated to occur, even when fluid and clad temperature conditions appropriate to rewet exist. This g.

a1 course is conservative. No less conservative position is justified by the record of the hearing.We concur with U

the Staffs proposal that the nucleate boiling term of the Westinghouse correlations not be used after DNB 4

is calculated to occur, and that other rnodels incorporate the equivalent assumption of stable film boiling i;

'j throughout the period after DNB.

Y The Groenewld Correlation: The Interim Policy Statement approved use of the Greeneveld conelation S.

j as conservative for calculating stable film boiling coefficients.The criticism of this choice on the grounds of S:

its derivation from steady state data has been discussed above. We have concluded that the use of a I

~

steady. state correlation for stable film boiling is conservative in this case, even takinginto account the use l

of realistic rather than bounding correlations in this application.

('

A difficulty has arisen as to which Groeneveld correhtion should be used.The Interim Policy Statement a

and the models accepted in this connection simply identify "the Groeneveld correlation" This situation yhs...

persists through the Staffs Testimony, the Staffs Supplementary Testimony, and in fact through most of t;

the record up till near the end. Under GE questioning, Mattson agreed that in all the foregoing the a

correlation referred to was that stated in equation 5.7 of D.C. Groeneveld,"An Investigation of Heat a

Q Transfer in the I.lquid Deficient Regime", AECl,3281, revised December 1969 (Transcript p.20696).

Mattson also agreed that the analysis of the acceptability of the Groeneveld cor elation that Slaughterbeck s

gg M

1 1118

1 i

had conducted at ANC had used the expression Groenewld had stated in Equation rather than that coefficient developed by Groeneveld in Equation 5.7. He aho said that heat t farms coefficients rnessured for rods and annuli should fall below those appropriate to rod bundles

+7 h cross flow effects. He said that this situation has been regularly seen in correlations tested The (Transcript, p.20563 5). Groeneveld's statemen of the same point was quoted from his J the was continued by the observation that the Groeneveld 5.7 conelation, derived using only singl

{

underestimates heat transfer in bundles that have been tested (20709).

. e of The Staffs concluding statement proposes acceptance of the Groeneveld 5.9 correlation, bu a ihe Groenewld 5.7. This posidan is not upheld by the record of the hearing,and we agree in gen sent argument rnade in this connection by GE in their Responsive Statement. lf there is any position main ilus throughout the hearing on the acceptability of a Groeneveld coreclation, it is that the Groene wnts correlation is acceptable.The rule must be in accord with this point for both PWR's and B near departure from this choice must be considered in hter reviews whether rule making or licen Je a Heat Trarrtfer at LowPressure It was observed a number of times that no non. proprietary been provided for heat transfer at pressures below 500 psl. Data developed at Columbia

ent Combustion Engineering sponsorship were provided to the Staff and the Hearing Board mh proprietary basis. The Staffs position in its Conclu6ng Statement was based in part on the existenc mes these data and their implications. GE has objected in their Responsive Concluding Stat

^

itier on GE reacton based on proprietary data provided by other vendors, as prejudicial to them

. I sss considered from other aspects in another section of this opinion. (see pp.10881089 aupra).

We note that CE data effectively fill the void in data below 500 psi used by Groenevel

. 'l

he correlation.

hng Dougall.Rohsenow Correlatiorm Combustion Engineering has used the Dougall.Rohsenow c

" i

!*us for calculating heat transfer coefficients in the stable film boiling regime. Babcock and Wilc "t>

inodined Dougall.Rohsenow correlation, which differs from the D.R through the use of McElig

{

T factor of (T r/Telad)h. In regions ofir}terest at low pressure, McEbgot's factor is signific

' 8, Thus, the heat transfer predicted by the D.R conelation is better than that of the modifie

.nr appreciable factor. CE has provided data and analysis showing that the D.R cortehtion leads to h i

transfer coeffscients for stable film boiling that are lower bounds to the range of values observ l.!

ori bebeves that their proprietary datajustif' se of the D.R correlation afte, CHF.

I

'w The Staff position has been to sewpt use of the rnodified D.R co; relation, but not the D R l'

me Presumably this has been done because the modified D.R correhtlon goes over in a smooth l

I Ditius.Boelter conelation for single phase steam at the limit of unit quality'(pure steam).

In their Response to the Staffs Concluding Statement, Combustion Engineering objects to cli ur of the D.R correhtion from the proposed Staff rule. On this point we believe the Staff wa evidence supplied by Combustion Engineering supports use of the Dougall.Rohsenow correlat d

course,1f the Dougall.Rohsenowis acceptable. W the modified Dougall Rohsenow.

6. Discussion of Pump Modeling e

l at For a postuhted cold les break in a PWR the flow througn the core would be reversed d f

part of the blowdown. This rewrse flow would be opposed by the pumps in the unbroken loops, s cateuhted flow rnagnitude and the resulting cooling would depend upon the model used some evaluation models it was assumed that the pump deliwred zero head as soon as the s e

was reduced to saturation pressure. The limited data available indicated, however, that pum would continue into the two. phase flow region. (Exh.132) Continued pump operation in the PW loops would diminish the core flow during the early part of blowdown and in some inst

.e t

calculated peak Ziscaloy temperature appreciably, sometimes in one direction, and sometim j

(Exh.113, p. 6 4).

t During the hearings pump modeling was criticized on the basis that pumping action might co the two phase region when it was assumed not to (Transcript pp. 631112 and 7477), and for y]

e applicable two phne performance data, at least in the public domain (Transcript p. 5643). The new addressesitself to theie criticisms.

j For the BWR's, the core flow would not reverse as a result of tne hypothetical accident. The 1

j recirculation line pumps would continue to drive the jet pumps and their continued operation would ten j

]

3 1119 1

w...

7;m

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i to prolong core flow and provide longer initial cooling of the core. GE's present model assumes tnat tne pump head would start to decrease as soon as saturation pressure is reached at the pump suction, goingi linearly to uro as the quality at the pump suction goes to one percent. In analyses reviewed by Regulatory to date, the jet pumps would become uncovered and the. core flow fall to uro before the one percent quality occurs at the pump suction (Exh 1113, p. 6 2). This situation is recogniud in the rule.

. In their responses to the Concluding Statement of the Regulatory Staff, Combustion Engineering and i

1, Westinghouse stated or implied that applicable dau do not yet exist for two. phase pump performance. We l

encourage obtaining such data. There is no disagreement with regard to the use of a dynamic pump model.

1 i

}

7. Discussion of Core Flow Dist-ibution During Blowdown.

'7.j The analytical models used in reviewing the course of a hypotheticalloss of coolant accident under the i

interim Policy Statement have all been one. dimensional, with no direct treatment of now redistribution in the core.The detailed flow in the reactor core following initiation of the hypotheticalloss of coolant would be complex, and would be different depending on the kind of reactor and fuel and the specific time durin the 14CA. The nature and dearee of flow redistribution were discussed at length during the hearing.

Flow redistribution between channels is a phenomenon prirnarily affecting Pressuriud Water Reactors, because they have no channel walls to restrict cross. flow. The principal forces affecting cross flow are friction, acceleration, drag in channeh, drag through spacer grids and fittings, and buoyancy (Morga Transcript pp.12678 9). The presence of a two. phase Guid affects buoyancy and frictional drag (thr the two. phase multipliers). In upflow the buoyancy effects tend to produce higher hot channel flow than average channel Dow (though frictional effects act in the opposite direction). In downflow the buoyancy and the friction act together to reduce hot channel flow relative to average channel flow (Morgan.

Transcript 12679).

The Interim Acceptance Criteria models have accounted for these effects by a requirement that the average channel flow during blowdown of a PWR be multiplied by a factor 0.8 to obtain the flow in the hot channel calculation. Westinghouse Testimony stated that calculations rnade using the THINC code are the basis for this choice of factor. The calculation assumed parallel channels, and uro cross. flow resistance

)

I between channels, so that the pressure w6s constant in every horizontal plane. These calculations did not unambiguouslylead to now reductions bounded by the factor 0.8.

Although the wndors' discussions of the effect of flow redistribution generally tended to support choice of the factor 0.8, there was little sympathy elsewhere for it. Its conservatism was questioned by Rosen (Testimony), l.awson (Transcript $755), and Ybarrondo (Transcript 6076. 6270,10255). and its continued use has not been proposed by the Staff.

i It appears that considerations of flow tedistribution prior to the hearing comprised only circumstances t

in which the clad is not deformed. Questioning during the hearing also dealt with effects of clad swelling.

fuel deformation,and partial blockage on flow redistribution. !! was less apparent that the factor 0.8 would be adequate if partial channel blockage occurred than if channels were undeformed.

The Staff's Supplementary Testimony recogniud this point, and proposed that models be developed t

[;.

and used that explicitly calculate the effect of flow redistribution during both the upflow and downflow Q

phases of blowdown. The view included use of models that calculate the flow redistribution resulting from M

flow blockage if that should be calculated to take place.

• )

We believe this is the correct course to follow. We believe the wording in the Staf!'s proposed rule j

adequately expresses the position supported by the record, with one exception. There is no basis in the record for continued use of the flow reduction factor of 0.8 after flow redistribution effects have calculated for the hot channel. We have not included this requirement in the Rule.

L

{

D. POST. BLOWDOWN PHENOMENA; HEAT REMOVAL BY THE ECCS 1.31agic Tellure Criterion. An analysis of possible failure modes of ECCS equipment and of their effects on ECCS performance must be made. In carrying out the accident evaluation the combination of ECCS subsystems assumed to be operative shall be those available after the most damaging single failure of

. ECCS equipment has taken placr.

2. Containment Prerrure. The containment pressure used for evaluating cooling effectiveness during seflood and spray cooling shall not exceed a pressure cafeu!ated conservatively for this purpose. The calculation shallinclude the effects of operation of allinstalled pressure reducing systems and processes.

-. = - m: * ~

? P. ;

~

3.Colcuktion of Reflood Rate forPressurited Water Reactors. The tefilling of the teactot vessel and ws s tne aion. g:ing the time and rate of reflooding of the core shall be calculated by an acceptable model that takes into Hegulatory consideration the thermal and hydraulic characteristics of the core and of the reactor system. The p

. eie percent system coohnt pumps shall be assumed to have locked impellers if this assumption leads to the maximum calculated chddmg temperature; otherwise the pump rotor shall be assumed to be running free. The ratio cenng and the total fluid flow at the core exit plane to the totalliquid flow at the core inlet plane (carryover fr shall be used to determine the core exit flow and shad be determined in accordance with applicab mance. we

.mp model.

experimental data (for example,"PWR FECHT (Full bngth Emergency Cooling Heat Transfer) Final Report," Westinghouse Report WCAP 7665, April 1971; "PWR Full bngth Emergency Cooling H l

Transfer (FLECHT) Group i Test Report," Westinghouse Report WCAP 7435, January 1970; "PWR FLECHT (Fuu kngth Emergency Cooling Heat Transfer) Group !! Test Report," Westinghouse under the WCAP7544 September 1970; "PWR FLECHT Final Report Supplement," Westinghouse Repo Arnonin WCAP 7931. October 1972).

. ant w uld The effects on reflooding rate of the compressed gas in the accumulator which is discharged

me dunns accumulator water discharge shall also be taken into account.

';e '

4.Ste0m }nteraction Mth Emergency Core Cooling Water in Pressurized Water Reactors. The M *' d' 5 -

thermalhydraulic interaction between steam and all emergency core cooling water shall be taken into

• 0"* 8

dhvpn.

account in calculating the core reflooding rate. During refill and reflood, the calculated steam flow in i

i unbroken reactor coolant pipes shall be taken to be aero during the time that accumulators are d nhoough water into those pipes unless experimental evidence is available regarding the realistic thermal hydraulic d,.m tlun interaction between the steam and the liquid. in this case, the experimentd data may bc used to s sua.s an.T alternate assumption.

j N'" F"'

5.RefillandRefloodHeat Tranxfer for Pressuri:ed Water Reactors. For reflood rates of one inch p second or higher, reflood heat transfer coefficients shall be based on applicable experimental data for

• thn dw unblocked cores including FLECHT results ("PWR FLECHT (Fuu bngth Emergency Cooling Heat q

']

Transfer) Final Report," Westinghouse Report WCAP 7665, April 1971). The use of a correlation deriwd )

I from FECHT data shau be demonstrated to be conservative for the transient to which it is appl presently avaihble FLECHT heat transfer correlations ("PWR Full bngth Emergency Cooling Heat JiJ n."

