ML20206E975
| ML20206E975 | |
| Person / Time | |
|---|---|
| Issue date: | 06/07/1985 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Minogue R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| NUDOCS 8606240062 | |
| Download: ML20206E975 (131) | |
Text
,
JUN 071985 MEMORANDUM FOR:
Robert B. Minogue, Director Office of Nuclear Regulatory Research FROM:
William J. Dircks Executive Director for Operations
SUBJECT:
CONTROL OF NRC RULEMAKING By memorandum of February 13, 1984, " Control of NRC Rulemaking by Offices Re arting to the EDO," Offices were directed that effective April 1,1984, (1 all offices under EDO purview must obtain my approval to begin and/or continue a specific rulemaking, (2) resources were not to be expended on rule.
makings that have not been approved, and (3) RES would independently review rulemaking proposals forwarded for nly approval and make reconnendations to me concerning whether or not and how to proceed with the rulemakings.
In accordance with n1y directive, the following proposal concerning rulemaking has been forwarded for my approval.
Proposed amendment to 10 CFR 50.46, Acceptance Criteria for Emergency Core i;
Cooling Systems for Light Water Nuclear Power Reactors.
(Sponsored by RES -
memorandum, Minogue to ED0 dated May 31,1985.)
I approve continuation of this rulemaking. The NRC Regulatory Agenda (NUREG.0936) should be modified to reflect the status of this rulemaking.
(sipesWilliam1.Dirclis William J. Dircks Executive Director for Operations cc:
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WASHINGTON, D.C. 20555 MAY 311985
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MEMORANDUM FOR: William J. Dircks Executive Director for Operations FROM:
Robert B. Minogue, Director Office of Nuclear Regulatory Research
SUBJECT:
CONTROL 0F NRC RULEMAKING: RES REVIEW 0F ONG0ING RES SPONSORED RULEMAKING Based on our review of the ongoing RES sponsored rulemaking to amend 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, RES recommends that NRC should continue with this rulemaking.
This reconnendation in draft form has been coordinated with the Office of Nuclear Reactor Regulation.
The basis for our recommendation is as follows:
1.
The NRC has sponsored research on ECCS performance since the original ECCS rule was adopted in 1974.
The results of experiments, computer code development, and code assessment now allow more accurate (along with reasonable estimates of uncertainty) calculations of ECCS performance during a loss-of-coolant accident for LWRs than is possible using present Appendix K evaluation models.
More realistic calculations of ECCS performance will provide plant licensees and the NRC with more reliable information with which to make decisions regarding safety systems and operator guidelines for response to LOCA and other transients.
2.
The effect of this modificd rule would be to ease the restrictions on total power output and allow more efficient use of fuel. This potential increase in power production could be worth several hundred million dollars per year to the industry without impacting public health and safety.
3.
The rule has been formulated so that industry could proceed under current regulations if it wishes, or else use the new procedures. Thus, though the cost of developing a new evaluation model would be on the order of several million dollars for generic models and several hundred thousand dollars for each application, industry would weigh these costs against the potential cost reductions.
The net effect should be significant cost savings.
g-EY.3.11985 William J. Dircks 2
4.
As a result of requests from industry, the NRC staff has already begun to
. implement a limited approach similar to that in the rule (SECY-83-472) and thus significant charges in planned NRC and industry resources are not anticipated.
The complete RES review package has been sent to OEDO (Attention: DEDR0GR) and to the Director of the Office of Nuclear Reactor Regulation.
W i d Rob t B. Minogue ' Director Office of Nuclear Regulatory Research
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MEMORANDtrl FOR: William J. Dircks Executive Director for Operations FROM:
Rotart B. Minogue, Director i
Office of Nuclear Regulatory Research
SUBJECT:
CONTROL OF NRC RULEMAKING:
RES REVIEW 0F ONGOING RES SPONSORED RULEMAKING Based on our review of the ongoing RES sponsored rulemaking to amend 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, RES recomends that NRC should continue with this rulemaking.
This recomendation in draf t fom has been coordinated with the Office of Nuclear Reactor Regulation.
The basis for our recommendation is as follows:
1.
The NRC has sponsored research on ECCS perfomance since th e original ECCS rule was adopted in 1974.
The results of experiments, computer code development, and code assessment now allow more accurate (along with reasonable estimates of uncertainty) calculations of ECCS performance during a loss-of-coolant accident for LWRs than is possible using present
' Appendix K evaluation models. 'More realistic calculations of ECCS 1
parfomance will provide plant licensees and the NRC with more reliable infomation with which to make decisions regarding safety systems and operator guidelines for response to LOCA and other transients.
i 2.
The effect of this modified rule would be to ease the restrictions on total power output and allow more efficient use of fuel. This potential l
I increase in power production could be worth several hundred million doliars per year to the industry without impacting public health and safety.
3.
The rule has been fomulated so that industry could proceed under current regulations if it wishes, or else use the new procedures. Thus, though the cost of developing a new evaluation model would be on the order of several million dollars for generic models and several hundred thousand dollars for each application, industry would weigh these costs against the potential cost reductions.
The net effect should be significant cost savings.
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i MY 311985 Willian J. Dircks
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As a result of requests frorn industry, thr NRC staff has already begun to implement a limited approach similar te that in the rule (SECY-83-472) and thus significant changr,r. in planned NRC and irdustry resourcrs are not anticipated.
The cmplete RES review package has been sent tc OEDO (Attention:
DEOR0GR) and to the Director of the Office of Nuclear Reactor Reguletion.
Robert B. Minogue, Director Office of Nuclear Regulatory Research Distribution RAMRB r/f Circ /Chron/Subj MFleishman JBelote JMalaro FGillespie Dross RMinogue WBesk +
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VOTING SHEET 10: RIRB FROM:
F. P. Gillespie, Chaiman, RICL TITLE OF RULEMAKING: Acceptance Criteria for ECCS (Part 50) v AGREE WITH RECOMENDATIONS REQUEST RIRB A
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Acceptance Criteria for ECCS (Part 50)
REQUEST RIRB
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AGREE WITH RECOMMENDATIONS MEETING.
IN RES RULEMAKING REVIEW PACKAGE MODIFY RECOMMENDATIONS IN NOT PARTICIPATING.
RES RULEMAKING REVIEW PACKAGE AS INDICATED BELOW COMMENTS AND SUGGESTIONS:
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. __.s RES INDEPENDEhi El VIEk' BOARD VOTING SHEE1 TO:
F. P. GILLESPIE, CHAIRMAN, RIRB FROM:
G. A. Arlotto, Member, RIRB TITLE OF RULEMAKING:
Acceptance Criteria for ECCS (Part 50)
X AGREE WITH REC 0lHENDATIONS IN RES RULEMAKING REVIEW PACKAGE MODIFY RECOMMENDATIONS IN NOT PARTICIPATING' RES RULEMAKING REVIEW PACKAGE AS INDICATED BELOW COMMENTS AND SUGGESTIONS:
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will be appreciated. RAMRB will use the voting sheets to assemble the complete RES review package for eventual transmittal to the OEDO and the Director of the user office.
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h Distributirn Subj2ct NZubsr Chron WDBeckner Cir WDBeckner R/F Branch R/F RBMinogue DFRoss OEBassett WMMorrison LMShotkin MEMORANDUM FOR: William J. Dircks Executive Director for Operations FROM:
Robert B. Minogue, Director Office of Nuclear Regulatory Research
SUBJECT:
CONTROL OF NRC RULEMAKING: RES REVIEW OF ONGOING RES SPONSORED RULEMAKING Based on our review of the ongoing RES sponsored rulemaking to amend 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, RES recomends that NRC should continue with this rulemaking. This recomendation in draft form has been coordinated with the Office of Nuclear Reactor Regulation.
The basis for our recomendation is as follows.
1.
The current requirements of Appendix K to Part 50 do not make use of the current knowledge of ECCS performance gained through research and, therefore, do not provide the most accurate calculations possible for use by NRC and plant operators in making safety related decisions.
2.
The current requirements of Appendix K are now known to be more conservative than required to protect the health and safety of the I
public and are unnecessarily restricting operation of some plants.
3.
This rulemaking is consistent with Comission guidance to reflect the results of research in the Commission's regulations.
The complete RES review package has been sent to OEDO (Attention: DEDROGR)and to the Director of the Office of Nuclear Reactor Regulation.
Robert B. Minogue, Director Office of Nuclear Regulatory Research g
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4, NRC REGULATORY AGENDA ENTRY l
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Acceptance Criteria for Emergency Core Cooling Systems (ECCS) for Light i
Water Nuclear Power Reactors CFR CITATION:
10 CFR 50 ABSTRACT:
- The Nuclear Regulatory Commission (NRC) staff will recommend to the Commission revisions to 8 50.46 related to the methods used to demonstrate the adequacy of the ECCS during a loss-of-coolant-accident (LOCA).
Currently, 9 50.46 requires that calculations of ECCS performance use features specified in Appendix K to Part 50.
The revisions would allow the applicant or licensee to perform these required calculations using best estimate models with cons'ideration of the uncertainty of the calculation as an alternativ.e to using the features of Appendix K.
The required and acceptable. features of Appendix K do not reflect the significantly improved knowledge'of ECCS performance during a LOCA that is now available as a result of research performed in the 10 years since Appendix K was written. As a result of this research, it is now known that the Appendix K features are highly conservative and that calculations performed using l
Appendix K features are, while conservative, not realistic. A revision to
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S 50.46 is needed to improve the realism of the calculations by allowing the calculations to be performed using the latest knowledge of ECCS performance during a LOCA. More realistic calculations would provide the NRC and the plant operators with more accurate information with which to make decisions regarding safety systems and operator guidelines for response to LOCA and other transients.
Realistic calculations also are consistent with other regulatory analysis requirements.
In addition, restrictions on reactor operation which are now known to not be necessary in order to protect the health and safety of the public could be relaxed.
Removal of unnecessary operating restrictions could result in significant reduction in the cost of operating certain plants and/or allow an increase in the electrical output of some plants.
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Since the proposed revision would represent a significant change in a regulatory requirement, the staff is currently preparing two supporting documents. The first document is a summary of ECCS research performed over the last 10 years which will serve as the technical basis for the proposed rule revision. The second document is a regulatory guide which will provide definition of what constitutes an acceptable best estimate model and acceptable methods of performing the uncertainty evaluation. The estimated cost to the NRC of this rulemaking is 2-3 staff years and $200K of contractor support.
The bulk of this cost is expected to be expended preparing the two reports.
TIMETABLE:
Draft Regulatory Guide July 1985 Draft Research Report August 1985 Commission Paper Making Recommendations August 1985 Regulatory Guide Issued for Comment September 1985 NPRM September 1985.
Final Research Report October 1985 NPRM Comment Period End December 1985 LEGAL AUTHORITY:
42 USC 2133; 42 USC 2134; 42 USC 2201; 42 USC 2232; 42, USC 2232 EFFECTS ON SMALL BUSINESS AND OTHER ENTITIES:
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AGENCY CONTACT:
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W. D. Beckner Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C.
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Acceptance Criteria for Emergency Core Cooling Systems for TITLE:
Light-Water-Cooled Nuclear Power Plants CFR CITATION:
10 CFR 50 ABSTRACT:
The advance notice of proposed rulemaking seeks comment on several questions concerning the acceptance criteria for in light-water-cooled Emergency Core Cooling Systems (ECCS) nuclear power plant commented on are (1) under what circumstances should corrections to ECCS models be used during licensing reviews without necessitating complete reanalysis of a given plant or an entire (2) what would be the impact of the proposed group of plants; procedure-oriented and certain specific technical rule cha is considering changing certain technical and nontechnical requirements within the existing ECCS rule. The technical changes would include consideration of new research information. The nontechnical changes would be procedure-oriented and would, among other things, allow for corrections to be made to vendor ECCS analysis codes during the construction review and during construction of the plant.
The changes would provide improvements to the ECCS rule and improve licensing evaluation in the light of present knowledge, while preserving a level of conservatism consistent with that knowledge.
TIMETABLE:
1 ANPRM 12/06/78 43 FR 57157 ANPRM Comment Period Begin 12/06/78 43 FR 57157 ANPRM Comment Period End 02/05/79 NPRM 04/00/85 LEGAL AUTHORITY:42 USC 2133; 42 USC 2134; 42 USC 2201; 42 USC 2232; 42 U No EFFECTS ON SMALL BUSINESS AND OTHER ENTITIES:
AGENCY CONTACT:
Morton R. Fleishman Office of Nuclear Regulatory Research Washington, DC 20555 301 443-7616 SO ~
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[10 CFR Part 50]
___ j Revisions to Acceptance Criteria for Emergency Core Cooling Systems AGENCY:
Nuclear Regulatory Commission.
ACTION:
Proposed Rule.
SUMMARY
The Nuclear Regulatory Commission is considering amending its regulations concerning acceptance criteria for emergency core cooling systems (ECCS). These amendments would change the methods used to demon-strate that emergency core cooling systems would protect the nuclear reactor core during a loss-of-coolant accident (LOCA).
Research performed since the current rule was written has shown that calculations performed in accordance with the current rule result in estimates of cooling system performance that are significantly worse than that which would be estimated based on the improved knowledge gained from this research.
In addition, the operation of some nuclear reactors is being unnecessarily restricted by the rule, resulting in increased costs of electricity generation.
The amended rule would allow use of the best information and knowledge avail-able to demonstrate that the ECCS would protect the reactor during a LOCA.
Procedural changes would also be made so that certain reanalyses which do not contribute to safety would not be required to be performed immediately.
The amendments would apply to all applicants for and holders of l
construction permits or operating licenses for light water reactors.
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DATES:
Comment period expires (60 days following publication in the f'
l Federal Register).
Comments received after that date will be considered 1
Enclosure E l
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if it is practical to do so, but assurance of consideraticn cannst be p
given except as to comments received on or before that date.
i ADDRESS: Written comments or suggestions for consideration in connection with the proposed amendments should be submitted to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention Docketing and Service Branch. Copies of comments received say be examined in the Commission's Public Document Room at 1717 H Street NW.,
Washington, DC 20555.
FOR FURTHER INFORMATION CONTACT:
L. M. Shotkin (301-427-4254), U.S.
Nuclear Regulatory Commission, Washington, DC 20555.
SUPPLEMENTARY INFORMATION:
Background
Section 50.46 of 10 CFR Part.50 provides " Acceptance Criteria for t
Emergency Core Cooling Systems (ECCS) in Light Water Nuclear Power Reactors."
This section requires that calculations of loss-of-coolant accidents (LOCA) be performed to show that the ECCS will maintain cladding temperatures, cladding oxidation and hydrogen generation to within certain specified limits.
It also requires that a coolable core geometry be maintained and I
that long term decay heat removal be provided. Appendix K to 10 CFR Part 50 sets forth certain required and acceptable features of the models used to perform these calculations.
The criteria of 10 CFR 50.46 and the calculational methods specified in Appendix K were formally issued in l
January 1974 after extensive rulemaking hearings and are based on the J.
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understanding of ECCS performance available at that time.
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In the t:n years following the rulemaking, ever $700 caillion has E
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been invested by the Nuclear Regulatory Commission on research investigat-p ing ECCS perfomance. The majority of this LOCA research is complete and
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has greatly improved the understanding of ECCS performance during a LOCA.
The methods specified in Appendix K are now known to be highly conserva-tive; that is, the actual temperatures during a LOCA would be much less than the temperatures calculated using Appendix X methods. The ECCS research has gone beyond showing that Appendix K is conservative, it has allowed quantification of that conservatism. The results of experiments, computer code development, and code assessment now allow more accurate calculations of ECCS performance during a LOCA than is possible using Appendix K procedures, along with reasonable estimates of uncertainty.
3 It is also known that some plants are being, restricted in operating flexibility by limits resulting from conservativ'e Appendix K calcula-tions.
These restrictions may be preventing optimal operation of some plants.
Based on research perfomed, it is now known that these restric-tions can be relaxed without affecting safety.
On December 6,1978, the NRC published an advance notice of proposed rulemaking (43 FR 57157) calling for a two phase approach to the revision of 10 CFR Part 50 and Apperdix K.
The first step would have been to make procedural changes and to permit minor model changes which would not have reduced the conservatism contained in Appendix K.
The second phase would have made further technical changes based on research results and operat-ing experience.
NRC activity on the ECCS rulemaking was severely curtailed as a result m
of the high priority efforts required by the THI-2 accident. This ECCS
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rvlemaking css:n'tially. sat dormant until July 1981, whin it was brought up e
again in the context of simplifying and streamlining the regulatory process.
The NRC has reviewecl the comments made by outside organizations on
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h the advance notice of proposed rulemaking, as well as a number of other
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comments received since that time.
