ML20246D450
| ML20246D450 | |
| Person / Time | |
|---|---|
| Issue date: | 05/04/1988 |
| From: | Mcdonald W NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| To: | Beckjord E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20245D160 | List: |
| References | |
| FRN-52FR6334, RULE-PR-50 AC44-2-08, AC44-2-8, NUDOCS 8905100228 | |
| Download: ML20246D450 (43) | |
Text
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A C 44 -1 P D It, f rea u o.,, UNITED STATES
[" g NUCLEAR REGULATORY COMMISSION r4 E WASHINGTON, D. C. 20565 '
k.....)
MEMORANDUM FOR:
Eric S. Beckjord, Director
..J 1988 ;
Office of Nuclear Regulatory Research FROM: William G. Mcdonald, Director Office of Administration and Resources Management
SUBJECT:
0FFICE CONCURRENCE: 10 CFR PART 50, FINAL RULE-REVISIONS TO ACCEPTANCE CRITERIA FOR THE EMERGENCY CORE COOLING SYSTEM (ECCS) REGULATIONS The Office of Administration and Resources Management (ARM) has reviewed the above subject rule that revises the ECCS provisions contained in 550.46 and Appendix K of Part 50. We concur subject to the modifications as contained on the enclosed marked copy and that are addressed in this memorandum.
We recommend that the summary paragraph be shortened by placing the paragraph that explains the use of the word shall" and "must" in a footnote in the Supplementary Information portion of the rule.
In response to our list review of this rule, we note that you set out in separate paragraphs the Commission's response to the comments received on this rule. We believe this format e.hances the readability of this rule.
We also recommend that a sentence be added in the text to indicate that a separate detailed analysis of the comments received on this rule has been prepared and is available in the NRC Public Document Room.
Throughout this rule, references are made to several documents such as Enclosure I, the regulatory analysis; SECY-83-472, the information report on " Emergency Core Cooling System Analysis Methot," November 17, 1983; several NULEGS; and Regulatory Guide publications; however, there is no indication of the availability of these documents. Since several documents are referenced and provide essential information in support of this rule, we recommend that you include on page 21 a separate paragraph antitiled " Availability oi Documents." In this paragraph, list each document referenced in the rule and note its availability.
If you have any questions or comments regarding this review, please have a member of your staff contact David L. Meyer, Chief, Rules and Procedures Branch, Division of Rules and Recoros, ARM on extension 27086 or Alzonia Shepard on extension 27651.
8905100228 890504 William G. Mcdonald, Director R
52 334 PDR Office of Administration and Rescurces Management
Enclosure:
As stated
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NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 Emergency Core Cooling Systems; Revisions to Acceptance Criteria AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule. ,
SUMMARY
The Nuclear Regulatory Commission (NRC) is amending its regulations to allow the use of alternative methods to demonstrate that the emergency core cooling system (ECCS) would protect the f
nuclear reactor core during a postulated design basis loss-of-coolant accident (LOCA). The Commission is taking this action because research, performed since the current rule was written, has shown that
' calculations performed using current methods and in accordance with the current requirements result in estimates of cooling system !
performance that are significantly more conservative than estimates While the based on the improved knowledge gained from this research.
existing methods are conservative, they do not result in accurate I calculation of what would actually occur in a nuclear power plant during a LOCA and may result in less than optimal ECCS design and l operating procedures, In addition, the operation of some nuclear i l reactors is being unnecessarily restricted by the rule, resulting in increased costs of electricity generation. This rule, while continuing to allow the use of current methods and requirements, also allows the use of more recent information and knowledge to demonstrate j This that the ECCS would protect the reactor during a LOCA.
amendment, which applies to all applicants for and holders of construction permits or operating licenses for light water reactors, also relaxes requirements for certain reporting and reanalyses which j do not contribute to safety.
I ENCLOSURE E Page 1
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- h. 4 Note that the text of this final pule adopts a standard convention for imposinganobligationorprohi[itionwhichisendorsedbytheOffice
/
of the Federal Register. This convention involves the use of the word
': "shall" when the subject' s a person or legal entity and the word "must" when the subjpct is an inanimate object. This wording has been
/
adopted solely for consistency with the convention ard in no way
- indicates that e NRC is adopting practices that differ from the past. Both e word "shall" and "must" are to be construed as
$) imposing egal requirements or prohibitions, g y F alt.
EFFECTIVE DATE: (TMrty day:-af ter Mi kauun irH.he--Feded Pe;ister)
FOR FURTHER INFORMATION CONTACT:
L. M. Shotkin, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301 92-3530.
SUPPLEMENTARY INFORMATION:
5 BACKGROUN [ _ ) ho/ee On March 3,1987, the Nuclear Regulatory Commission published in the Federal Register proposed amendments (52 FR 6334) to 10 CFR Part 50 end append;x K. These proposed amendments were motivated by the fact that since the promulgation of Section 50.46 of 10 CFR part 50, j
" Acceptance Criteria for Emergency Core Cooling Systems (ECCS) in light Water Power Reactors," and the acceptable and required features and models specified in Appendix K to 10 CFR Part 50, cor.siderable research has been performed that has greatly increased the understanding of ECCS performance during a LOCA. It is now confirmed that the methods specified in Alapendix K, combined with other anaiysis methods currently in use, are highly conservative and that the actual cladding temperatures which would occur ducir.o e LOCA would be much In so7iciting lower than those calculated using Appendix K methods.
the public's comments on the proposed rule, the NRC specifically requested its views on cuestions posed by Commissioner Asselstine and The ACRS the Advisory Committee on Reactor Safeguards (ACRS).
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requested that the Commission solicit the public's coments on whether the existing rule should be " grandfathered" indefinitely. That is. '
1 I
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- 1. Should the conservative ECCS evaluation method of Appendix K.
be permitted indefinitely or should this- aspect of the ECCS -l L
rule be phased out after some period of time?
Commissioner Asselstine requested the public's coments on the following:
- 2. Should this rule change include an explicit degree of conservatism that must be applied to the evaluation models?
- 3. This rule change would allow a 5 to 10 percent increase in the fission product inventory that could be released from any core meltdown scenario. Should this rule change explicitly prohibit any increase in approved power levels until all severe accident issues and unresolved. safety
' issues are resolved?
- 4. Should the technical basis for this proposed rule change be reviewed by an independent group such as the American Physical Society? ~
5 swsqc/A McC.vnw/f ?
The comment period for tha proposed rule revision and the draft regulatory guide (52 FR 11385) expired on July 1,1987. Twenty-seven letters addressing the proposed rule were received by the expiration date, as well as nine responses to the request for comments on questions in the regulatory p. side. A number of late comats were 1 also received. These were also considered to the extent that new and substantial comments were provided.
The public comment on the proposed rule revisions have been divided into thirteen categories and are summarized in tne following paragraphs. Catagories one through four represent the responses to the specific questions posed by the ACRS and Commissioner Asselstine.
In general, consideration of the public comments resulted in no substantive revision to the proposed rule.
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design basis performance and a better understanding of small break behavior is a desirable safety goal from a risk perspective. Therefore, the grandfathering provision has been ,
retained in the final rule.
- 2. Specification of Explicit Degree of Conservatism (Question 2).
The majority of the responses to this question indicated that the proposed rule already contains conservatism in both the required uncertainty evaluation and in the acceptance criteria.
The use of additional conservatism would be inconsistent with the objective of the rule which is to provide a realistic evaluation of plant response during a LOCA. The NRC has not included an additional explicit degree of conservatism in this rule.
- 3. Resolution of all Safety Issues Prior to Allowing power Level Increases (Question 3). Some commenters pointed out that fission product inventory is not a direct function of total power, but rather it is the rate of fission product formation that is a direct function of power. Fission product inventory available for release during a core meltdown would be a function of burnup, not total power.
Actually, the inventory of fission products is a complex function of both time and power and not as simple as described by the commenters. Short lived isotopes, such as xenon and iodine, quickly reach an equilibrium inventory and total steady state inventory of these fission products is a direct function of power. Inventories of long-lived isotopes, sucr. as strontium and cesium, are functions of total fuel burnup, as described by the commenters. Intermediate-lived isotopic inventories are cou. plex functions of time, power, and integrated power. In an independent study documented in chapter XII of NUREG 1230, the staff determined that the change in risk due to a 5% power increase is negligible. The arguments above do not alter the Commission's position that the increase in fission products available for release during a core meltdown caused by a 5% power Page 5 ENCLOSURE E
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- 1. Grandfathering of Conservative ECCS Methods of Appendix K
-(Question 1).
l 1
Twenty of the commenters specifically addressed the ARCS question concerning the grandfathering of the current Appendix K approach.