Transfer (FECHT) Group 1 Test Report," Westinghouse Report WCAP.7544, September 1970;"PWR FLECHT Final Report Supplement," Westinghouse Report WCAP.7931. October 1972)are not j

'"PP'"'

New correlations or modifications to the FLECHT heat transfer correlations are acceptable only!

• • J D are demonstrated to be conservative, by comparison with FLECHT data, for a range of pararneters

. and w.

consistent with the transient to which they are applied.

During refill and during reflood when reflood rates are less than one inch per second, heat transfer d

calculations shall be barn on the assumption that cooling is only by steam,and shall take into ac

• E"8 -

flow blockage calculated to occur.as a result of cladding swelling or rupture as such blockage mig

• m.m!J both local steam flow and heat transfar.

3

6. Convectin Heat Trentfor Coefficients for Bolling Water Reactor FuelRods Under Spray Cooling Nd l

Following the blowdown period, convective heat transfer shall be calculated using coefficients based on

""0"'

appropriate experimental data. For reactors with jet pumps and having fuel rods in a 7 x 7 fuelassemblyl "8U""

array, the following conweti e coefficients are acceptable:

(a) During the pmod followingloose plenum flashing but prior to the core spray reaching rated flow

,,j,g conwetive heat transfer coefficient of aero shal be applied to D fuelrods.

(b)During the period after core spray reaches rated flow but prior to reflooding, convective heat transfer coeflicients of 3.0,3.5,1.5,and 1.5 Blu-ht"8-ff ?F8 sha!! be applied to the fuel rods in the outer 8

corners, outer row, next to outer row, and ta those ternaining in the interior, respectively, of the assembly.

1 (c) After the two phase teflooding fluid reaches the level under consideration,a convective heat transfer l shall be applied to all fuel rods.

l coefficient of 25 Blu hr' ff /F8 8

s then

7. The Bolling Water Reactor ChannelBox UnderSpray Cooling. Following the blowdown period.hs un on transfer from, and wetting of, the channel box shs11 be based on appropriate experimental data. For a:uie. f reactors with jet pumps and fuel rods in a 7 x 7 fuel assembly array, the fo!!owing heat transfer coefficients and wetting time correlation are acceptable.

(a)During'the period after lower plenum flashing, but prior to core spray reaching rated flow, a se conva.ctive coefficient of zero shan be applied 1o the fuel assembly channel box.

~

.es

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1121 1

. =. v =. : -- ~ - r = - ~~ ~ -

ri

., s (b)During the period after core spray reaches rated flow, but prior to wetting of the channel, a convective heat transfer coefDeient of 5 Btu.hr-8-ft*8 *F3 shall be applied to both sides of the channel box.

(c) Wetting of the channel box shall be assumed to occur 60 seconds after the time determined using the correlation based on the Yamanouchi analysis (" Loss of Coohnt Accident and Emergency Core Co Modeh for General Electric Boiling Water Reactors," General Electric Company Report NEDO.10329, April 1971).

DISCUS $10N Containment Pressure. One of the anomalies of the LOCA is that, although one would normally think that it would be desirable to have a low containment pressure to reduce leakage, a high containment pressure would be advantageous for the operation of the ECCS. A high containment pressure after the accident would terminate the blowdown sooner and would improve the capability of the ECCS because '

conwetive heat transfer coefficients are higher at high pressures. For the PWWs,a high ambient pressure t

would reduce the steam binding (through the higher steam density) and would lead to high renood rates, which would further improve the heat transfer coefScients. Thus a guide is needed for the containment pressure that can be used in the aralysis of the hypothetical LOCA.

He Interim Policy Statemerit specified that the maximum containment pressure allowed for the

~

calcuhtien of the effectiveness of the ECCS should be the original pressure (presumably atmospheric)plus 80% of the pressure increase estimated to be brought on by blowdown. For containment systems used so far for P%Ts (dry containment) this prescription was shown to be conservative by the Regulatory Staff (Exhibit 1113, Sec.15). However it is not too difncult to calculate the actual pressure, allowing for the various cooling devices placed in the containment structure to limit its pressure,and the new requirement is i

e to make such calculations. Impioving the generality of the requirement has made it applicable to other j

i types of containment, such as ice condenser containment.

ne importance of containment pressure has been less for BWR's than P%Ts because steam binding is less of a problem and because the heat transfer coefficients available for use in BWR analysis were derived g

at atmospheric pressure and so are conservative for elevated presures.

ti in their response to the Staffs Concludmg Statement, Combustion Engineering states a preference for j

the old formuh. As the nrw rule is stated the old formula may still be used, provided it is shown that the i

g pressure so calcuhted continues to be less than that obtained by a detailed calculstion, Both Combustion Engineering and Westinghouse pointed out in their responses to the Staffs a

h

- l Concluding Statement that the LOCA analysis is rnade assuming some loss of power, under which condition some of the pressure reducing systems might not be operable. They therefore suggest that for conshtency

. d!

j one should not assume that all of the pressure reducing systeins would be operating. Since it is possible for a

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the LOCA to occur with no loss of power, the Commission rejects this suggestion.

Calculation ofRefloodRatefor Pressurized Water Reactors. The renooding tste for pressurized water

(- l q

tl,

i reactors would be controlled to a large extent by steam binding, the phenomenon by which the resistance 2

to flow through the reactor system (steam generators, pumps, etc.) of the ernuent from the reactor core st E;

limits the rate of renood and, indirectly, the rate of heat removal from the fuel rods. The pumps in their

,M h

locked rotor condition would typically provide more than half of this resistance to flow so that the stipuhtfon of their being locked is a serious limitation. If the pump rotors were not locked, their resistance ar i 7:i to flow would be reduced by 60% (Exhibit 1113, p 1410). In their Concluding Statement, Combustion es; Engineering states that if the pumps were free running during renood the calculated maximum temperature pr l i

of the zircaloy cladding would be reduced by 75'F (CE Concluding Statement, p 3-61).

es,

'I The stipulation of locked pumps during renood is unchanged from the Interim Policy Statement,and th i no new experimental information was provided during the hearing justifying a change in this part of the

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rule, be -

The Reguhtory Staffin their Concluding Statement proposed the development of more sophisticated We refill.renood computer programs, including those capable of predicting the expected oscillatory flow of water into the reactor core. The Staff sho proposed that the calculation should consider the carry over of be -

fluid from the top of the reactor core to be based on experimental data, principally the FLECHT tests',

dif i which were carried out with fixed flow rates. Combustion Engineering, who have a sophisticated code, sue PERC, that predicts oscillations, pointed out a difficulty in their Comments on the Staffs Concluding an.

Etatement, in that use of the experirnental carry over data would snake major portions of their dif 1122

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T.. --~Em.

.-.e

o.. -...... -.-

.. -.. +-

o

.. -=

NN.fl@%!lt#E2Eh 7

t channel, a Sophisticated code superfluous. Since the heat transfer coefficients must come from the Fl.ECHT tests,at annel box.,

least for now,it seems logicaland consistent to use those data to determine the s. mount of fluid leaving the ined using core and passing out through the system. The Commission believes with the Staff that improved and more ce Cooling realistic models are desirable, but realizes that the full benefit of sophisticated models that predict the yo.30329, oscillatory flow cannot be obtained until there are more suitable experiments with which they can be compared.The rule, as written, allows for the use of new data and of more sophisticated codes when they are avallable.

As sever 31 participants haw observed, the method chosen for calculating the mass of fluid leaving the top of the reactor overestimates it, at least for the FECHT tests from which it is derived, by the amount t:lly think of two phase fluid retained in the core above the quench front. To the extent that this carryever fraction is ntainment overestimated, the amount of steam binding is overestimated and the reflood rate is underestimated, a after the providing an additional rnodest conservatism. The inethod that we have chosen for calculating the 3 becauM carry over fraction is,howver,as realistic as the FLECHT data allow.

Pituute Steam Interaction with Emergency Core Cooling Water in PWR 's. The steam flow coming from the core lood rates.

through the cold legs of Combustion Engineering and Westinghouse PWR's would be subject to increased ntainment resistance as a result of the high accumulator injection flow rates (thousands of gallons per minute)into those legs. The restriction of no steam flow in the non ruptured legs during ECC inje: tion is conservative;it d for the was part of the Interim Policy Statement. Combustion Engineering has already proposed a new prediction a

herie) plus model based on experiment;.I evidence (Concluding Statement,p 3 64; Exhibit i144, pp 810 to 818),and ns used so Westint, house also proposed subrnitting a new model (Concluding Statement, p 80).

i tory Statf Repood Heat Transfer for Pressurized Water Reactors. The convective heat transfet coefficients used ng for the for calculating the cooling of the fuel rods during the reflood phan are derived from the PWR FLECHT test frement is program (Exhibit 150). In these tests electrically heated rods, simulating fulllength fuel rods,were cooled f

e 13 other by bottom flooding at various rates. Characteristically, after the very bottom of the heated section of the rods had been wetted, the rapid generation of steam led to an entrainment of water droplets and the s

generation of a two phase fluid that swept upward through the rod bundle. The mechanisms of heat te sed transfer to this two phase fluid have been postulated to be (1) convective cooling to the steam,(2) droplet impingement on the rods and (3) radiation to the water droplets (and to the steam). Alms at least part of erencefrr the length of the rods the two phases would not be in thermal equilibrium,it, the steam would be at a a that the higher temperature than the water droplets,and therefore the water droplets would be heated by the steam and evaporate as they pass up the column (Exhibit !$0, pp 3-69-3 77).This complicated heat transfer se Staffs mschanism was interpreted by calculating heat transfer coefficients, which, when multiplied by the conditi:n difference between the measured surface temperature of the rods and the saturation temperature of water esisterey at the test pressure, yielded the measured heat flux from the rods to the coolant.

issible for The FLECHT tests and their calculated heat transfer coefficients have been criticized on several bases (Transcript, pp 6868,19,489; Exhibit 1041, Sec. 6).The principalitems questioned were: the effect of the zed wat:r flow housing, the use of stainless steel for the cladding of the heater rods instead of zircaloy, the use of resistance steady flow instead of allowing the flow to oscillate as it would in a reactor,and the probability of enors.

etor co88 Each of these will be discussed in tum.

s!2 their

- The FIECHT tests were carried out with bundles of either 49 or 100 rods, arranged in a square array that the and surrounded by a steel housing about % inch thick.The peak temperature of the housing was typically esistance 750*F, even when the rods next to it were at temperatures in the vicinity of 2000*F. he concern was mbustion expressed that this housing did not suitably simulate the surrounding rows of fuel rods that would be iperature present in the reactor. The Commission believes that this question has been adequately explored by experirnent and examined in the record (Exhibit 1113, pp 17 2; Exhibit 1078, pp 4659), and concludes nent, and that the heat transfer rnessured for the inside rods of the bundles was not affected in any major way by the at of the housing. The effect of radiation to the housing on the calculated heat transfer coefficients was estimated to be less than 5% by Aerojet Nuclear Corporation (Exhibit !!!3, p 17 3) and about three percent by sisticated Westinghouse (Exhibit 107, pp 4652).