In general, the commenters support a rule change that would permit greater flexibility in meeting the regula-tions and would incorporate the use of presently available research infor-nation.
Many felt that the Phase 1 scope should be expanded to allow the use of additional inf,>rmation available from the ECCS research that has
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been performed.
Because of the delay in changing the ECCS rule, the NRC has adopted an interim approach described in SECY-83-472,1 to accommodate requests for I
improved evaluation models, submitted by. reactor, vendors, for the purpose of reducing reactor operating restrictions. This interim approach requires a best estimate calculation with an evaluation of the uncertaifity in the calculation to hq demonstrate that.an adequate conservatism or safety factor exists.
'The NRC has decided to proceed with the rulemaking, but in the form of a more comprehensive rule amendment based on (1) the comments received since the publication of the 1978 notice of proposed rulemaking, (2) the additional research conducted and experience gained since the 1978 notice, and (3) recent experience using host estirate calculations and uncertainty evaluations to support licensing decisions.
I A report providing the technical basis for revision of the ECCS rule has been prepared.
It summarizes both the extensive ECCS research 7..-
l 25ECY-83-472, " Emergency Core Cooling System Analysis Methods," November 13 1983, is available for inspection and copying for a fee at the Commission's
, I Public Document Room at 1717 H Street NW., Washington, DC.
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that has been c:nducted cnd the cxperience gained using this information
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SLMtARY OF PROPOSED RULE CHANGES y
$ 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors:
Section 50.46(a)(1) would be amended so as.to include require 3that trie evaluation model include sufficient supporting justification to show f
that the analytical technique describe the behavior of the reactor system during a LOCA. Deleted was the requirement that the features of Appen-1 dix K to 10 CFR Part 50 be used to develop the evaluation model.
Before the evaluation model could be used it would have'to be accepted by the NRC staff. The staff will expect that the analytical technique will, to the extent practicable, utilize the best available methods and data. The
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amended rule would also require that t'he uncertaint of the calculation be evaluated and considered when comparing the results of the calculation to the temperature limits and other criteria of S 50.46(b). The rule would not specifically state the analytical methods or uncertainty evaluation techniques that shall be used but guidance is being provided 2 to indicate methods that the NRC staff in the form of a Regulatory Guide would consider acceptable.
It should be noted as discussed in SECY-83-4721, that the NRC has in the past found acceptable a method for estimating the E
3 Regulatory Guide xx (issued for comment), " Acceptable Methods for Evaluating T
ECCS Performance During a LOCA," is available (insert boilerplate on Regula-tpru Guides)
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unccrtainty that was judged to be at least ct the 95% probability level.
j Appendix X Section II, " Required Docu.nentation" would remain generally
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applicable, with only minor revisions made to be consistent with the amended rule.
Sections 50.46(a)(2) and (3) would be totally revised to eliminate those sections concerned with historical implementation of the current rule and will be replaced as follows:
Section 50.46(a)(2) would be revised to allow the features of Appen-dix K to be used in evaluation models as an alternative to performing the uncertainty evaluation specified in the amended S 50.46(a)(1).
This method would remain acceptable because it is believed that the conservatism in the Appendix K features is sufficient and furthermore, is conservative with respect to the new method as proposed in th,e amended S 50.46(a)(1).
ThiswouldallowcurrentandfutureapplicantsaIidlicenseestouseexist-ing evaluation models if they did hot need or desire relief from current
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operating restrictions, The revised S 50.46(a)(3) would state specific requirements for reanalyses when errors in evaluation models are discovered or changes are made to evaluation models. This section would require that all errors affecting peak cladding temperature and any changes or combination of changes such that the sum of the absolute magnitudes of the calculated cladding temperature changes exceeds 50*F shall be reported to the NRC.
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If as a result of the error correction or change, the limiting transient l
calculation exceeds the criteria of S 50.46(b), then immediate steps would have to be taken to achieve compliance.
If the criteria of S 50.46(b) are not exceeded, a reanalysis shall be performed on a schedule either y
l developed using an approved integrated scheduling system, if available,
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or one established by the NRC after receipt of a proposed schedule.
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.ujg Appendix K ECCS Evaluaticn Models g.-
-t Minor amendments would be made to Appendix K, I.(C)(5)(b) to modify post-CHF heat transfer correlations listed as acceptable.
The "McDonough" 1-1 7'
reference would be replaced with a later paper which is more generally available and which includes additional data.
The "Dougall-Rohsenow" correlation, which has been found to be non conservative in some situa-s.
l tions, would be eliminated as a generally acceptable model.
Appendix K, II(1)(b) would De removed since this requirement would be clarified under -
the amended S 50.46(a)(3).
Likewise, Appendix K, II(5) will be 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 y
Appendix K should not effect any existing approved evaluation models y
since the changes are either " house keeping" changes or only changes to
" acceptable features," not " required features."
REGULATORY ANALYSIS t
The Commission has prepared a regulatory analysis for this proposed regulation.
The analysis examines the costs and benefits of the rule as considered by the Commission.
A copy of the regulatory analysis is avail-able for inspection and copying for a fee at the NRC Public Document Room, 1717 H Str.eet NW., Washington, DC.
PAPERWORK REDUCTION ACT The proposed rule will be submitted to the Office of Management and j.'
Budget for clearance of the application requirements that may be appro-priate under the Paperwork Reduction Act (Publ. L.96-511).
The SF-83 7
Enclosure E
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" Request for Clsaranca,* supporting Statement, and related documentation
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submitted to OMB will be placed in the NRC Public Document Room at 1717 H. Street NW., Washington, DC 20555.
The material will be available p
for inspection and copying for a fee.
REGULATORY FLEXIBILITY ACT In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission hereby certifies that this rule will not, if
. promulgated, have a significant econ'omic impact on a substantial number of small entities.
This proposed 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 o,f "small entities" set a
forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Adminis-tration in 13 CFR Part 121.
Since these companies are dominant in their service areas, this ' proposed rule does not fall within the purview of the Act.
LIST OF SUBJECTS IN 10 CFR PART 50*
Nuclear Power Plants and Reactors Accordingly, notice is hereby given that, pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as
- Comparative text has been used to indicate deletions and additions by dashing through and underlining.
4-l
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8 Enclosure E
~ _ _ _ _.
u-C II amended and S:cticn 553 ef Titio 5 of the United States Code, adoption of
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the following amendments to 10 CFR Part 50 is contemplated.
PART 50-DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES 1.
The author.ity citation for Part 50 reads as follows:
AUTHORITY:
Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2133, 2134, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201,
~
202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C. 5841, 5842, 5846), unless otherwise noted.
Section 50.7 also issued under Pub. L.95-601, sec.10, 92 Stat.
2951 (42 U.S.C. 5851). Sections 50.58, 50.91 and 50.92 also issued under Pub. L.97-415, 96 Stat. 2073 (42 U.S.C. f239).
Section 50.78 also 1ssued 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).
Sectio'ns.50.100-50.102 also issued under.sec. 186, 68 Stat. 955 (42 U.S.C. 2236).
For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C.
2273), SS 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));
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 SS 50.55(e), 50.59(b), 50.70, 50.71, 50.72, and 50.78 are issued under sec.161o, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).
I 5
9 Enclosure E
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2 j?
2.
In 5 50.46(e)(1), (a)(2), and (a)(3) are amended to read as
- j follows
~
Acceptance criteria for emergency core cooling systems for light S 50.46 water nuclear power reactors.
(a)(1) [Except-as provided-in paragraph-(a)(23-and-(33-of-this section-] Each boiling and pressurized light water nuclear power reactor fueled with uranium oxide pellets within cylindrical Zircaloy cladding shall be provided with an emergency core cooling system (ECCS) which shall 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 shall be calculated in accordance with an [acceptabie] evaluation model that has been accepted by the NRC staff and shall be calc. plated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the entire spectrum of postulated loss-of-coolant accidents is covered.
[ Appendix-K--EEES Evainstion-Modeis--sets-forth certain-required-and-acceptable-features of evaination-modeis-] Except as provided in paragraph (a)(2) of this section, the evaluation model must include sufficient supporting justi-fication to show that the analytical technique describes the behavior of the reactor system during a loss-of-coolant accident.
Comparisons to applicable experimental data should be made and uncertainties in the analysis method and inputs should be identified and assessed so that the uncertainty in the calculated results can be estimated.
This uncertainty shall be accounted for when the calculated ECCS cooling performance is compared to the criteria of paragraoh (b) of this section. Appendix K, j
4 10 Enclosure E
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k Part II, Required Documentatien, sits forth the documentatir.n requira-j
}
ments for each evaluation model. One acceptable method for developing 7 I n
, - - I J
the evaluation model, calculating the ECCS cooling performance and performing the uncertainty analysis is giv k in Regulatory Guide XX, Conformance with the criteria set forth in paragraph (b) of this s'ction with'ECCS cooling performance calculated in accordance e
with an [acceptabie] evaluation model that has been accepted by the NRC staff, may require that restrictions be imposed on reactor operation.
(2) Alternatively, an ECCS evaluation model may be developed in conformance'with the required and acceptable features of Appendix K, ECCS Evaluation Models.
(3)(i) Each applicant for or holder of an operating license or
~
construction permit shall estimate the effect of'any change to an evalua-tion model that has been accepted by the NRC staff to determine if the change is significant.
For this purpose, a significant change.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 latest model that was accepted, or is a cumulation of changes such that the sum of the absolute magnitudes of the respec-l l
tive temperature changes is greater than 50*F.
(ii) Each significant change to, or any error discovered in, an accepted evaluation model which affects the temperature cylculation shall be documented and the effect of the change or correction of the error on the limiting ECCS analysis reported to the Director of Nuclear Reactor Regulation within 30 days.
Each applicant or licensee shall y
also include with the report a proposed schedule for completing actions
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11 Enclosure E l
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needed to comply with applicable requirements. This sch dule may be j
l developed usino an intearated scheduling system previously approved for
]
the facility by the NRC.
For those facilities not having an NRC approved integrated scheduling system, a schedule for achievino compliance will be established by the NRC staff within 60 days of receipt of the proposed schedule.
For any change or error correction that results in a calculated ECCS cooling performance that does not conform to the criteria set forth in paragraph (b) of this section, the affected applicant or licensee must notify the NRC promptly and propose immediate steps to bring plant design or operation into compliance.
(4) (Boilerplate on Reg. Guide incorporation by reference.)
3.
In 10 CFR Part 50 Appendix K, paragrap) I.C.5.b is amended J
to read as follows:
[
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-Bougali-Rohsenew flow-fiim-beiling-correlation-(R--S--Bougail-and-W--M--Rohsenew-uFiim Boiling-en-the-inside-of-Verticai-Tabes-with-Upward-Fiow-of-the-Finid-at l
tow-Qualities--MiT-Report-Namber-9079-26--Eambridge--Massachusetts-September-19633-] and the Westinghouse Correlation of steady-state transition boiling (" Proprietary Redirect / Rebuttal Testimony of Westing-house Electrical Corporation," USNRC Docket RM-50-1 page 25-1, October 26, 1982) are acceptable for use in the post-CHF boiling regimes.
In addi-tion the transition boiling correlation of McDonough, Milich, and King b
(J. B. McDonough, W. Milich, E. C. King, [uPartiai-Film-Boiling-with
+
i 12 Enclosure E
y-3'.
[
thter-ct-2000 psig-in-a-Round-Verticai-Tobe 8-MSA-RIs: arch-EIrp---Tech-nicai-Report-62-(NP-69763--(19563] "An Experimental Study of Partial p
Film Boiling Region with Water at Elevated Pressures in a Round Vertical
. -1 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 shall be restricted as follows:
a a
a a
In 10 CFR Part 50 Appendix K, paragraph II.1.b. is deleted.
4.
l 5.
In 10 CFR Part 50 Appendix K, paragraph II.1.c. is redesignated II.1.b. and is amended to read as follows:
b.
A complete listing of each computer program, in the same form as used in the evaluation model, shall be furnished to the Nuclear Regulatory Commission upon request.
6.
10 CFR Part 50 Appendix K is amended by revising Section II.S.
to read as follows:
5.
General Standards'fo'r Acceptability - Eleme'nts of evaluation models' reviewed will include tech'nical adequacy of the calculational methods, including compliance with_ required features of Section I of I
this Appendix K, if applicable, ank provision of a level of safety and I.
margin of conservatism comparable.to other acceptable evaluation models, 1
1 taking into account significant differences in the reactors to which they apply.
Dated at Washington, DC this day of
, 1985.
For the Nuclear Regulatory Commission.
b t
Samuel J. Chilk i
l Secretary of the Commission 13 Enclosure E l
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TASK LEADER EVALUATION O
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4 TITLE:
Acceptance Criteria for Emergency Core Cooling Systems (ECCS) for Light Water Nuclear Power Reactors CFR CITATION:
10 CFR 50 TASK LEADER:
W. D. Beckner (7-4311)
SCREENING:
a.
The issue to be addressed.
Should the requirements of 6 50.46 be revised to allow the latest knowledge gained through research to be used to demonstrate the adequacy of the ECCS during loss-o.f-coolant-accidents (LOCA) in place of the current methods specified in Appendix K to Part 50?
b.
The necessity for addressing the issue. The current Appendix X requirements result in unnecessary restriction on the operation of reactors and do not provide the most accurate calculations possible for use by NRC and plant operators in making safety related decisions.
The Commission has directed the staff to make use of research results in the Commission's regulations.
c.
Alternatives to rulemaking.
The staff have considered a number of l
alternatives, all of which involve rulemaking except the option of doing nothing. This would result in continued use of the licensing approach discussed in SECY-83-472. This approach is viewed as an interim approach to be used until the rule can be revised.
The disadvantages of the SECY-83-472 approach are:
(1) two separate calculations are required to meet both requirements of the rule and staff conditions for use of this approach, and (2) the use of this approach risks case-by-case litigation due to lack of a basis in rulemaking.
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i d.
How the issue would be addressed through rulemaking.
The required and I
acceptable features of Appendix K do not reflect the significantly improved knowledge of ECCS performance during a LOCA that is now available as a result research performed in the 10 years since Appendix K was written.
As a result of this research, it is now known that the Appendix K features are highly conservative and that calculations performed using Appendix K features are, while conservative, not realistic. A revision to S 50.46 is needed to improve the realism of the calculations by allowing the calculations to be performed using the latest knowledge of ECCS performance during a LOCA. More realistic calculations are also consistent with other regulatory analysis requirements.
The staff will recommend to the Commission revisions to S 50.46 related to the methods used to demonstrate the adequacy of the ECCS during a LOCA.
Currently, S 50.46 requires that calculations of ECCS performance use features specified in Appendix K to Part 50. The revisions would allow the applicant or licensee to perform these requirsi calculations using best estimate models with consideration of the uncertainty of the calculation as an alternative to using the features of Appendix K.
Since the proposed revision would represent a significant change in a regulatory requirement, the staff' is currently preparing two supporting documents. The first document is a summary of ECCS research performed over the last 10 years which will serve as the technical basis for the proposed rule revision. The second document is a regulatory guide which will provide definition of what constitutes an acceptable best estimate model and acceptable methods of performing the uncertainty evaluation.
e.
How the public, industry and NRC will be affected by the rulemaking.
More realistic calculations would provide the NRC and the plant operators with more accurate information with which to make decisions regarding safety systems and operator guidelines for response to LOCA and other transients.
In addition, restrictions on reactor operation which are now known to not be necessary in order to protect the health and safety of the public could be relaxed.
Removal of unnecessary operating restrictions could result in a significant reduction in the cost of operating certain plants (several million dollars a year) and/or 2
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I allow an increase in the electrical output of some plants (~5%). Best estimate
{
analyses of ECCS performance would require that the licensee or applicant develop and submit for NRC approval a new evaluation model.
This would require industry resources estimated to be on the order of several million dollars for generic models and several hundred thousand dollars for each application.
However, use of best estimate analyses would only be an alternative that the applicant or licensee could use if the potential benefits justified such an expenditure. The existing Appendix K requirements would remain available for r
use by licensees or applicants who would not benefit from the improved analyses.
It should also be noted that the industry and the NRC staff are already working on best estimate approaches as part of'the SECY-83-472 approach and, therefore, significant changes in planned NRC and industry resource requirements are not anticipated.
3, i
f.
NRC resources and scheduling for rulemaking. The estimated cost to the NRC of this rulemaking is 2-3 staff years and $200K of contractor support. A major portion of these resources will_be used to prepare the supporting documents, particularly the summary of research.
