ML20154M033
| ML20154M033 | |
| Person / Time | |
|---|---|
| Issue date: | 09/13/1988 |
| From: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| To: | |
| References | |
| FRN-52FR6334, RULE-PR-50 PR-880913, NUDOCS 8809270038 | |
| Download: ML20154M033 (35) | |
Text
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DOCKET NUMBER
, PROPOSED RULE SO Ni octstter w.n ,
(SRPA033y)1)
[7590-0 l
'88 SEP 13 All :46 l NUCLEAR REGULATORY COMilSSION
[0Ch. '. 'c 10 CFR Part 50 Emergency Core Cooling Systems; Revisions to Acceptance Criteria AGENCY: Nucle.r Regulatory Commission.
ACTION: Final rule. [
SUf94ARY: The Nuclear Regulatory Comission (NRC) is amending its '
regulations to allow the use of alternative methods to demonstrate that the emergency core coolino system (ECCS) would protect the i nuclear reactor core during a postulated design br. sis loss-of-coolant accident (LOCA). The Commission is taking this action because i research, pe? formed since the current rule was written, has shown that I calculations performed using current methods and in accordance with ;
the current requirements result in estimates of cooling system f performance that are significantly more conservative than estimates !
I based on the improved knowledge gained from this research. While the existing methods are conservative, they do not result in accurate l calculation of what would actual)y or. cur in a nuclear power plant l during a LOCA and may result in less than optimal ECCS design and ;
operating procedures. In addition, the operation of some nuclear reactors is being unnecessarily restricted by the rule, resulting in l increased costs of electricity generation. This rule, while '"
, l continuing to allow the use of current methods and requirements, also allows the use of more recent information and knowledge to demonstrate that the ECCS would protect the reactor during a LOCA. This amendment, which applies to all applicants for and holders of l
construction permits or operating licenses for light water reactors, also relaxes requirements for certain reporting and reanalyses which do not contribute to safety. :
8809270038 800913
[52b6334 PDR Page 1 D 3/ 0
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EFFECTIVE DATE: October 17, 1988 FOR FURTHER INFORMATION CONTACT: L. M. Shotkin, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington, DC 20555, telephone (301) 492-3530.
SUPPLEMENTARY INFORMATION:
BACKGROUND On March 3, 1987, the Nuclear Regulatory Commission published in the Federal Register proposed amendmenta (52 FR 6334) to 10 CFR Part 50 and Appendix K. Thest proposed amendments were motivated by the fact that since the proinulgation 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 Fart 50, considerable 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 Appendix K, combined with other analysis :
methods currently in use, are highly conservative and that the actual l claddin; temperatures which would occur during a LOCA would be much
- lower than those calculated using Appendix K methods. In soliciting I
the public's coments on the proposed rule, the NRC specifically ,
requested its views on questions posed by Comissioner Asselstine and j the Mvisory Comittee on Reactor Safeguards (ACRS). The ACRS j requested that the Comission solicit the publir's coments on wr. ether the existing rule should be "grandfathered" indefinitely. That is: i
- 1. Should the conservative ECCS evaluttion method of Appendix K be permitted indefinitely or should this aspect of the ECCS rule be phased out after some period of time?
Comissioner Asselstine requested the public's coments on tne following: )
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- 2. Should this rule change ine'aude an explicit degree of {
l conservatism that must be applied to the evaluation models? !
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- 3. This rule change would allow a 5 to 10 percent increase in i the fission product inventory that could be released from i any core meltdown scenario. Should this rule change
- f j explicitly prohibit any increase in approved power levels l until all severe accident issues and unresolved safety i
! issues are resolved? ;
i' l
- 4. Should the technical basis for this proposed rule change be L
l reviewed by an independent group such as the American i
- Physical Society? [
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j $UMARY 0F PUBLIC Com ENTS l I
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The coment period for the proposed rule revision and the draft l
) regulatory guide (52 FR 11385) expired en July 1, 1987. Twenty-seven l l 1etters addressing the proposed rule were received by the expiration I date, as well as nine responses to the request for coments on l ;
j questions in the regulatory guide. A number of late coments were i I also received. These were also considered to the extent that new and l substantial coments were provided.
4 i I l The public coment on the proposed rule revisions have been l divided into thirteen categorics and are sumarized in the following l l I paragraphs. Catego*ies one through four represent the resp,nses to l the specific questions posed by the ACRS and Comissionr: Asselstine.
In general, consideration of the public coments resulted in no I substantive revision to the proposed rule.
i l
} I j 1. Grandfathering of Conservative ECCS Methods of Appendix X l
l Question 1). l l Twenty-one of the comenters specifically addressed the ARCS i question concerning the grandfathering of the current Appendix K a
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- 9. l 1 i j' approach. Seventeen of these comenters recomended indefinite !
- grandfathering of the existing Appendix K evaluation models. 3 l Most cited the known conservatism as the basis of their i recomendation. In addition, several commenters stated that in ,
i light of the known conservatisms not allowing continued use of l i
existing Appendix K evaluation models would be unfairly s
burdensome to licensees who determine that they would not derive I an economic benefit by performing realistic analysis of ECCS l performance. The position of an additional comenter is unclear !