1. flow of Stainless steel was usti instead of zircaloy as the cladding material for nearly all of the FLECHT tests Y

' of because it is more durable under the test conditions. Although it is not usual to expect significant differences in convective heat transfer coefficients from different solid material surfaces,the possibility of such differences was considered, perhaps resulting from such factors as differences in thermal conductivity 4 -

'ncluding and differences in wetting properties. The reasonable conclusion was reached that the effect of the of their difference between zircaloy and stainten steel,if any, would be small. There is a difference, of course,in

\\

1123

. i f 1.;. j _..

y -

y the rate of heat generation from stam oxication, but this heat is deposited within the metal under the i

surface of the oxide film. The presence of this heat source should not affect the heat transfer coefficients, v

which depend on conditions in the coolant outside the rod, s

The few FLECHT runs made with zircaloy clad rods provide uncertain and conflicting evidence.

t Westinghouse pointed out that all of the zircaloy runs except one (run 9573) yield higher heat transfer c

coefficients than were obtained with steel (Westinghouse Concluding Statement, pp C 74 to C-76; Exhibit 150, pp 3 98 & ff). Consohdated National Intervenors pointed out that inost of these runs were made at i

unreasonably high Goodmg rates, and that a different result was obtained from run 9573 where the f

flooding rate was about one inch per accond. In the first 18 seconds of this run, before multiple heater rod v

failures occurred, the aircaloy clad rods heated up faster than predicted from the stainters steel based 5

correlations (Exhibit 1041, pp 6.7 & ff). His anornalous result has been attributed to experimental error, c

or possibly to an unusually skewed initial temperature distribution along the length of the rod (Exhibit t

4 Ill3 pp 17 6-17 7).

t, t

On balance, the Commiannon sees no basis for concluding that the heat transfer mechanism is different cl

' for aircaloy and stainless steel, and believes that the heat transfer correlations derived from atainless steel clad heater rods' are suitable for use with zircaloy clad fuel rods. It is apparent, however, that more c

experiments with zircali. cladding are needed to overcome the impression left from run 9573.

vj At a number of places in the record mention is made of the osctilatory now that would be expected in c) reflooding a reactor (Exhibit i113, Sec.17, pp 1,2. I1,12,13; Exhibit i144,Sec. 9)and of an enhanced I

heat transfer to be expected from it (Trans.p 6838). In the P%R FLECHT tests presently available, the a

water entered the bottom of the rod bundle at a predetermined rate,without provision for the development i

of an oscillatory flow. Westinghouse is carrying out another group of tests called FLECHT SET with a hydraube system that more closely simulates a reactor, which is expected to allow oscillatory flow to take a

place. he results of one run, run 5,are discussed in some of the above references.This run did exhibit the c

oscillatory flow. The initial reflood rate, averaged over the oscillations, was quite high, but settled down to s

a value of about two inches per second within 20 seconds. After allowance is made for the effects of the r

high initial flow, the heat transfer coefScient observed is very similar to that which would have been i

predicted from the prior FLECHT tests. (Exhibit 1 I13, p 1712). Thus, until more experimental evidence is e

available, heat transfer coefficients should continue to be based o: the FIICHT tests,and caution should s

be exercised in asserting the existence of a conservatism because th.y are baud _on steady flow.

s he accuracy of the FLECHT determined heat transfer coefficients has been examined several times.

(Cf. the review in the Babcock and Wilcox Concluding Statement, pp 202 204.) Westinghouse estimated a J

5l possible urcertainty of 12% in the coerncients. (Trans. page 6878). The Aerojet Nuclear Company s

concluded "that the FLECHT data currently represent a best estimate of the heat transfer that will occur in c

a large undistorted core." They also concluded that an allowance of up to 20% may be needed "to bound 1

the data due to experimental and inferential errois."(Exhibit il13, p 1714)The Commission apptaves of cl the un of the FLECHT data for calculating P%R reflood heat transfer, but notes that these will be more r

nearly "best estimate" calculations than bounding calculations.

5 T.4e PWR FLECHT Final Report Supplement,WCAP 7931, gives revised formulae for the calculation of t

heat transfer coefficients, and,in a series of curves, compares both the old and new calculations with the c

experiments. Then curves indicate that the calculations, both old and new, predict greater heat transfer c

than would actually occur in the early part of the renood transients for low reflood rates.(Exhibit 1113, p 1714). As a result,we requite that new coef0cients be used in this region, together with a demonstration t

that they represent the data in a conservative inmaaer.

s The FLECHT tests simulated now blockage in a number of runs by the insertion of perforated I

horizontal plates. With reflood rates of one inch per second or higher, imptovement was found in the rate f

of heat transfer as far as tw feet upstream and four feet downstream of the blockage.The improved heat i

b transfer was shown to be caused by break up of the entrained droplets and increased turbulence.(Exhibit t

I 1006a). The blockage in these tests ranged up to complete blockage over several channels with 75%

(

blockage in other channels. For the flow blockage tests at a reflood rate of 0.6 inches per second, heat transfer was degraded by blockage. Presumably the poor results at the low re0md rate were the result of a 1

3 lack of entrained water droplets,lesving only single phase steam cooling. (Exhibit Ii13, p 17 5).

s The FLECHT Dow blockage tests were criticized on the basis that the flat plates were not typical'of bulging of the cladding. Hawever, Davis tried blockage with sleeves versus plates and found litile difference.

1 (Trans.p 4130),

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As a result of these tests it appears that heat transfer coefficients based on undistorted rod geometry

fficients, would provide a reasonable approach to estirnating core temperature behavior during renood, for rerlood rates above one in/sec. For lower reflood rates blockage would have a deleterious effect and one must

evidence, j

resort to calcuhtion with single phase steam cooling, taking into consideration the effects of blockage on

(

transfer core flow distribution.

.; Exhibit Convective Heat Transfer Coefficients for BWR FuelRods UnderSpray Cooling. The time history of a made at hypothetical design basis Boiling Water Reactor accident can be divided into six periods (Exhibit 1001, here the pp 4-7 & ff). A flow coast.down period would end when the water level fell below the jet pump inlet. There satn :d would then be a short period of flow stagnation in the core undl more water escaped from the pressure vessel. When steam started to escape through the break there would be a rapid drop in the system pressure,

{

ni based tal grr:r, causing the water in the lower plenum to flash to steam, creating a two phase flow through the core. After

{

l (Exhibit the end of flashing there is assumed to be a short " core heatup period" during which no core coolirt would j

take place before the ECCS came into action.The ECCS would start first with a spray of water into the top i

different of the core and later flood the core from the bottom.

ies steel Flooding the core is said to occur within three minutes of the piping break in the design basis accident, at more or within awo and a half minutes after the start of the core spray. (Exhibit 1113. p 16 36).The reflood rate would be quite high, typica!!y 3.7 inches per ucond (Exhibit 137, p 23) and would terminate the excursion

- <cted in i

quite rapidly once the reflood water started entering the bottom of the core. (Exhibit 1069, p 15). Thus

.nhanced the function of the core spray is to keep the cladding temperature from rising too rapidly during the abie,the approximately two and a half minutes between the core heatup period and the quenching setion provided opment by the core reflood.

f with a The BWR fuel rods are in 7 x 7 arrays, with each array contained in its own channel box. With this to take arrangement each group of 49 fuel rods is largely independent of the rest of the core as far as coohng is ilbit the concerned. The chsenel boxes would not becorne as hot as the fuel rods, since their only source of heat i

Iownto would be absorption of thermal radiation and gamma rays from neighboring fuel rods,and they would be s{

more easily quenched by the core spray. They would serve as a convenient heat sink for radiation from the

<e\\

hotter fue! sods, especially those near the outside of the 7 x 7 array. There would be a diffusive flow

.ience is outward of heat from the inner rods, through radiation. The fuel rods would s!so lose some heat to the should water droplets and associated steam by convection and radiation. Through these mechanisms the core spray would limit the rate at which the core heats up.(Exhibit 1069, p 15).

I times.

From the BWR FLEC}fr tests there is information on the heat transfer coefficients for both the nord a convective heat flow to the water droplets and steam and for the reflood phase. The FLECHT tests were mpany rnade with an electrically heated mock-up of a 7 x 7 rod array complete with its channel box. The ccur in convective heat transfer coefficients were determined from the residue of a thermal balance after allof the bould known inputs and outputs were calculated. The factors considered were the electrical heat input, the rate of aveser change of the heat content of the rods as calculated from their temperature history,and the calculated radiation from the rods to each other and to the channel walls.The residue from these inputs and outputs e more was ascribed to convective heat transfer. The convective heat transfer coefficier ts so determined could not

son of be very accurate because their calculation involved taking the difference between two large numbers.The ith the coefficients so obtained are small and are about what one would expect from the mechanisms of natural ansfa convection and radiation :o steam. (Exhibit 1113, p 1614).

1113 The values of the calculated convective heat transfer coefficients depend to some extent upon the value ration i

used for the thermal emissivity of the stainless steel, since the conveedve heat transfer is obtained after subtracting the radiative heat transfer from the total. Theoretically a high value of the emissivity leads to a arated low calculated convective heat transfer coefficient. Values of the emissivity measured after the tests ranged t2 rate from 0.6 to 0.9 (Exhibit 461,p 81 and Exhibit li13,p 1614),and to add conservatism to the calculation, d heat the Interim PolicyStatement required the use of the highest measured emissivity.0.9,for the calculation of shibit the convective but transfer coefficients. However it turned out that this resulted in a higher coefficient i 759 i

(less conservative) for the criticat inner rods, with a higher estimated standard error. (Exhibit 461. Table 2.)

, heat After reviewing the derivation of the coefficients as given in Exhibit 461, we believe that those originally itofa listed as best estimates by General Electric are the most credible and should be used. The effect.of this change on the peak cladding temperature will be small, about five degrees according to Exhibit 461.

a There has been a great deal of criticism of the BWR FLECHT tests, particularly by the Consolidated enu.

National Intervenors (Exhibit 1041, Chapter 5), and both General Electric and the Regulatory Staff have defended them (Closing Statements). However, for the purpose of calculating the rnaximum cladding 1125

. ]L

..__......~}.~

Z. q--

i temperature, only the derived heat transfer coemeients are of any great importance. The values obtained have always been known to have a high statistical error;furthermore the values are low and reasonable, and ten there seems little to be gained by renewing the controversy over the manner of conducting and inter mo all features of the tests.

The high but inevitable statistical error of the coemeients for the inner rods (1.5 s 1.0 BTU /hr-ft'.*F) sha h bothersome and leads to an estimated error band of as much as 1200*F in the calculated peak temperature in some circumstances (Exhibit 1113, p 1636). The test bundle SS2N was used to derive the mo heat transfer coemetents; another test bundle, SS4N, resulted in cladding temperatures 200*F higher than those of the bundle used as a standard; one half of this discrepancy could be explained by test differences on with the other half left to be attributed to statistical variations. (Exhibit 1113, p 16 38). The problem of inct these large statistical errors in the convective heat transfer ooemeients is compensated to some exten whi the fact that the coemcients were determined at atmospheric pressure, whereas the reactor would be at

'Y(

some elevated preuure at which the heat transfer would be improved (Exhibit 1113, p.16 26).

con "9

The evidence for the value 25 BTU /hr.ft'.*F of the two phase reflooding heat transfer coemeient is sketchy (Exhibit 1032, p.II-6.3 51), but it is applied for only a short time because the high reflood rate ades would quickly quench the core,and the exact value is oflittie significance, Apg The Elm ChannelBox under Spary Cooling. Radiation to the channel box would be an important ottu snechanism for cooling the outer fuel rods of a 7 x 7 array, and the magnitude of the radiative coolingapp I

would depend to some e.~ tent upon the temperature of the channel box. During the LOCA the channel box would be heated by radiation from the fuel rods and by absorption of psmma rays, and later it would be DRS cooled by the core spray, its temperature would drop sharply after it is wetted by the core spray and I

subsequently it would become a better heat sink.

be t The time of wetting has been calculated by an extension of a theory developed by Yamanouchi 669 (Yamanouchi, A., Journal of Nuclear Science and Technology, 5, pp.547 558, Nov.1968). This theory calculates the progress of the wetting front for spray cooling by considering the longitudinal thermal N

conductance of the cliannel wall. Although the direct application of this theory over. predicted the wettiry N

times observed in the FLECHT tests,it was possible to correlate the data from the stainless steel runs with a i

group of channel parameters that were involved in the theory. The data for the runs with zircaloy claddingseks had more scatter, with both positive and negative deviations from the stainleu steel correlation line. By som adding one minute to the witing times predicted by the correlation, all but one of the quench times e

Id'"

observed in the FLECHT tests were encompassed. (Exhibit 461, g6) Modified in this way,the calculated y

quench times seem to be adequately conservative.