Eighteen of these commenters recommended allowing continued use l
of the existing Appendix K evaluation models. Most cited the known conservatism as the basis of their recommendation. In addition, several commenters stated that in li0ht of the known conservatism not allowing continued use of existing Appendix K evaluation models would be unfairly burdensome to licensees who determine that they would not derive an economic benefit by performing realistic analysis of ECCS performance. The position of an additional commenter is unclear concerning grandfathering.
The remaining commenter was not opposed to grandfathering but v ,
thought the question is premature. This commenter believes that
- indefinite use of existing ECCS evaluation methods should be Q
considered when significant experience has been gained with the E' l d" s implementation of the new features of the rule but makes no 9.; recommendation as.to what policy the Commission should pursue in I- x { .the meantime.
- s_ The commission agrees with the majority of the r.o,mmenters that
} i existing Appendix K evaluation models should be permitted !
l '. . y, -
indefinitely. The Commiss!on also believes that the decision to j
'u j ,'
, ; permit continued use of suc.h models can and should be made at -
c ~. this time because it believes that both reethods provide acequate X ss "
protection of the public health and safety. Asdescribed3Mthe
.]
regulatory analysis (Enclosurejy, the probability of a large break is so low, that the choice of beet estimate versus g_y '
Appendix K has little effect on public risk. The TMI action plan calls for industry to improve their small break LOCA evaluation models to be more realistic when evaluating the more probable i small break accident scenario. This has been done within the "
context of M f50.46 and Appendix K compliance and was entirely appropriate since small breaks are not limiting in !
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increase is negligible compared to the uncertainty in fission product release. The Commission has decided not to delay the proposed rule revision pending resolution of all unresolved safety issues or severe accident issues4 and,%,t/w cvM4eproceedwiththis/
rulemaking, as planned.
4, Independent review of Technical Basis (Question 4). Several commenters indicated that the technical basis for the proposed rule has had adequate review as the research was being performed.
A number of commenters stated that it was the role of the ACRS to parform any review the proposed rule revision oecause it is uniquely qualified due to its familiarity with the research.
The Commission agrees that the technical basis has.had adequate review, except for the uncertainty methodology which is new and untried except for the General Electric Company's use of an uncertainty evaluation of their SAFER code. As a proof of principle and demonstration of feasibility, the ACRS and a second independent peer group has reviewed the uncertainty methodology developed by the NRC for use in quantifying the uncertainty of NRC developed thermal hydraulic transient codes. Based on the paucity of negative response concerning the technical basis for the proposed rule revision and favorable review of the NRC uncertainty methodology, the Commission plans no further review of the technical basis.
- 5. General Comnents on proposed Rule. Twenty-one commenters made comments of this nature. The majority of the comments came from the nuclear industry of which 19 expressed support of tne proposed rule. The intiustry also strongly supports the specific ECCS rule approach proposed by the NRC. One commenter neither supported nor opposed the proposed approach. One negative comment was received from an anonymous individual within the nuclear industry who implied, without specifics, that the ECCS rule is not sound and that public comment is not a fair hearing because expert insiders would be afraid to comment.
Page 6 ENCLOSURE E
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l Based on the absence of any supporting justification fer the negative response and the unprecedented amount of research supporting the rule revision, the NRC does not consider this comment to be valid and has proceeded with this rulemaking with no major revisions.
One commenter suggested that fuel reload suppliers should not be required to complete full LOCA/ECCS analyses because the hydraulics are not changed by a fuel change.
Although this point is valid, the Commission believes that it is an unworkable situation to allow fuel suppliers to make use of previous analyses performed by others. It is believed that serious questions of accountability would arise in cases where errors are discovered in evaluation models, requests are made to revise plant technical specifications, or some other questions regarding the analyses are raised. The NRC believes that sharen responsibility for evaluation models would not be in the best interest of the public health and safety and therefore has not implemented the suggestion of this commenter.
. The NRC received two requests for an extension of the comment
! period to allow time for review of NUREG-1230, which describes the research supporting the proposed rule revision.
The NRC believes the comment period was sufficient since most of l
I the research is not new and has been extensively reviewed in the l past. Both commenters were contacted and told that comments received after tne comment period would be censidered if time l 1
permitted. Comments from both parties were received late and were indeed considered by the NRC.
l l
! 6. Reportino Requirements. Some commenters viewed the proposed I reporting procedures as new requirements needing consideration in the backfit analysis while others stated that they are a major l relaxation and clarification of existing reporting requirements.
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The NRC position is that the reporting requirements are new only in the sense that they will now appear in the Code of Federal Regulations. In practice, these reporting requirements are indeed a clarification and relaxation over the current interpretation of the existing requirements and therefore the Commission does not view them as a backfit.
A number of commenters requested that only significant errors or changes in the non-conservative direction or only those thht result in exceeding the 2200# F limit be required to be reported.
In addition, a number of commenters suggested that the NRC require only annual reporting of significant errors or changes.
i The NRC considers a major error or change in any direction a cause for concern because it raises potential questions about the adequacy of the evaluation model as a whole. Therefore, the NRC requires the reporting of significant errors or changes, in either direction, on a timely basis so that the Commission may make a determination of the safety significance. Thus, the final rule contains no change in this requirement.
One commenter recommended that the word "immediate" be deleted 1
from the requirement to propose steps to be taken to demonstrate l
compliance in the event that the criteria in Aaea9eapW{50.46(b) are exceeded.
The Commission considers this a very serious condition in which the plant is not in compliance with the regulations and may be operating in an unsafe manner. The word "immediate" reflects this seriousness and is further defined by reference in other sections of Part 50.
l Several commenters questioned the need to report minor or f inconsequential errors.or changes, even on an annual basis, as l
required in the proposed rule.
Page 8 ENCLOSURE E
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While errors or changes which result in changes in calculated 0
peak clad temperatures- of less than 50 F are not considered to be ,
1 of imediate concern, the NRC requires cognizance of such changes .
or corrections since they constitute a deviation from what previously has been reviewed and accepted. The proposed annual reportin'g is believed to be a fair compromise between the burden of reporting and the Commission's need to be aware of changes and error corrections being made to evaluation models. Therefore, the annual reporting of minor errors remains,in the final rule.
One commenter interpreted the use of the words."or in the application of such a model" as requiring reporting when facility changes (already reportable under 50.59), resulting in model input changes, occur.
The regulatory language referred to is intended to ensure that applications of models to areas not contemplated during initial review of the model do not result in errors by extending a model s beyond the range that it was intended. The Commission does not believe that further clarification of this requirement is necessary and has not done so in the final rule.
Several commenters requested a further, relaxation of the reporting requirement by changing the definition of significant C
code errors from 50 F to 100'F.
While justification for the 50 F criteria is largely judgmental, l the NRC believes that it is sufficiently large to screen the code f error corrections and changes which have little safety significance while providing a mechanism for timely reporting of more serious errors and change.s. Since 50 F is a threshold for reporting and no further action is required pending NRC l
determination of safety significance, the Commission has retained this criteria in the final rule.
1 One commenter requested consideration for allowing that the cumulative effect of several errors and corrections be applied towards the 50 F threshold.
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'The requirement, which states that the 50 F criteria applies to the sum of the absolute magnitudes of temperature changes from ~
numerous error corrections or model changes was formulated specifically because the Commission requires knowledge of serious deficiencies in evaluation models in use by licensees.
Allowing errors or corrections which offset one another to relieve a licensee of the thirty-day reporting requirement, would be counter to this objective. If this recommendation were accepted, two errors or changes, having a large impact on the calculated peak cladding temperature but in the opposite direction, would not be repor_ table if the net magnitude of their difference was less than 50 F. For this reason, and the fact that no further action (beyond reporting within thirty days) is required, the Commission retained this requirement in the final n rule.
N .
' N.h 7. Continued Use of Dougall-Rohsenow. Five comments that addressed N this aspect of the proposed rule were received. One commenter
~
h believed that this correlation should not be permitted without j further verification'and should be phased out. Other commenters L w. supported continued use of the correlation subject to the M - :n. provisions of the proposed rule.
j -
f Ra The NRC position is that no safety concern is created by continued use of the correlation, as long as the evaluation model
,f[ .p h fj is overall conservative. Therefore, the Commission can not
., ? ., justify the burden of requiring licensees to modify their evaluation models and to ),erform reanalysis. As discussed in hhj
)3 SECY 83-472, current evaluation models contain more conservatism dan jdt those required by Appendix E. However, error
- $ [j corrections or changes could alter the conservatism of the model.