! concerning grandfathering. The remaining comenter was not !
opposed to grandfathering but thought the question is premature, f This commenter believes that indefinite use of existing ECCS j evaluation methods should be considered when significant l j 9perience has been gainad with the implementation of the new l 9 features of the rule but makes no recomendation as to what !
policy the Comission should pursue in the meantime. [
l The Comission agrees with the majority of the comenters that I existing Appendix K evaluation models should be permitted I indefinitely. The Comission also believes that the decision to l permit continued use of such models can and should be made at !
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this time because it believes that both methods provide adequate protection of the public health and safety. As described in the l l regulatory analysis, the probability of a large break is so low, !
l that the choice of best estimate versus Appendix K has little !
l effect on public risk. The TMI action plan calls for industry to f l improve their small break LOCA evaluation models to be more [
realistic when evaluating the more probable small break accident
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j scenario. This has been done within the context of I 50.46 and j
! Appendix K compliance and was entirely appropriate since small l l breaks are not limiting in design basis performance and a better i l understanding of small break behavior is a desirable safety goal I l l from a risk perspective. Therefore, the grandfathering provision !
l has been retained in the final rule. I J
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- 2. Specification of Exalicit Dearea of Conservatism (Question 2).
i The majority of the responses to this question indicated that the ,
I proposed rule already contains conservatism in the required !
uncertainty evaluation.
I i The use of additional conservatism would be inconsistent with the ;
4 objective of the rule which is to provide a realistic evaluation l l of plant response during a LOCA. The NRC has not included an l 1 additional explicit degree of conservatism in this rule. l
] 3. Resolution of all Safety Issues Prior to Allowine power Level I l Increates (Question 3). Some commenters pointed out that f j fission product inventory is not a direct function of total [
j power, but rather it is the rate of fission product formation l
] that is a direct function of power. Fission producc inventory i i available for release during a core meltdown would be a function f f of burnup, not total' power, f
I i
l Actually, the inventory of fission products is a complex function of both time and power and not as simple as described by the I l
! comenters. Short lived isotopes, such as xenon and iodine, quickly reach an equilibrium inveatory and tot.a1 steady state ;
inventory of these fission producM is a direct function of l power. Inventories of long 1tved isotopes, such as strontium and '
cesium, are functions of total fuel burnup, as described by the comenters. Intermediate-lived isotopic inventaries are complex i functions of time, power, and integrated power. In an f
independent study, documented in chapter XII of NOREG 1230, the (
staff determined that the change in risk due to a 5% power l l increase is negligible. The arguments above do not alter the I I Comission's position that the increase in fission products l available for release during a core meltdown caused by a 5% power l increase is negligible compared to the uncertainty in fission product release. The Comission has decided not to delay the l l
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'f proposed rule revision pending resolucion of all unresolved safety issues or severe accident issues and therefore will l proceed with this final rulemaking, as planned. ,
) 4. Independent review of Technical Basis (Question 41. Several ,
l comenters indicated that the technical basis for the proposed t i rule has had adecJate review as the research was being performed.
A number of comen;ers stated that it was the role of the ACR$ to j perform any review of the proposed rule revision because it is f l uniquely qualified due to its familiarity with the research. :
i l !
! The Comission 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 j unedrtainty evaluation of their SAFER code. As a proof of i principle and demonstration of feasibility, the ACRS and a second I f independent peer group has reviewed the uncertainty methodology i developed by the NRC for use in quantifying the uncertainty of NRC developed thermal hydraulic transient codes. Both the ACRS and the peer group made generally favorable coments concerning j the methodology; however, both groups recognized that a complete l l demonstration (i.e., application to small break LOCA and the I reflood portion of large break LOCA) has not , vet been ,
accomplishtd and certain reviewers questioned whether such a i
! demonstration could be performed successfully. The only !
objectives of the NRC methodology demonstration are to l
demonstrate feasibility, to develop an audit tool, and to provide j the necessary experience to audit licensee submittals. The staff
! does not believe that an NRC demonstration of the methodology is
! a prerequisite to this rulemaking. Licensees wishing to adopt ;
l
] the best estimate approach permitted as a result of this rule are j neither required to use this m thodology nor to model their own ,
l methodologies af ter it. This methodology will play an important f l part in the best-estimate modei review process. The NRC has l l determined through twenty years of experience that independent i j analysis with independent methodologies is the most effective way f
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! to intelligently review new vendor or licensee methodologies. It :
- is therefore appropriate that this new methodology be subjected i to stringent technical scrutiny, as directed by the Comission, j The NRC staff is comitted to coe$1eting this demonstration by j
{ the time that it will be needed to review licensee submittals and ;
) is confident that such a demonstration will be successful. Based f i on the paucity of negative response concerning the technical j i basis for the proposed rule revision and generally favorable j review of the NRC uncertainty methodology, the Comission plans !