'5 With regard to the convective heat transfer coefficient to be used during core spray but prior to wetting, 3

I General Electric's present practice is to use a convective heat transfer coeffielent of 20 BTU /hr.ft'.*F d

applied to only one side of the channel box. This number was derived from the FI ECHT experiments as a'

%I best fit to the data. (Exhibit 461, p8) The Regulatory Staff points out that the geometry in the Fl.ECHT

1. P

^

M experiments differed from that in a reactor,in that there was an insulated outer channel surrounding the h

channel box in the experiment. They calculated that radiation from the channel box to the outer channel heari

- may have contributed up to half of the heat transferred from the channel box in the FLECHT experiments.

3*'"

3 (Exhibit 1113, p.16 8) They therefore recommend reducing the convective heat transfer coefficient by a

)

j factor or two. Although General Electric objected (Responsive Closing Statement, Vol. 2, pp.80 81) on the gl

.(

basis tha'. the cateulation is already over. conservative by virtue of the minute added to the channel wetting i

7 i

time, the Comtnission supports the position of the Regulstory Staff on the basis that no single coolingI mechanism should be counted upon to exceed its expected performance.

y l

h W. REQUIRED DOCUMENTATION i

mo&i i

techr. !

1.(s) A description of each evaluation snodel shall be fumished. The description shall be sufficiently the Rl complete to permit technical review of the analytical approach including the equations used, their B!

approximations in difference form, the assumptions made,and the values of all parameters or the procedure, of th of th for their selection, as for example,in accordance with a specified physicallaw or empirical correlation.

I (b)The description shall be sufficiently detailed and speelfic to require significant changes in the the et evaluation model to be specified in amendments of the description. For this purpose, a significant change is a change that would result in a calculated fuel cladding temperature different by more than 20*F from the t

1126

n

.- --, -- T: ~~.~ ::,= -- -

g f~

M I

s,b temperature calculated (as a function of time) for a pouulated LOCA using the last previously accepted

model, reting (c) A complete listing of each computer program,in the same form as used in the evaluation model, shall be furnished to the Atomic Energy Commission.

8?F) 2.For each computer program, solution converpace shan be demonstrated by studies of system Peak modeling or noding and calculational time steps.

9the

3. Appropriate arnaitMty studies shall be performed for each evaluation model, to evaluate the effect than on the calculated results of variations in noding, phenomena assumed in the calculation to predominate,

nees, I

including pump operation ce locking, and values of parameters over their applicable ranges. For items to sn of which results are shown to be sensitive, the choices made shall be justified.

tt by be at 4.To the extent practicable, predictions of the evaluation model, or portions thereof, shall be compared with applicable experimentalinformation.

5. General Standards for Acceptability-Elerrdnts of evaluation modeh reviewed willinclude technical ntis adequacy of the calculational methods, including compliance with required features of Section I of rate Appendix X to 10 CFR Part 50 and provision of a level of safety and margin of conservatism comparable to other acceptable

• valuation modeh, tPing into account signiDeant differences in the reactors to which they tant

apply, g

sling box DISCUSSION:

d be l

and Previous Experience with the Interim Policy Statement has shown that additional doeurnentation would be useful (Exhibit 103), page 2; CN1 Concluding Staternent, page 4.16; and Transcript pages $643; 6675; uchl 6691;10,879; to 10,883).

,,,y Considerable hearing tirne was devoted to consideration of the adequacy of codes and analysis method

{

m3 (see Exhibit 1043 and Transcript pages 8294; 8386; 11,06511, !!2: 11,156). Time would be saved in the

. hearing process,in generic reviews, and in case reviews, if for each evaluation model a detailed description ting I

tg*

were provided which denned the analytical approach und equations, the assumptions, the references, the 33g selection and justillcation for the input parameters, and the mathematical symbolism used to establish the j

i By c rresponding computer prograrns. A complete description and listing of the computer programs, in i !

identical form to those approved and being used (at a specific time) for LOCA analyses,is needed l

nes' Reguhtory Staff so that they can be certain that the codes used for asfety analyses always correspond to

' {

ted the approved, published evaluation model.

,g, it is recognized that revisions in the evalustion modeh w!U be made from time to tirne,within the

,a p restrictions imposed by the section on required and acceptable features. When these revisions constitute a t

"significant change" as defined in paragraph 1.(b) above, the changes must be described in detail and an l

gy updated tevision of the computn codes provided.

he The need for noding and sensitivity studies for the computer programs is clearly reflected by the

.,,3 hearing record (e.g., Exhibits 2006, 1043, 1044, 1001, 1113, !!48). This rule formalizes the scope and intent of such studies.

n The need for comparisons of the calculations of analytimi models with experimental data is discussed f

and the value is recognhed in the written testimony of nearly all of the participants, including the he ng Regulatory Staff (Exhibits 1001 & 1113). Westinghouse has stated the existence of some problems in i

ng i

lat'rpreting the requirement for such comparisons. It is reasonable to restrict these comparisons to those

!j that the proponents of evaluaSon modeh have msde of their own volition to check out certain features and

.I to cornparisons requested by the Regulatory Staff.

In their comments Babcock and Wilcox suggested omission of the technical review of the evaluation models. it is the Commission's opinion that,with the changes being made by this rule it is necessary that a technical review of the evaluation modeh be made by the Commhsfon; this review is the responsibility of ly the Regulatory Staff, e

Both Westinghouse and General Electric objected that the subject taf computer codes should not be part of this rule. As indicated in the above references, the codes were discussed at length in the record in terms re 8

of their adequacy and content; that there should be this much question is deemed sufficient reason to have

{ o g-the computer codes revealed to the Commission.

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CONCURRING OPINION OF COMMISSIONER ANDERS i

Though ! join in the Commission's rulemaking decision issued today, a staternent of the basis for my concunence may be helpful to the participants. As to the substantive rule itself,I agree generally that the record leads to our conclusions. In my view, however, a weakness of the present record is that it does not provide an adequate basis for a thorough analysis of the benefits and penalties of actual compliance with

,.l i

the rule's operational requirements. I am nonetheless able to concur with the prescribed implementation g

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procedure because it offers the potential for development of information now lacking.

I Compliance with the new rule's operational limits within any short time period wiU entad certain i

penalties. At the same time,such compliance willachieve some degree of safety benefit.These factors were summarized by the Regulatory Staff (with respect to its recommendation) thusly:

, h.--

(a) decrease in an already low radiological risk to the natural environment and public health and safety from postulated lossef. coolant accidents;(b) potential derating averaging about 5 percent of capacity

[

of nuclear power plants for about two years, with attendant increase in air pollution and economic costs from increased operation of fossil-fueled generators;(c) increased nuclear generation costs due to cost of fuel design changes permitting resumption of full. power operation. (Environmental Statement r

P.1).

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While these benefits and some penalties were identified, they were not quantified to the degree of

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approximation that I believe possible and clearly warranted by this important issue.Moreover,the criteria by which the balance was struck were undefined. in short, the staff's cost. benefit analysis (Final

_1 Environinental Statement, pp 99149) was neither adequate not persuasive. The Commission has properly

~

f rejected the precipitous implementation schedule recommended by the staff since the burden ofjustifying It was not mit.

Today's decision does not exclude the possibility of an exemption from the requirement that a reauer be brought into compliance with the new rule six months after its effective date. Hopefu!!y, few,if any.

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exemptions will be found warranted after careful review.in this regard, as stated in Section 50.46(a)(1)(vi, vil) of the new rule, a licensee must show to the satisfaction of the Commission that any such exemption would be in the pubile interest. This procedure leaves room for submission and Commission consideration of appropriately detailed showings and quantification and analysis of factors affecting the implementation

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w schedule such as those I now discuss briefly.

9 First,it is undisputed that the hypotheticallosser. coolant accident is itself a highly unlikely event.The Q

staff repeatedly acknowledged on the record that the probability of such an accident, coupled with a 4 }7 simultaneous failure of an ECCS conforming to the Interim Acceptance Criteria,is" negligible"(Ex.2023.

75*:3 d,

Tr. 22179; Transcript-Oral Argument Before. AEC, p 20). The probability of these simultaneous 1

occurrences has been estimated at approximately 10 per reactor year or one in 10 million reactor years 4

^

(Tr. 22177, 22181, 22185). Even if such a highly unlikely coincidence would occur, the reactor r.

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containment structure would still provide substantial additional margins of safety for public protection.Of course, there must be reasonable assurance of public health and safety, and the new rule unquestionably i

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affords a somewhat increased degree of assurance than the old. But, without attempting to prejudge the outcome of any particular exemption request, the advantage of a further reduction in an already S

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" negligible" risk must be weighed critically against potential adverse impacts of rapid implementation.This 4

'j is consistent with the Commission's statutory responsibility to regulate "in the national interest" (Section 2e of the Atomic Energy Act of 1954,as amended).

d 3

Second, compliance with the rule's operational requirements prior to fabrication and loading of h

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I redesigned fuel will apparently require a derating of some plants. For example, the Regulatory Staff

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estimated that its proposed rule could produce a temporary average derating of nuclear generating capacity of about 5% (Final Environmental Statement,pp 108,130).Such deratings could conceivab!y cause power J

reserves to fall below acceptable limits with increased probability of power outages. Occurring at a time when the nation faces an unpualleled energy crisis which already threatens to' produce such disruptions.

4 y

deratings would exacerbate a critical situation. The potential adverse impacts of these deratings certa z

would not be " negligible."

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Third, derating also would force utilities to resort to attemate fuels to make up the lost energy.

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International conditions involving uncertain oil supplies strongly suggest the reduced availability of this y " dp fuel. To the extent that utilities must tum to coal, derating could cause some adverse environrnental J

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consequences. In the Chicago region, for example, the staff has stated that "there would be a significant impact on air pollution"if a 5% derating were to occur (FES,p 126). Air pollution increases could result in adverse effects on the health of persons living in affected areas,thereby tendmg to nullify the public health and safety considerations which were submitted to support the recommendation for rapid implementation, my The presence of these factors shows the need for more information and better analysis so as to provide the the Commhston a more rational basis for an assessment of the true impact of rapid compliance with the not operational requirements of the new rule. it may be that the incremental safety benefit to be gained will uth justify the costs. On the other hand,it may be that the benefit 'of rapid compliance is outweighed by the th18 various dertting penalties-penalties which might be avoided by requiring compliance with the rule in accordance with a reactor's normal refueling cycle. In any event, showings can be made in support of-or sain in opposition to-.-requests for exemption submitted under the procedures described in our decision. It is 4ere precisely because we have devised these procedures to obtain the information now lacking that I am able to

.1rty concur in today's dedslon.

In addition,I wish to add a few words conceming the question of proprietary data. Protection of AHY proprietary matters legitimately preserves for industry the benefits of its own research. This fundamental

'nSC principle of our free enterprise system fosters private investment in safety.related R&.D which in turn has 8e m resulted in enhanced safety margins for our nation's nuclear power plants. On the other hand, the public

.ent interest clearly requires some disclosure of data which provides the underpinning of a particular safety I

requirement applicable to an entire industry. The public has a legitimate right to know and examine such data for the purpose of determining its vahdity.But this right must not be allowed to be abused especially

.,,,(

as IJ.S. industrial participants are able to provide substantial support to safety.related RAD in an tena environment of rising domestic and foreign competition. Balancing these conflicting considerations is a a mal y,gy difficult task, in the special circumstances of the case which is before us, I would agree that the Combustion Engineering item should be made public; and that this action should not be considered as mn, precedent. Hopefully, procedures which can satisfy all legitimate interests will be developed in the near

,{

future.