Therefore, the Commission believes that it is necessary to ensure continued overall conservatism in the evaluation models as a basis for continued use of the correlation. Therefore, the final
' rule does not modify this requirement except for the correction of a typographical error identified by one commenter.
Page 10 ENCLOSURE E
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'8. Uncertainty Evaluation. The comments received on the uncertainty >
evaluation support the proposed rule, particularly the .
flexibility provided by a non-prescriptive requirement.
Therefore, the Commission is publishing the final rule without modification of this requirement.
- 9. Acceptance Criteria. The'three comments received on this topic l I
were all supportive of the existing criteria, as contained in 550.46(b), and thur the Commission did not give consideration to altering them in the final rule.
- 10. , , Cladding Materials. Three commenters requested that the Commission consider broadening the language of the rule to allow the use of a range of zirconium based alloys for cladding material.
The Commission believes that this modification is beyond the scope of the current rule revision and should be considered in a separate rulemaking action in which it would receive appropriate In addition, public review and comment prior to implementation.
zircaloy cladding material is specified in other portions of the Code of Federal Regulation, su'ch as 550.44, making a change of 3- .
thistype{'smere suitable in a broacer regulatory context.
Therefore, the Commission ic not broadening the definiticn of cladding materials within this rulemaking.
I
- 11. Other Suggested Expant. ions to Rule Scoce. One commenter believes l
that hydraulic loads occurring during a LOCA could cause steam generator tubes to rupture and that the NRC should resolve steam generator tube integrity safety issues prior to publishing this rule.
Steam generator tubes are designed to withstand LOCA loads at allowed thinning, and there is no evidence to contradict this. l If anything, the problem would be with inspection techniques to Page 11 ENCLOSURE E !
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r detect the actual tube thinning and whether there .is an unacceptably high probability that a tube rupture during a LOCA due to tube thinning _is in excess of-the design basis. However, the risk from LOCA with concurrent tube rupture will not be greatly affected by the proposed rule change. As a result of the commenter's concerns, this issue has been assigned as a generic issue.(GI-141) to be prioritized by the NRC staff. The L results of the prioritization process will determine if further-action is required.
A second commenter believes that the ECCS rule does not i.
adequately address a plant's'long term decay heat removal capability, and recommends a "short/long term integrative analysis approach." Both the existing requirements and the-proposed rule contain the requirement to provide for long term cooling subsequent to a LOCA. Small increases in power that may result from the proposed rule should not greatly change decay heat removal requirements following a LOCA or any other accident or transient.- Thus, the issue of decay heat removal is not materially impacted by this rulemaking. Moreover, any proposed .
increase in power resulting from this rule promulgation would be approved only after the licensee demonstrates that decay heat removal capacities remain adequate. The Commission is planning no further action with regard to this issue.
x.l
- 12. Acceo_tability of Models Aporoved Under: SECY-83-472. One commenter requests that the rule Tangua D F F5M fie) to state exnlicitly that ECCS evaluation models that have been previously approved under(SECY-83-472 continue to be acceptable under this rule. - - - - - -
SECY 83-472 provides an alternative, acceptable method for
'defiliping ECCS evaluation models. Licensees were still required, however, to demo,n g te that evaluation modeis developed using thes SECY-83-472 ,ppoach complied with the requirements of Appendix K lo a0CF 50. This final rule e
page 12 ENCLOSURE E
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i explicitly finds that ECCS evaluation models, which have been previously approved as satisfying the requirements of Appendix K, remain acceptable. Therefore, the Comission sees no need for -
i.-
further clarificV ion of this issue.
Six letters commenting
- 13. Concents Received After Comment Period.
on the proposed rule were received subsequent to the end of the comment period. The Ccmmission considered these coments to the extent that the coments provided substantive information not previously considered.
One commenter believes that the proposed f50.46(a)(2) expands the discretion of the Director of the Office of Nuclear Reactor Regulation (NRR) by allowing imposition of immediate effective restrictions on reactor operation without a prior determination that such action is required to protect the public health, safety, or interest. NRC's intent is not to alter the responsibilities of the Director of NRR but to simply retain the 5 description of the scope of the authority that is currently found in 950.46(a)(1)(v). Furthermore, the provisions of 950.46(a)(2) do not specify the procedure followed by the Director of NRR. These procedures are g .out in Part 2 and remain unchanged by this rulemaking.
One commenter believes that the rule is illegal because it is based solely or cost savings considerations and that there is nothing wrong with large conservatism.
The Commission disagree.s With this assessnienc.
Safety factors are required to p otect the health and safety of the public when uncertainties in plant response exist. As these uncertainties are reduced, it is appropriate to modify these safecy factors to l The provide more realistic evaluation of actual plant response.
large conservatism of Appendix K served the public well in 1974 However, when there was great uncertainty in ECCS performance.
these conservatism are now known to be very large, and there is Page 13 ENCLOSURE E 4
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ao need to "over regulate" by maintaining this unnecessary margin. This type of activity can often result in the
- expenditure of resources that would be better spent improving safety in other areas. As indicated in the Regulatory Analysis, 7 the staff does not minimize these potential, benefits to safety of f -
the proposed rule revision. Rather, these benefits, while /
difficult to quantify, are believed to be substantial. While cost savings may have been one factor resulting in the rule change, the Commission believes that the conservatism contained in the acceptance criteria themselves, as well as those required in the uncertainty evaluation required in this rule, are adequate.
'to protect _ the health and safety of the public.
This commenter also cites portions of the 1975 General Electric Company's Nuclear Reactor Study (Reed Report), which claims that there is a lack of understanding of phenomena and small safety margins.
Many of the conclusions of the " Reed Report" were valid in 1975-when it was written and due to this fact it was difficult to show that sufficient safety margins existed. Most of the research discussed in NUREG-1230 has been conducted since the " Reed Report" was written and has resulted in significant improvement in understanding LOCA phenomena. We now know that significant margin to the ECCS acceptance criteria exists, particularly for the BWR/6 which was of concern in the " Reed Report." The contents of this report have been reviewed by the Commission on several occasions, most recently in NUREG-1285, and the finding has been made that no new significant safety issues are identified. For these reasons, the NRC is proceeding with this rulemaking, as proposed.
The same commenter also recommends that credit for ECCS margins be taken in the Individual Plant Examinations (IPE) and not through generic rulemaking.
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The Commission agrees that plant specific differences may justify the application of different margins and that these may be addressed through Individual. plant Examinations. However, the .
requirement for licensees to evaluate ECCS performance and meet l
the acceptance criteria specified in 10 CFR 50.46(b) is generic.
l- The Commission believes that margins that may be reduced due to a' better understanding of a reactor's response to a LOCA should be applied through a generic rulemaking action because it allows a l
broad range of technical review of the issues, enhances public participation in the process, and provides a complete public ;
record. Therefore, the Commission has decided to proceed with the rulemaking as planned.
Finally, this commenter questions the experimental basis for this rule because full-scale ECCS bypass data is not yet available.
i The 20/3D tests which will provide this important data represent a small portion of the total researe upon which this rule ,
relies. Significant research on EC bypass has already been completed in small scale vessels and the full-scale work is required only to confirm the smaller scale results and quantify p any uncertainty due to scale effects. One full scale ECC bypass test has already been completed under the 2D/3D program which showed that more margin exists than expected from the small scale I
tests. Completion of the full-scale tests only affects the "
uncertah+ies in the calculations, and reduces them,,Mr v Uncertainties must be addressed by licensees in any analysis I
under the revised rule whetner 20/3D results are available or not. The Commission concludes that there is no need .to delay the final rule, while awaitirig these data. .
i
SUMMARY
OF RULE CHANGES 5 50.46 Acceptance Criteria for Emergency Core Cooling Systems for :
Light Water Reactors:
Page 15 ENCLOSURE E
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Section 50.46(a)(1) is amended and redesignated S 50.46(a)(1)(i) todeletegherequirementthatthefeaturesofSectionIofAppendixK to O CF Part 50 be used to develop the evaluation model. This sect on now requires that an acceptable evaluation model have suf-ficient supporting justification to show that the analytical technique
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realistically describes the behavior of the reactor system during a LOCA. The NRC expects that the analytical technique will, to the extent practicable, utilize realistic methods and be based upon applicable experimental data. The amended rule also requires that the uncertainty of the calculation be estimated and accounted for when comparing the results of the calculation to the temperature limits and other criteria of f 50.46(b) so that there is a high probability that the criteria would not be exceeded. The Commission expects the realistic evaluation model to retain a degree of conservatism consistent with the uncertainty of the calculation. The final rule does not specifically prescribe the analytical methods or uncertainty evaluation techniques to be used. However, guidance ha_s..be_en provided .
in the form of a Regulatory Guide . As discussed in Y-83-472]the NRC has, in the past, found acceptable a method for estimating'the uncertainty that was judged to be at least at the 95% probability level. This probability level of 95% is considered adequate to meet ;
the high level of probability required by the rule. It is also recognized that the probability cannot be determined using totally rigorous mathematical methods due to the complexity of the calculations. However, the NRC requires that any simplifying assumptions be stated so that the Commission may evaluate them to ensure that they are reasonable. The NRC has independently developed and exercind a niethodology to estimate the uncertainty associated l
with its own thermal-hydraulic safety codes. This methodology is
' 1 Regulatory Guide, "Best Estimate Calculations of Emergency gl Core Cooling Systems Performance," was issued to all licensees. _
' '~~
- I- E T'ifeguests for single copies of the guide, which may be reproduced, I
f should be made in writing to the U.S. Nuclear Regulatory
' Commission, Washington, DC 20555, Attention: Director, Division~~~
of Technical Information and Document Control.