no further review of the technical basis. (
l I i
4 5. General Coments on proposed Rul_m. Twenty-one comenters made f coments of this nature. The majority of the coments came from f l j the nuclear industry of which 19 expressed support of the j j proposed rule. The industry also strongly supports the. specific ;
} ECCS rule approach proposed by the NRC. One comenter neither i supported nor opposed the proposed approach. One negative-l 3
coment was received from an anonymous f adividual within the j nuclear industry who implied, without specifics, that the ECCS l rule is not sound and that public coment is not a fair hearing i
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j because expert insiders would be afraid to comment.
l l Based on the absence of any supporting justification for the f l negative response and the unprecedented amount of research l
! supporting the rule revision, the NRC does not consider this j coment to be valid and has proceeded with this rulemaking with t l no major revisions. l 1 I One comenter suggested that fuel reload suppliers should not be l required to complete full LOCA/ECCS analyses because the
! hydraulics are not changed by a fuel change.
i Although this point is valid, the Comission 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 casms where j i
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1 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 shared i
responsibility for evaluation models would not be in the best j l interest of the public health and safety and therefore has not t
) implemented the suggestion of this commenter.
i The NRC received two requests for an extension of the comment l
4 period to allow time for review of NUREG-1230, which describes
! the research supporting the proposed rule revision. l i I' J The NRC believes the coment period was sufficient since most of
- the research is not new and has been extensively reviewed in the ;
past. Both comenters were contacted and told that coments l l received after the coment period would be considered if time ,
permitted. Coments from both parties were received late and
) f
- were indeed considered by the NRC. j l
i l Some comenters viewed the proposed
- 6. Rooortino Reautrements.
reporting procedures as new requirements needing consideration in l the backfit analysis while others stated that they are a major j j relaxation and clarification of existing reporting requirements. ,
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! The NRC position is that the reporting requirements are new in ;
l the sense that they will now appear in the Code of Federal f I J j Regulations. However, in practice, these reporting requirements l are indeed a clarification and relaxation over the current l
- inter,retation of th. .xisting requir.m.nts and th.r. fore the not l
j effect of these requirements will be to reduce the frequency for l l reporting and reanalysis. !
I !
! A number of comenters requested that ontj significant errors or l changes in the non-conservative direction or only those that (
} result in exceeding the 2200'F limit be required to be reported. !
In addition, a number of comenters suggested that the NRC l require only annual reporting of significant errors or changes. I l -
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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 arrors or changes, in either direction, on a timely basis so that the Comission may make a determination of the safety significance. Thus, the final rule contains no change in this requirement.
One comenter recomended that the word "imediate" be deleted from the requirement to propose steps to be taken to demonstrate compliance in the event that the criteria in $ 50.46(b) are i exceeded.
The Comission 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 "imediate" reflects this seriousness and is further defined by reference in other sections of part 50.
l j
Several comenters questioned the need to report minor or
- inconsequential errors or changes, even on an annual basis, as
] reauired in the proposed rule.
) While errors or changes which result in changes in calculated !
l peak clad temperatures of less than 50'F are not considered to be ;
l of imediate concern, the NRC requires cognizance of such changes l or corrections since they constitute a deviation from what previously has been reviewed and accepted. The proposed annual reporting is believed to be a fair compromise between the burden I of reporting and the Comission's need to be aware of changes and j error corrections being made to evaluation models. Therefore, ;
the annual reporting of minor errors remains in the final rule.
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- i One comenter interpreted the use of the words "or in the application of such a model" as requiring reporting when facility e
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changes (already reportable under i 50.59), resulting in model input changes, occur.
The regulatory language referred to is intended to ensure that ;
l applications of models to areas not contemplated during initial !
review of the model do not result in errors by extending a model ,
beyond the range that it was intended. The Comission does not beliese that further clarification of this requirement is ;
necessary and has not done so in the final rule.
i Several commenters requested a further relaxation of the i reporting requirement by changing the definition of significant i code errors from 60'F to 200'F.
i l While justification for the 50'F criteria is largely judgmental,
) the NRC believes that it is sufficiently large to screen the code error corrections and changes which have little safety l
significance while providing a mechanism for timely reporting of more serious errors and changes. Since 50'F is a threshold for ;
- reporting and no further action is required pending NRC l determination of safety significance, the Comission has rotatiad
< this criteria in the finsi rule, q
One commenter requested consideration for allowing that the r
- cumulative effect of several errors and corrections be applied towards the 50'F threshold.
l The requirement, which states that the 50'F criteria applies to ,
] the sum of the absolute magnitudes of temperature changes from j numerous error corrections or model changes was formulated j specifically because the Comission requires knowledge of serious ,
l deficiencies in evaluation models in use by licensees. !
3 Allowing errors or corrections which offset one another to
- relieve a licensee of the thirty-day reporting requiremer.t would ,
! be counter to this objective. If this recomendation were accepted, two errors or changes, hwing a large impact on the
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. 1 calculated peak cladding temperature but in the opposite direction, would not be reportable if the not 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 Comission retained this requirement in the final rule.