Finally, as we recognize in today's decision, the inquiry does not end here.'Ihere are areas in which j

i funher research is accessary, in particular, the record shows conflicting estimates as to the acceptable maximum cladding temperature. Though I have accepted the recommendation for a limit of 2200*F (reflecting a conservative interpretation of the available experimental data),I am inclined to believe that g,,,

there is a high probability that this interpretation is overly conservative. But,the limitations of the present

.g.

record do not junify any course other than that which we have taken today. For the future, however,!

emphasize our instruction to the Director of Reactor Safety Research (supra, page 1088) to give priority the attention to this important area. In my view, the experimental procedures to be used should be developed

,n, with sufficient rigor to be acceptable to the nuclear manufacturing industry and intervenor experts as well c,

as to the Commission. More information will be developed as this research continues. Where appropriate, these data will enable us to refine those assumptions now buih into the rule which may prove to be overly r

1 conservative. The end resuh of this and the other factors noted above should be a fuller a

.. s.,

benefits of nuclear power, while maintaining a policy of meticulous attention to matters of public health e th

..m and safety.

,e iI:r

.e ak APPENDIX

)

On November 30, 1971, the Atomic Energy Commission published in the Federal Register (36F.R. 22774) a notice scheduling a legislative. type public rule snakfng hearing on January 27,19 before a hearing board consisting of Nathaniel H. Goodrich, Esq., Chairman, Dr. Lawrence R. Quarle

,u Dr. John H. Buck, concerning its interim staternent of policy establishing acceptance criteria for emergency p

light wates. cooled nuclear power reactors, published June 29,1971 jl

,p coee cooling systems for (36F.R.12247). Amendments to the interim criteria were published in the Federal Register on

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I 4

December 18,1971 (36 F.R. 24082) in a notice that stated that the amendments would also be cons at the rule making hearing.

Participation in the rule snaking hearing was extensive. The primary participants included the e

Commission Regulatory Staff, four reactor manufacturers, a consolidated group of electric utility

~

companies, and the Consolidated National Intervenors (CNI), a group of about 60 organiza 1129

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? u - _.:.

P individuals In addition, three states, the I.Joyd Harbor Study Group,and severalindividuals participated to I

owr a lesser degree. The hearings lasted a total of 125 days and generated a record of more than 22,000 pages of not t:

transcript and thousands of pages of written direct testimony and exhibits. Oral argument from the seven

(;

principal partidpants was heard by the Commission on October 9,1973.

mant.

In implementation of the National Environmental Policy Act of 1969, (P.L.91190), a Draft not )

Environmental Statement concerning the proposed rule making was forwarded to the Council on a con Environmental Quality on December 6,1972, and circulated for comment to participants in the hearing be ex and interested Federal Agencies on December 7,1972. Notice of public availability of the Statement and an

(:

invitation for comment was also published in the FederalRegiarer at that time. Comments on the Draft Statement were received and a Fma! Environmental Statement was published on May 9,1973.

prods smeer The Commission noted in the interim Policy Statement:

powe Protection against a highly unlikely lossof coolant accident has long been an essential part of the sange chfense-in. depth concept used by the nuclear poner h.dustry and the AEC to assure the safety of oxida nuclear poner plants. In this. concept, the primary assurance of asfety 's accident prevention by p

correctly designing, constructing, and operating the reactor. Extensive and systematic quality assurance to 22 l practicas are requirad and applied at ewry step to achieve this primary assurance of sefety.

prowl Nevertheless, deviations from expected behavior are postulated to occur, and protectin systems are the at instatied to take corrective action as recuired in auch events. Notwithstanding all this, the occurrence of 0

y serious accidents is postulated, lo spite of the fact that they are highly unlikely, and engineered safety Unite features are installed to mitigate the consequences of these unlikely ownts.The loss.of. coolant accident Part

• is such a postuisted improbable accident; the emergency core cooling system is one of the enginnred theF safety features installed to mitigate its consequences.

3, The Commission has adopted new regulations, set forth below, dealing with the effectiwness of ECCS.

I In a 140 page opinion issued on December 28,1973, the Commission discussed the changes from the (a

interim acceptas.se criteria and the technical reason for them. Copies of this opinion are avalhble for U

inspection and copying at the Commission's Public Document Room,1717 H. Suset, N. W., Washington, I

• • • I' D. C.

'8"*

The prindpal changes from the Interim Policy Statement are as follows.ne old criterion number one, 1

specifying that the temperature of the Zircaloy chdding should not exceed 2300*F,is rephced by two I

criteria, lowering the allowed peak Zircaloy temperature to 2200*F and providing a limit on the maximum (a

allowed local oxidation. The other three criteria of the IAC are retained, with some modification of the U

wording. These three criteria limit the hydrogen generation from metal. water reactions, require Ui maintenance of a cochble core geornetry, and provide for long. term cooling of the quenched core.

• • • I' The most irnportant effect of the changes in the required features of the evaluation models is that E*'""

i swelling and bursting of the cbdding must now be taken into consideratiwn when they are calculated to 1 I occur, and that the maximum temperature and oxidation criteria must be applied to the region of clad sweihng or bursting when the maximum temperature and oxidation are calculated to occur there. Another important change is the requirement that,in the steady state operation just before the postulated accident.

I*,

the thermal conductance of the gap between the fuel pellets and the chdding should be calcuhted taking into consideration any increase in gap dimensions resulting from such phenomena as fuel densification,and g

should also consider the effects of the presence of fission gases. When these effects are taken into considtration a higher stored energy may be calculated.Other changes in the evaluation models are mostly in the direction of repheing previous broad conservative assurnptJons with more detailed calculations where new experimentalinformation is available or where better calcuhtlonal methods have been developed.

'I'***

The wording er the definition of a loss.of coolant accident has been modified to conform to its long, accepted usage, limiting it to breaks in pipes. The new regulations also require a more complete documentation of the evaluation models that are used.

The Commission believes that the implementation of the new regulations will ensure an adequate

"I' {

' margin of performance of the ECCS should a design bests LOCA ever occur.This marpn is provided by O

conservative features of the evaluation models and by the criteria themselves.Some of the major points that OPe 1

highest estimated thermal resistar'ce between the UOs N

contribute to the conservative nature of the evaluations and the criteria are as follows:

(1) Stored Heat. The assumption of 102% of maximum power, highest allowed peaking factor, and begin and the cladding provides a calculated stored heat that is possible but unlikely to occur at the time of a hypothetical accident. While not necessarily a margin s

t 1130

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over the extreme constion, it represents at least an assumption that an acadent happens at a time wnich is eages cf not typical.

(2) Blowdown.The cateuhtlon of the heat transfer during blowdown is made in a very conservative e newn rnenner. There is evidence that more of the stored heat would be removed than calcuhted, although there is Draft mot yet an accepted may of calculating the heat transfer more accurately.lt is probable thee this represents scil on a conservatism of several hundred degrees F in atored energy after blowdown, most of whici an renonably varing be expected to carry owr to a reduction in the calculated peak temperature of the Zircaloy cladding.

and an (3) Rate ofHeat Generation. It is assumed that the heat generation rate from the decay of fission

Draft a

products is 20% greater than the proposed ANS standard. This represents an upper limit to the degree of i

uncertainty. The assumption that the fission product levelis that resulting from operation at 102% of rated

{

power for an infinite time represents an improbable situation, with a conservatism that k probably in the t

of tb range of 5 to 15%. De use of the BakerJust equation for calculatirg the heat generation from the steam

]

ety of oxidation of t.ircaloy should aho provide some conservatism, but the factor is un.:ertain.

i

'on by (4) The feak Tempenrture Criterion. De limitation of the peak calculated temperature of the cladding i

to 2200*F and the stipulation that this criterion be applied to the hottest region of the hottest fuel rod urance

safety, provide a substantial depee of conservatism. They ensure that the core would suffer very little damage in

]

ms are the acddent.

a(* of Pursur.st to the Atomic Energy Act of 1954, as amended, and Sections 552 and $53 of Title 5 of the asfetY United States Code, the following amendments to Title 10, Chapter I, Code of Federal Regulations, cident Part 50, are pubbshed as a document subject to codification to be effective on [30 days after publication in wred the FederalRnrister).

1. A new sentence is added to Section 50.34(a)(4) of 10 CFR Part 50 to read as follows:

ECCS.

850.34 Contents of applications: technical information m the (a)"

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(4)"* Analysis and evaluation of ECCS ccoling performance following postulated losser. coolant

,,/

-l accidents shaB be performed in accordance with the requirements of $50.46 for fscilities for which

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construction permits may be issued after December 28,1974.

2. A new sentence is added to Section 50.34(b)(4) 10 CFR Part 50 to read as follows:

,. m' 550.34 Contents of applications; technical information.

(a) *"

. mum

.fthe (b)"*

quire (4)"* Analysis and evaluation of ECCS cooling performance following postulated baser. coolant accidents sha!! be performed in accordance with the requirements of $50.46 for facilities for which a gi license to operate may be issued after December 28,1974.

ed 13

3. A new $50.46 is added to 10 CFR Part 50 to read as follows:

stad

$50.46 Acceptance Criteria for Emergency Core Ceoling Systems for IJght Water Nuclear Power

,,g, Reactors.

Jent.

(a)(1)Except as provided in subparapsphs (2) and (3) of this pararraph, each boiling and pressurized

,gg light. water nuclear power reactor fueled with uranium oxide pellets Mthin cylindrical r.ircaloy cladding

.and shall be provided with an emergency core cooling system (ECCS) which shcu be designed stah that ita m

calculated cooling performance following postulated loss.of. coolant accidenti conforms to the criteria set t

i sstly forth in paragraph (b). ECCS cooling performance shall be calcuhted in accordance with an acceptable kn evaluation model, and shall be cateulated for a number of postulated lossef. coolant accidents of different sizes, locations,and other properties sufficient to provide assurance that the entire spectrum of postulated

., g, losser. coolant accidents is covired. Appendix K. ECCS Evaluation Modeh, sets forth certain required and

i 4,,,

seceptable features of evaluation modek. Conformance with the criterin act forth in paragraph (b), with ECCS cooling perforn.ance caletlated in accordance with an acceptable evaluation model, may require that m,

restrictions be imposed on reactor operation.

j i by (2)With respect to reactors for which operating licenses have previously been issued and for which-operating licenses rnay issue on or before December 28,1974:

g i

(1)The time within which actions required or permitted under this subparagraph (2) must occur shall e

begin in run on [30 days after publication of the rule in the FederalRigister).

(ii)Within six months following the date specified io subparagraph (i) of this subparagraph (2), an rgm' -

evaluation in accordance with subparagraph (1) of this paragraph (a) shall be submitted to the Director of t'

1131 1

_--_7

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kgulation. "Ihe evaluauon shall be accompanied by such proposed changes in technical specifications or I

liceme amendments as may be necessary to bring reactor operation in conformity with subparagraph (1) of this paragraph.

(111) Any licensee may request an extension of the six. month period referred toin subparapaph(ii)of this subparapaph (2) for good cause. Any such request shall be submitted not lers than 45 days prior to expiration of the six-month period, and shaU be accompanied by affidavits showing precisely why the

,i ;[

evaluation is not complete and the minimum time believed necessary to complete it. The Director of

(

Regulation shaU cause notice of ruch a request to be published acomptly in the FederalRegister; such notice shsil provide for the submission of comrnents by interesteo persons within a time period to be I

k established by the Director of Regulation. If, upon reviewing, the foregoing submissions, the Director of 1

Regulation concludes that good cause has been shown for an extension, he may extend the six. month Q.;

period for the shortest additional time which in this judgment will be necessary to enable the licensee to 1

)

furnish the submissions required by subparapaph (ii) of this subparagraph (2). Requesta for extensions of t

y the six-month period, submitted under this subparagraph, shall be ruled upon by the Director of Regulation prior to expiration of that period.

.m (iv)Upon submission of the evaluation required by subparapaph (ii)of this subparapsph 2)(or under subparapaph (iii),if the six-month period is extended) the facihty shall continue or commence operation g

only within the limits of both the proposed technical specifications or license amendments submitted in accordance with this subparapaph (2) and all technical specification or license conditions previously imposed by the Commission, including the requirements of the Interim Policy Statement (June 29,1971, 36 F.R. I2248),as amended (December 18,1971,36 F.R. 24082).

a, (v) Further restrictions on reactor operation will be imposed by the Director of Regulation if he finds that the evaluations submitted under subparagraphs (ii)and (iii)of this subparagraph (2)are not consistent with subparapaph (1) of this parapaph (a) and as a result such restrictions are required to protect the public health and safety.