This summary of research,is a task that would be performed with or without the rulemaking and contractor support funding was previously budgeted for this task..The major change in this' effort l
as a result of the rulemaking is to accelerate the effort to expend most of the resources over the next few months.
i QUALITY CONTROL EVALUATION:
Review against expt.ip NRC procedures and guidelines.
Because of the a.
major scope of th's Ners ing, high level staff from several NRC offices have been involved very earb % the process.
Draft Commission papers have already been circulated and reviewed by RES, NRR and ELD staff at the branch chief level or higher.
Several meetings between NRR, RES and ELD staff at various levels have also been held. An informational meeting is also planned for the CRGR. The goal of these meetings has been to ensure that agreement on j
objectives cf the proposed rule revision is achieved and that the rule as proposed will meet those objectives.
l 3
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.e b.
Adequacy of technical data to support rule. The technical basis for this rule revision has resulted from over ten years of research conducted at a cost of $700M. This research is summarized in hundreds of formal documents.
The very size of this data base is in itself a problem in that it must be made clearthat-thelargedatabaseiNdeedsupportssucharulechange.
For this reason, a major summary document will be prepared which will attempt to put these overall research results in perspective.
RECOMMENDATIONS:
The RES task leader recommends that the rulemaking action continue and that the staff paper be forwarded to the Commission as soon as good drafts of the supporting material are available. The support documents are the critical path items.
The regulatory guide'is especially critical since it will be referenced by the rule and must be published as a draft for comment when the NPRM is issued.
Specific recommendations are as follows.
~
First priority should b'e developing the regulatory guide and ensuring that NRR and ELD concur with the guide.
Second priority should be the research summary report which will be developed in parallel by RES staff and contractors.
The staff paper and other supporting documents have been reviewed extensively by staff from other offices and are in near final form.
Minor work on these items should continue.
l 4
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m SUPPORTING DOCUMENTATION e
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y:
Ie M 2 81985 For:
The Commissioners From:
William J. Dircks 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 Commission approval to publish a notice of proposed rulemaking revising the ECCS Rule contained in Appendix K and Section 50.46 of 10 CFR Part 50.
Category:
This paper covers a major policy matter.
Resource estimates are Category 1.
3 Issues:
(a) Whether the results of NRC-sponsored research, other research, and licensing experience should be reflected in the ECCS rule.
(b) Whether the ECCS rule should be revised to allow certain changes or corrections to be made to the ECCS evaluation models without requiring a complete reanalysis by the applicant or licensee.
Summary:
Section 50.46 of 10 CFR Part 50 requires that calc.ulations be performed to show that the emergency core cooling systems (ECCS) will adequately cool the reactor in the event of a loss-of-cool-ant accident (LOCA).
Appendix K sets forth certain required and acceptable features that the evaluation models, used to perform these calculations, shall contain.
The results of these calcula-tions are used to determine the acceptability of the ECCS
Contact:
L.M. Shotkin, RES 427-4254
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The Commissigners 2
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s performance.
In many instances, these calculations result in technical specification limits on the reactor operation (e.g.,
peak local power) in order to comply with the 2200'F cladding temperature limit of S 50.46.
These limits restrict the total power output and optimal operation of some reactors in terms of efficient fuel utilization, maneuvering capability and surveil-lance requirements.
It is estimated that removing these restric-tions on cperation may allow increased U.S. electricity produc-tion, worth several hundred million dollars a year, without loss of benefit to the public health and safety.
The NRC has spent over $700M sponsoring research on ECCS perfor-mance since the ECCS rule was written.
This extensive research has shown that these restrictions are too stringent. Thus, the staff recommends that the ECCS rule be amended to reflect the results of this research and to remove unnecessary operating v
restrictions.
A number of alternative approaches have been con-sidered by the staff and each approach evaluated in terms of safety, impact on the industry, NRC and industry resources required, and risk of litigation both during the rulemaking process and during application of the rule.
As a result, the staff recommends eliminating the requirement to use Appendix K features.
It is believed that the prescriptive nature of Appen-dix K is no longer necessary in view of the current improved knowledge of ECCS performance.
Best estimate analyses, combined with uncertainty evaluations, could be used as an alternative method to demonstrate conformance with S 50.46 criteria for ECCS l~
performance.
Alternatives:
The staff has considered the following options for amending the ECCS rule:
A.
Retain the existing rule with its present conservatism (no i
change).
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B.
Modify the rule as stated in the advance notice of proposed rulemaking published in the Federal Register on December 6, 1978 (Enclosure "A").
C.
Modify certain models contained in Appendix K, for which research investigations have been completed and a well documented data base exists.
These changes have been sel-ected in areas for which new experimental data has shown that the existing models contain a larger degree of conser-vatism than justified by current data uncertainties or ar,e obviously unrealistic.
]
D.
Eliminate the requirement to use Appendix K models and allow best estimate models to be used.
Reduce the 2200 F and 17%
a oxidation limits of f 50.46 appropriately to ensure that sufficient conservatism exists to cover uncertainties in the best estimate calculation.
E.
Eliminate the requirement to use Appendix K models and allow best estimate models combined with an evaluation of the uncertainty in the overall calculation, similar to that discussed in SECY-83-472. Thef50.46limitsof2200Fand 17% oxidation would be unchanged.
Background:
10 CFR 50.46 provides " Acceptance Criteria for Emergency Core Cooling Systems (ECCS) in Light Water Nuclesr Power Reactors."
This section requires that calculations of loss-of-coolant acci-dents (LOCA) be performed to show that the ECCS will maintain l
cladding temperatures, cladding oxidation, and hydrogen generation to within certain specified limits.
It also requires that a coolable core geometry be maintained and that long term decay heat removal be provided.
Appendix K sets forth certain rules on how these calculations shall be performed. The criteria of S 50.46 and the calculational methods specified in Appendix K were formally issued in January 1974 after extensive rulemaking
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0 The Commissigners 4
4R hearings and are based on the understanding of ECCS performance available at that time.
In the ten years following the rulemaking, over $700M has been spent by the NRC on research investigating ECCS performance.
It is estimated that a similar amount has been spent by DOE (includ-ing AEC and ERDA), the U.S. industry, and foreign researchers; resulting in a total estimated expenditure of over $1.5 billion.
The majority of this LOCA research is complete and has greatly improved the understanding of ECCS performance during a LOCA.
The methods specified in Appendix K are now known to be highly conservative; that is, the actual temperatures during a LOCA would be much less than the temperatures calculated using Appen-dix K methods. This fact is best illustrated by comparisons showing temperatures during LOCA simulat bns in LOFT which are more than 600*F lower than calculations p'erformed using Appendix K procedures.
The ECCS research has gone beyond showing that Appendix K is conservative, it has allowed quantification of that conservatism.
The results of experiments, computer code development, and code assessment now allow more accurate calcula-tions of ECCS performance during a LOCA than is possible using Appendix K procedures, along with reasonable estimates of uncertainty.
In order to highlight the contributions or research toward improving understanding of LOCA phenomena, it is useful to dis-cuss two categories of research.
The first category of research included phenomena for which Appendix K requires specific calcu-lational models to be used.
Examples include decay heat, metal-water reaction rate, discharge model, reflood heat transfer at low reflood rates, and other heat transfer phenomena.
Enclosure l
"B" provides a discussion of these research findings.
In some l
cases research has shown the required models to be inadequate 1
(e.g., discharge model) and in one case nonconservative.
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The Commissieners 5
.e7 However, most of the requirements of Appendix K have been shown to be more conservative than needed to protect public health and safety. The best example is the decay heat calculation which is now known to be conservative by over 20% during the initial phases of the LOCA.
The second category of research was more general and was directed toward a best estimate understanding of the overall performance of ECCS.
This research included many investigations into specific phenomena in the areas of heat transfer and two phase flow.
Many facilities such as the THTF, FLECHT and the foreign CCTF and SCTF facilities have been used to obtain specific heat transfer data.
This research data has been used to develop major new computer codes such as TRAC and RELAP5, which account for complex phenomena such as non equilibrium and multidimensional effects and provide 4
an ability to perform best estimate calculations of ECCS perfor-mance.
In addition, major integral test facilities such as LOFT, Semiscale, TLTA, MIST, FIST and a number of foreign facilities provide complete simulations of LOCA and other transients for comparison with the calculations of the new computer codes.
This allows assessment of the overall accuracy of the calculations and identification of needed improvements.
This research, further described in Enclosure "B", now allows significantly improved calculations of ECCS performance over those possible when Appen-dix K was developed.
The staff is currently preparing a report which will document in detail the ECCS research which has been performed and the current knowledge of ECCS performance.
It is also known that some plants are being restricted in operat-ing flexibility by limits resulting from conservative Appendix K calculations.
In addition, Appendix K requires reanalyses to be performed in the event that errors are discovered or certain changes are made to approved evaluation models.
Very often, the
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The Commissien:;rs 6
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reanalyses contribute little to safety, but require significant staff and industry resources.
These restrictions may be prevent-F ing optimal operation of some plants. Based on research performed, it is now known that these restrictions can be relaxed without affecting safety. Some research results have been used in licensing calculations, but many important results cannot be i
used since Appendix K specifically requires certain methods now known to be overly conservative.
Thus a modification to the ECCS rule is desirable to relax unnecessary operating restrictions.
On December 6, 1978, the Commission published an advance notice of proposed rulemaking (43 FR 57157) calling for a two phase approach to the revision of 10 CFR Part 50 and Appendix K (Enclo-
[
sure "A").
The first step would have been to make procedural changes and to permit minor model changer which would not have a
reduced the conservatism contained in Appendix K.
The second phase would have made further technical changes based on research results and operating experience.
Staff activity on the ECCS rulemaking was severely curtailed as a result of the high priority efforts required by the TMI-2 acci-dent. This ECCS rulemaking essentially sat dormant until July 1981, when it was brought up again in the context of simplifying and streamlining the regulatory process.
l The staff has reviewed the comments made by outside organiza-tions on the advance notice of proposed rulemaking, as well as a number of other comments received since that time.
In general, the commenters support a rule change that would permit greater flexibility in meeting the regulations and would incorporate l
the use of presently available research information.
Many felt that the Phase 1 scope should be expanded to allow additional model changes such as use of the new decay heat standard.
The most consistent comment received from the industry, licensees,
7 y
2g The Commissigners 7
NRC contractors, and other government agencies was that a hearing g
like that required to support the original rulemaking should not be used to accomplish needed changes to the requirements. Many commenters indicated that they would not support a revision to the ECCS rule included in 10 CFR Part 50 and Appendix K if a lengthy hearing was required.
The ACRS has also supported a revision to the ECCS rule, as most recently stated in the ACRS annual report to Congresc (NUREG-1105).
Based on the comments received from the advance notice of proposed rulemaking and gene-ral support for a revision to the ECCS rule the staff reactivated the effort to modify the ECCS rule.
?
Because of the delay in changing the ECCS rule, the staff has adopted an interim approach, described in SECY-83-472, to accom-modate requests for improved evaluation models like that received from the General Electric Co.
This interim approach requires a best estimate calculation with an evaluation of the uncertainty in the calculation, to demonstrate that an adequate conservatism or safety factor exists.
Discussion:
Many options can be proposed for revisin the ECCS rule.
L Based on staff discussions, taking into account the numerous comments received and the industry efforts in response to SECY-83-472, the staff developed several different options for revising the ECCS rule.
In this section, each option is dis-cussed and evaluated.
Alt. A:
Retain the existing ECCS rule with its present con-servatism (no change).
1 PRO: a.
The current well-established and stable l
licensing process would be retained.
l b.
No staff resources would be required for rulemaking.
EL
{:[
The Commissitners
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o CON: a.
Some plants would continue to be unneces-sarily restricted in operation by the cur-rent rule.
b.
This would be contrary to Commission policy (NUREG-0885) to incorporate the results of research into the licensing process.
c.
Many licensees would continue to seek relief from restrictions through requests for exemptions or by using the approach discussed in SECY-83-472.
Both these approaches are interim measures and should not substitute for revising the rule to makeitconsistentw{thcurrentknowledge and practice.
Alt. B:
Modify the ECCS rule as stated in the advanced notice of proposed rulemaking published in the Federal Register on December 6, 1978.
PRO: Consistent with previously-stated plans and would allow minor changes to be quickly implemented.
CON: a.
Substantial changes would be delayed until a later phase.
b.
Does not resolve comments received on the i
advanced notice recommending more substan-t tial changes.
c.
Would not allow all data from NRC sponsored research to be used.
z - -- -.
E 15 The Commissitners 9
7 r
Al t. C:
Modify,certain models contained in the ECCS rule, for which $esearch investigations have been completed and a well-documented data base exists.
These changes would be selected in areas for which new experimental data has shown that existing models contain a larger degree of conservatism than justified by current data uncertainties or are obviously unrealistic.
PRO: a.
Plants would no longer be limited in opera-tion by current ECCS rule restrictions.
b.
Research results would be reflected in the licensing process.
?
2-
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TherevisedECCSrulgmightnotcontain CON: a.
sufficient conservatism to account for calculational uncertainty.
Additional analyses would be required to demonstrate that sufficient conservatism remained in the calculation.
b.
The ECCS rule would have to be changed in the future to make use of research results from the 20/3D program or other information which may become available.
Alt. 0:
Eliminate the requirement to use Appendix K models and l-allow best estimate models to be used.
Reduced the l
2200 F and 17% oxidation limits of S 50.46 appropriately to ensure that sufficient conservatism exists to cover uncertainties in the best estimate calculation.
PRO: a.
Maximum use of completed research could be made in licensing to relax unnecessary operating restrictions.
l l
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"1,.S The Commissicners 10 b
b.
Licensing models would provide more realistic calculations to allow more accurate determination of the effect of equipment changes or failures and operat-ing procedures.
CON: a.
The conservatism used to account for uncer-tainties would be fixed through the revised S 50.46 limits and could not be varied to account for more accurate calculations of uncertainty which may be available in the future (i.e., little incentive for further
^
improvement).
b.
Additional staff resources would be required to establish fixed co'nservatisms applicable to all plant types.
c.
The elimination of Appendix K requirements and the introduction of best estimate models in the licensing process would represent a substantial change in the licensing process and may increase the risk of a legal chal-lenge.
l Alt. E:
Eliminate the requirement to use Appendix K models and allow best estimate models combined with an evaluation of the uncertainty in the overall calculation, similar i
to that discussed in SECY-83-472, to be used.
The S 50.46 limits of 2200 F and 17% oxidation would be unchanged.
l l
PRO: a.
Maximum use of completed and future research could be made in licensing to relax unneces-sary operating restrictions.
L.,S
}9:.
J.-.:
The Commissicners 11 b.
Licensing models would provide more realistic calculations to allow more accurate determina-tion of the effect of equipment changes or failures and operating procedures.
c.
The uncertainty evaluation would quantify the conservatism in the calculatichs which could change as the accuracy of the calcula-tions improved.
d.
The industry and NCR staff are already investing effort to follow this approach.
CON: The elimination of Appendix K requirements and the introduction of best estimate models into the licensing process would represent a substantial change in the licensing process and may increase the risk of a legal challenge.
In all alternatives considered, the current Appendix K would remain available for those applicants or licensees not desiring to use a revised evaluation model.
The staff believes that Alternatives A and B, which would provide little or no change in the ECCS rule, are unacceptable.
The ECCS rule should be changed because:
(1) A data base now exists that supports relaxation of the ECCS rule.
(2) A revised ECCS rule would remove unnecessary operating restrictions on plants.
i
-e
.n,,-._--_
y E
9 The Commissicners 12 (3) Almost all U.S. research on LOCAs has been completed. The remaining portions of the MIST, Semiscale and 2D/3D programs are expected to provide valuable information for assessment of models, but should not affect the proposed rule change.
(4) Nuclear reactor vendors are currently working on future plant designs which would be influenced by the revised ECC5 rule.
The staff has also considered Alternative C which would modify certain models in Appendix K, for which research investigations are completed.
The revisions considered under Alternative C include:
a.
Reanalysis requirements, b.
Post-CHF heat transfer, c.
Return to nucleate boiling, d.
Refill and reflood heat transfer (steam cooling below reflood rate of one inch per second),
e.
Fission product decay, f.
Metal-water reaction, and g.
Discharge model.
Based on recent supporting analyses performed by vendors and national laboratories, the staff has determined that if Appen-dix K were to be revised according to Alternative C, it is poss-ible that the remaining overall conservatism in the evaluation models would be on the same order or less than the uncertainty
,e H
[
The Commissitners 13 b
- f of the calculation, as discussed in Enclosure "C".
This would
^
be unacceptable since one could no longer assume that Appendix K contained sufficient conservatism to account for the total uncer-tainty in the calculation.
Thus, use of Alternative C without supporting oncertainty analysis was dropped from consideration.