)
Page 16 ENCLOSURE E
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described in the " Compendium of ECCS Research."2 While this method has not been reviewed for acceptability from the standpoint of safety licensing, it may provide additional guidance on how the uncertainty may be quantified. In addition to providing guidance to industry, j this work was undertaken to provide a proof of principle and a tool to independently audit industry submittals. Appendix K,Section II,
" Required Documentation," remains generally applicable, with only minor revisions made to be consistent with the amended rule.
A new paragraph (ii) has been added to 650.46(a)(1) to allow the features of Section I of Appendix K to be used in evaluation models as an alternative to performing the uncertainty evaluation specified in the amended 9 50.46(a)(1)(i). This method would remain acceptable because Appendix K is conservative with respect to the realistic method proposed in the amended 9 50.46(a)(1)(i). This would allow both current and future applicants and licensees to use existing evaluation models if they did not need or desire relief from current operating restrictions.
In 50.46 j paragraphs (a)(2) and (3) have been revised to eliminate portions of those paragraphs concerned with historical implementation of the current rule. These provisions have been replaced as describea in the following paragraphs:
Section 50.46(a)(2) has been revised to indicate that restrictions on reactor operation may be imposed by the Director of Nuclear Reactor Regulation, if the ECCS cooling performance evaluations are not consistent with the requirements of 69 50.46(a)(1)(i) and (ii). This section has been added to retain -
similar requirements that have been deleted fro 50,46(a)(1)(i) by this rule revision. This section does not spec"1fy the procedures to ~
be followed by the Director. These procedures are found in l'O b ' ~ ~
Part 2 and are unchanged by this rulemaking.
2 " Compendium of ECCS Research for Realistic LOCA Analysis "
NUREG-1230, TBp.
page 17 ENCLOSURE E
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The current rule contains no explicit requirements concerning re-porting and reanalysis when errors in evaluation models are discovered or changes are made to evaluation models. However, current practice has required reporting of errors and changes and reanalyses with the revised evaluation models. This final rule explicitly sets forth requirements to be followed in the event of errors or changes. The definition of a significant change is currently taken from Appendix K,Section II.1.b which defines a significant change as one which changes calculated cladding temperature by more than 20 F.
The revised 5 50.46(a)(3) states specific requirements for reporting and reanalyses when errors in evaluation models are discovered or changes are made to evaluation models. It requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported. Thereafter, although reanalysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected. The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the significance of the error or change and to maintain cognizance of modifications made subsequent to NRC review of the evaluation model. Past experience has I shown that many errors or changes to evaluation models are very minor and the burden of immediate reporting cannot be justified for these l minor errors because they do not affect the immediate safety or operation of the plant. The Mtc therefore requires periodic reporting '
to satisfy NRC's hfie'ed to be apprised of changes or errors without imposing an unnecessary burden on the applicant or licensee. This ,
report is to be filed within one year of discovery of the error and must be reported each year thereafter until a revised evaluation model or a revised evaluation correcting minor errors is approved by the NRC staff.
1 Page 18 ENCLOSURE E
Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as ,
to the adequacy of the overall evaluation model. This final rule
- @ Q defines a significant error or change as one which results in a calculated peak fuel cladding temperature different by more than 500 F, or an accumulation of errors and changes such that the sum of the absolute magnitude of the temperature changes is greater than 500F. More timely reporting (30 days) is required for significant errors or changes. This definition of a significant change is based on NRC's judgement concerning the importance of errors and changes typically reported to the NRC in the past. This final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements. Errors or changes that result in the calculated plant performance exceeding any of the criteria of 6 50.46(b) mean that the plant is not operating within the requirements of the regulations and require immediate reporting as required by 9 50.55(e), 5 50.72 and 4
5 50.73 and immediate steps to bring the plant into compliance with <
9 50.46.
kppendixKECCSEvaluationModels:
Amendments have been made to Appendix K,Section I.C.S.b, to modify the post-CHF heat transfer correlations listed as acceptable.
The "McDonough" reference has been replaced with a more recent paper by the same authors entitled "An Experimental Study of Partial Film Boiling Region With Water at Elevated Pressures in a Round Vertical Tube" which is more generally available and which includes additional date.
The heat transfer correlation of Dougall and Rohsenow, listed as an acceptable heat transfer correlation in Appendix K, paragraph I.C.5.b, has been removed, since research performed since Appendix K was written has shown that this correlation overpredicts heat transfer Page 19 ENCLOSURE E
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coefficients under certain conditions and therefore can produce nonconservative results. A number of applicants and licenrees currently use the Dougall-Rohsenow correlation in approved evaluation models. The NRC has concluded that the continued use of this correlation can be allowed. This is. appropriate (even though parts of the approved evaluation model, Dougall-Rohsenow, are known to be nonconservative) because the existing evaluation models are known to contain a large degree of overall conservatism even while using the Dougall-Rohsenow correlation. This large overall conservatism has been demonstrated through comparisons between evaluation model calculations and calculations using NRC's best estimate computer codes. Thus requiring that the applicants and licensees remove the Dougall-Rohsenow correlation from their current evaluation models cannet be justified as necessary to maintain safety.
A new Section I.C.5.c has been added to Appendix K to state the Commission's requirements regarding continued use of the Dougall-Rohsenow correlation in existing evaluation models.
Evaluation models which make use of the Dougall-Rohsenow correlation and have been approved prior to the effective date of this rule
- )may continue to use this correlation as long as no changes are made to the evaluation model which significantly reduce the current overall conservatism of the evaluation model. If the applicant or licensee submits propcsed changes to an approved evaluation model, or submits corrections to errors in the evaluation model which significantly reduce the existing overall conservatism of the model, continued use of the Dougall-Rohsenow correlation under
' conditions where nonconservative heat transfer coefficients result would no longer be acceptable. For this purpose, a significant reduction in overall conservatism has beeh defined as a " net" reduction in calculated peak clad temperature of at least 50 F from that which would have been calculated using existing evaluation models. A reduction in calculated peak clad temperature could potentially result in an increase in the actual allowed peak power in the plant. An increase in allowed plant peak power with a known Page 20 ENCLOSURE E
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nonconservatism in the analysis would be unacceptable. This definition of a significant reduction in overall. conservatism is based on a judgement regarding the size of the existing overall conservatism in evaluation model calculations relative to the conservatism required to account for overall uncertainties in the calculations.
Appendix K,Section II.1.b, has been removed since this requirement has been clarified in the amended 9 50.46(a)(3).
Likewise, Appendix K, Section 11.5, has been amended to account for
~ the fact that not all evaluation models will be required to use the features of Appendix K,'Section I. These minor changes to Appendix K do not affect any existing approved evaluation models since the changes are either " housekeeping" in nature or are changes to
" acceptable features," not " required features."
i ) kiWud2277Y e r Docwn nv1 :
FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY d l.
The Commission has determined under the National Environmental p Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule is not a major Federal -
action significantly affecting the quality of the human environment
-and therefore an environmental impact statement is not required. The primary effe'ct of the rule is to allow an increase in the peak local power in the reactor. This could be used either to tailor the power shape within the reactor or to increase the total power. Changing the power shape without changing the total' power has a negligible effect on the environmental impact. The total power could also be increased,
~out is expected to be increased by no more than about 5% to 10% due to hardware limitat-ons in existing plants. This 5% to 10% power increase is not expected to cause difficulty in meeting the existing environmental limits. The only change in non-radiological waste will be an increase in waste heat rejection commensurate with any increase in power. For stations operating with an open (once through) cooling system, this additional heat will be directed to a surface water body.