(vi) Exemptions from the operating requirements of subparagraph (iv)of this subparagraph (2) may be 8

granted by the Commission for good cause. Requests for such exemption shall be submitted not leu than sy 45 days prior to the date upon which the plant would otherwise be required to operate in accordance with

'?

the procedures of said subparapaph (iv). Any such request shall be filed with the Secretary of the Commission, who shall cause notice ofits receipt to be published promptly in the FederalRegister;such j

notice shall provide for the submisalon of comments by interested persons within 14 days following Federal j

Register publication. The Director of Regulation shall submit his views as to any requested exemption gg within five days fo!!owing expiration of the comment period.

y.

(vii) Any request for an exemption submitted under subparagraph (vi) of this subparapaph (2) rnust g'

show, with appropriate affidavits and technical subrnissions, that it would be in the public interest to allc6w the licemee a specined additional period of time within which to alter the operation of the facilityin the manner required by subparapaph (iv) of this subparapaph (2). 'the request shall also include a discussion of the ahernatives avalhble for establishing...npliance with the rule.

(3) Construction permits may be issued after December 28,1973 but before December 28,1974 j

subject to any applicable conditions or restrictions imposed pursuant to other regulations in this choter fw and the Interim Acceptance Criteria for Emergency Core Cooling Systerns published on June 29,1911 (36F.R.12248) as amended (December 18,1971,36F.R. 24082): horlded, homer, that no operating l

['q.

license shall be issued for facilities constructed in accordance with construction permits issued pursuant to this subparapaph, unless the Commission determines,smong other things, that the proposed facility meets 1

the tequirements of subparapaph(!)of this parapaph.

q (b)(1) Peak Clodding Temperatsac. The calculated maximum fuc* element cladding temperature shall not exceed 2200*F.

(2) Maximum Cladfing Oxidation. The calcuhted otal oxidation of the cladoing shall nowhere exceed 0.17 times the total cladding thickness before oxidstion. As used in this subparapaph total oxidation means the total thickness of cladding metal that would be locally converted to oxide if all the oxygen

. ~'

1 absorbed by and reacted with the chdding locally were converted to stoichiometric rirconium dioxide.lf a

cladding rupture is calculated to occur, the b. side surfaces of the cladding shall be included'in the oxidation,beginning at the calculated time of rupture. Cladding thickness before oxidation means the radial distance from inside to outMe the cladding, after any calculated rupture or twelling has occurred but before sigilficant oxidaGon. Mere the calculated conditions of transient pressure and.emperature lead to I

i 1132 t

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I a prediction of cbdding swelling, with or without cladding rupture, the unoxidized cladding thickness shall 3,

, g,gg be defined as the chdding cross sectional area,taken at a horizonta! plane at the elevation of t e rupture if h

it occurs, or at the elevation of the highest cladding temperatu;e if no rupture is calculated to occur,

, ggygg divided by the average circumference at that elevation. For ruptured chdding the circumference doe.not include the rupture opening.

,gg hy the (3)Marimum #phogen Generation. The calcuhted total amount of hydrogen generated from the

.: tor cf chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount r;s ch that would be generated if s!! of the metalin the cladding cylinders surroundmg the fuel,exclu&ng the I o be chdding surounding the plenum volume, were to react.

etm cf (4)Coolable Geometry. Calcuhted changes in core geometry shall be such that the core remains

-month amenable to cooling.

g,g (5)Long-Term Cooling. After any calculated sucassful initial operation of the ECCS, the calculated gg core temperature shall be rnaintained at an acceptably low value and decay heat shall be removed for the e;lation extended period of tirne required by the long lived radioactivity remaining in the core.

(c) As used in this section:

r under (1)less of ccmlant accidents (LOCA's) are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system,from breaks in

,,g,,

~ tied in P Pes in the reactor coolant pressure boundary up to and including a break equivalent in size to the i

double ended rupture of the largest pipe in the reactor coolant system.

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• 9.j *

(2) An evaluation model is the calcuhtional framework for evaluating the behavior of the reactor synem during a postulated loss of-coolant accident (LOCA). It includes one or more computer programs (sds and all other inforrmtion necessary for application of the calculational framework to a specific LOCA, such as mathematical models used, assumptions included in the programs, procedure for treating the program l'nput anf output information, specification of those portions of analysis not included in computer utthe programs, values of parameters, and all other information nemssary to specify the calculational procedure.

"(

(d)The requirements of this section are in ad& tion to any other requirements applicable to ECCS set t

forth in this Part. The criteria set forth in paragraph (b), with cooling performance calculated in accordance with an acceptable evaluation model, are in implementation of the general requirements with respect to a with ECCS coohng performance design set forth in this Part, including in particular Criterion 35 of Appendix A.

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4. A new Appendix K is added to 10 CFR Part 50 to read as follows: Appendix K-ECCS Evaluation c such Models.

I' * ##

1. Required and Acceptable Features of Evaluation Models.

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II. Required Documentation.

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1. REQUIRED AND ACCEITABl.E FEATURES OF THE EVALUATION MODELS j

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• 8dr A. SOURCES OF HEAT DURING THE LOCA i

6 For the heat sources listed in Paragraphs I to 4 belowit shall be assumed that the reactor has been gan

,wpw, operating contin.nously at a power level at least 1.02 times the licensed power level (to allow for such

, pg uncertainties as instrumentation error), with the maximum peaking factor allowed by the technical

,, g m, specifications. A range of power distribution shapes and peaking factors representing power distributions J

that may occur over the core lifetime sha!! be studied and the one selected should be that which results in una so meen the most severe calculated consequences,for the spectrum of postulated breaks and single failures analyzed.

1.TheInfrialStored Energy in the Fuel The steady-state teraperature distribution and stored energy in

, gy the fuel before tha Vpothetical accident shat be calculated for the burn up that yields the highest calcuhted cisoding temperature (or, optionally, the highest calculated stored energy). To accomplish this,

,.%,;g she thermal conductivity of the UO shall be evaluated as a function of burn up and temperature, taking i

,s,,,,

into consideration differences in initial density, and the therms! conductance of the gap between the UOa

.ge3 and the chdding shall be evaluated as a function of the burn.up,taking into consideration fuel densification uJe 3:

and expansion, the composition and pressure of the gases within the fuel rod, the initial cold gap dimension irg with its tolerances,and chdding creep.

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2.Fis:fon fear. Fission heat shall be calcuhted using reactivity and reactor kinetics. Shutdown i

A.

reactivities resuhing from temperatures and voids shall be given their minimum plausible values, including eu to l

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i allowanse for uncertainties, for the range of power distribution shapes and peaking factors in I

studied abow. Rod trip and insertion rnay be assumed if they are calculated to occur.

3. Decay of Actinides. The heat from the ndioactin decay of actin! des, including neptun plutonium generated during operation, as well as hotopes of uranium, shall be calculated in with fuel cycle calcuhdons and known radioactin properties. The actinide decay heat chosen sppropriate for the time in the fuel cycle that yields the highest calculated fuel temperature IDCA.

4.Faraton hoduct Decay. The heat generation rates from radioactive decay of fission produ

's assumed to be equal to 1.2 times the values for infmite operating time in the ANS Standard American Nuclear Society Standard " Decay Energy Release Rates Following Shutdown o

'l

'g Fueled Thermal Reactors", Approwd by Subcommittee ANS 5, ANS Standards Committe 1971). The fraction of the locaUy generated gamma energy that is deposited in t!.e fu3

,t,g 1:.

cladding) may be different from 1.0; the value used shall be justified by a suitable calculation.

's 5.MetalWater Reaction Rare. The rate of energy release, hydrogen generation,and chdd from the snetal/ rater reaction shall be calculated using the BakerJust equation (Baker, L

" Studies of Metal Water Reactions at High Temperatures, !!!. Experirnental and Theoretica Zirconium Water Reaction," ANI.r6548, page 7 May 1962).The reaction shall be assurned n 8

limited. For rods whose cladding is calcuhted to rupture during the LOCA,the inside of g

sho be assumed to react after the rupture.The calcuhtlon of the reaction rate on the shall aho follow the BakerJust equation, starting at the time when the chdding is calcu I

and extending around the chdding inner circumference and axially no less than 1.S inches the location of the rupture, with the reaction assumd not to be steam limited.

a

6. Reactor Internal Heat Teant(er. Heat transfer from Iiping, vessel nik, and non-fuel interna hardware shall be taken into account.

7. Pressurized ! Vater Reactor Prinwry.to3econdary Heat Dansfer. Heat transferred be and secondary systems through heat exchangers (steam generators) shall be taken into a I

I' applicable to Boiling Water Reactors.)

y B. SWELLING AND RUPTURE OF THE CLADDING AND FUEL ROD k

q THERMAL PARAMETERS j

Each evaluation model shall include a provision for predicting cladding swelling and ruptu{

f consideration of the axial temperature distribution of the cladding and from the difference in p E

}

between the inside and outside of the cladding, both as functions of time.To be acceptabl inl f

rupture calculations shall be based on applicable data in such a wy that the degree of sw!!ing a incidence of rupture are not underestimated. The degree of sw!!ing and rupture shall be taken int sa

}f in calculations of gap conductance, chdding oxidation and embrittlement,and hydrogen generation sn.

The alcuhtions of fuel and chdding temperatures as a function of time shall me values fri g

conductance and other thermal parameters as functions of temperature and other applicable tim en.

f dependent variables. The gap conductance shall be varied in accordance with changes in gap ex!

.h any other applicable variabisa.

y 4.i N

C. BLOWDOWN PHENOMENA I

act I. Break Charseteristia and Flow Po j

a.!n analyses of hypothetical lossof. coolant accidents, a spectrum of possible pipe breaks shall be Per; 4

considered. This spectrum sha!! include instantaneous doubic. ended breaks ranging in cror2. section up to and including that of the hrgest pipe in the primary coolant system. The analysis shall aho include the.

the effect* oflongitudinal spilts in the hrgest pipes, with the split area equal to the cross.sectionalarea o l

the p'pe.

I

b. Discharge Model For a!! times after the dischargns fluid has been calculated to be two phase composition, the discharge rate shall be calculated by use of the Moody model (F.J. Moody, " Maximum C Flow Rate of a Single Component, Two-Phase Mixture," Journal of Heat Tranrfer, Transactions of

?g American Society ofMechanicalErgineers, 87, No.1, Febrcary 1965). The calculation shall be conducted L

\\

1134

[

=-

I udtde with at least three values of a discharge coefficient cpplied to the postulated break area, these values apanning the range from 0.6 to 1.0. If the results indicate that the maximum chd temperature for the alum and hyp thetical accident is to be found at an even lower value of the discharge coefficient, the range of I

cordance discharge coefncients shan be extended until the. maximum chd temperature calculated by this variation 11 be that has been schieved.

rring the c.Ead ofBlomt:fown. (Applies Only to Pressurized Water Reactors.) For postulated cold leg breaks,all emergency cooling water injected into the inlet lines or the reactor wssel during the bypass period shallin s shall be the calcuhtions be subtracted from the reactor wssel calculated inventory. This may be executed in the i

Uranium-

'l calculation during the bypass period, or as an alternative the amount of emergency core cooling water Pr: posed calcuhted to be injected during the bypass pedod may be subtracted later in the calculation from the water Oct&r remalrdog in the inlet lines, downcomer, and reactor wesel lower plenum after the bypass period. This ading ee bypassing shall end in t!'e calcuhtlon at a time designated as the "end of bypass," after which the expuhlon or entrainment rnechanisms responsible for the bypassing are calculated not to be effeed.e. The oxidation end of bypass definition used in the calcuhtion shall be justified by a suitable combination of analysis and att, LC.

eXPuirnental data. Acceptable methods for defining "end of bypass" include, but are nn limited to, the a dee foll wing: (1) Prediction of the blowdown calculation of downward flow in the downcomer for the be steam remainder of the blowdown period;(2) Pre &ction of a threshold for droplet entrainment in the upward

'g g,33

)

wlocity, using local fluid conditions and a conservative critical Weber number.

chdding d.Noding Near the Break and the ECCIn/cetion Polars. The noding in the vicinity of and including rupture the broken or split sections of pipe and the points of ECCS injection shall be chosen to permit a reliable

,y g,,'

analysis of the thermodynamic history in these regions during blowdown.