The staff also considered revising Appendix K in a manner similar to Alternative C, but requiring an additional best estimate uncertainty analysis to ensure that the evaluation model contained sufficient conservatism.
This option would require two calculations, a best estimate with uncertainty analysis and an evaluation model calculation.
This option was also dropped from considerati.on since the licensee would be required to perform two calculations, of which one, the evaluation model, would provide little benefit to safety.
The staff has decided that, based on our current understanding of ECCS performance, the approach of a prescriptive Appendix K with fixed conservatism is no longer appropriate and that a best estimate calculation, taking into account the overall uncertainty in the analyses, is the correct approach to ensure the safety of the public without unnecessarily restricting applicants and licen-sees.
Thus, the staff has concluded that the ECCS rule should be revised accordingly and the requirement to use Appendix K eliminated. Therefore, Alternatives D and E were considered, both of which use best estimate calculations.
The difference between the alternatives is in the treatment of uncertainties.
Alternative D would reduce the S 50.46 limits of 2200 F and 17%
cladding oxidation to cover uncertainties in the calculation and the point at which substantial core damage would occur.
Alter-native E would require an uncertainty factor to be added to the l
best estimate calculation.
Alternative D was not selected because l
(1) the 6 50.46 limits of 2200*F and 17% cladding oxidation are believed to already be appropriate and conservative limits below which l
y
=
n.
FF The Commissitners 14 y
Vi substantial core damage will not occur, as discussed in I
Enclosure "C", (2) the conservatism used to account for uncer-tainties would be fixed and could not be varied to account for more accurate calculations, and (3) further staff analyses would be required to support establishing these limits.
The staff recommends that Alternative E be adopted.
This alter-native would require that the licensee show that the criteria of 5 50.46 are met using a best estimate calculation combined with an evaluation of the uncertainty of the overall calculation.
This uncertainty evaluation, combined with the low probability of a large LOCA and the additional conservatism in the 2200*F criteria, should ensure a negligible risk to the public.
This approach to licensing is consistent with the interim method discussed in SECY-83-472exceptthattheadditionalA9endixKcalculation, which contributes little to safety, would not be required.
Fur-ther, Appendix K would remain available (with minor modifica-tions) as an alternative.
Therefore, licensees and applicants which do not need nor desire relief from current operating restrictions would have no new requirements and could continue being licensed under the existing Appendix K requirements.
The burden of performing new calculations would only be placed on those applicants and licensees who elect to gain relief from LOCA restrictions.
The proposed ECCS rule change would also provide relief from the reanalysis requirements which do not contribute substantially to safety, as well as allow use of research data that has been obtained since the current rule was written.
The modification would allow applicants and licensees relief from unnecessary operational restrictions resulting from loss-of-coolant accident (LOCA) analyses and still result in an adequate level of conserva-tism in the ECCS analyses.
This revision in the analysis methods is expected to reduce the calculated peak clad temperature in a typical plant and allow an increase in the local peak power or
w z
=-
The Commissioners 15 e
total power, as discussed in Enclosure "C".
The additional flexibility could also be used to improve safety (e.g., reduce the neutron flux on the reactor vessel wall to alleviate the pressurized thermal shock problem) and improve the efficiency of reactor operations.
These changes would allow some plants to increase total power, improve fuel burnups, have longer fuel cycles, accommodate steam generator tube plugging and reduce equipment surveillance requirements, thereby reducing operating costs.
Enclosure "0" is a Regulatory Analysis describing these potential cost savings.
Resource NRC staff resources to implement the proposed rule change are Estimates:
thought to be negligible under the assumption that no unusual or special rulemaking procedures (e.g. hearings) will be estab-lished by the Commission.
IftheCommisponchoosestohold hearings, resources would have to be diverted from other high priority activities. Given that the rule is implemented, the impact of the changes on resource requirements will depend on the number of applicants or licensees which make use of the rule change.
The major staff resources required under the proposed rule change would be to review the best estimate calculation and l
uncertainty analysis required by the revised ECCS Rule.
Based on previous experience with the General Electric Co.
SAFER model and the learning that has resulted from these efforts, it j
is estimated that approximately one staff year would be required l-to review each generic model submitted. There are three major l
reactor vendors (GE already has a revised evaluation model approved under the existing Appendix K) and several fuel sup-pliers and utilities which perform their own analyses and poten-tially might submit generic models for review.
However, it is expected that only 2 or 3 generic models would be submitted since not all plants would benefit from the rule change.
Thus, about I
i
g _ _...... = _ _ _ _ -
I-The Commissicners 16 2-3 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 best estimate models in conjunction with the use of SECY-83-472.
Therefore, staff resources would be expended to review these models. 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 accurately estimate, we expect that the proposed rule change will have a small overall impact on NRC resources.
~
Recommendations:
That the Commission:
1.
Approvethepublicationofproposed; amendments,assetforth in Enclosure "E", which would allow certain changes and cor-rections to be made to the ECCS evaluation models without requiring an immediate reanalysis by applicants or licensees and would permit the use of best estimate calculations, along with uncertainty analyses, to be used in the evaluation of ECCS performance.
Evaluation models based on Appendix K features could also continue to be used.
I 2.
Note that:
a.
The notice of proposed rulemaking in Enclosure "E" will be published in the Federal Register, allowing 60 days l
for public comment, b.
Pursuant to S 51.21 and 51.31 of 10 CFR Part 51 of Part 51 of the Commission's regulations, a preliminary environmental assessment and finding of no significant impact is attached as Enclosure "F".
W.
Ilf The Commissi:ners 17 9
c.
Pursuant to the Regulatory Flexibility Act of 1980, s.
the proposed rule contains a statement that the Commis-sion 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 certifica-tion will be forwarded to the Chief Counsel for Advocacy, SBAbytheDivisionofRulesandRecords,ADy; d.
The subcommittee on Nuclear Regulation of the Senate Committee on Environment and Public Works, the Subcom-mittee on Energy and the Environment of the House Com-mittee on Interior and Insular Affairs, the Subcommittee on Energy Conservation and Power of the House Committee on Energy and Commerce, and the Subcommittee on Envir-onment, Energy and Natural Reseurces of the House Com-mittee on Government Operations will be informed.
e.
That a Regulatory Analysis is attached as Enclosure "D";
f.
A public announcement will be issued (Enclosure "G");
and g.
Copies of the Notice of Proposed Rulemaking will be distributed by the Office of Administration, Division 7
l of Technical Information and Document Control to each affected applicant, licensee, and other interested parties.
h.
The ACRS has been regularly consulted concerning this proposed rule change and has been provided this infor-mation for comment.
l l
l l
l
y
^5-SP?-
The commissicners 18 x
E-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 term.
William J. Dircks Executive Director for Operations
Enclosures:
"A" - Advanced Notice of Proposed Rulemaking, 12/06/78 "B" - Summary of ECCS Research "C" - Conservatism in Appendix K and 50.46 "D" - Regulatory Analysis "E" - Notice of Proposed Rulemaking "F" - Environmental Assessment "G" - Draft Public Announcement a
s' e
v.
l[
- ?
+
ENCLOSURE "A" a
ADVANCED NOTICE OF PROPOSED RULEMAKING DECEMBER 6, 1978 88 e
t O
9 0
9 m
4
.e ft0MHz alAaB S1187 S. Os mese ses47 en the madme gyygo et ds) and ser wa, safety evahsaaass eslumn, the 4th pernavaph. tale 2nd were verleved oc a ense 4prasase bests.
and 3rd aantences abould rund. "The ORMIM RMAATORY By int anuch angetin=nt rummarsh
]
maamasanas kniorinauen en the EDCA/EtX3 had 1rttness cased that Eraft pians to been obtained, and the Regulatory evert mila freom its peel sigply plant 188 8's em sol 1 anaff had noeutred estaessve esperi-I at Springftau, aso le as eenpool planta at Bentonytue. Dettyytile and gentesTIC utseesose or recouenose ase emer to its licensing revtew of over 34 i
Springfield bestaning in the faU wnutances racants BCC8 designs. IncJuded la this espert-i amonths of IMS Under the present once was the Intse amount of staff r
I eeder provisions, anilk diverted to Ben.
A"e**""o Crewee See &===esert Cne Case" time spent in todividual BCC5-related tenviDe and Bererv1Ut freen Spring 8'8 Sr**"* 8"' We'-W***4* ease sanse.'
bloensing rettess. In some snaca new h
8'""
evaluation anodels, assumptaceu, and field would be priced at Springfield be-eause these two locauona are within AagNCY: UA. Nuclear Regulatory parameters were peeposed for each
- 130 mDes of Springfield."
Cornmidon.
8"**==19e plant.
To alleviate this altunuon. an inter-
- 9. On page 84648 in the 1st solumn. ACTION. Advance nottee of proposed im Pobey Statement (IPSI prm6 ding the 4th paragraph should read. One ruinnams spectf6e guidance on ECC8 evaJustions of proponent's concerns was that snitt which is diverted to its planta at Sen-
SUMMARY
.The US Nuclear Regula and based on the then current state of f
tonytDe and Berryvtue be priced at the tory Commiasion is considering knowledge of IDCA/ECQ5 sas detel-emme level. The present prvvisions pro-amendlns its regulat6ons to change oped. The IPS sas lasued by the Com-
?
vide that smilk everted to nonpool sertain technical as seu as nontechnt. misalon lamediately effecure on June plants within 130 mues of Springfield es! requirementa tithin the existing
- 39. IMI. Hosever, fouestng pub 1.e er St. Emuts shall be priced at the emergency core ecolms system rule. comment, the Commission announced plant from wheen diverted Smee the Modifications under consideration its decialon on November 30.1911. to gentonytue and Serryvule planta are would take into acrount til espertence hold a rulemallas hearing to deter.
M within 130 miles of Springf6 eld, any gajned lo the tacerWng proosaa. tSi nes amine whether the IPS should be re-
=
aank diverted by Eraft frorn the researth informauon. and (31 operst. Sained as issued or whether different Springfida p? ant to such locations will ins empertence. This notice is to antite criteria should be adopted advice and recommendauons on sever-A rulemanins hearing. contened in receive the amme price."
- 10. In the middle column, the 1st al quesuona concerning the acceptance January Igi2. gett vated an estensive paragraph, the 15th and Stat lines. the artterta for emergency core cooling record of discuamics. and evaluation of worets. " producers" and " previous" erstems in light sater<ooled nuclear the available evndence (Le., esperimen-should replace. ' procedures" and " pro. power platita There till be a later op-tal results and analrucal unodelst per.
portunity for public comrnent in con tinent to 14CA/ECCS The complete enous". respecurely.
asetton with any proposed rules that heding record was cer* lined to the 1
- 11. ne sniddle column. the and een.
tence abound read. **1'he money is de.
may be hveloped by the Commission. Coramission for the Commianon's use In maktna its determinsuon of polley ducted by the anartet administrator in DATEEL Comment period espires Feb. en ECCS Based on this Neord, the the computauon of the blend price ruary 8.1979 Cometaalon Opinion of December 3s.
and is turned over to an agency com. ADDRMSM Intertsted perwu an 1973 (C1J.73 39. 8 AEC 10851 was posed of pro fducer representauves.1". km to submit en comments hed, prodom W baala for W 13 In the 3rd column. the and part. and suggesuons to the Secretary of ECCS rule 10 CPR 6 60 46. " Accept.
graph. the 1st sentence should read. the Commtaston. UA Nue! ear Regula-anee Criteria for Ernergency Core "At the request of the cooperative. a tory t'ammtanton. Washington. D C.
Cooling Systems for IJght Water Nu.
s representauve of the United Dairy In' 30655. Attention' Docketing and Serv-clear Poser Reactors" and Appendts dustry Aanociauon presented data in ice BrancA Copies of cornmenta re-E. "ECCS Evaluauon Modds." to 10 support of the coopersuve a position eetved by the Commisalon may be ea' CPR Part to (published in the Panam-that additional funds am mee&d bF amined in the Commtaslon's Public AL Ractersa January 4.1FI4).
Llw agency.
Document Room at 1717 H Street.
In Ita Opinion, the Commisalon care-
- 13. Alao, in the 3rd column. the 4th N.W., Washingon. D.C.
fully eensidered the el.frerent views FOR PURDEER DfFORMATION produced by the record and decided on f
that w h wa the criteria and the required and ac.
CONTACT-not opposed to advertastng and promo-eeptable features of the ECCS evalua-tJon he felt that the bisher ma=aa-Mr. James A. Norberg. Offler of taon snodels The Commtanton stated snent to producers would be paaned on Standards Deselopment. UA. Nucle-its bebel that the margin provided by to Eroger through the price that pro-sr Regulatory Comrniasion. Wash-these criteria and their inherent con.
ducers chargv Eroger for milk ".
Inston, D C. 30455 phone 301-443* aervative features would be adequate
- 14. On page 64649 in the 1st column.
Sett.
to easure core cooling should a design the 1st fun paragraph, the last een. SUPPLEMENTARY 1NFORMATION: basta lDCA evet oecur.
tence should read. "The funding rate one Commission in its Optalon aho adopted herein is in line with these Eastomiest BAeEGeotrND stated Ita intent to provide lautude for ease incnames for bewpapa and radio Emergency core ecoling systema change when new research informa.
adverusing.
(ECCSI were recogntr.ed in 1966 as Im.
tion became available.
15 Aho. In the 3rd full paragraph. portagengineered safety featurts for For sanny years, the Ceramtaalon tfje sth itne should read. aequal to.73 enttigattna the consequences of a tend its predecessor agency, the percent of the welshted...
postulated leap 4f. coolant. accident Atomic Energy Commtmalons have had IS. On page M660 in the salddle (IDCAI in light water. cooled nuclear programs of experimental and theo-oolumn, 31041.hd. 39 the word, power planta. During the period 1966 reucal research related to ECCS per.
" plan" should read. " plant".
Le 1911 entensive research programs formarxe. The rulemaking Opinion
- 17. Also, the last sentence in that vs.,e tatuated to better understand noted the onsonna research programs solumn th zuid road. " 3.
Section the 1DCA and eeneral comprehensive of the Commission and the nuclear in-1962.13 le rettsed to read as foUows.".
reviews were made to evaluate the ade-dustry and their potential for kn.
quacy of ECC3. ECCS dealsna submit-proved knowledge (8 AEC 1084.1988.
Samanas esers7ta, vot. es, eso. ass-wussenAT, gecamensa 4 ters
.mn og - e e
e v
[
57158 scoposso mutas 3404.1188.1130.11M.1130). The goe-eenaervause of the rule of the pro-3 and 4 above. ne resutu of m amoresment ernment and ladustry ECCS research paced Phase ) changes Concurrently, wtuch show that these chutes have a secu.
Seegmas tien produesd signLfkant the sta.ff le developing the methodolo. gente tapart en the overs 11 conservsuam of new informauen in the years slace the gy for nam *==fng the techructHy com. une rule will be snaar ersalab6e as part of "h' '"3' **AIA8 8'"
rulesnaking Purthermore. the Com-plex Phase 3 changes. It is espected amission has noguimd algnif6 cant espe-that the Phase I assessment wtD be yuast 3 (1450 Turas ) acLE CMANCts reence In using the ECCS rule. completed within als rnonths A ccm-aAsas on artw surosnL471on rkott aE-
-1 through !!d review of saany ECC8 pletion tame for the Phase 3 ma-esAncu aan orunatans sxrrattnet evaluation models and Mcensing of ment has not been estabushed but The foUowing char ges to the FCCS I)CCS designs.
this aanessment is expected to take
""*I""""
I#""
In order to utlhte mes teci.nical in-severaJ years to complete In each
~
i fh j
co der eC on 1
3 p the err s. en d
e sonaldering modifytr.g the EtlS rule me nt'
- 1. Rea' sed Isoessersos e, Pusten Prod.
r With both procedura; thanges and act Decey West Asia The changes saald Ln.
technical changes. The procedural Erseme Cossscenatross wohe ren.tns pararemph I A 4 of Appen&s changes provide improvernents to the E ehich na.s rr es a, heat senerstmn rate he foDosW aWie areas are froen ramoartive deca of a 3 umes the oc.
1 rule wh!ch aould elimir.ste ddficulties previously encountered in appl 3 ins under considerat!on by the Commts. sober ten A*ss Proposed standard' to an.
the rule De technical chanIes aould sfon for proposed niemaking-othn pcM dway hnt w mament with present knoe nedse Conasacrauen wiu he in the direction of impros1.ig the Pstatt I (6ssoar Tzmas) raoczscamoon. he then to the combination of uneena.ntles res! Lam of ECCS Deensing essfustion asrm Asro cur:sta srreme Tucups. In dersy heet wnh uncertatnues in inlual a
in the light of preser.t k not ledge-cat stu CMAnGE.s heat n't.
e while preserving a level of censerva.
t ruralee ortdation Rate The chanses tasm consistent tith that kriosletse.