Discharge of this heat is regulated under the Clean Water Act Page 21 ENCLOSURE E
AVAILABILITY OF DOCUMENTS
- 1. Copies of NUREGS'1230 and 1285 may be purchased from the Superintendent of. Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, D.C. 20013-7082. Copies are also available from the National Technical Information Service, 5285 Port Royal Road, Springfield YA 22161. A copy is a'.so available.for public inspection and/or copying at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555.-
- 2. Copies of SECY-83-472, an information report entitled " Emergency Core Cooling Systems Analysis Methods", dated November 17, 1983 is available for inspection and copying at the NRC Public Document Room,1717 H Street NW., Washington, DC 20555. Single copies of this report may_be obtained by writing ( give name and address of the contact).
- 3. Regulatory Guide, "Best Estimate Calculations of Emergency Core Cooling Systems Performance" may be obtained by writing to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory
- Commission, Washington, DC 20555.
- 4. The Summary of Public Comments Analysis received in response to this rule is available for public inspection at the NRC Public Document Room 1717 H Street NW., Washington, DC 20555 1
? '- / I
. c . .,
administered by the U.S. Environmental Protection Agency (EPA) or designated state agencies. It is not intended that NRC approval of increased power level affects in any way the responsibility of the licensee to comply with the requirements of the Clean Water Act. The environmental assessment and finding of no significant impact on which this determination is based are available for :nspection at the NRC Public Document Room, 1717 H Street NW, Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from L. M. Shotkin, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington DC. 20555, 492-3530.
telephone (301)f f
PAPERWORK REDUCTION ACT STATEMENT j (-n W kip I This rule amends information collection requirements that are ,
3 subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These reporting requirements were approved by the Office of Management and Budget (Approval Number 3150-0011).
REGULATORY ANALYSIS _ l'- ~ h, The Commission has prepared a regulatory aralysis for this final regulation. The analysis examines the costs and benefits of the f
I alternatives considered by the Commission. The regulatory analysis is available for inspection and copying for a fee at the NRC Public Docu-ment Room, 1717 H Street NW, Washington, DC. Single copies of the analysis may be obtained from L. M. Shotkin, Office of Nuclear ~
Regulatory Research, Washington, DC. 20555, telephone (301 43-7825.
I'~ "
REGULATORY FLEXIBILITY CERTIFICATION-- ,
As required by the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this rule will not have a l significant economic impact upon a substantial number of small I
entities. This rule affects only the licensing and operation of 1
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nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in rtigulations issued by the Small Business Administration in 13 CFR Part 121. Since these companies are dominant in their service areas, this rule does not fall within the purview of the Act.
BACKFIT ANALYSIS _ >
Although a backfit analysis is not required by 10 CFR 50.109 because the rule does not require applicants or licensees to make a change but only offers additional options, the factors in 10 CFR 50.109(c) have been analyzed as indicated below. More detailed information relevant to this backfit analysis may be found in the regulatory analysis referenced above.
- 1. Statement of the specific objectives that the backfit is designed to achieve.
The objective of the rule is to modify 10 CFR 50.46 and Appendix K to permit the use of realistic ECCS evaluation models.
More realistic estimates of ECCS performance, based on the improved knowledge gained from recent research on ECCS performance, may remove unnecessary operating restrictions.
- 2. General description of the activity that would be reouired by the licensee or applicant in order to complete the backfit.
The amendment allows alternative methods to be used to demonstrate that the ECCS would protect the nuclear reactor core during a postulated design basis loss-of-coolant accident (LOCA).
While continuing to allow the use of current Appendix K methods and requirements, the rule also allows the use of more recent information and knowledge currently available to demonstrate that l
the ECCS would perform its safety function during a LOCA. If an l
page 23 ENCLOSURE E
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applicant or licensee elects to use a new realistic model they will be required to provide sufficient supporting justification to vclidate the model and include comparisons to experimental data and estimates of uncertainty. In accounting for the uncertainty, the analysis would have to show, with a high level of probability, that the ECCS performance criteria are not exceeded.
3.
Potential change in risk to the public from the accidental offsig a release of radioactive materials.
s a
The rule could result in increased local power within the reactor core and possibly increases in total power. Power in-creases on the order of ' 10% will have an insignificant effect on risk. One effect of increased power could be to increase the fission product inventory. A five percent power increase would result in a less than five percent increase in fission products.
Thus, less than five percent more fission products might be released during core melt scenarios and potentially released to the environment during severe accidents.
The rule still requires that fuel rod peak cladding temperature (PCT) remain below 2200 F. Reactors choosing to increase power by about five percent will be operating with less margin between the PCT and the 2200 F limit than previously. The increased risk represented by this decrease in margin and increase in fission product inventory is negligible and falls within the uncertainties of PRA risk estimates. In addition, other safety limits, such as departure from nucleate boiling (DNB), and operational limits, such as turbine design, will limit the amount of margin reduction permitted under the rule. The rule could also potentially reduce the risk from pressurized thermal shock by allowing the reactor to be operated in a manner which reduces the neutron fluence to the vessel.
Page 24 ENCLOSURE E
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- 4. Potential impact on radiological exposure to facility employees, j Since the primary effect of the rule involves the calculational methods to be used in determining the ECCS cooling performance, it is expected that there will be an insignificant impact on the radiological exposure to facility employees.
Because of the reduced LOCA restrictions resulting from the new calculations it is possible for the plant to achieve more
- efficient operation and improved fuel utilization with improved maneuvering capabilities. As a result, it is conceivable that there could be a reduction in radiological exposure if the fuel reloads can be reduced. This effect is not expected to be very significant.
- 5. Install & tion and continuing costs associated with the backfit, including the cost of facility down times or the cost of construction delay. <
LOCA considerations resulting from the present rule are re-stricting the optimum production of nuclear electric power in some plants. These restrictions can be placed into the following three categories:
(1) Maximum plant operating power, (2) Operational flexibility and operational efficiency of the ,
plant, and (3) Availability of manpower to work on other activities.
The effect of the rdle will vary from plant to plant. Some plants may realize savings of several million dollars per year in fuel and operating costs. Significantly greater economic benefit would be realized by plants able to increase total power as a result of this final rule. The regulatory analysis cited above indicates that the total present value of the energy replacement i
Page 25 ENCLOSURE E
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cost savings for a five percent power upgrade would vary between 18 and 127 million dollars depending on the plant. Additional.
information concerning these potential cost savings are included ,
in the regulatory analysis.
- 6. The potential safety impact of changes in plant or operational complexity including the effect on other proposed and existing regulatory requirements.
There are safety benefits derivable from alternative fuel management schemes that could be utilized. The higher power peaking factors that would be allcwed with the final rule provide greater flexibility for fuel designers when attempting to reduce neutron flux at the vessel wall. This can result in a corresponding reduction in risk from pressurized thermal shock.
The reduced cladding temperatures that would be calculated under the revised rule offers the possibility of other design and operational changes that could result from the lower calculated temperatures. ECCS equipment numbers, sizes or surveillance -
requirements might be reduced and still meet the ECCS design criteria (if not required to meet other licensing requirements).
Another option may be to increase the diesel / generator start time duration.
In summary, the effect of this rule on safety would have both potential positive and negative aspects. The potential for 7
reductionofECCfsystemsinexistingornewplantsispresent.
However, several positive aspects may also be realized under the final rule. The net effect on safety would be plant specific.
However, the probability of a large break LOCA is so low that the choice of best estimate versus Appendix K would have little effect on public risk.
page 26 ENCLOSURE E l
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- 7. The estimated resource burden on the NRC asso'ciated with the pro-posed backfit; and the availability of such resources.
The major staff resources required under the final rule are ,
to review the realistic models and uncertainty analysis required by the revised ECCS Rule. Based on previous experience with the General Electric Company's SAFER model and the learning that has resulted from these efforts, it is estimated that approximately i
one staff year would be required to review each generic model submitted. There are four major reactor vendors (GE already has a revised evaluation model approved under the existing Appendix K
~for both jet pump and non-je ump plants and may update their methodology under this new ule) and several fuel suppliers and utilities which perform their own analyses and potentially might submit generic models for review. However, it is expected that ,
only 3 or 4 generic models would be submitted since not all plants would benefit from this rule. Thus, about 3-4 staff years c would be required to review the expected generic models. Once a generic model is approved, the plant specific review is very short. In addition, several vendors are currently planning to submit realistic models in conjunction with the use of SECY-83-472. Therefore, staff resources would be expended to review these models in any event. Since these models would not change as a result of the revised ECCS rule, there should be no net increase-in resources required over that already planned to be expended. In summary, while it is difficult to estimate accurately, it is expected that the rule change will have a small overall impact on NRC resources.