2. Frictional Pressure Drops. The friedonal losses in pipes and other components including the reactor internal core shall be calculated using modeh that include reabstic variation of friction factor with Reynolds number, and realistic two. phase friction multipliers that have been adequately wrified by comparison with experimental data, or modeh that prove at least equally conservative with respect to maximum clad

t' temperatute calculated during the hyrothetical accident.The mod fled Baroczy conelation (Baroczy,C.J.,

"A Systernstic Correktion for TwoPhase Pressure Drop," Chem. Erigiv. Prog. Symp. Series. No.64, Vol. 62,1965) or a combination of the Thom correlation (Thom, J.R.S.," Prediction of Pressure Drop During Forced Circuhtlon Bolling of Water," Int. J. of Hear A Mau Dans/cr, 7, 709 724,1964) for pressures equal to or greater than 250 psia and the Martinelli Nelson correlation (Martinelli, R.C., Nehon, D. B., " Prediction of Pressure Drop During Forced Circuhtion Bolling of Water," Dansactions ofASME, we from 695 702,1948) for pressures lower than 250 psia is acceptable as a basis for calculating realistic two phase pressure friction multipliers.

!!ing and

3. Momentum Equation. The foilowing effects shall be taken into account in the conservation of ling and rnomentum equation: (1) temporal change of momentum, (2) momentum convection, (3) area change account momentum flux, (4) momentum change due to compressibility, (5) pressure loss resulting from wall friction, (6) pressure loss resulting fron.

..a change, ano (7) gravitstional acceleration. Any omission of for gap one or more of these terms under stated circumstances shall be justified by comparative analyses or by le time.

er.perimental data.

Mms and

4. Critical Heat Flux

s. Correlations developed from appropriate steady. state and transient state experimental dats are a

acceptable for use in piedicting the critical heat flux (CHF) during LOCA transients. The computer prograns in which these correhtlons are used shaB corr'ain suitable checks to assure that the physical parameters are within the range of parameters specified for use of the correlations by their respective shall be authors.

nelarea b.3 cady stase CHF conehtions acceptable for use in IDCA transients include,but are not limited to, include the foBowin5:

g,,, g (1) WJ. L. S. Tong," Prediction of Departure from Nucleate Boiling for en Axhlly Non uniform Heat Flux Distribution,"JournalofNuclear Energy, Vol. 21,241 248,1967.

(2)Bd W.2. J. S. G<!!ctstedt, R. A. Lee, W.J. Oberjohn, R. H. Wilson, L J. Stanek," Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water " 7ko Phase flow and# car Tranzferin Rod 3

s of the Bundles, ASME,NewYcrk !969.

nducted 1135

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= n m.....__.

1 m-. P -wvv Tm e n yu em-

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(3)#ench-Levy. J. M. Heaher, J. E. Hench, E.Janssen, S. Levy," Design Basis for Critical Heat Flux Condition in Bolling Water Reactors," APED SI86, GE Company Private report, July 1966.

two-1 (4)Macherh. R. V. Macbeth,"An Apprsisal of Forced Convection Burr.out Dats," Proceedings of the head Institure ofMechanicalErnrinters.19651966.

sucti.

(5)Samert. P. G. Barnett,"A Correlation of Burnout Data for Uniformly Heated Annull and its Uses PefC' for Predicting Burnout in Uniforrnly Heated Rod Bundles " AEEW.R 463,1966.

7.Co (6)#ughes. E.D. Hughes,"A Conehtlon of Rod Bundle Critical Heat Flux for Water in the Pressure a

Range 150 to 725 psia,: IN 1412, Idaho Nuclear Corporation, July 1970.

time.

fue! 4

c. Correlations of appropriate transient CHF data may be accepted for use in IDCA transient analyses if betwi comparisons between the data and the conelations are provided to demonstrate that the conelations or ru l

..y, predict values of CHF which allow for uncertainty in the experimental data throughout the range of than

{*;

parameters for which the correlations are to be used. Where appropriate, the comparisons shall me b

"i statistical uncertainty analysis of the data to demonstrate the conservatism of the transient correhtfon, heatt

d. Transient CHF correlations acceptable for use in LOCA transients include, but are ac! lirrdted to, the cr.hu

[

following:

i (1)GE Deersient Off. B. C. Slifer, J. E. Hench,"lessof Coolant Aeddent and Emergency Core D.PC Cooling Modeh for General Electric Bolling Water Reactors," NEDO-10329, General Electsic Company, g

Equation C 32, April 1971.

,grec

e. After CHF h first predicted at an axial fuel rod location during blowdown,the m!culation shall not

. ECCf I.

use nucleate boiling heat transfer correlations at that location subsequently during the blowdown enn if ECCf the calcuhted local fluid and surface conditions would apparently justify the reestablishment of nucleate 2

boiling. Heat transfer assumptions characteristic of return lo nucleate boiling (rewetting) shall be permitted 1efim when justified by the calcuhted local fluid and surface conditions during the reflood portion of a LCCA.

salcu 3

the

5. Post.CHF Heat Transfer Correlations a.Conelations of heat transfe, from the fuel chdding to the sunounding fluid in the post CHF regimes syste-f of transition and film boiling shall be compared to spplicable steady. state and transient strie data using alcu statistical correlation and uncertainty analyses. Such comparison shall demonstrate that the conelations theti g

predict values of heat transfer coefficient equal to or less than the mean value of the applicable ahnli experimental heat transfer data throughout the range of parameters for which the conelations are to be exper I

• h used. The comparisons shall quantify the relation of the correlations to the statistical uncertainty of the

• Po applicable data.

Ty, b.The Groeneveld flow film boiling conehtion (Equation 5.7 of D.C. Groeneveld,"An investigation ggg of Heat Transfer in the Liquid Deficient Regime," AECle3281, revised December 1969), the geg

?

Dougall Rohsenow flow film boiling conehtion (R. S. Dougall and W. M. Rohsenow," Film Bolhng on the y

L Inside of Vertical Tubes with Upward Flow of e Fluid at Iow Qualities," M1T Report Number 9079 26, accur r

Cambridge, Massachusetts, September 1963), and the Westinghouse correhtion of steady. state transition 4)

,1 boiling (" Proprietary Redirect /Rebutta! Testimony of Westinghouse Electric Corporation," U.S.A.E.C.

therst

[

Docket RM.501,page 251, October 26,1972) are acceptable for use in the post CHF boiling regimes. In accet l

")

addition the transition boiling correhtlon of McDonough, Milich, and King (J. B. McDonough, W.Milich, unbrc l "i

E.C. King," Partial Film Boiling with Water at 2000 psig in a Round VerticalTube," MSA Research Corp.,

water l i[

Technical Report 62 (NP 6976), (i958) is suitable for use between nucleate and film boiling. Use of all inters; io these correlations shall be repricted as follows:

akerr -

E Y

0) The Groeneveld conelation sh ll not be used in the region near its low pressure singularity,

[

(2)the first term (nucleate) of the Westinghouse correhtlon and the entire McDonough, Milich, and t

King conelation sha!! not be used during the blowdown afler the temperature difference between the clad

""g ;I and the saturated fluid first exceeds 300*F.

Transi i

(3) transition boilig heat transfer shall vst be reapplied for the remainder of the LOCA blowdown, I"

even if the clad superiaat retums below 300*F,except for the reflood portion of the LOCA whenjustified

{ll

~

by the calcuhted local fluid and surface conditions, pg

6. Pump Modeling. The chnacteristics of rotating primary system pumps (exial flow, turbine; or y,, f centrifugal) shall be derived from a dynamic model that includes momentum transfer between the fluid and

,,, g 4 the rotating member, with variable pump speed as a function of time.The pump model resistance used for

,an,i.

analysis should be justified. The pump model for the two-phase region shall be verified by applicable DI calcu!:

j 1136

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o

[*

J L' two phase pump performance data. For BWR's after scturation is calculated at the pump suction.the pump c/the head may be assurned to very linearly with quality, going to zero for one percent quality at the pump suction, so long as the analysis shows that core flow stops before the quality at pump suction reaches one is Uses perc nt.

7. Core Flow Distribution During Blowdown. (Applies only to pressurized water reactors.)

ressure a.The flow rate through the hot region of the core during blowdown shall be calculated as a function of ti ? For the purpose of these calculadons the hot region chosen shall not be i;teater than the sin of on'.

fuel assembly. Calculations of average flow and now in the hot region shas take into account cross flow i'

hsesif be ween regions and any flow blockage calculated to occur during blowdown as a result of cladding melling ladons or rupture. The calculated flow shall be smoothed to eliminate any caledated rapid oscillations (period less 888 Cf thari 0.1 secoads).

an use

b. A method shall be specified for octermining the enthalpy to be used as input data to the hot channel hestup nalysis from quandtles calculated in the blowdown analysis, consistent with the flow distribution 88.the calculations.

e Cere D. POST. BLOWDOWN PHENOMENA; HEAT REMOVAL BY THE ECCS

3Pany,

l. Single failure CWrerion. An analysis of possible failure modes of ECCS equipment and of their effects on ECCS performance must be made. In carrying out the accident evaluation the combination of all get ECCS subsysterns assumed to be operative shan be those available after the most damaging single failure of rver. !f ECCS equipment has taken place.

ichste 2.Contsinment Pressure. The containment pressure used for evaluating cooling effectivenen during tuttId reflood and spray cooling shall not exceed a pressure calculated conservatively for this purpose. The A,

calculation shsU include the effects of operation of allinstalled pressure. reducing systems and processes.

3. Calculation of Reflood Rate for Pressurized Water Reactors. The reillling of the reactot vessel and the time and rate of renooding of the core shall be calculated by an acceptable model that takes into si[g consideration the thermal and hydraulic characteristics of the core and of the reactor system.The primary system coolant pumps shall be assumed to have locked impellers if this assumption leads to the maximum usu.a calculated cladding temperature;otherwise the pump rotor shall be assurned to be running free. The ratio of stions the total fluid flow at the core exit plane to the totalliquid Dow at the core inlet plane (carryover fraction)

.ssble shaU be used to deterrnine the core exit flow and shall be determined in accordance with spolicable to be experimental data (for example, "PWR FLECHT (Full kngth Emergency Cooling Heat Trmfeff Final

,g gg,,

Report," Westinghouse Report WCAP.7665, April 1971; "PWR Full bngth Emergency Cool,y Heat Transfer (FLECHT) Group i Test Report," Westinghouse Report WCAP.7435, January 1970; "PWR FLECHT (Full bngth Emergency Cooling Heat Transfer) Group !! Test R port," Westinghouse Report g

WCAP.7544, September 1970; "PWR FLECHT Final Report Supplermat," Westingl.ouse Report the WCAP-7931. October 1972).

"8k The effects on reflooding rate of the compressed gas in the accumulator which is discharged following A26 accumulator water discharge shall also be taken into account.

il!0*

4. Steam interaction with Emergency Core Coolim Water in Pressurized Water Reactors. The l

.E.C.

thermal. hydraulic interaction between aream and all ernergency core cooling water shall be taken into

s. In account in calculating the core renooding rate. During refill and reflood, the calculated steam now in

lich, unbroken reactor coolant pipes shaU be taken to be acro during the time that accumulators are discharging orp.,

water into those pipes unleu experimental evidence is available regarding the realistic thermal-hydraulic if all interaction between the stesm and the liquid. in this case, the experimental data may be used to support an ahemate assumption.