1 Arenaleru Acentrements a Resmal6 u would twolte revisins Darnersph 1A8 of Acewemenu /be Ctrustn.tries Permit dp.
Apren&s K from the BakerJust equauen'
.x SennsAmt ce Peoroser RrLa Cnan:cs pricetiona chanses to no Cnt g sc se sound of adar los: to a calru:stian method based allom for eer1.ain correcuens to be made to en present knossedse and aseeed conserve.
The Commission is conAldering the,endor acCs computer annJ rsa enees tasm In adawr6 the basic performance re-intlation of rulemaking in tuo phases. dartna the construeuen peneet renes or guirement act forth in 3 60 44 t 3* that as fouost during corwtrucuen of the plant without a "the enaculated total osaasuce of the clad.
earnplete reanalys&s of ECCS tierformance eing sha 1 nethere esteed 0 If times the FEAaB $ (68toef TsAas) in compliance enh le CTR 9 to 34 untu the toLal claddans thickness befort oaJdauon" would maso be reesamined to ensure er+nstat.
Inluate rulemaking for procedure-e,s,,ersuna IAcenar rettee cruerts would be
, teed to erftne the bounds withLn wtuch encF with the new data en steensus and ettented and certain speelfac technical gne eery,etsons could be saepted without ducuaity of partialls esidised strenacy.
changes in the ECCS rule These rule plant spectric remnaJysia
- c. peditional Da2a ne chanses wiu in.
changes are espected to have little h Aesmateru Acewemnts fbe Operarent etude any changes to the ECCS rule needed kopact on the overall conaerentism of Lacesse applicesions and Licenard planta to take into amunt see Information that the rule. and such changes are anuel. The changes to 10 CTR l50 34 would das mdacanes the present rule is less conaerva-
"#"a"r o*f e'eners'ency'enciarit injecuan caused
""d
" # '"** *8'I'
- h N 8"f""
- 3'"3
- i pated to require a minimum of time Mo's'to'"the event of correeu"^on'a to vendor' del saan i
and effort to implement.
Ecm eoenputer anaJasta codes tf 8L to dem-by heat transfer to the ecotant from hot enstraM en a emne basis that W moeel M W W h fa m a W M bu m n W PEast 8 (1050 TEEat) shanges reduce the pena claMme temapers SWR ECCS cort spray.
In partDel with Phase 1, initiate de.
ture and Lt no change in plant techrutal
& Oppstins Erpresenca TW chanses wiu velopment of the beaes for a enor, spectf umare is teveed metuer ans newoes to w Ecm re 3 Actuem ao Nacleste So*h as The Peeded to eacunt for phenomena not epe-temprehergve rulefraking accon to ehanges would allow an eas unpuon of a enfertily identtfled at the Strae the rule was Incorporate new knowledge and oper-evturn to mucJeste totuna dartne the bloo.
prormulanted but that base etnoe been teen-ating esperience into the ECG rule. sown pl.aer of the tDCA shen supported tafled through operauns espertence as I
This effort would include manaming the br aselkaalt esta. T?us change wouJd in.
hartna a slanificant effect on ECG per-J knpact of proposed changes on the volve modifir nans to to cra Part so. As-formance Such rettsaons wiu te identified evers11 conservatism of the rule. As pendia E parsaraph I C e e ne oajecuve eurtne the developonent et the proposed part of the overtu *===== ment of con-of lats charue is to snow use of rewnt eaAa rule.
servatism, a systematic rettee of all en rmtuns During the detelopment of the pro-3 no as.,
rule changes an ma== ament ttu relevant information wiu be pet.
g p
fortned to ensure that it is approprt-emanese would delete the reguliement (Ap made of me impact of Ge proposed stely considered Ne information on pendia K. parserson ID 45 that heat tiana.
changes on the overaH conservatism of decay best and sirtonium-tster reac-fer enarulauons tw based on the assumption the ECCS rule.De impact aanessment
,?
Uon would be considered together that sooling to only by steam for flooding will include a reassessment of the re-with all other new information, includ-rotas below one Inch per aerond and reptace quirements presently specified in Ap-
?
Ing any adverse resulta sfor example, at enn a regairement that heat transfer esJ pendia E in light of current inform-diacrepannes in the pretest predaction culauons be tsaed on apeticable espertmen. Llon te.g., statistical combination of tal esta acoconnately enoununs for fnow heat sources) as well as consideration of significant research test results, un-bi'88 *4' Lf 8t l8 PN#5'd L* *tmr-of other phenomena of importance to certaintles namoclated with the predse.
d, to the transluon toutna earntat6on in fhans Nd eI.
C8 pruance that han been Uon of counter <urrent flow phenom
- el the ena and core sprey distribution, and ene the pons:buity of steam generator tube Append:s K. marsaraoh I C & wtth an im-
' Proposed American Nuclear Soem y 5-faUureal. If, during this review. It is proved reference in e later puburanaa by StandardWDecar Release Rates Feuertna determined that any information re-the same authors Shutdess of 17ranJum Pueled nermal Re.
quires more spectf6c treatment than is stems 3 and 3 above constitute eertain spe.
actors. approved 67 suticornmjuee ANS-8,
'f ***h''*J 88*^8 '* th' 8*ent rule. AJus standarda coaunsttee october len.
presendy proetoed in Appendia K of "*E**""
""' "s*e"s"do n'ot sta"ruficantti
'"' *
- 8 ' ' 'Ad' -
'Bahn.1. c. -mudas of mtaJ Water Re-In present licensing practices appro. este that these than acuens at ILah Temperstures. In Espers.
prtate rulemahtas acuon wiu b'e taken.
atfact W everall consensusm of the rule. unental and necevucar ph of m nr.
De Cornmtanton staff is presenuy nerefore. these chaners stu be mnstdered eenJum water Ranctor." 4 Mb.4643 pase T*
asseeming the impact on the overtu alans with the procedarsJ changes of 6tems naar spot.
~
fMIRAA teelt?M. WOL. 4, Is0. 336-treestoAY, peGaeam 6, Ifrt
. ww w w.
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k 8toPosso num mise tematafled almos the pr==ais=Ha= et Fer the Nasdear Regulatory Casa-the Dahed States law as aptoumatas Appendts E tea, asw esmaammae and pa-sa.
7, WT test semita, stansa ponerator SaarvuL J Cam.a.
Secuan 8 et the Ast regelrus eplo.
sube rupturea, asuntereurrent Osw ger,efert ef me Osamumen.
maue salsstons. = a=ahi=e of me-la==
phem=mana. WWR sere spray estrfts-their faannian and eenior effidala M EPR Dec. TE.3aets pued 13-4-Ts. tes asal the United Nauens who are entlued le tion. subcooled treak flowL A namehad.
elogy vul be developed for an==== lag amplamarie teamunity to have and maintala hability tasuranee against the impact of proposed emehawn
[gry m )
chanate on the escraD senservaussa of rtaka attains trean their operauon of the rule.
SEPARTMDfT of STATI smotor vehicles, eennela, or aartraft.
no Prealdent is erected to estabush Adetoe and r=aa-maaadatta-a en the Em 838 PAaf tell law requiresments for thh haMuty M.
proposed areas of rettsson to the ao.
eurance by regulauco. Ezeeuure 4
orpunce enterta for mergency sore speaket sea so-pol Order 13101 (43 FR H195) delegates 1
eso!1ng systems k Mght water <oooled
.uclear power planta are invited freen ObaW'tA30ST HAnstrY essWs ApeC3 PCs to the Secretary of State the authod.
ftC 8tnascent A88 Pm60esst 87 to prescribe these regulauona.
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& Bow should stets marense to eunau. et aD dipla-aHe amasalona mhesabers of j
and and how ena esemptame sniets matoins masalone and their famnian and offl. Ess-8 88 '
88 ID"'""*'e't" tsaaremme esserese but he seseaneet ainas of the United Mauoos seuued to
& what pher===aa have home lesettned eplomaue haununity, including the SA&J es.
' sense prenansaume et uhe 30C5 rule that Mm!ts a habiuty. and descHbe the evt.
are sasntnmat to acts performanse ans drnce af GDeumnor nemmanry before 33 giaalattaa thafu fer anotar enharte gn.
that are est sessentatt emmsedered as the aussuns BCCE rule, en asht et surrent the Department of State endorses ap.
- arease.
taowtsese and espertsame er as sevent u.
Pasennaaa for diplomaue automobue U.
SSLA==== amend Damas for amater eeN.
esmsene erneusest esame plates er ts-p* Lama fream regte.
age tasurance.
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1st e Authertmed kuwer.
the esclussen et are sammarch haformause 141.T PeDer series omnsestant stub the Act, DAN Written mammanta must be 181J Briermee et hauurnace for molar eeN.
and apartaams esportsmar? Cha er should this to eene en a esmunuais taatst Bow received by February 3.1979. In add!.
etua j
eheuW prortstas et emsspiame marssas he toon, kiterested persons snar efter talJ Evidense of huwanes reeutrod for baneed to such a present comments orally at a puhuc moeung "P-am alcenne pentes and waiver et 2
i to be held at le s.aa, February 8.1979 ggg74, m g f
he th=l-saa has sooeluded pre.
g, at Recon 1913. Department of State, erasn and/w emmeem MaunarQF that the NWd 3301 C Street N.W. Waahingsm,e D.C.
lat.at_asunnus,e et o.w,nere.eds, ma.amse.
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ADDRESS: Written comments and UAC. Meet Eo. tales sea Fet telesL ren and wSI not re ihe p-pa,au-
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tempact statement pursuant to Secues sent. David P. Stewart. W As.
ar akHakaa regulauens re.
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103(3 NC) of the Nauomal Envtrensman. assunt. Offlee of the Emeal Adviser.
L 4 of the Dtplo, tal Foucy Act of 1980 (NEPA).
5toom 6433.3301 C Street NW., Wash.
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In view of the uncertainty regarding Inston. D.C. 30430.
M ML These regulations require aD a ho6ona, ehembers of missions and I
the poes!ble tachnical changes to the M M INMMATION SCG rule based on new informauen their fam1Ues, and those efficials of U
from research and operaung expert.
the Walted Nadons who are enUUed to enee (Phase S t.
the rhml== san David P. Stewart. Department of diplounaus kamunity to have and cannot snake a determinadon at this State. Washington D.C., telephone smaintain liab(uty kasuranos against thee concerning the posalble need for 303-433-3168.
the rteks of bodily injury, including an emetroruneetal kapact statement. SUrrimtnfTARY INFORMATION: death, and property damage, laeludtag Any proposal for rutamaking acuoc he Diploenaue Retsuons Act. Pub.1.
thus of use, artsing from the owner.
along these linse wC1 include a Coma.95-393. Sepumber 30. Ip18 (33 UA C.
ship. maintenance, or name in the mission deteretnattan whether or not S H a et see SS DA C.
1944 wug United States of any ashotor sehicle, j
haan=a effecure December 29, 1978. Teesel. et aircraft.
an sottrotunental amapare ps a t ament should be prepared for that acuan.
As of that date, previous statutes se I ' O m.
aplounaue hamualty daung frcen the i
to cae. sat ase. lat. puk 1. es.ses, es menteenth eentury win be repealed 1
sneans the Depwaii Re.
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""*8" etsions of the 1961 Ynenna Convenuce 54a 98 eso,30 U AC.1304).
I Dated at Waahington. D.C., this se Dtplaniaue PLelauens (33 UET 3337 (b) *' Persons subject to the Aeta j
SOLb day of Novesaber 1978, 000 UNT18 M); wul be estab!! abed as aqsene the -hase of misalons vtbe fumA& 000s575, vet,41. see as6-essesseAY Becamme e, ters d
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6 E' CLOSURE "E"
(,,..
m
w j
i L
SUPMARY OF ECCS RESEARCH In the ten years following the ECCS rulemaking, over $700M has been spent by the NRC on research investigating ECCS performance.
It is estimated that a similar amount has been spent by DOE (including AEC and ERDA), the U.S. indus-try and foreign researchers; resulting in a total estimated expenditure of
- .A, $1.5 billion.
The majority of this LOCA research is complete and has greatly evee improved the understanding of ECCS performance during a LOCA.
The methods speci-fied in Appendix K are now known to be highly conservative; that is, the actual temperatures during a LOCA would be much less than the temperatures calculated using Appendix K methods.
This fact is best illustrated by comparisons showing temperatures during LOCA simulations in LOFT which are more than 600*F lower than calculations performed using Appendix K procedures as shown in Figure 1.
The ECCS research has gone far beyond simply showing that Appendix K is conser-vative, however, and has allowed quantification of that conservatism.
The results of experiments, computer code development, and code assessment now allow more accurate calculations of ECCS performance during ad.0CA than is possible J
using Appendix K procedures, along with reasonable e.;timates of uncertainty.
In order to highlight the contributions of research toward improving understand-ing of LOCA phenomena, it is useful to discuss two categories of research.
The first category of research included phenomena which are specifically addressed in Appendix K and for which Appendix K requires specific calculational models to be used.
Examples include decay heat, metal-water reaction rate, discharge model, reflood heat transfer at low reflood rates, and other heat transfer pheno-In some cases research has shown the required models to be inadequate mena.
(e.g., discharge model) and in one case nonconservative.
However, most of the requirements of Appendix K have been shown to be more conservative than needed to protect public health and safety.
The best example is the decay heat calcu-lation which is now known to be conservative by over 20% during the initial phases of a LOCA.
This category of research has been highlighted because it represents completed research that can not be used in licensing until the ECCS rule is modified.
The second category of research was more general and was directed toward a best estimate understanding of the overall performance of ECCS.
This research includes 1
Enclosure "B"
T g_
.y e:
e W
2500 I
I I
I 1 DATA 2 NRC 2000 3 EXION 4 WEST b
1500
)!
f1000 [
7 F
l i 'j 500
+
I I
I l
O 0
25 50 75 100 125 T=.
M e.-
m,e. c 3
l l
Figure 1.
Comparison of cladding temperatures measured during LOFT test L2-3 with calculations using several vendor evaluation models and an NRC calculation using conservative Appendix K assumptions.
The large differerce between the data and the calculations illustrate the conservatism of these Appendix K models.
(EGG-CAAP-5134) 2 Enclosure "B"
w 7-L; y
many inv;stigaticns into spicific phenomena in the arcas of heat transfer and two phase flow. Many facilities such as the THTF, FLECHT and the foreign CCTF and SCTF facilities have been used to obtain specific heat transfer data.
This
^
research data has been used to develop major new computer codes such as TRAC and RELAP5 which account for complex phenomena such as non equilibrium and multidimensional effects and provide an ability to perform best estimate calcu-lations of ECCS performance.
In addition, major integral test facilities such as LOFT, Semiscale, TLTA, MIST, FIST and a number of foreign facilities provide complete simulations of LOCA and other transients for comparison with the calcu-lations of the new computer codes.
This allows assessment of the overall accu-racy of the calculations and identification of needed improvements.
This research now allows significantly improved calculations of ECCS performance over those possible when Appendix K was developed.
9 4
3 Enclosure "B"
E
?
RESEARCH DIRECTLY APPLICABLE TO APPENDIX K FEATURES I
POST-CHF HEAT TRANSFER [ Appendix K, I.C.5.b]
)
r se The McDonough paper referenced in Appendix K presents heat flux results obtained at one pressure point, 2,000 psig.
Application of boiling predic-tions at lower pressures results in inaccurately (possibly non-conservative) calculated cladding surface temperatures.
The extended pressure range corre-lation, given in a second reference,(1) provides a more realistic indication of fuel cladding temperature for transition boiling at low system pressures.
Research performed since the writing of Appendix K has also shown that the Dougall-Rohsenow correlation, which is specifically listed as suitable in Appendix K, may be nonconservative under certain conditions.(2,3) Thus, it is desirable to delete this reference as an acceptable post-CHF heat transfer
^
correlation from Appendix K.
Extensive post-CHF heat transfer research has been performed since g
Appendix K was written which is now available for use irt analyses of LOCA.
This receht research was reviewed and summarized at a recent meeting in Salt Lake City.(4) 1.
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).
2.
G. L. Yoder et al., " Dispersed Flow Film Boiling in Rod Bundle Geometry-Steady State Heat Transfer Data and Correlation Comparisons," NUREG/
CR-2435, March 1982.
3.
D. G. Morris et al., "An Analysis of Transient Film Boiling of High-Pressure Water in a Rod Bundle," NUREG/CR-2469, March 1982.
1 4.