- 7. Continued Use of Dougall-Rohsenow. Fiva coments that addressed
< this aspect of the proposed rule were received. One comenter I believed that this correlation should not be permitted without i further verification and should be phased out. Other comenters supported continued use of the correlation subject to the l provisions of the proposed rule. ,
i The NRC position is d at no safety concern is created by l
continued use of the correlation, as long as the evaluation model j is overall conservative. Therefore, the Comission can not justify the burden of requiring licensees to mcdtfy their i j evaluation models and to perform reanalysis. As discussed in SECY 83-472, current e uluation models contain more conservatisms than just those required by Appendix K. However, error corrections or changes could alter the conservatism of the model. ,
j Therefore, the Comission believes that it is necessary to ensure j continued overall conservatism in the evaluation models as a
) basis for continued use of the correlation. Therefore, the final i rule does not modify this requirement except for the correction of a typographical error identified by one connenter.
) 8. Uncertainty Evaluation. The coments received on the uncertainty :
I evaluation support the proposed rule, particularly the l j flexibility provided by a non-prescriptive requirement. ;
j Therefore, the Comission is publishing the final rule without modification of this requirement, i
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90 Acceptance Criteria. The three coments received on this topic '
were all supportive of the existing criteria, as contained in .
I 50.46(b), and thus the Comission did not give consideration to l altering them in the final rule.
1 j, 10. Claddina Materials. Three commenters requested that the Comission consider broadening the language of the rule to allow {
the use of a range of zirconium based alloys for clariding ;
material.
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- The Comission believes that this modification is beyond the ;
) scope of the current rule revision and should be considered in a ;
separate ruleslaking action in which it would receive appropriate ;
public review and coment prior to implementation. In addition, j zircaloy cladding material is specified in other sortions of the :
I Code of Federal Regulation, such as i 50.44. Making a change of f this type is more suitable in a broader regulatory context. f j Therefore, the Comission is not broadening the definition of !
cladding materials within this rulemaking, j
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f j 11. Other Sunnested Expansiorp M vp fiop,3,. One comenter believes i l that hydraulic loads occurring during a LeCA could cause steam generator tubes to rupture and that the NRC should resolve steam f
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l generator tube integrity safety issecs prior to pub'lishing this f
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rule. [
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j Steam generator tubes are designed to withstand LOCA loads at l
j allowed thinning, and there is no evidence to contradic' this. l If anything, the problem would be with inspection techniques to l I detect the actual tube thinning and whether there is an i 1
1 j unacceptably high probability that a tube rupture during a LOCA r l due to tube thinning is in excess of the design basis. However, I l the risk from LOCA with concurrent tube rupture will not be greatly affected by the proposed rule change. As a result of the i 1
l comenter's concerns, this issue has been assigned as a generic !
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- I I issue (GI-141) to be prioritized by the NRC staffo The !
results of the prioritization process will determine if further ;
i* action is required. '
i A second commenter believes that the ECCS rule does not l
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adequately address a plant's long term decay heat removal l
} capability, and recomends a "short/long term integrative i 1
analysis approach." Both the existing requirements and the ;
l propcsed rule contain the requirement to provide for long ters ;
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 j l or transient. Thus, the issue of decay heat removal is not i
- materially impacted by this rulemaking. Moreover, any proposed increase in power resulting from this rule promulgation would be j approved only after the Itcensee demonstrates that decay heat i
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j removal capacities remain adequate. The Coenission is planning no further action with regard to this issue. ,
j! I
) 12. Acceptability of Models Accroved Under SECY-83-472. One comenter requests that the rule language be modified to state -
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explicitly that ECCS evaluation models that have been previously approved under SECY-83-472 continue to be acceptable under this j rule. !
l !
- i q SECY 83-472 provides an alternative, acceptable method for i
- developing ECCS evaluation models. Licensees were still l required, however, to demonstrate that evaluation models f l '
developed using the SECY-83-472 approach complied with the 3
l requirements of Appendix K to Part 50. This final rule !
l 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 further clarification of this issue.
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i 13. Coments Received Af ter Coment Period. Six letters comenting
- on the proposed rule were received subsequent to the end of the coment period. The Comission considered these coments to the extent that the coments orovided substantive information not oreviously considered.
One comenter believes that the proposed i 50.46(a)(2) expands
! the discretion of the Director of the Office of Nuclear Reactor
! Regulation (NRR) by allowing imposition of imediate effective i
restrictions on reactor operation without a prior determinatica j that such action is required to protect the pubile health,
! safety, or interest. NRC's intent is not to alter the
! responsibilities of the Director of NRR but to simply retain the ;
1 description of the scope of the authority that is currently found f in i 50.46(a)(1)(v). Furthermore, the provisions of
} l 50.46(a)(2) do not specify the procedure to be followed by the l Director of NRR. These procedures are set out in Part 2 and l
! remain unchanged by this rulemaking. I
! One comenter believes that the rule is illegal because it is based solely on cost savings considerations and that there is nothing wrong with large conservatisms. !
l
- The Comission disagrees with this assessment. Safety factors
) are required to protect the health and safety of the pubite when i uncertainties in plant response exist. As these uncertainties are reduced, it is appropriate to modify these safety factors to i provide more realistic evaluation of actual plant response. The 4 1arge conservatisms of Appendix K served the public well in 1974 when there was great uncertainty in ECCS performance. However.
l these conservatisms are now known to be very large, and there is j j no need to "over regulate" by maintaining this unnecessary 1 margin. This type of activity can often result in the ;
expenditure of resources that would be better spent improving f
- safety in other areas. The benefits to safety, while difficult l to quantify, are believed to be substantial. While cost savings j
may have been one factor resulting in the rule change, the 1
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Comission believes that the conservatisms 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 comenter also cites portions of the 1975 General Electric Company's Nuclear Reactor Study (Reed Report), which claims that j there is a lack of understanding of phenomena and small safety margins. ,
t 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 l 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 "Road Report." The contents of this report have been reviewed by the Comission on several occasions, most recently in NUREG-1285, and the finding [
has been made that no row significant safety issues are identified. For these reasons, the NRC is proceeding with this rulemaking, as proposed. !