$.Refilland RefloodHeat Trarnfer for Pressurized Water Reactors. For reflood rates of one inch pet and second or higher, reflood heat transfer coefficients shall be based on applicable experimental data for clad unblocked cores including FLECHT results ("PWR FECHT (Full bngth Emergency Coo:ing Heat Transfer) Final Report," Westinghouse Report WCAP.7665, April 1971). The use of a correlation derived from FLECHT data shall be demonstrated to be conservative for the transient to which it is applied;

,wn, ified presently available FLECHT heat transfer correlations ("PWR Full Length Emergency Cooling Heat Transfer (Fl.ICHT) Group i Test Report," Westinghouse Report WCAP.7544, September 1970; "PWR FLECHT Final Report Supplement," Westinghouse Report WCAP.7931, October 1972)are not acceptable.

New correlations or modifications to the FLECHT heat transfer correlations are acceptable only after they are demonstrated to be conservative, by comparisort with FLECHT data, for a sange of parameters

, k' consistent with the transient to which they are applied.

During refill and during reflood when reflood u. tis sw bss than one inch per second, heat transfer calculations shall be based on the assumption that co%ngt sly by steam,and shall take into account any

\\

1137

. =

A e-flow blockage calculated to occur as a result of claddmg swelling or rupture at such blockage might affect both local stem ' law and heat transfer.

6. Convective Heat Transfer Coefficienn for Boiling Water Reactor FuelRods Under Spray Cooling.

Following the blowdown period, convective heat transfer shau be calculated using coefficients bued on appropriate experimental data. For reactors with jet pumps and having fuel rods in a 7 x 7 fuelassembly array,the following convective coefficients are acceptable:

a.Dudng the period following lower plenum flashing but prior to the core spray reaching rated flow, a conwctiw heat transfer coemcient of aero shall be applied to a!! fuel rods.

b. During the period after core spray reaches rated flow but prior to reflooding, conwetive heat transfer coefficients of 3.0,3.5,1.5, and 1.5. Btu-hr*8.ff.*F8 shall be applied to the fuel rods in the outer corness, 8

outer row, next to outer row, and to those remaining in the interior, respectively, of the assembly.

c. After the two-phase reflooding fluid reaches the level under consideration, a convective heat transfer coemcient of 25 Stu-hr-8 ff8 *F8 shall be applied to all fuel rods.

en thi

7. The Boiling Water Reactor ChannelBox Under Spray Cooling. Following the blowdown period, hest CON:

1 transfer from, and wetting of, the channel box shall be based on appropriate experimental data. For seactors withjet pumps and fuel rods in a 7 x 7 fuelassembly array, the following heat transfer coemeients I

and wetting tirne correlation are acceptable, e.During the period after lower plenum flashing, but prior to core spray reaching rated flow, a convective coemcient of zero shall be applied to the fuel assembly channel box.

b.During the period after core spray reaches rated flow, but prior to wetting of the channel, a convective heat transfer coemeient of 5 Btu-hf -ff8.*F8 shall be applied to both sides of the charu,4 box.

8

c. Wetting of the channel box shall be assumed to occur 60 seconds after the time determined using the correhtlon based on the Yarrraouchi analysis (%ss of Coolant AccMent and Ernergency Core Cooling Models for General Electric Boiling Water Reactors," General Electric Company Report NEDO 10329, April 19,71).

H. REQUIRED DOCUMENTATION w

1.a. A description of each evaluation model shall be furnished. The desedption shall be sumciently jurisdic complete to permit technical review of the analytical approach including the equations used, their la approxirnations in difference form, the assumptions made, and the values of all parameters or the procedure amran.,

for their selection,as for example,in accordance with a specified physical hw or empirical conelation.

f aH con b.The description shall be sufficiently detailed and specific to require significant changes in the wm ao evaluation model to be specified in amendments of the description. For this purpose,a significant change is alisedy a change that would result in a calculated fuel cladding temperature different by more than 20*F from the would temperature calculated (as a function of time) for a postulated LOCA using the last previously accepted

funher, model.

our mar

c. A complete Esting of each computer program,in the same form as used in the evaluation model,shall equaHyi be fumished to the Atorr.ic Energy Commlasion.

The:

2.For each computer program, solution convergence shall be demonstrated by studies of system Ci'cunst modeling or noding and calculational time steps, their fin l

3. Appropdate sensitivity studies shad be performed for each evaluation model, to evaluats the effect endissa. I on the calculatsd suults of variations in poding, phenomena assumed in the calculation to predominate, This i including pump operation or locking, and values of parameters over their applicable ranges. For items to organissi which results are shown to be sensitive, the choices made shad be justified, remedy ;

4.To the extent practicable, predictions of the evaluation model, or portions thereof, shad be order to I compared with appUcable experimentalinformation.

to compi ;

5. General Standards for Acceptability-Elements of evaluation models reviewed willinclude technical and 2),

adequacy of the calculational methods, including compliance with required fes.tures of Section I of this Commias Appendix X and provision of a level of safety and margin of conservatism comparable to other acceptable f88Pectl' evaluation models, taking into account signific:mt differences in the reactors to which they apply.

ED NOTE: There wRI be forthcoming en official pubtlearlon containing the materialin the ECCg proceedinge.

M7Z - U ?)

1

+

J, 1138 I

Generic Letter 88-16 1

/ au:\\

UNITED STATES I

NUCLEAR REGULATORY COMMISSION o

g wAssworow.o. c.sossa

\\,

00T 0 4 W B TO ALL POWER REACTOR LICENSEES AND APPLICANTS

SUBJECT:

REMOVAL OF CYCL -SPECIFIC PARAMETER LIMITS FROM TECHNICAL SPECIFICATIONS (GENERIC LETTER 88-16)

License amendments are generally required each fuel cycle to update the values of cycle-specific parameter limits in Technical Specifications (TS). The processing of changes to TS that are developed using an NRC-approved method-ology is an unnecessary burden on licensee and NRC resources.

A lead plant proposal for an alternative that eliminates the need for a license amendment to update the cycle-specific parameter limits each fuel cycle was submitted for the Oconee plant with the endorsement of the Babcock and Wilcox Owners Group. On the basis of the NRC review and approval of that proposal, the.en-closed guidance for the preparation of a license amendment request for this alternative was developed by the NRC staff.

Generally, the methodology for determining cycle-specific parameter limits is documented in an NRC-approved Topical Report or in a plant-specific submittal.

As a consequence, the NRC review of proposed changes to TS for these limits is primarily limited to confirmation that the updated limits are calculated using an NRC-approved methodology and consistent with all applicable limits of the safety analysis. These changes also allow the NRC staff to trend the values of these limits relative to past experience. This alternative allows continued trending of these limits without the necessity of prior NRC review and approval.

I Licensees and applicants are encouraged to propose changes to TS that are consistent with the guidance provided in the enclosure.

Conforming amendments will be expeditiously reviewed by the NRC Project Manager for the facility.

Proposed amendments that deviate from this guidance will require a longer, more detailed review.

Please contact the. Project Manager if you have questions on this matter.

Sincerely, 991^^59058~

'g Denn s M. Crutchfiel Acting Associate Di ector for Projects Office of Nuclear Reactor Regulation

Enclosure:

i As stated j

a

x Generic Letter 88-16 Enclosure o

GUIDANCE FOR TECHNICAL SPECIFICATION CHANGES FOR CYCLE-SPECIFIC PARAMETER LIMITS INTRODUCTION A number of Technical Specifications (TS) address limits associated with reactor physics parameters that generally change with each reload core, requir-ing the processing of changes to TS to update these limits each fuel cycle.

If these limits are developed using an NRC-approved methodology, the license amendment process is an unnecessary burden on the licensee and the NRC.

An alternative to including the values of these cycle-specific parameters in in-dividual specifications is provided and is responsive to industry and NRC efforts on improvements in TS.

This enclosure provides guidance for the preparation of a license amendmen't request to modify TS that have cycle-specific parameter limits.

An acceptable alternative to specifying the values of cycle-specific parameter limits in TS was developed on the basis of the review and approval of a lead 'lant proposal p

for this change to the TS for the Oconee units.

The implementation of this alternative will result in a resource savings for the licensees and the NRC by eliminating the majority of license amendment requests on changes in values of cycle-specific parameters in TS.

DISCUSSION This alternative consists of three separate actions to modify the plant's TS:

(1) the addition of the definition of a named formal report that includes the values of cycle-specific parameter limits that have been established using an NRC-approved methodology and consistent with all applicable limits of the safe-ty analysis, (2) the addition of an administrative reporting requirement to sub-mit the formal report on cycle-specific parameter limits to the Commission fot-information, and (3) the modification of individual TS to note that cycle-specific parameters shall be maintained within the limits provided in the defined formal report.

In the evaluation of this alternative, the NRC staff concluded that it is essential to safety that the plant is operated within the bounds of cycle-specific parameter limits and that a requirement to maintain the plant within the appropriate bounds must be retained in the TS.

However, the specific values of these limits may be modified by licensees, without affecting nuclear safety, provided that these changes are determined using an NRC-approved method-ology and consistent with all applicable limits of the plant safety analysis that are addressed in the final Safety Analysis Report (FSAR).

Additionally, it was concluded that a formal report should be submitted to NRC with the values of these limits.- This will allow continued trending of this information, even though prior NRC approval of the changes to these limits would not be required.

The current method of controlling reactor physics parameters to assure conform-ance to 10 CFR 50.36 is to specify the specific value(s) determined to be with-in specified acceptance criteria (usually the limits of the safety analyses) using an approved calculation methodology.

The alternative contained in this guidance controls the values of cycle-specific parameters and assures conform-ance to 10 CFR 50.36, which calls for specifying the lowest functional

Generic Letter 88-16 Enclosure performance levels acceptable for continued safe operation, by specifying the calculation methodology and acceptance criteria.

This permits operation at any specific value determined by the licensee, using the specified methodology, to be tsithin the acceptance criteria.

The Core Operating Limits Report will docu-ment the specific values of parameter limits resulting from licensee's calcula-tions including any mid-cycle revisions to such parameter values.

.The following items show the changes to the TS for this alternative.

A defined formal report, " Core Operating Limits Report" (the name used as an example for the title for this report), shall be added to the Definitions section of the TS, as follows.

[ CORE] OPERATING LIMITS REPORT 1.XX The [ CORE] OPERATING LIMITS REPORT is the unit-specific document that provides [ core] operating limits for the current operating reload cycle.

These cycle specific [ core] operating limits shall be determined for each reload cycle in accordance with Specification 6.9.X.

Plant

. operation within these operating limits is addressed in individual specifications.

A new administrative reporting requirement shall be added to existing reporting requirements, as follows.

CORE] OPERATING LIMITS REPORT

6.9.X] [ Core] operating limits shall be established and documented in the

[ CORE] OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

(If desired, the individual specifications that address [ core) operating limits may be referenced.) The analytical methods used to determine the [ core) operating limits shall be those previously re-viewed and approved by NRC in [ identify the Topical Report (s) by number, title, and date, or identify the staff's safety evaluation report for a plant-specific methodology by NRC letter and date).

The [ core) operating limits shall be determined so that all applicable limits (e.g., fuel therm-al-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The [ CORE] OPERATING LIMITS REPORT, in-cluding any mid-cycle revisions or supplementt thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

Individual specifications shall be revised to state that the values of cycle-specific parameters shall be maintained within the limits identified in the defined formal report.

Typical modifications for individual specifications' are as follows.

The regulating rods shall be positioned within the acceptable operating range for reculating rod position provided in the [ CORE] OPERATING LIMITS REPORT. (Used where the operating limit covers a range of acceptable operation, typically defined by a curve.)

The [ cycle-specific parameter] shall be within the limit provided in the

[ CORE] OPERATING LIMITS REPORT.

(Used where the operating limit has a single point value.)

O i

Generic Letter 88-16 3-Enclosure

SUMMARY

The alternative to including the values of cycle specific parameter limits in' individual specifications includes (1) the addition of a new defined term for the formal report that provides the cycle-specific parameter limits. (2) the addition of its associated reporting requirement to the Administrative Controls section of the TS, and (3) the modification of individual specifications to re-place these limits with a reference to the defined formal report for the values of these limits.

With this alternative, reload license amendments for the sole purpose of updating cycle specific parameter limits will be unnecessary.

G 4

4 9

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