"The First International Workshop on Fundamental Aspects of Post-Dryout Heat Transfer," NUREG/CP-0060, December 1984.
4 Enclosure "B"
g___-
n Tvt CRITICAL HEAT FLUX (Appendix K, I.C.4 e)
This criteria require that only transition or film boiling heat transfer x
correlations be used once CHF is predicted.
For a large break LOCA, CHF is usually predicted to occur during the initial blowdown flow reversal as a result of momentary low flow during the reversal.
This occurs very soon after the break is initiated.
Experimental data, however, have demonstrated that for certain break sizes and fluid conditions, cladding rewet actually occurs during the blowdown as shown in Figure 2.
Therefore the heat transfer conditions which exist are characteristic of nucleate boiling.
This has been clearly supported by the nuclear bundle tests performed in the Loss of Fluid Test (LOFT) facil-ity(1,2) and Power Burst Facility (PBF)(2) ; by various non-nuclear tests per-formed at Two Loop Test Apparatus (TLTA)(3), Thermal Hydraulic Test Facility (THTF)(4) and Semiscale(2); and the numerous data presented at the Denver Work-shop on Rewet Phenomena (5),
The use of experimental data to justify a return to nucleate boiling duringtheblowdownisnowwarrantedinviewofthelar{eexperimentaldata base showing that this phenomenon exists.
Furthermore, the past Commission policy has generally been to allow the use of valid applicable data as seen in other parts of section I.C.4.
1.
Paul D. Bayless and Janice M. Divine, " EXPERIMENTAL DATA REPORT FOR LOFT LARGE BREAK LOSS OF COOLANT EXPERIMENT L2-5," NUREG/CR-2826, August 1982.
2.
A.M. Eaton & E.L. Tolman, "A Summary and Assessment of Return to Nucleate Boiling Phenomena During Blowdown Tests Conducted at the INEL," EGG-LOFT-5127, April 1980.
~3.
L.S. Lee, et al., "BWR LARGE BREAK SIMULATION TESTS - BWR BLOWDOWN /
EMERGENCY CORE COOLING PROGRAM," NUREG/CR-2229, April 1982.
4.
W.G. Craddick, et al., "PWR BLOWDOWN HEAT TRANSFER SEPARATE-EFFECTS PROGRAM DATA EVALUATION REPORT - HEAT TRANSFER FOR THTF TEST SERIES 100,"
NUREG/CR-0105, September 1978.
5.
MINUTES OF THE DENVER WORKSHOP ON REWET PHENOMENA, April 11-12, 1979, as compiled by L.B. Thompson and Y.Y. Hsu.
Transmitted to attendees by May 15, 1979 letter of L.S. Tong.
Available in U.S. Nuclear Regulatory Public Document Room.
5 Enclosure "B"
i i
900.
FUEL ASSEMBLY 6 (Hl5-261 FUEL ASSEMBLY 5 (H7-26)
FUEL ASSEMBLY 9 (H15-263 B00.
r_ _::: q
,/
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700.
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V i
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(
I
)
600.
g 4
i W
4 500.
os 400.
O.0 2.0 4.0 6.0 8.0 10.0 TIME AFTER RUPTURE (s)
E n
o E
2 Figure 2.
Peak cladding temperature histories for LOFT test L2-2 showing the cladding rewet which occurred after the initial CHF. (LTS 20-99)
=
- =
u T
E C
1 REFILL AND REFLOOD HEAT TRANSFER FOR PRESSURIZED WATER REACTORS 1
(Appendix K, I.D.5)
The present Appendix K requirement for an instantaneous switch from two-phase flow at core reflood rates above one inch per second to steam flow only at core reflood rates below one inch per second is not supported by experimental data obtained from the cooperatively funded Full Length Emergency Cooling Heat Transfer (FLECHT) program or data from the international 2D/3D program initiated after the rule making hearing. The intent of the Commission when it promulgated the rule was to ccnservatively bound the reflood heat transfer in calculations at core reflooding rates of less than one inch per second and in the presence of blockage.
This was primarily due to the fact that a very small number of applicable data with core reflooding rates of less than one inch per second were available at that time.
In practice, the requirement to switch to steam cooling only has produced difficulty in making mechanistic models account for this non physical transition.
Earlyeffortstomakethemodelsswitchtosteamcoolinqonlyresultedinsuch artificially high steam flows that the calculated steam cooling was higher than would actually occur with two phase flow.
To correct this problem, the heat transfer had to be artificially lowered in the models.
Since then, a large data base on core reflood rates of less than one inch per second for both unblocked and blocked core geometries (especially the FLECHT Program, references 1 to 6) has become available.
Experimental data and visual observations have shown that the fluid state in the hot bundle is a non-equilibrium mixture of superheated steam and liquid droplets.
Hence, pure steam cooling has not been observed during reflood for reflooding rates less than one inch per second.
Furthermore, core reflooding rates near or below one inch per second did not effect reflood heat transfer in any unexpected way and no particular significance can be associated with a reflooding rate of one inch per second or below.
The data base, as illustrated in Figure 3, shows a smooth transition in heat transfer from reflooding rates above one inch per l
second to those below one inch per second.
The FLECHT results are also sup-i ported by findings from the 2D/3D program.
The large scale data obtained from the Cylindrical Core Test Facility (CCTF) and Slab Core Test Facility (SCTF),
for reflood rates in the range of 0.8 inches per second, indicate that the heat transfer coefficients during reflood are typically an order of magnitude greater 7
Enclosure "B"
i--
T o
y-
[
FLECHT SEASET Peak power = 0.7 kW/ft Pressure = 40 psia 50 Flooding rate (in/s) 0.81 0
0.97 0
1.58 A
A
/
40
/
I I
I I
I 1
o i
I g
i I
I l
30 f,
l j
I b
e
/
,/
Lax
/
G Q
,/
,f
/
z 20 9'
/
o A'
/
w
/
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/
/
g'a#
D "A 10 G -D,','
,o
/*
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/ /
///
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0 0
100 200 300 400 500 t(s)
Figure 3.
FLECHT SEASET data show no unusual behavior below reflood rates of opp inch per second that justify special treatment required by Aphndix K.
8 Enclosure "B"
I#
than that which would occur with pure steam cooling only( -15)
This increase in heat transfer appears to be caused by significant amounts of liquid entrained into upper parts of the core.
The liquid fraction above the quench front is typically in the range of 10-15% rather than 0-1% previously estimated.
Since a large data base exists to eliminate the need for conservatively bounding the reflood heat transfer coefficient for core refloodin0 rates less than one inch per second, there is no need for a distinction between heat transfer calculations for reflooding rates greater than, equal to or less than one inch per second.
The use of reflood heat transfer correlations based on applicable experimental data or alternative models that result in reflood heat transfer coefficients which do not exceed those obtained from applicable experi-mental data for both blocked and unblocked conditions would be acceptable and, in fact, would be consistent with current practice.
r 1.
"PWR FLECHT COSINE Low Flooding Rate Test Series Evaluation Report,"
Westinghouse Report WCAP-8838, March 1977.
2.
"PWR FLECHT Skewed Profile Low Flooding Rate Test Series Evaluation Report," Westinghouse Report WCAP-9183, November 1977.
3.
"PWR FLECHT-SEASET Unblocked Bundle, Forced and Gravity Reflood Task:
Data Evaluation and Analysis Report", NUREG/CR-2256, February 1982.
4.
"PWR FLECHT-SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Report, Vol. 1 & 2," NUREG/CR-1532, June 1980.
5.
"PWR FLECHT-SEASET #1 Rod Bundle flow Blockage Task Data and Analysis Report, Vol. 1 & 2, NUREG/CR-2444, September 1982.
6.
"PWR FLETCH SEASET 163-Rod Bundle Flow Blockage, Task Data Report,"
NUREG/CR-3314, October 1983.
~.
Hiromichi Adachi et al., "SCTF Core-I Reflood Test Results," Japan Atomic 7
Energy Research Institute, Eleventh Water Reactor Safety Research Infor-mation Meeting, NUREG/CP-0048, January 1984.
8.
"SCTF Core-I Test Results (System Pressure Effects on Reflooding Phenomena)",
JAERI-M 82-075, July 1982.
9.
" Droplets Flow and Heat Transfer at Top Region of Core in Reflood Phase",
JAERI-M 83-022, February 1983.
10.
" Evaluation Report on CCTF Core-I Reflood Test CI-19 (Run 38) - Experimental Assessment of the Evaluation Model for the Safety Analysis on the Reflood Phase of a PWR LOCA", JAERI-M 83-029, February 1983.
l 9
Enclosure "B" l
t xy I
11.
" Examination of Repeatability in Reflood Phenomena Under Forced Flooding in SCTF Core-I Tests", JAERI-M 83-237, January 1984.
12.
" Evaluation of CCTF Core-II Second Acceptance Tests C2-AC2 (Run 052) -
Investigation of Difference in Reflooding Behaviors between Core-I and Core-II Facilities", JAERI-M 84-036, March 1984.
13.
Y. Murao et al., " Status of CCTF Test Program", Eleventh Water Reactor Safety Research Information Meeting, NBS, Gaithersburg, MD, October 24-28, 1983.
14.
H. Adachi et al., "SCTF Core-I Reflood Test Results", Eleventh Water Reactor Safety Research Information Meeting, NBS, Gaithersburg, MD, October 24-28, 1983.
15.
Y. Murao et al., " Status of CCTF/SCTF Test Programs", Twelfth Water Reactor Safety Research Information Meeting, NBS, Gaithersburg, MD, October 22-26, 1984.
FISSION PRODUCT DECAY (Appendix K, I.A.4)
The existing ANS 1971 decay heat standard required by Appendix K is a simplified calculation which was based upon the assumption that fission products of all fissioned nuclides produced the same decay power as those products of U235 fissions.
It was alsc based upon a simplified view of several other phe-nomena.
The infinite operating time assumption and the 1.2 multiplier required by the existing rule were used to bound all the uncertainties including:
(1) experimental uncertainty, (2) uncertainty in the actual fuel nuclide content, (3) uncertainty due to simplification of the phenomena, and (4) operating i
history.
In contrast, the 1979 standard is a more detailed and complete calcu-lation which includes specific calculations of the contribution of three fission-able nuclides (U235, U238, and Pu239) based on the actual content of the fuel.
The contribution of fission product neutron capture is also included in the l
1979 standard. While Appendix K currently requires the use of an infinite operating history with the 1971 standard, it is desirable to use the actual operating history with the 1979 standard.
The 1979 standard also includes uncertainty estimates for each contributor to the decay heat.
Uncertainty upper limits can, therefore, be determined for each specific case and as a function of time, instead of including one large uncertainty covering all cases as was accomplished by using the 1.2 multiplier in the existing rule.
10 Enclosure "B"
i sf Y
The 1979 standard, with its more accurate calculation and reduced uncer-tainty, was developed using an extensive data base developed through research programs sponsored by NRC, ERDA, DOE, and EPRI. The new data base includes the results of five experimental studies on the short term decay of U235 with much 4
smaller uncertainties; new data for Pu239; new summation calculations for U235, U238, and Pu239 using an improved data base and uncertainty evaluation; and summation calculations of neutron capture effects. The foreword to the 1979 standard (I) provides a good bibliography and summary of this research.
Other I3) summaries include RIL #8(2), a review by Shrock
, and the proceedings of an American Nuclear Society conference (4).
The effect of using the 1979 standard varies depending on the operating time and the actual nuclide content.
Using typical BWR nuclide contents, the 1979 standard yields decay heat production 5 to 10% lower than the 1971 stand-ard, depending on the operating time and the time after shutdown. When compar-ing the 1971 standard using an infinite operating time and the 1.2 multiplier to the 1979 standard using the actual operating time and the two standard devia-tionupperbound,asshowninFigure4,the1979standayyieldsdecayheat values 10 to 30% lower.
While it was generally agreed that the 1971 standard was conservative when Appendix K was written, this new research was required to determine the accuracy and improved uncertainty estimates of the 1979 standard.
This new data base and the calculational method of the 1979 standard has had extensive review by the technical community through the American Nuclear Society and the American National Standards Institute, by the NRC and NRC sponsored research programs, and by the Electric Power Research Institute.
1.
American Nuclear Society Standard, ANSI /ANS-5.1-1979, "American National Standard for Decay Heat Power in Light Water Reactors," August 1979.
2.
Ltr. S. Levine to B.C. Rusche and R.B. Minogue, "Research Information Letter (RIL)-8, Decay Heat Data Applicable to LOCA Evaluation," January 31, 1977.
11 Enclosure "B"
- s m:+ s Q r
13 -
12 -
11 -
-ANS-5.1 + 20%
n19 AwS S. t 34eyeusikt>998 Standard g g _*
-Note-i i
4
(
Uncertainty Bars are Sa ed
~.
8-MeV/F l
7-
./-
-1973 ANS-5.1 Draft Standard 4-
-NOTE-Shaded area is + 20% - 40% to IOOos; + 10%, - 20% > 1000s 8-M 4-2-
PANS-5.1 + 10%
2-
)
3-I y I I I III[
I I I I I I II]
I I
I I IIIIl I
I I I IIII s
108 10s Its 10' le E3 Cooling Time (s) 8 C
2 Figure 4.
Comparison of the revised 1979 ANS decay heat standard with l
the 1973 ANS standard nlus 20% required by Appendix K.
=
- =
p-ik 5
3.
V.E. Schrock, " Evaluation of Decay Heating in Shutdown Reactors," PROG. IN
[y NUCLEAR ENERGY, 3, 125-156, 1979.
1 4.
Proceedings of the American Nuclear Society Topical Conference on Thermal Reactor Safety, Session on Decay Heat, Sun Valley, Idaho, August 1977.
METAL-WATER REACTION RATE (Appendix K, I. A.5)
At the time that the existing rule was written, available data concerning
^
zirconium oxidation was very limited.
The Baker-Just equation (1) was obtained by making use of their own data at high temperatures and the available data of others at low temperatures. The data of Baker and Just can be questioned for a number of reasons, but was used in Appendix K because it was believed to be conservative at high temperatures and represented the only data available at that time.
Since that time, a large amount of data concerning zirconium oxidation has g
become available, as illustrated in Figure 5. The AEC/NRC sponsored a program at the Oak Ridge National Laboratory from 1973 to 1977 Which resulted in the Cathcart-Pawel equation.
This research is reported in reference 2 and summarized in RIL #9(3)
Research on zirconium oxidation has also been conducted indepen-dently by the Federal Republic of Germany, the Japanese and by the Electric Power Research Institute.
These data were obtained by a variety of experimental techniques and using different analytical methods. When placed on a common basis, however, these data are in good agreement with the Cathcart-Pawel equa-tion.
Reference 4 provides a good technical summary of the large zirconium oxidation data base that currently exists. The Cathcart-Pawel equation indicates that the reaction rates are only 76% of those indicated by the Baker-Just equa-tion at 2200'F.
At 1600*F, the two equations cross and the Cathcart-Pawel equa-l tion falls above the Baker-Just equation, giving reaction rates that may be too high. This temperature range is below the range of conditions tested in the
~
Cathcart-Pawel study, however, and is below the temperatures where significant oxidation occurs.
Appendix K was written at a time when the zirconium oxidation data base was very limited and the Baker-Just equation was chosen to be very conservative to reflect the lack of data.
It was the intent at that time that the data base would be improved and that Appendix K be modified at a later date to reflect improved understanding of zironium oxidation rates.
This research has now been 13 Enclosure "B"
Sr.
N t.7 y
+
r e
s-TEMPERATURE l*Cl 1500 1400 1300
$200 1100 1000
~
-5 c Arne a br. heuti-NA 5
BAMER-JUST
=
\\
LEMMON E,
\\
BIEDERMAN s
b
\\
-o
_A 2
g 3,
g
\\
. \\,.
a e
g 8 10
^
\\
u
' N s
z x
s W
- 5 s
E 5
' b N g
% \\.
.\\
.)
y 2
rr h
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ss W
\\' h s w.
a:
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9
\\ \\.\\
y XVy a
\\'
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~5h '
16 5.5 6.0 6.5 7.0 7.5 8.0 tO OOCVT(*K) h Figure 5.
comparison of metal-water oxidation rate constants determined by Cathcart-Pawel with the data of other investigators and the Baker-Just equation required by Appendix K.
(ORNL/NUREG-17) 14 Enclosure "B"
w f..
completed and a very large and consistent data base is now available.
This T
improved understanding of zirconium oxidation rates is reflected in the Cathcart-Pawel equation and it is appropriate to incorporate it in licensing n
calculations.
1.
L. Baker and L.C. Just, " Studies of Metal-Water Reactions at High Temperatures:
III.