The same comenter also recommends that credit for ECCS .s ins be taken in the Individual Plant Examinations (IPE) and i.ot i through generic rulemaking.
The Commission agrees that plant specific differences may justify l the application of different margins and that these may be addressed through Individual Plant Examinations. However, the ;
requirement for Itcensees to evaluate ECCS performance and meet !
the acceptance criteria specified in 10 CFR 50.46(b) is generic.
The Comission 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 o**.:ric rulemaking action because it allows a broad range of technical review of the issues, enhances public ;
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particiostion in the process, and provides a complete public
. record. Tnertfore, the Comission has decided to proceed with the rulemakin) as planned.
- r Finally, this commenter questions the experimental basis for this 4
rule because full-scale ECCS bypass data is not yet available.
The 20/30 tests which will provide this important data represent l a small portion of the total research upon which this rule relies. Significant research on ECCS 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 I any uncertainty due to scale effects. One full-scale ECCS bypass l test has already been completed under the 20/3D program which showed that more margin exists than expected from the small scale 4
tests. Cwpletion of the full-scale tests only affects the
! uncertainties in the calculations, and reduces them.
- Uncertainties must be addressed by licensees in any analysis under the revised rule whether 20/30 results are available or
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not. The Comission concludes that there is no need to delay the final rule, while awaiting these data,
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! SUtHARY OF RULE CHANGES I
t Section 50.46 Acceptance Criteria for Emergency Core Cooling Systems ,
for Light Water Reactors Section 50.46(a)(1) is amended and redesignated i 50.46(a)(1)(1) to delete the requirement that the features of Section ! of Appendix K to Part 50 be used to develop the evaluation model. This section new requires that an acceptable evaluation model have sufficient l
- supporting justification to show that the analytical technique realistically Jescribes the behavior of the reactor system during a 3
LOCA. The NRC expects that the analytical technique will, to the
- extent practicable, utilize realistic methods and be based upon l
applicable experimental data. The amended rule also requires that the l
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uncertainty of the calsulation be estimated and accounted for when comparing the results of the calculation to the temperature limits and ,
other criteria of l 50.46(b) so that there is a high probability that the criteria would not be exceeded. The Comission 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 has been provided in the form of a Regulatory GuideI . In SECY-83 472, the NRC has found acceptable an approach for estimating the 95th percantile of the I
l probability distribution. Thir,percential is considered adequate to meet the high level of probability required by the rule. It is also l recognized that the probability cannot be determined using totally l rigorous mathematical methods due to the complexity of the
- calculations. However, the NRC requires that any simplifying l assumptions be stated so that the Commission may evaluate them te ensure that they are reasonable. The NRC has independently developed I and exercised i 'Aethodology to estimate the uncertainty associated !
with its owr. thermal-hydraulic safety codes. This methodology is l I described in the "Compendium of ECCS Research. 2 This document also provides reference to the large body of relevant thermal hydraulic research, documents NRC studies on the effects of reactor power 1 increases on risk, and provides background information on the ECCS rule. While this method has not been reviewed for acceptability from
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the standpoint of safety licensing, it may provide additional guidance on how the uncertednty may be quantified, In addition to providing guidance to industry, this work was undertaken to provide a proof of l principle and a tool to independently audit industry sut'mittals.
Appendix K. Section !!, "Required Document,ation," remains generally applicable, with only minor revisions made to be consistent with the l j amended rule. !
1 1
Regulatory Guide, "Best Estimate Calculations of Emergency Core Cooling Systems Performance," RG 1.157, 2"Compendium of ECCS Research for Realistic LOCA Analysis," (
NUREG-1230. TBP. j i i i !
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A new paragraph (ii) has been added to i 50.46(a)(1) to allow the features of Section I of Appendix K to be used in evaluation models as l an alternative to performing the uncertainty evaluation specified in the amended i 50.46(a)(1)(1). This method would remain acceptable !
because Appendix K is conservative with respect to the realistic method proposed in the amended i 50.46(a)(1)(1). This would allow both current and future applicants and licensees to use existing l evaluation models if they did not need or desire relief from current operating restrictions. ,
t In i !0.46, 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 described 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 f Nuclear Reactor Regulation, if the ECCS cooling performance !
evaluations are not consistent with the requirements of f ll 00.46(a)(1)(1) and (ii). This section has been added to retain l similar requirements that have been deleted from i 50.46(a)(1)(i) by j this rule revision. This section does not specify the procedures to l be followed by the Director. These procedures are found in Part 2 and j are unchanged by this rulemaking. !