- 8. The potential impact of differences in facility type, design or age on the relevancy and practicality of the backfit.
The degree to which the rule would affect a particular plant depends on how limited the plant is by the LOCA restrictions. I General Electric Company (GE) plants do tend to be limited in Page 27 ENCLOSURE E
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operation by LOCA restrictions ~ and would benefit from relief from (
LOCA restrictions. 'However, this relief is already available for most GE plants through the recently approved SAFER evaluation .
model. Any additional relief due to a rule change would be of little further benefit. Westinghouse (W) plants would appear to directly benefit from relaxation of LOCA limits. W plants represent the largest number of plants, with 47 plants operating and 10 additional plants being constructed. W indicates that most of these plants are limited by LOCA considerations. The potential benefit for plants of B&W and CE design is uncertain at this time.
- 9. Whether the proposed backfit is interim or final and if interim, the justification for imposing the oroposed backfit on an interim basis.
The rule, when made effective, will be in final form and not interim form. It will continue to permit the performance of ECCS c cooling calculations using either realistic models or models in accord with Appendix K.
(Er*!Of LIST OF SUBJECTS IN 10 CFR PART 50 ;)
Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, j Reporting and Recordkeeping requirements.
For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the ssz Ener,gy a,v Reor-ganization Act of 1974, as amended, and 5 U.S.Cg553, the NRC is adopting the following amendments to 10 CFR Part 50.
Page 28 ENCLOSURE E
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PART 50-DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES t 0Nmurs C- 1. The authority citation for Part 50 h r=iset to read as Secs. 102, 103, 104, 105, 161, 182, 183, 186, follo g UTHORITY:
189, 68 Stat. 936, 937, 938, 948, 953, 954,- 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246, (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L.95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs.101, 185, 68 Stat. 936, 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.23, 50.35, 50.55, 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C.
2235). Sections 50.33a, 50.55a, and Appendix Q also issued under sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat 1245 (42 U.S.C. 5844).
Sections 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat. 2073, (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Section 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C.
2138). Appendix F also issued under sec. 187, 68 Stat. 955 (42 U.S.C.
2237).
For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C.
2273), g6 50.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54, and 50.80(a) are issued under sec. 161b, 68 Stat. 948, as amended I (42 U.S.C. 2201(b)); ss 50.10(b) and (c) and 50.54 are issued under sec. 161i, 68 Stat. 949, as amended (42 U.S.C. 2201(i)); and s 50.9, 50.55(e), 50.59(b), 50.70, 50.71, 50.72, 50.73, and 50.78 are issued l
under sec. 1610, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).
(
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\Q In,9 50.46, paragraph (a) is revised to read as follows: H 50.M ,
KAcceptancecriteriaforemergencycorecoolingsystemsforlight-water
- nuclear power reactors.
(a)(1)(1) [Except-as-previded-in-paragraph-faff23-and-fB)-ef this-sectient] Each boiling and pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical Zircaloy cladding must be provided with an emergency core cooling system (ECCS) that must be designed such that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated i loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated. [ Appendix-K; E6GS-Evaluatien-MedelsT-sets-ferth-eertain-required-and-acceptable .
features-of-evaluation-medels:] Except as provided in para gaph C.i AY (a)(1)(ii) of this section, the evaluation model must include sufficient supporting justification to show that the analytical technioue realistically describes the behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the' calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded. Appendix K, Part II, Recuired Documentation, sets forth the documentation requirements for each evaluation model. [6enfermance-with-the-eriteria-set-ferth-in paragraph-fb)-ef-this-sectien-nith-EG6S-eeeting-performance-eaiemiated in-accordance-with-an-acceptable-evaluation-medeit-may-require-that l
restrictions-be-imposed-en-reneter-eperation;]
l.
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Page 30 ENCLOSURE E
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1 (ii) Alternatively, an ECCS evaluation model may be develooed in conformance with the required and acceptable features of Aopendix K ECCS Evaluation Models.
(2) The Director of Nuclear Reactor Regulation may impose restrictions on reactor operation if it is found that the evaluati g of ECCS cooling performance submitted are not consistent with paragraphs (a)(1)(i) and (ii) of this section.
(3)(i) Each applicant for or holder of an operating license or construction permit shall estimate the effect of any change to or error in an acceptable evaluation model or in the application of sug a model to determine if the change or error is significant. For this purpose, a significant change or error is one which results in a calculated peak fuel cladding temperature different by more than 50 F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the resoective temperature changes is greater than 500F.
(ii) For each change to or error discovered irt an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or licensee shall recort the nature of the change or error and its estimated effect ,cn the i
limiting ECCS analysis to the Commission at least annually as specified in 6 50.4. If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a f l reanalysis or taking other action as may be needed to show compliance l
with 6 50.46 recoirements. This schedule may be developed using an integrated scheduling system previously accroved for the facility by the NRC. For those facilities not using an NRC approved integrated scheduling system, a schedule will be established by the NRC staff within 60 days of receipt of the prooosed schedule. Any change or Page 31 ENCLOSURE E
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9 error correction that results in a caiculated ECCS performance that does not conform to the criteria set forth in paragraph (b) of this section is a reportable event as described in 66 50.55(e), 50.72 and 6 50.73. The affected applicant or lico_nsee shall propose immediate steps to demonstrate compliance or bring plant design or operation into compliance with 6 50.46 requirements.
i
- 3. In 10 CFR Part 50 Appendix K, paragraph II.1.b is deleted, paragraph II.1.c is redesignated II.1.b, the text of paragraph I.C.5.b and paragraphs II.1.b and 11.5 are revised, and a new section I.C.S.c is added to read as follows:
APPENDIX K - E'CCS EVALUATION MODELS I. REQUIRED AND ACCEPTABLE FEATURES OF THE EVALUATION MODELS***
C. Blowdown Phenomena ***
- 5. Post-CHF Heat Transfer Correlations.***
b
- b. The Groeneveld flow film boiling correlation (equation 5.7 of D. C. Groeneveld, "An Investigation of Heat Transfer in the Liquid Deficient Regime," AECL-3281, revised December 1969), the Dougall-Rohsenow flow film boiling correlation (R:-S;-Bengail-and-W:
M:-Rehsenewr.uFilm-Beiling-en-the-inside-of-Vertical-Tabes-with-8pward Flen-of-Fluid-at-Eew-Qua+4tiesi .MIT-Repert-Namber-9079-267-Eambr+dge7 n i Massachusetts-September-1963)T] and the Westinghouse correlation of steady-state transition boiling (" Proprietary Redirect / Rebuttal Testimony of Westinghouse Electric Corporation," USNRC Docket RM-50-1, page 25-1, October 26, 1972) are acceptable for use in the post-CHF boiling regimes. In addition j the transition boiling correlation of j McDonough, Milich, and King (J. B. McDonough, W. Milich, E. C. King,
[uPartial-Film-Beiling-with-Water-et-2000-psig-in-a-Reund-Verticai Taberu -MSA-Research-Gerp:7-Technicai-Repert-62-fNP-6976)T-fi958)] "An Experimental Study of partial Film Boiling Region with Water at Elevated Pressures in a Round Vertical Tube," Chemical Engineering l
Page 32 ENCLOSURE E
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Progress Symposium Series, Vol. 57, No. 32, pages 197-208,!(1961) is suitable for use between nucleate and film boiling. Use of all these correlations i,s [ shah-be] restricted as follows: ,
k Q. jQ i c. Evaluation models approved after ( ) which make use et the Dougall-Rohsenow flow film boiling correlation (R. S.
, $, Dougall and W. M. Rohsenow, " Film Boiling on the Inside of Vertical x
i.
j Ng t t.iv Tubes with Upward Flow of Fluid at Low Qualities, MIT- Report Dk
. ,. t Number 9079 26, Cambridge, Massachusetts, September 1963) may not use this correlation under conditions where nonconservative predictions of 4
,{[f
. heat transfer result. Evaluation models that make use of the Dougall-x _'<
- 1. Rohsenow correlation and were approved prior to (
)
% ^ 9 continue to be acceptable until a change is made to, or an error is
( ? 'ENf corrected in, the evaluation model that results in a significant jM
- reduction in the overall conservatism in the evaluation model. At
% that time continued use of the Dougali-Rohsenow correlation under e conditions where nonconservative predictions of heat transfer result will no longer be acceptable. For this purpose, a significant reduction in the overall conservatism in the evaluation model would be a reduction in the calculated peak fuel cladding temperature of at least 50 F from that which would have been calculated on
( ) due either to individual changes or error corrections or the net effect of an accumulation of changes or error corrections.