Experimental and Theoretical Studies of Zirconium-Water Reaction," ANL-6548,1962.
2.
J.V. Cathcart, R.E. Pawel, et al. " Zirconium Metal-Water Oxidation Kinetics:
IV.
Reaction Rate Studies, "0RNL/NUREG-17, August 1977.
3.
Ltr. S. Levine to B.C. Rusche and R.B. Minogue, "Research Information Letter (RIL) No. 9, High Temperature Oxidation of Zircaloy Fuel Cladding in Steam," March 14, 1977.
4.
Ltr. M.L. Picklesimer to M. Silberburg, " State-Of-The-Art Review of The
~
Oxidation of Zircaloy by Steam to 1700 F)," February 1,1982.
a J
DISCHARGE MODEL (Appendix K, I.C.1.b)
Since the publication of the ECCS Rule in 10 CFR 50.46 and Appendix K, a considerable number of theoretical and experimental studies of critical flow discharge from a break or leak in the primary system of light water reactors have been reported.
The results of these studies indicate that during the early stages of blowdown, the water is first subcooled, then saturated, and during depressurization it may become superheated and result in substantial thermal non-equilibrium.
The mass flux of discharge, in general, depends upon the break configuration (e.g., entrance geometry, pipe length, pipe diameter and length "to diameter ratio), the upstream pressure, and thermodynamic properties (e.g.,
thermal non-equilibrium phenomena) of the coolant, and it tends to be choked and reaches a maximum critical flow rate value.
The Moody (1) model required by Appendix K, which is based on a non-homogen-eous equilibrium (or equilibrium slip) model, is not an adequate flow discharge model as shown in Figure 6.
Reviews done by Saha(2) and Abdollahian et al.( )
revealed that theoretical treatments of the Moody model are not completely cor-rect, and the experimental data on relative velocities show much lower values than those predicted with the Moody model.
Like the homogeneous-equilibrium l
t l
15 Enclosure "B" i
(
m m,yosf c,
i I
i, j
i I
5 10-I
'Q -
p = 20 bor sss
\\s\\
l 42 '/. A Experiment j
\\
o
\\\\
HENRY FAUSKE Model j
\\,
MOODY Model
.~.
\\
s 3 2.510'-
i s
s o
i
\\
\\
4
\\
s i
o i
g\\
\\
s \\
'Q,__
-a.,_C':
i
- :. 2.%.
t 0
.=
-40
-20 0
20 40 60 m
8 AT(K)
X ( */. )
C E
2 Figure 6.
Comparison of critical flow models with experimental data shows that the Moody model required by Appendix K is inadequate.
=
"=
E I
D model, the Moody model also can not predict the critical flow rate data for short pipes.
~
In other words, effects of thermal non-equilibrium for short pipes were not taken into account by Moody.
It should be pointed out that the pipe length (L), pipe diameter (D), and/or L/D are important in determining the effects of thermal non-equilibrium.
Recent Marviken Full-Scale Critical Flow Test (4) data did show that (1) the critical mass flux is not only dependent on nozzle length but also on nozzle diameter since the nozzle mass flux increased with decreasing diameter, and (2) when L/D is less than 1.5, the nozzle mass fluxes seem to be dependent on the L/D ratio.
Based on the above discussions, a more general model or correlation for critical flow than the Moody model should be used for licensing calculations.
1.
Moody, F.J., " Maximum Discharge Rate of Liquid-Vapor Mixtures from
^
Vessels," ASME Symposium Vol. 1 Non-Equilibrium Two Phase Flows, 1975, pp. 27-36.
g 2.
Saha, P., "A Review of Two-Phase Steam-Water tritical Flow Models with Emphasis on Thermal Non-Equilibrium," NUREG/CR-0417, September 1978.
3.
Abdollahian, D. et. al., " Critical Flow Data Review and Analysis:
Final Report," 5. Levy, Inc., Report SLI-8103, March 1981.
4.
"The Marviken Full Scale Critical Flow Tests, Volume 1:
Summary Report,"
EPRI Report NP-2370, Vol. 1, Research Project 956-1 Marviken MXC-301, May 1982.
l l
i l
i l
l 17 Enclosure "B" l
l l
l
y h
GENERAL ECCS RESEARCH The previously discussed research was highlighted because Appendix K specifically addresses these phenomena.
Further, the evaluation models can not currently make use of the research results in these areas because Appen-dix K prescribes how these phenomena must be calculated in approved evalua-tion models.
The research performed since 1974 has been much broader, however, and has been directed toward an overall understanding of all phenomena important to ECCS performance.
Much of this research has already been introducgd in licensing in areas where Appendix K does not specify the use of specific models.
In addition to the decay heat and metal-water reaction rate research pre-viously discussed, other important fuel research has been performed.
Research on material properties and fuel-cladding gap conductance has greatly improved the understanding and ability to calculate the initial stored energy in the fuel rods and the rapidity with which that energy escapes during the initial phases of the large-break LOCA as shown in Figure 7.
T9seresultsareless important for current evaluation models since Appendix K calculations result in peak temperatures late in the reflood phase of the transient when initial con-ditions are less important. We now, however, have abundant evidence that the peak temperatures during a large-break LOCA could actually occur during the first few seconds of the blowdown and are highly dependent on initial fuel con-ditions.
This fuel research has been extremely valuable in improving our abil-ity to accurately calculate these initial conditions and the actual peak temper-atures which would occur during a large LOCA.
These research results are used in the FRAP computer code series (1,la,2,3) used to determine initial fuel condi-j tions and fuel behavior during the transient.
These models are used to bench-mark more simple fuel models used in thermal hydraulic computer codes discussed below.
Early research in the area of thermal hydraulics during a LOCA concentrated on heat transfer during specific phases of the transient.
The Two-Loop Test l
Apparatus (TLTA)4 and the Thermal Hydraulic Test Facility (THTF)5,6 were separate effects tests facilities which investigated heat transfer during the blowdown phase of the BWR and PWR LOCA, respectively, under temperature and t
pressure conditions similar to an actual reactor. A major finding of these l
programs was that departure from nucleate boiling would indeed occur very early l
l 18 Enclosure "B" l
l l
E w
0-Ly 0
'l.
L..
GENERIC BWR COMPARISON AT 32 kW/m, EOL s.s ",
ST1
{
j j,
s.o 3,
N y
s\\g 42 E N
=.
34 19oo g
N a
r s
a.s 3
\\
=
\\\\
14oc 9.
g FAAPCON 2
\\
(FAACAS)
\\
\\
\\
toco i
I i
I t
I t
I t
no ao so so 100
% OF FUEL R ADeus l
l i
Figure 7.
Comparison of rod temperature profiles and stored energy calculated with 1973 models (G-T-1) versus FRAPCON predic-l tions.
The lower stored energy of the later calculations is due to better calculations of the effects of fuel relocation.
l 19 Enclosure "B" l
l
>e
'61 in the transient cs essumed in Appendix K models.
Howevsr, as previously J
F discussed, rewets, which are not allowed in Appendix K calculations, also occur-red after the initial flow reversal resulting in better than expected cooling
[
of the rods.
Research in the Full Length Emergency Core Cooling Heat Transfer (FLECHT) facility investigated heat transfer during the reflood phase of the LOCA transient.(7,8) A key recent finding is that flow blockages due to clad swelling, of which Appendix K requires consideration, do not degrade heat trans-fer during reflood.(,9a) Much of the results of the FLECHT program have been used in evaluation models, with the exception of low reflood rate research which is specifically excluded from evaluation models by Appendix K.
ECCS Bypass research has also been performed which investigated the potential for escaping steam to prevent ECCS water from entering the vessel.
This concern arose prior to the writing of the ECCS Rule due to early tests in the semiscale facility during which all the injected ECCS fluid was expelled from the vessel.(10) An extensive research program on ECCS Bypass has greatly improved the understanding of this phenomena.(11,12) It is now known that the results of the semiscale tests were due to the extremelf,small size of the tests and the timing of the ECC injection, and that this is not expected in a reactor.
This research program is essentially complete, except for final proof of the scaling of this phenomena which will be demonstrated by tests in the full scale Upper Plemun Test Facility in 1986 as discussed below.
These major separate effects programs have been supported by a number of model development programs.
These programs are usually small research programs performed at e'iversities, designed to look at certain phenomena in great detail and serve as a link between analysis and experiments.
These research programs have studied heat transfer and two phase flow in the detail required to develop models or correlations used to predict the phenomena.
Most of the models and correlations for heat transfer and fluid flow used in both evaluation models and best estimate codes were originally developed from very small test facilities operating at low temperatures and pressures in which it was practical to conduct l
tests of the detail and number required to understand the physical phenomena.
These models and cortelations were then checked against a more limited number of tests in larger and more expensive separate effects facilities described above. Model development programs have also helped develop instruments required to study these phenomena in large test facilities.
Examples of model development programs include the mist-flow studies at SUNY(34) and flow heat transfer work at Lehigb(35),
20 Enclosure "B"
'L.
Y
[
Wher;as separate offccts tests were used to investigate c rtain phases E
of the LOCA transient or concentrate or certain specific phenomena, integral tests have been conducted to simulate the entire LOCA transient to ensure that the overall process is understood. The NRC has sponsored three major integral test facilities, each of which have undergone a number of upgrades over the years. The Loss of Fluid Test (LOFT) facility was the major NRC integral test facility.
LOFT is an actual nuclear reactor operating at temperatures and pres-sures similar to commercial reactors, although only about 1/5 as large as the PWR which it simulates.
The LOFT program conducted three large break LOCA tests, two intermediate LOCA tests, nine small LOCA tests and numerous separate effect tests.(13,14,15) The LOFT results demonstrated that actual temperatures during the large break LOCA tests were significantly lower than those predicted using evaluation model calculations.
One of the reasons for this was the re-wets that uccurred in the LOFT tests after the initial rod heat-up which are not allowed in Appendix K calculations.
LOFT results also provided data to compare with best estimate calculations to allow identification of areas where the cal-culations required improvement.
AnotherkeyfindingofpheLOFTprogramwas t
l that the actual nuclear fuel rods which were used in LOFT behave differently 1-from the electrical heater rods used to simulate nuclear rods in other tests.
The actual nuclear rods tend to be cooled better than the electrical rods used in most experiments.
The Semiscale facility is another NRC sponsored integral PWR test facility.
l Semiscale operates at full reactor pressures and temperatures, but is much smal-ler than LOFT and uses electrical heaters to simulate the nuclear rods.
As a smaller, non-nuclear facility.
Semiscale is less expensive to operate and many more tests could be conducted.
Semiscale has been and continues to be the " work-horse" test facility.
Tests were conducted to investigate scaling of the LOFT
'results and to assist in planning of LOFT tests.
Larger numbers of tests to investigate specific blowdown, refill and reflood phenomena, alternate ECCS con-cepts and the influence of various assumed initial conditions and failures were conducted over a long period.(16,17,18)
The TLTA served as the NRC's BWR integral facility.
Initially a separate effects facility studying only the blowdown phase of the large-break LOCA, the TLTA was upgraded to include ECCS injection and to simulate the entire BWR large-break LOCA transient.(19,20) TLTA tests also showed temperatures significantly lower than evaluation model calculations.
Several reasons for 21 Enclosure "B"
y h
this were offGctive cooling by stsam befera th2 ECCS injection started, a counter-
~
current. flow limit at the bottom of the core which allowed the core to be covered before totally filling the lower plemum, and the effect of leakage holes (a BWR I
design change) at the bottom of the core which allow liquid to flow into the 7
bottom of the bundle.
Figure 8 shows a comparison of heat transfer coefficients measured in the TLTA versus heat transfer coefficients used in BWR evaluation models.
Evaluation models did not include these phenomena at that time, but have since been improved to include some of these phenomena.
The TLTA was later used to simulate small break LOCAs.(21) 'Later, the TLTA was replaced by the Full Integral Simulation Test (FIST) facility.
FIST was a major upgrade of TLTA using information learned from TLTA to improve the facility simulation capability.
A number of large, intermediate and small-break LOCA tests were conducted to confirm the TLTA results.(22)
The TLTA and FIST facilities were small test facilities and the scaling of the LOCA test results required investigation.
Particularly, a number of multidimensional phenomena exist which cannot be simulated in these small facil-ities.
SeparateeffecttestingwasconductedundertheyWRRefill/Refloodpro-gram investigating these multidimensional phenomena in a nearly large-scale facility.(23,23a) A " steam binding" phenomena had been postulated which might prevent ECCS fluid injected above the core in a BWR to flow into the core and cool it.
The Refill /Reflood tests demonstrated that this postulated phenomena does not exist.
Multidimensional flow patterns allowed liquid to easily flow into the core, as shown in Figure 9, and cooling in an actual BWR is expected to be better than even that observed in small test facilities.
The remaining experimental work to be completed involving ECCS research is the international 2D/3D program involving the U.S., Germany and Japan. This program is investigating multidimensional behavior in PWRs a during large break LOCAs.
Much of the testing in the Japanese facilities has been completed with only the full-scale tests in the German Upper Plenum Test Facility (UPTF) remaining.
These tests will be performed in 1986-1987, along with tests in the 3rd core of the Japanese Slab Core Test Facility (SCTF). The completed Japanese tests, along with the U.S. analyses of these tests, have shown excellent core cooling during a large-break LOCA, even with highly conservative experimental conditions.(24,25) Although " steam binding" was shown to occur, the actual 1
cooling of the simulated fuel rods was much greater than allowed by Appendix K.
l l
22 Enclosure "B"
M' I
t sessa Y
JET Puur suction
( \\[uNeov8Rv + i.e oc
- g g
TEST M21 RUN 3 MO ECC)
\\
.- TEST M22 RUN 3 IAVE ECC)
- TEST M23 RUN 3 4 Low ECCL g' '
- "" -1 i
l'\\
l
/
4 i
3 l-si
]
t 5
t I y
!- - 1 p...
)1
{
i i
3 lik
\\
.l
=
("
l jk.j T_____
Ygg SuNDLE REFLOODS
- = 1 l
\\P M ft' l
io i
COLLAPSE g
l CORNER l
l
'}
I i
--i E
p Em gg[E l
CENTER RODS RATED PLOW I
I i
a a
u
,,.E,,,,,
u l
Figure 8.
Heat transfer coefficients determined from large-break LOCA i
simulations in the TLTA are shown to be much larger than that specified in Appendix K.
(NUREG/CR-2229) i 23 Enclosure "B"
bC e;-
K' 4
STPRE WS WWAT93 T3 ALL CMANe0ELS h
h sesLET ORIFICE MOLDS UP WATER tas ALL CMANNELS ALL C94ANNELS PROe#PTLY RE FLOOD L90UtO CONT 000U0U5 REGI04 IM UPPER PLENUM 800 CCFL AT TOPCF4YPASS e SVPASS REGION FILLS R APIDLY h PERIPMERAL UPPER PLENUM SUSCOOLING e UPPER Tit PLATE CCFL GRE AKDOWN b
p,,,-'~
,, /
/
UPPER PLENUM LEVE L w e _.
@i @. n, m _ y
.::.v
-K
.1_
, _.1 l /,"'/t'.".'$5 'Iv'%. - --- -
I'
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i'h O
n
(
(c) 8.)
(b)
CHANNEL LEVELS e
a T
l I
i I
t
/
l>
=
i a
/
i
'/
~~
l a
p' \\
l y
m i
/
/
LOWER PLENUM LEVEL f
_a
._n_n e
e*
i i
i i
i i
0 20 40 80 30 100 120 140 TIME lascl Figure 9.
Large scale separate effects tests conducted under the BWR Refill /Reflood program showing that multi-dimensional effects enhance refilling of the core region by ECCS fluid injected above the core.
(NUREG/CR-3223) 24 Enclosure "B"
up y
- r v
i..
r F:
I600 i
i
~1400
- C.s WRAP
_ Q '.s ~, j ' % p E 1200
~
WREM N:.N-( pre - test)
$.1000 WREM (post-test)
~
9 S
CCTF 800 8
o
't 600 S
3 400 o
2000 100 200 300 400 Time offer flood (s)
Figure 10.
Comparison of cladding temperatures during a reflood test in the l
Cylindrical Core Test Facility (CCTF) with calculations using typical evaluation model assumptions.
The significantly improved cooling observed in the test was due to more water accumulating in the core than calculated by the conservative models.
(JAERI-M 83-029) l l
l 25 Enclosure "B"
y b
A special tsst(26) was run in the Cylindrical Ccra Test Facility (CCTF) yj using initial and boundary conditions specified by the USNRC/NRR to simulate evaluation model conditions. The resulting rod temperatures, shown in Figure 10,
+
were much lower than those predicted by Appendix K modeling.