L i
The current rule contains no explicit requirements concerning re-
{
porting and reanalysis when errors in evaluation models are discovered !
or chanbes are made to evaluation models. However, current practice I has required reporting of errors and changes and reanalyses with the I revised evaluation models. This fina) rule explicitly sets forth requirements to be followed in the event of errors or changes. The f definition of a significant change is currently taken from Appendix K, l Section !!.1.b which defines a significant change as one which changes !
calculated cladding temperature by more than 20'F. I Page 18
i l
)- !
) The revised i 50.46(a)(3) states specific requirements for reporting and reanalyses when errors in evaluation models are j discovered or changes are made to evaluation models. It requires thn.. ;
l j all changes or errors in approved evaluation models be reported at l
- least annually and does not require any further action by the licensee i until the error is reported. Thereafter, although reanalysis is not 1 required solely because of such minor error, any subsequent calculned i evalustien of ECC5 performance requires use of a model with such l
- error, and any ,rier errors, corrected. rhe N C needs to e a,pri.ed ;
of even miner errors or changes in order to ensure that th6y agree [
l j with the applicant's or licensee's assessment of the significance of l l the error or change and to maintain cognizance of modifications made i subsequent to NRC review of the evaluation model. Past experience has f l
shown that many errors or changes to evaluation models are very minor i and the burden of imediate reporting cannot be justified for these !
minor errors because they do not affect the immediate safety or operation of the plant. The NRC therefore requires periodic reporting l to satisfy NRC's need to be apprised of changes or errors without f imposing an unnecessary burden on the applicant or licensee. This {
report is to be filed within one year of discovery of the error and l j must be reported each year thereafter untti a revised evaluation model {
j or a revised evaluation correcting minor errors is approved by the NRC l
staff. l l'
t l
significant errors require more timely attention since they may j
- be important to the safe o;,eration of the plant and raise questions as I
! to the adequacy of the overall evaluation model. This final rule d
defines a significant error or change as one which results in a !
i calculated peak fuel cladding temperature different by more than 50'F. ,
or an accumulation of errors and changes such that the sum of the j absolute magnitude of the temperature changes is greater than 50'F. !
More timely reporting (30 days) is required for significant errors or changes. This definition of a significant change is based on NRC's ;
judgement concernine the importance of errors and changes typically l reported to the hRC in the past. This final rule revision also allows )
the NRC to determine the schedule for reanalysis based on the l page 19
._._U
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 l 50.46(b) mean that the plant is not operating within the requirements of the regulations and require inrnediate reporting as required by i 50.55(e), f 50.72 and i 50.73 and immediate steps to bring the plant into co;npliance with i 50.46.
Appendix X ECCS Evaluation Models Amendments have been made to Appendix K,Section I.C.5.b, to modity the post-CHF heat transfer correlations listad as acceptable.
The "McDonough" reference has been repisced with a more recent paper by the same authors entitled "An Experimental Study of Partial Film doiling Region With Water at Elevated Pressures in a Round Vertical Tube" which is more generally available and which includes additional data.
The heat transfer correlation of Dougall and Ro'nsenow, listed as an acceptable heat transfer correlation in Appendix K, paragraph 1.C.5.b, has been removed, because research performe~ since Apoendix K ras written has shown that this correlation overpredicts heat transfer coefficients under certain ccaditions and therefore can produce nonconservative results. A number of applic. ants and licensees currently use the Dougall-Rohsenow correlation in approved evaluwtion 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 ltrge 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, requ'. ring that the applicants and licensees remove the ,
Dougall-Rohsenow correlation from their current evaluation models l cannot be justified as necetsary to maintain safety. The stipulation 1
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_ _ , - . ,. - _ . __l
that the Dougall-Rohsenow correlation will cease to be acceptable for
. previously approved evaluation models applies only when changes to the model are made which reduce the calculated peak clad temperature by 50 F or more. However, the requirement to report any changes or culmination of changes, such that the sum of the absolute magnitudes 0
of the respective temperature changes is greater than 50 F, still applies.
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 aie made to the evaluation model which significantly reduce the current overall conservatism of the evaluation model. If the applicant or licensee submits proposed changes to an approved evaluation model, or submits corrections to errors in the evaluation model which significtntly reduce the existing overall conservatism of the model, continued use of the Dougall-Rohsenow correlation under conditions where noncenservative heat transfer coefficients result would no longer be acceptable. For this purpose, a significant reduction in ovto all conservatism has been defined as a "net" reduction in calculated peak clad temperature of at least 50 F0 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 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 requiremont has been clarified in the amended i 50.46(a)(3),
t.ikewise, Appendix K. Section !!.5, has been amended to account for 1
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j 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."
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, VA 22161. A copy is also available for public inspection and/or copying at the NRC Public Document Room, 2120 L 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 inspottion and copying at the NRC Public Documents Room,2120 L Street NW., Washington, DC 20555. Single copies of this report may be obtained by writing L. M. Shotkin, Office of Nuclear Regualtory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555,
- 3. Regulatory Guide, "Best Estimate Calculations of Emergency Core Cooling Systems Performance," Task RS 701-4, may be obtained by writing to the Division of Information Support Services, U.S.