II. REQUIRED DOCUMENTATION 1.a. ***
- b. A complete listing of each computer program, in the same form as used in the evaluation model, must be furnished to the Nuclear Regulatory Commission upon reouest.
Page 33 ENCLOSURE E
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- 5. General Standards for Acceptability - Elements of evaluation models reviewed will include technical adequacy of the calculational ,
methods, including : for models covered by 6 50.46(a)(1)(ii),
compliance with required. features of Section I of this Appendix K [and l previsien-of-a-level-ef-safety-and-margin-of-conservatism-eemperable te-ether-acceptable-evaluation-models--taking-inte-accennt-significant differences-in-the-reneters-to-which-they-apply-] : and, for models covered by 6 50.46(a)(1)(i), assurance of a high level of probability that the performance criteria of 6 50.46(b) would not be exessded.
hock'/N, A1E Dated at-WashinvLwn, 00 this day of , 1988.
For the Nuclear Regulatory Commission.
Samuel J. Chilk, Secretary of the Commission.
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Page 34 ENCLOSURE E L _ _ . _ _ . _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ -
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For: The Commissioners From: Victor Stello, Jr.
Executive Director for Operations ,
Subject:
REVISION OF THE ECCS RULE CONTAINED IN APPENDIX K AND SECTION 50.46 0F 10 CFR PART 50
Purpose:
To obtain Comission approval for publication in the Federal Register of final amendments revising the ECCS rule contained in Appendix K and Section 50.46 of 10 CFR Part 50.
Cateaory: This paper covers a major policy question.
Issue: Should the final text of the proposed ECCS rule be approved by the Commission. The proposed amendments would:
- a. allow the use of best-estimate evaluation methods through the utilization of more recent information gained on the performance of ECC systems,
- b. relax certain reanalysis requirements that do not contribute to safety,
- c. permit the continued use of current methods for'those licensees and applicants wishing to do so, and
- d. delete from Appendix K the reference to the Dougall-Rohsenow heat transfer correlation as an acceptable model.
Summary: Section 50.46 of 10 CFR Part 50 requires that calculations be performed to show that the emergency core cooling systems (ECCS) will acequately cool the reactor in the event of a loss-of-coulantaccident(LOCA). Appendix K :.ets forth certain required and acceptable features that the evaluation models, used to perform these calculations, must contain.
Contact:
Louis M. Shotkin, RES 49-23530
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R The Commissioners 2 The results of these calculations are used to 'detemine the
. acceptability of the ECCS' performance. In many_ instances,
- these calculations result in technical specification limits
. on reactor operation (e.g., peak local. power) in order to comply with the 2200'F cladding temperature limit and other limits of $50.46. These limits may restrict the total power output and optimal o Westinghouse plants)peration in terms. ofofefficient many reactors (e.g., most fuel utilization, maneuvering capability ano surveillance requirements.
Removing unnecessary restrictions on operation will allow increased U. S. electricity production, worth several hundred L million dollars a ' year, without loss of benefit to the public.
health and safety. j
~
N'RC, DOE (includingAECandERDA),U.S.nuclearindustryand foreign research on ECCS performance since the present ECCS ,
I rule was issued provides a technical understanding which
... shows that the existing ECCS rule. restrictions are more stringent than necessary for safety. ,Thus, the staff'
. recomends that. the ECCS rule be amended to reflect -this more realistic safety assessment and to remove unnecessary operating restrictions. This is consistent with the 1973 Comission opinion oublished with the existing rule. As a result of.the large body of research into the behavior of emergency core cooling systems during a loss-of-coolant accident, the Nuclear Re amend. its requirements52(gulatory FR 6334)Comission has proposed to allow licensees hnd to applicants to use best-estimate calculations accompanied by an uncertainty quantification to demonstrate compliance with the acceptance criteria specified in 10 CFR 50.46(b). The rule changes were published as a proposed rule on March 3, 1987. Based on the generally fasorable public response to the proposed rule, the staff recomends.that the Comission approve the final rule for publication with no changes. "
Discussion: On March 3,1987, the Nuclear Regulatory Comission published in the Federal Register proposed amendments (52 FR 6334) to l 10 CFR Part 50 and Appendix K (Enclosure A). These proposed amendments were motivated by the fact that since the promulgation of Section 50.46 of 10 CFR Part 50, " Acceptance Criteria for Emergency Core Cooling Systems (ECCS) in Light Water Power Reactors", and the acceptable and required features and models specified in Appendix K to 10 CFR.Part 50 l- for performing ECCS performance analyses, considerable research has been performea that has greatly increased the understanding of ECCS performance during a LOCA. We have now confinned that the methods specified in Appendix K, combined with other analysis methods currently in use, are highly conservative and that the actual cladding temperatures which '
would occur during a postulated LOCA would be much lower than those calculated using Appendix K methods. In addition, the large body of research available has provided a method to
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The Commissioners 3 both estimate the degree of conservatism in Appendix'K.
calculations and, to determine to a reasonable extent, the uncertainty associated with that estimate. In light of these factors, the Comission approved the publication of the proposed rule which would permit licensees and applicants to make realistic calculations of ECCS performance during a LOCA in the regulatory process, as well as to require an estimate of the uncertainty of the calculation to assure that there is a high probability that the acceptance criteria in 10 CFR 50.46 (b) (e.g., calculated peak cladding temperature shall ;
not exceed 2200*F) would not be exceeded. Additional 1 discussion regarding inherent margin in the 2200"F limit is provided in Enclosure J.
In considering the staff's recommendation to approve publica-tion of the proposed rule, the Comission directed the staff to subject the methodology for evaluating the uncertainty in
, NRC codes that was developed by NRC to both peer and ACRS review and to solicit public coment on several specific questions (Enclosure B). The staff was directed to seek public coment on an ACRS question concerning the indefinite grandfathering of plants with acceptable Appendix K models.
The staff was also instructed to submit three questions posed by Commissioner Asselstine for public coment. Namely:
- 1. Sh'ould this rule change include an explicit degree of conservatism that must be applied to the evaluation models?
- 2. Should this rule change explicitly prohibit any increase in approved power levels until all severe accident issues and unresolveo safety issues are resolved?
- 3. Should the technical basis for this proposed rule change be reviewed by an independent group; such as'the American Physical Society?
In accordance with the Commission's request, the ACRS reviewed the NRC methodology for ECCS code uncertainty evaluation on September 10, 1987 (Enclosure C). In addition, review of the NRC methodology by an independent group of experts was also conducted in January 1988. The sumary report of this independent panel, which was chaired by Professor Todreas from MIT, has been received and is also provided in Enclosure C. The recommendations of these two {
reviews have been considered by the staff for incorporation in the NRC uncertainty methodology.
The comment period for the proposed rule revision expired on July 1, 1987. Twenty-seven letters addressing the proposed rule change were received. In addition, six letters commenting on the proposed rule were received after the
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expiration of the coment period. These were also considered by the staff to the extent that new and substantial coments were provided. A list of the rule commenters, and a sumary of the coment letters are provided in Enclosure D. '
The bulk of the coments received on the rule came from the nuclear industry and were largely supportive of'the action proposed. These coments suggested a number of minor revisions to the rule, which are not discussed here, but can be found in Enclosure E. One negative coment was received '
from an anonymous commenter but specific objections were not' provided. . Two commenters recomended that the rule not be implemented until other safety issues are resolved. A detailed analysis of the public comments may be found in the Federal Register Notice (Enclosure E) including proposed-responses. to the public coments on questions posed by the ACRS and Commissioner- Asselstine, as well as several other
. comments that recommended major changes.
. . In conjunction with the publication of the proposed amendments to 10 CFR 50.46_and Appendix.K the NRC staff prepared a regulatory guide to set forth the staff's views ~-
concerning acceptable procedures for compliance. This regulatory guide, entitled,"Best Estimate Calculations of Emergency Core Cooling System Performance," was also released for public coment to assist the public in understanding the proposed revisions and to allow public participation in their development. This guide describes features that a realistic ECCS evaluation model should contain and guidance.on performing the uncertainty evaluation. The regulatory guide lists a number of models and corresponding experimental data
~ that are considered suitable for use. The comenters on the i regulatory guide largely supported the inclusion of !
experimental data that was acceptable and models and !
correlations that fit the data acceptably well. Therefore, these references have been included in the final guide with the statement that other models and correlations will be considered but must be justified with appropriate experimental data. This regulatory guide is provided as Enclosure F and a list of comenters on the guide and a paraphrased sumary of guide coments is provided in Enclosure G.