Nuclear Regulatory Comission, Washington, DC 20555,
- 4. The Paraphrased Sumary of Public Comments on the ECCS Rule is available for public inspection at the NRC Public Documents Room, 2120 L Street NW., Washington, DC 20555.
FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY The Comission has determined under the National Environmental j Policy Act of 1969, as amended, and the Commission's regulations in ]
l Page 22
Subpart A of it 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 effect 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, but is expected to be increased by no more than about 5% due to hardware limitations in existing plants. This 5% 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 w,th 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 administered by the U.S. Environmental Protection Agency (EPA) or designated state agencies. It is not intended that NRC approval of increased pcwer 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 inspection at th6 NRC Public Document Room. 2120 L Street NW, Washington, DC. Single egies 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, telephone (301) 492-3530.
PAPERWORK REDUCTION ACT STATEMENT This final rule amends information collection requirements that are 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).
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REGULATORY ANALYSIS The Conunission has prepared a regulatory analysis for this final regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The regulatory analysis is available for inspection and copying for a fee at the NRC Public Document Room, 2120 L 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) 492-3530.
REGULATORY FLEXIBILITY CERTIFICATION As rtquired by the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this rule will not have a significant economic impact upon a substantial number of small entities. This rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration 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 A backfit analysis is not required be, 10 CFR 50.109 because the rule does not require applicants or Itcensees to make a change but only offers additional options and p+ovides a clarification and relaxation of existing reporting requirements. Nonetheless, the factors in 10 CFR 50.109(c) have been analyzed for th" entire rule.
- 1. Statement of the specific objectives that the backfit is designed to achieve.
I 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, bab i on the l
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improved knowledge gained from recent research on ECCS performance, may remove unnecessary operating restrictions. Also experience with the previous version of f 50.46 has demonstrated that a clearer definition of reporting requirements for changes and errors is very desirable.
- 2. General description of the activity that would be required by the licensee or applicant in order to complete the backfit.
The amendment allows alternative methods to be used to deru istrate that the ECCS would prctect the nuclear reactor cors 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 applicant or licensee elects to use a new realittic model they j will be required to prov . sufficient supporting justification to validate the model and include comparisons to experimental j data and estimates of uncertainty. In accounting for the l uncertainty, the analysis would have to show, with a high level of probability, that the ECCS performance criteria are not exceeded. Whether or not a licensee or applicant chooses to use realistic analysis, complete with an uncertainty analysis, each licensee must comply with the requirement to report changes to their evaluation models (i.e., less than 50'F change in calculated peak cladding temperature) annually to the NRC. In addition, significant changes (those which have a greater than 500 F change in calculated peak cladding temperature) have to be ,
reported within 30 days. !
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- 3. potential change in risk to the public from the accidental
- offsite release of radioactive materials.
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l The rule could result in increased local power within the !
reactor core and possibly increases in total power. Power increases on the order of 5 will have an insignificant effect on I
Page 25
l, 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 ,7roducts 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 .
l margin between the PCT and the 2200'F limit than previously. The i 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. i
- 4. potential impact on radiological exposure to facility employees.
1 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 I impact on the radiological exposure to facility employees.
Because of the reduced LOCA restrictions resulting from the new l 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 l l reloads can be reduced. This effect is not expected to be very significant, 1
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, - - - - - . , w, y-,,,,-y-._ , - . -, - _,,,m,.,,,, ---. _ ~ .emm, , _ , . ~ , . _ _ _ ,__.m ,,__,,.,_..-_,m.._,-y.__,___. -.,. ,._,.,,,, .
- 5. Installation 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 activitics.
The effect of the rule 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 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.
The costs associated with the new reporting requirements are deemed to be minimal. Although the existing Appendix K has no official reporting requirements, paragraph II.1.b was interpreted by the staff to require a reanalysis and report to NRC when significant changes are made which change the peak cladding l
temperature by more then 20 F. Therefore, this rule change, by 1 changing the definition of significant changes to 50'F, is actually a relaxation of current, practices. The annual reporting 1 of changes that are not significant is not viewed by the NRC as a niajor burden since no other action is required, i
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- 6. The potential safety impact of changes in plant or operational complexity including the effect on other proposed and existing regulatory recuirements.
There are safety benefits derivable from alternative fuel management schemes that could be utilized. The higher power peaking factors that would be allowed 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. i The reduced cladding temperatures that would ne 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 reduction of ECCS system capability in existing or new plants is present. 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 l 1
little effect on public risk. j i
- 7. The estimated resource burden on the NRC associated with the proposed 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 i
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General Electric Company's SAFER model and the learning that has resulted from these efforts, it is estimated that approximately 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-jet pump plants and may update their methodology under this now rule) and several fuel suppliers and utilities which perform their own analyses and potentially might submit generic models for re<iew. 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 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 sumary, 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.
General Electric Company (GE) plants do tend to be limited in operation by LOCA restrictions and would benefit from relief from i 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 (M) plants would appear to ,
directly benefit from relaxation of LOCA limits. M plants l represent the largest number of plants, with 47 plants operating l
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and 10 additional plants being constructedo 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 .iustification for imposing the proposed 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 cooling calculations using either realistic models or models in accord with Appendix K.
LIST OF SUBJECTS IN 10 CFR PART 50 l
Antitrust, Classified information, Fire prevention, Incorporation l by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, 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 Energy l
Reorganization Act of 1974, as amended, and 5 U.S.C 552 and 553, the NRC is adopcing the following amendments to 10 CFR Part 50.
PART 50-00MESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES
- 1. The authority citation for Part 50 continues to read as follows:
AUTHORITY: Sees. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938, 948, 953, S54, 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).
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Section 50.7 also. issued under Puba L.95-601, seco 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, i
. 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). Section.s 50.23, 50.35, j 10.55, 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. 97-4.15, 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), $$ 50.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54, and 50.80(a) are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); $5 50.10(b) and (c) and 50.54 are issued under sec. 1611, 68 Stat. 949, as amended (42 U.S.C. 2201(i)); and $$ 50.9, 50.55(e), 50.59(b), 50.70, 50.71, 50.72, 50.73, and 50.78 are issued under sec. 161o, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).
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- 2. In i 50.46, paragraph (a) is revised to read as follows:
$ 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.
(a)(1)(i) 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 Page 31
performance must be esiculated in acco-dance with an acceptable evaluation model and must be calculated for a number of postulated 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. Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must include sufficient supporting justification to show that the analytical technique 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, Required Documentation, sets forth the documentation requirements for each evaluation model. l (ii) Alternatively, an ECCS evaluation model may be developed in conformance with the required and acceptable features of Appendix K ECCS Evaluation Models, t
(2) The Director of Nuclear Reactor Regulation may impose restrictions on reactor operation if it is found that the evaluations of ECCS cooling performance submitted are not consistent with paragraphs (a)(1)(i) and (ii) of this section.
l (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 such !
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 l calculated peak fuel cladding temperature different by more than 50'F from the temperature calculated for the limiting transient using the Page 32 l
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O last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 500F.
(ii) For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in i 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 sch6dule for providing a reanalysis or taking other action as may be needed to show compliance with i 50.46 requirements. This schedule may be developed using an integrated scheduling system previously approved 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 proposed schedule. Any change or error correction that results in a calculated ECCS performy,nce that does not conform to the criteria set forth in paragraph (b) of this section is a reportable event as described in il 50.55(e), 50.72 and i 50.73. The affected applicant or licensee shall propose immediate
- steps to demonstrate compliance or bring plant design or operation into compliance with i 50.46 requirements.
- 3. In 10 CFR Part 50, Appendix X, paragraph II.1.b is deleted, mm (4. w 'wl paragraph II.1.c is redesignated 11.1.b thef ext t of p7a agraph I.C.5.b 9j/f j'f y and paragraphs !!.1.b and !!.5 are revised, and a new section I.C.S.c 9 f a- 70 5/
is added to read as follows:
APPENDIX X - ECCS EVALVATION NODELS
!. REQUIRED AND ACCEF ABLE FEATURES OF THE EVALVATION MODELS***
C. Blowdown Phenomena ***
- 5. Post-CHF Heat Transfer Correlations.***
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- b. The Groeneveld flow film boiling correlation (equation 5.7 of O. C. Groencveld, "An Investigation of Heat Transfer in the Liquid Deficient Regime," AECL-3281, revised December 1969) and the Westinghouse cortelation 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, the transition boiling correlation of McDonough, Milich, and King (J. B.
McDonough, W. Milich, E. C. King, "An Experimental Study of Partial Film Boiling Region with Water at Elevated Pressures in a Round Vertical Tube," Chemical Engineering 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 is restricted as follows:
- c. Evaluation models approved after ' October 17, 1988, ' which j make use of the Dougall-Rohsenow flow film boiling correlation (R. S.
Dougall and W. M. Rohsenow, "Film Boiling on the Inside of Vertical Tubes with Upward Flow of Fluid at Low Qualities," MIT Report Number 9079 26, Cambridge, Massachusetts, Septemtar 1963) may not use this correlation under conditions where nonconservative predictions of heat transfer result. Evaluation models that make use of the Dougall-Rohsenow correlation and were approved prior to : October 17, 1988, .
l continue to be acceptable until a change is made to, or an error is :
corrected in, the evaluation model that results in a significant reduction in the overall conservatism in the evaluation model. At that time continued use of the Dougall-Rohsenow correlation under conditions where nonconservative predictions of heat transfer result
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i 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 0F from that which would have been calculated on October 17, 1988, due either to individual changes or error I corrections or the net effect of an accumulation of changes or error corrections. l l
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4 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 request.
- 5. General Standards for Acceptability - Elements of evaluation models reviewed will include technical adequacy of the calculational methods, including: for models covered by i 50.46(a)(1)(ii),
compliance with required features of Section I of this Appendix K; and, for models covered by 5 50.46(a)(1)(1), assurance of a high level of probability that the performance criteria of 5 50.46(b) would not be exceeded.
Dated at Rockville, MD this day of 7V1988. !
For the Nu ear Regulato y Commission.
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'J. Chil CT 'Ns
, Secretary of t 1e Commission.
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