In order to apprise licensees and applicants of the large
[' body of research which supports realistic calculations of ECCS performance, the staff has prepared NUREG-1230,
" Compendium of ECCS Research.for Realistic LOCA Analysis." i This report identifies the relevt.nt ECCS research performed and describes the NRC oeveloped methodology for the estimation of the uncertainty of thermal-hydraulic safety analysis codes. The staff feels that this document will I provide valuable guidance to licensees and applicants wishing
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. 1 to avail themselves of the benefits of performing realistic LOCA analyses-to evaluate ECCS performance.
Based on the coments received and considered and the large j experimental data base available at this time, the staff recommends that the Commission approve for publication the final rule as contained in the Federal Register Notice (Enclosure E). This final rule does not alter the requirements found in the notice of proposed rulemaking (Enclosure A). A regulatory analysis of the final rule is 1
'provided in Enclosure I.
Resource The NRC staff resources to implement this rule are thought Estim_ates: to be negligible under the assumption that no unusual or special rulemaking procedures (e.g., adjudicatory hearings) will .be established by the' Comission. . If the Comission chooses to hold hearings, resources would have to be diverted
. from other high priority activities'.
. The major staff resources required under the rule change will be to review the realistic models and uncertainty analysis that may be submitted by licensees or applicants wishing to utilize the revised ECCS Rule. Based on previous experience with the General Electric Co. SAFER model and the learning that has resulted from this effort, it is estimated that approximately one staff year will be required to review each
. generic model submitted. There are four major reactor vendors (GE already has a revised evaluation model approved under the existing Appendix K for both jet-pump and non-jet pump plants that is consistent with SECY 83-472 and meets the requirements of the revised ECCS rule proposed herein) and several fuel suppliers and utilities which perform their own analyses and potentially might submit generic models for review. However, it is expected that only 3 or 4 generic models would be submitted since not all plants would benefit from the rule. Thus, approximately three to four staff years will be required to review the expected generic models. Once a generic model is approved, the plant specific review is expected to be very straightfomard.
Recommendations : That the Comission:
- 1. Approve the publication of final amendments, as set forth in Enclosure E, which would permit the use of realistic evaluation models and the accompanying estimate of the uncertainty of the calculation to demonstrate that the acceptance criteria contained in Sectinn (b) of 10 CFR 50.46 are not exceeded.
- _m
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.The Commissioners 6
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- 2. Note:
a.
That the notice of final rulemaking in Enclosure E will be published in the Federal Register to be effective 30 days after publication.
- b. That the regulatory guide in Enclosure F will be published concurrent with this rule.
- c. That pursuant to $5 51.21 and 51.31 of 10 CFR Part
' 51 of the Comission's regulations, an-environmental assessment and finding of no significant impact is attached as Enclosure H.
- d. That the reporting requirements in cornection with analyses required by the rule (Enclosure E) impose information collection requirements.that are
, subject to the Paperwork Reduction Act. The requirements were approved by OMB.
- e. That pursuant to the Regulatory Flexibility Act of 1980 the rule contains a statement that the Comission _ certifies that the rule will not, if promulgated, have a significant economic impact upon a substantial number of small entities and a copy of this certification will be forwarded to the Chief Counsel for Advocacy, SBA by the Division of Rules and Records, ADM.
- f. That the Subcommittee on Nuclear Regulation of the Senate Comittee on Environment and Public Works, the Subcommittee on Energy and the Environment of the House Comittee on Interior and Insular Affairs, the Subcommittee on Energy Conservation and Comerce, and the Subcommittee on Environment, Energy and Natural Resources of the House Committee on Government Operations will be informed.
- g. That a Regulatory Analysis is attached as Enclosure I.
- h. That a public announcement will be issued.
- i. That copies of the Notice of Final Rulemaking will be distributed by TIDC, ADM to each affecteo licensee and other interested parties.
- j. That the staff recommends the paper be placed in the PDR.
- k. That this paper has been reviewed with the ACRS Subcommittee on Thermal-Hydraulic Phenomena on April 21,1988, and they have indicated tnat they
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4 ThF Commissioners 7 have no objection to this rule. .The ACRS has indicated that it wishes to be apprised of the progress made on.the demonstration of the uncertainty methodology. The staff will continue to brief the ACRS until the demonstration is complete.
Scheduling: Recommend affirmation at an open meeting. No specific circumstance is known to the staff which would require Commission action by any particular date in the near future.
Victor Stello, Jr.
Executive Director for Operations
Enclosures:
A. Notice of Proposed Rulemaking B. Memorandum Chilk to Stello, dtd 1/9/87 C. ACRS and Peer Review Summaries D. Summary of Public-Comment E. Notice of Final Rulemaking F. Draft Regulatory Guide G. Summary of Public Comment on Guide H. Environmental Assessment I. Regulatory Analysis J. Margin Inherent in 2200*F Limit
NRC/ACRS M SHINGTON DC P.02
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'71 AM a NUCLEAR REGULATORY COMMIS810N ADVISORY COMMITTEE ON REACTOR SAFEOUARDe WASHINGTON, D, C. 20505 May 10, 1988 The Honorable Lando W. Zech, Jr.
Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Chairman Zech:
SUBJECT:
PROPOSED REVISION OF THE ECCS RULE CONTAINED IN 10 CFR 50.46 AND APPENDIX X During the 337th meeting of the Advisory Committee on Reactor Safe-guards, May 5-7, 1908, we met with members of the NRC Staff and reviewed the final version of the proposed revision to the emergency core cooling Our system (ECCS) rule contained in 10 CFR 50.46 and Appendix .K.
Subcommittee on Thermal Hydraulic Phenomena met on April 20,1988 to '
discuss this matter. We also had the benefit of discussions with the NRC Staff and of the documents referenced. The ACRS previously comment-ed on the proposal to issue this rule for public coment in a letter dated September 16, 1986.
The proposed revision to the ECCS rule will eliminate the requirement to use the models specified in Appendix K and allow use of realistic models combined with an uncertainty analysis of the overall calculation.
Certain criteria in 10 CFR 50.46, such as 2200*F peak cladding tempera-l ture e.nd 17% cladding oxidation, would be maintained. The regulatory guide which will acccmpany the revised rule describes features of a realistic evaluation model acceptable to the NRC Staff and contains guidance on performing the necessary associated uncertainty evaluation.
No changes have been proposed to the final rule version as a result of the public comments received. The regulatory guide has been modified somewhat ,o clarify tha NRC Staff's intent in certain areas.
The ACRS has long advocated use of best estimate or realistic evalua-tions for safety analysis. We believe the proposed rule We is awish major step to note forward in this effort, and we support its adoption.
the following points:
- Work to demonstrate the Code Scaling, Applicability, and Uncer-tainty (CSAU) method for the peak cladding temperature calculated to occur in the reflood phase of a large break LOCA has not been completed. This will be needed to establish guidelines While for Staff the review of future licenseereasonably submittalsdemonstrated under the newfor rule.
the so-called j- CSAU mathod has been y o- a m oo #
n ov2 Vl v ? ' "
. M A CRS m SHINGTON 1. P.03 The Honorable Lando W. Zech, Jr. May 10, 1988 blowdown peak, application to the reflood demonstration will be more cifficult. We do not object to plans to proceed with promulgation of the rule change, but we would like to be kept informed about the development of and allowance for uncertainty in the reflood peak temperature.
- We note that the draft Federal Register notice provided to support the rule change has eliminated reference to any claimed safety advantages for the rule. We believe the safety advantages are substantial.
Additional comments by ACRS Member Harold W. Lewis are presented below.
Sincerely, W. Kerr Chairman Additional Comments by ACR5 Member Harold W. Lewis I have no quarrel with the Committee's letter, but want to seize the opportunity to reinforce a point that has been made before. It is stimulated by unsatisfactory answers to questions at the presentation to the Committee.
The CSAU " methodology" purports to be a systematic procedure for esti-mating the uncertainty in code calculations. That is a laudable objec-tive, and its achievement would be even more laudable. It would be helpful if, in so doing, there were less confusion between the concepts of uncertainty and a probability distribution, and less misuse of the term " confidence limits." These objectives will not be reached unless In this case, it is of some professional statisticians become involved.
more than usual importance, since the uncertainty is directly.related to the acceptable level of conservatism which must be added to the realis-tic calculations. ..
References:
- 1. U.S. Nuclear Regulatory Comission, Draft SECY paper for the Commissioners from V. Stello, EDO, " Revision to the ECCS Rule contained in Appendix K and Section 50.46 of 10 CFR Part 50,"
provided to the ACRS, April 20, 1988.
- 2. U.S. Nuclear Regulatory Commission, Oraft NUREG-1230, " Compendium of ECCS Research for Realistic LOCA Analysis," Office of Nuclear Regulatory Research, dated April 1987.
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