Core blockage tests in the full-height SCTF showed enhancement of cooling near the blocked bundles (2 out of 8 bundles blocked) rather than the degradation of cooling required by Appendix K.
The previously discussed experimental research is only part of the over-all ECCS research.
The final product of this research is computer codes which can accurately calculate ECCS performance with well defined uncertainties.
Computer code development has been closely linked to the experimental programs.
Separate effect and model development tests have been used to develop and improve the computer codes and integral tests used to evaluate the accuracy and identify areas requiring improvement.
The TRAC computer code has been developed for both PWR(27,28) and BWR(29,30) applications.
TRAC has multidimensional calculational capabilities and is capable of calculating two phase non-equilbrium phenomena which occur during a LOCA.
TheREtyP5(31)computercode is a one-dimensional code which has gained popularity for use in evaluation of one-dimensional experiments and for other applications where multidimensional i
effects are not important.
These computer codes have reached sufficient maturity that no further major improvements are planned. Assessment of these codes has shown that they l
can accurately predict LOCA phenomena.( 2,33,36) Figure 11 shows a comparison between system pressure and rod temperatures calculated by BWR TRAC and measure-ments in the FIST Facility.
Figure 12 compares rod temperatures calculated by PWR TRAC with the measured temperatures from LOCA simulations in a number of different facilities.
Future work with the codes involves making the codes faster and easier to use and assessment under an international program.
Over 12 international participants have agreed to provide data and/or assessment calculations.
The goal of this program is to determine the effects of the user on the accuracy and to better quantify the uncertainty of the calculations.
These computer codes have reached the state where they can be confidently used to calculate ECCS performance within reasonable uncertainties.
These codes now form an important part of the regulatory process and are used extensively to evaluate a wide range of reactor transients.
Their use as ECCS evaluation models 26 Enclosure "B"
yg.y has be:n more liaitsd, however, due to the restricticns en Gyaluation models I
imposed.by Appendix K.
The nuclear industry is, however, now starting to make 3
use of these codes to benchmark evaluation models and to evaluate the conserva-tism of the licensing calculations.
l l'
4 e
l l
27 Enclosure "B" l
g.
e,-
- g g h
i'I 11M
=
a TRAC e
--- DATA
~
o' M
=
i
- 4E NO m
I 1
l
/
\\
Y s
y\\ \\
- l/
~
/
l Y
f
/
- % ~ - % A -
!s\\/
w m
I I
I I
I f
400 0
60 100 ISO 300 360 m
l TIME isect 8-s-
J MSIV CLOSURE 7
TRAC g LOWER PLENUM FLA$NING
== = = DATA g
(
\\ JET PUMP EXIT PLANE UNCOVERY
\\
5 n\\
w 4
I,
,,,C. \\ '"c5 1
\\ Il PCI L\\{
2 1
=
I g
0 I
0
'w 300 s00 1
TIME tenci ed %k W Figure 11.
Comparison of cladding temperatures mdasured during TLTA 3
large-break LOCA simulation with best estimate calculations i
)
using the TRAC-BWR computer code.
(NUREG/CR-2571) 28 Enclosure "B"
., mqqens y e
4 TRAC-PD2 LARGE LOCA ASSESSMENT 1500 g
I
^
l a
4 i
i g
1300 -
l I
I W
6 O
5
+
0 1100 -
b
+
A o
N O
D j
900 -
s 5
/
t 700 700 900 1100 1300 1500 MEASURED PEAK cud TEMPERATURE D
LOFT
+
SEMISCALE o
PKL a
FLECHT T2 O
E Figure 12.
Comparison of TRAC-PD2 calculated peak clad temperatures with experimental data 2
from large LOCA simulations in various facilities.
Assessment work such as this helps to determine the uncertainty in the code calculations.
(derived from
=
=
l
w I
Rafaranc'2s 1.
Letter Levine to Case and Minogue, "RILf25, FRAP-53," March 24, 1978.
la.
G. Berna, et al., "FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods,"
NUREG/CR-1845, January 1981.
2.
Letter Levine to Denton and Minogue, "RIL#59, Transient Fuel Rod Behavior Code:
FRAP-T4," September 21, 1979.
3.
L. Siefken, et al. "FRAP-T6: A Computer Code for the Transient Analysis of Oxide Fuel Rods," NUREG/CR-2148, May 1981.
4.
R. Muralidharan, et al., "BWR Blowdown Heat Transfer Final Report,"
GEAP-21214, General Electric Co., February 1976.
5.
W. G. Craddick, et al., "PWR Blowdown Heat Transfer Separate Effects Program Data Evaluation Report-Heat Transfer for THTF Test Series 100,"
NUREG/CR-0105, September 1978.
6.
G. Yoder, et al., " Dispersed Flow Film Boiling in Rod Bundle Geometry-Steady State Heat Transfer Data and Correlation Comparisons," NUREG/CR-2435, March 1982.
7.
Letter Levine to Denton, "RIL#67, Reflooding of Simulated PWR Cores at Low Reflood Rates," November 6, 1979.
8.
Letter Minogue to Denton," RIL#135, Heat Transfer and Hydraulics in Full Length 17 x 17 Rod Bundle During Reflood Phase of PWR LOCA," September 13, 1983.
9.
"PWR FLECHT-SEASET Final Report," NUREG/CR-4167, to be published.
9a.
"PWR FLECHT-SEASET 163-Rod Block Bundle Data Analyses Using COBRA-TF,"
NUREG/CR-4166, to be published.
10.
D. Olson, "Semiscale Blowdown and Emergency Core Cooling (ECC) Project Test Report:
Test 845 (ECC Injection)" ANCR-1014, Aeroject Nuclear Company, January 1972.
11.
Letter Levine to Denton, "RIL#57, Small Scale ECC Bypass Research Results," August 10, 1979.
12.
Letter Minogue to Denton," RIL#128, PWR Lower Plenum Refill Research Results,"
December 8, 1981.
13.
Letter Levine to Denton, " RIL#37, LOFT Reactor Safety Program Research Results through October 1, 1978," September 29, 1978.
30 Enclosure "B"
g..
i
&7 14.
Letter Lovine to Dent:n," RILf63, LOFT Reactor Safety Prograa R2ssarch t
Results from Nuclear Loss of Coolant Experiments L2-2 and L2-3,"
F November 1, 1979.
15.
C. Nalezny, " Summary of Nuclear Regulatory Commission's LOFT Program
~
f Research Findings," NUREG/CR-3005 (Draft For Review), June 1983.
16.
D. Solberg, " Introduction to the Semiscale Program," Twelfth Water Reactor Safety Research Information Meetina, NUREG/CP-0058, Vol. 1, page 33.
17.
C. Dobbe, " Code Assessment and Applications Program-United States Standard Problem 8 Preliminary Report," CAAP-TR-036, INEL, December 1978.
18.
T. Boucher and R. Dimenna, "Semiscale Mod-2A Intermediate Break Test Series-Test Results Comparison," NUREG/CR-3126, January 1983.
19.
L. S. Lee, et al., "BWR Large Break Simultion Tests-BWR Blowdown / Emergency Core Cooling Program, " NUREG/CR-2229, April 1982.
20.
Letter Minogue to Denton, "RIL#127, Heat Transfer and Hydraulic During a BWR Large-Break LOCA," December 4,1981.
21.
Letter Minogue to Denton, "RIL#126, BWR Small-Break Tests," December 4,
~
1981.
J 22.
W. S. Hwang et al., "BWR Full Integral Simulation Test (FIST) Phase I Test Results," NUREG/CR-3711, September 1984.
23.
L. L. Myers, "BWR Refill-Reflood Program Final Report," NUREG/CR-3223, September 1983.
23a. Letter Minogue to Denton, "RIL#134, Multidimensional Phenomena During Refill Phase of a BWR LOCA," April 20, 1983.
24.
"Research Information Report on the results of the Core I Test Series at I
the JAERI CCTF," MPR-863, to be published March 1985.
25.
M. Sobajima, et al., " Characteristics of Lower Plenum Injection Reflood Tests in SCTF Core-1," JAERI-M-84-223, Japan Atomic Energy Research Institute, December 1984.
26.
Y. Murao, et al., " Evaluation Report on CCTF-Core I Reflood Test Cl-19 (Run 38)-Experimental Assessment of the Evaluation Model for the Safety Analysis on the Reflood Phase of a PWR LOCA," JAERI-M-83-29, Japan Atomic Energy Research Institute, February 1983.
27.
Letter Minogue to Denton, "RIL#125, TRAC-PD2-An Advance Best-Estimate Computer Program for PWR LOCA Analysis," September 1, 1981.
l l
31 Enclosure "B" l
t
w A.
lb;
- 28. " TRAC-PF1: An Advanecd Best-Estimate Comput:r Pr: gram for Pressurized
[~
Water Reactor Analysis," NUREG/CR-3567, February 1984.
i 29.
Letter Minogue to Denton, "RIL#132, TRAC-BD1 Computer Program,"
August 30, 1982.
- 30. " TRAC-801/M001:
An Advanced Best-Estimate Computer Program For Boiling Water Reactor Transient Analysis," NUREG/CR-3633, April 1984.
31.
"RELAPS/ MOD 2 Code Manual," EGG-SAAM-6377, INEL, April 1984.
32.
M. Aluegir, "BWR Refill-Reflood Program Task 4.8-TRAC-BWR Model Qualification For BWR Safety Analysis Final Report," NUREG/CR-2571, October 1983.
33.
" TRAC-PD2 Independent Assessment," NUREG/CR-3866.
34.
S. Lee et al., "A Study of Droplet Hydrodyamics Across a Grid Spacer,"
NUREG/CR-4034, November 1984.
35.
D. Evans et al., " Measurement of Axial Varying Non-Equlibrium in Post-Critical-Heat-Flux Boiling in a Vertical Tube," NUREG/CR-3363, June 1983.
36.
L. N. Kmetyk, et al., "RELAPS Assessment:
Quantitative Key Parameter and Run Time Statistics," NUREG/CR-3802, to be published.
1 l
l 32 Enclosure "B"
n_
E i
E l
V 4
ENCLOSURE "C" CONSERVATISM IN APPENDIX K AND 50.46 a.
i l
l l
y n
- g CONSERVATISM IN APPENDIX X AND 50.46 e
y Prior to revising the ECCS rule, two major questions must be addressed:
7 l
1.
What is the current conservatism in evaluation models (i.e., the differ-ence between peak cladding temperatures calculated by evaluation models and the peak cladding temperatures actually expected during LOCAs) and is this conservatism overly large so as to justify revisions to the ECCS rule?
2.
What would the conservatism be if the ECCS rule were to be revised and would that remaining conservatism be large enough to cover the overall uncertainty in the calculated temperatures?
The answer to these questions must be treated on a case-by-case basis since they are plant specific and will also depend on the specific accident in question and the time in life of the plant.
Nevertheless, it is desirable to obtain answers to these questions for several typical cases prior to proceeding with a proposed rule change.
RES coordinated a calculational study to address these questions making use of calculations performed at LANL and INEL using the TRAC computer code and calculations performed by the four major reactor vendors using the vendor evaluation models.
Four basic types of calculations were performed for large break LOCAs:
1.
Best estimate calculations using TRAC assuming that the plant was operating in the most probable manner (BE).
2.
Best estimate calculations using TRAC assuming that the plant was operating at the worst state allowed by technical specifications (e.g., beginning of life power distribution, minimum equipment available) (BE/EM).
3.
Calculations using the existing approved vendor evaluation models (EM).
1 Enclosure "C"
w--
.w L
r 4.
C31culcticns using the existing approved vandor Gvaluation models with modifications in areas where Appendix K Models are known to be highly conservative (EM').-
m These calculations are reported in references 1-6.
EXISTING CONSERVATISM IN APPENDIX K Current best estimate TRAC (1) calculations of PWR plants operating at technical specification limits (BE/EM) estimate peak cladding temperature during large break LOCAs to range from 1403-1851*F depending on the plant.
Comparing these numbers with existing evaluation model calculations (EM) results in current conservatisms of 775-240*F.
Further conservatism (as high as 1000*F) exists during most of the life of the plant since plants do not normally operate at technical specification peaking factors for any length of time. This wide variation in calculated temperatures and conservatism is the result of different evaluation models, different plant types, and varying technical specification power limits.
From these calculations, it can be seen that some plants (e.g., Westinghouse) are operating under highly conservative technical specifications.
Best estimate TRAC calculations (2) of a BWR large break LOCA yield peak cladding temperatures of 645-1079 F.
Thus, current BWR evaluation models are also highly conservative.
However, BWRs have the option of using the newly approved SAFER model which reduces the conservatism to a value more consistent with the uncertainty in the calculations.
-CONSERVATISM WITH REVISED APPENDIX K MODELS**
The PWR vendors performed calculations of a large break LOCA using their approved evaluation models with modifications to certain Appendix K require-ments.(3,4,5) These modifications were in areas that research has shown that Appendix K models are highly conservative or unrealistic.
The most important change made was the decay heat model.
Peak cladding temperature reductions of 341-562 F resulted from use of the changes in the different vendor evaluation
- These revisions to Appendix K do not represent the ECCS rule change that is currently being recommended.
2 Enclosure "C"
3=p mod 21s. The offset of this reducticn en the conservatism of the r2 vised I
evaluation models is highly plant dependent.
For the Westinghouse plant calculation, a minimum conservatism of 300*F
~
would remain.
An even greater conservatism of 540*F would be present if it was assumed that the plant was operating at the most probable state rather than at technical specification limits. While a formal analysis has not been performed, it is expected that this safety factor is sufficient to cover the uncertainty in the calculations.
The B&W and CE comparisons are not completely valid comparisons gince a number of inconsistences are known to exist between the TRAC and vendor calcu-lations. These calculations have not been repeated, however, because it is not expected that the known inconsistencies would significantly change the con-clusions of the study.
The current calculations indicate that little or no conservatism would remain in B&W and CE evaluation models if the alternative Appendix K requirements were adopted.
It should again be noted, however, that the calculations are known to contain inconsistencies.
GE also performed calculations on the effect of mod'ifications to Appendix K requirements using SAFER.(6) The rule change would not benefit BWR's beyond the benefit already available through use of SAFER.
The calculations (proprie-tary) are of great interest, however, since GE did perform a detailed uncer-tainty calculation as part of the SAFER analysis.
This analysis indicates that if SAFER made use of the modified Appendix K requirements, the remaining conser-vatism would be less than the uncertainty in the best estimate calculation.
However, the conservatism would only be slightly less and the same order of magnitude as the uncertainty.
We conclude that if the ECCS rule were to be revised as assumed in this study, the remaining conservatism or safety factor would be of the same order of magnitude as the uncertainty in the calculations for some plants, but too little conservatism may remain for other plants.
However, both these con-clusions need to be supported by further analysis.
Since these analyses will be highly plant specific, it is appropriate that the licensee perform the analysis and it is therefore recommended that such an analysis be part of any I
revised ECCS rule.
l 3
Enclosure "C" t
3 i
CONSERVATISM IN 50.46 CRITERIA The ECCS rule requires that calculations be performed to show that the ECCS will maintain the cladding temperatures, cladding oxidation and hydrogen v
generation within certain limits. The conservatism of the calculation was previously discussed.
Now the limits of 50.46 are examined. These limits are 2200*F cladding temperature and 17% cladding oxidation which must not be exceeded.
It is believed that these criteria are appropriate and should not be changed.
Data from the Power Burst Facility (PBF) has shown that at temperatures of 2600 F, the steam-zircalloy reaction becomes sufficiently rapid to produce an autocalalytic temperature excursion (i.e., cladding oxidation is a self-sustaining process).I7) Thus 2600*F is a real limit, above which significant fuel damage would occur.
The current 50.46 criteria of 2200 F has a conserva-tism of about 400 F below the point that this temperature excursion would be expected.
This 400 F conservatism is thought to be sufficient, but not overly conservative.
Therefore, no change to the 2200 F criteria is recommended.
Cladding oxidation will also occur at temperatures below 2200 F, although at a slower rate.
The degree of oxidation is a function of both time and tem-perature.
Even at lower temperatures, the cladding could be significantly dam-aged if the temperature was maintained for long periods.
Thus a second criteria of maximum allowable cladding oxidation is required to ensure that the cladding maintains sufficient strength to rerrain intact during reflood.
Research has shown that the current 50.46 criteria of 17% maximum allowed cladding oxidation, combined with the 2200 F limit, is an appropriate limit, as illustrated in Figure 1, which will ensure survival of the cladding following reflood.(0) l USE OF ADDITIONAL MARGIN l~
Concern has been expressed that the 2200 F limit of 10 CFR 50.46 would be l
used as a " speed litait" if the ECCS rule were to be revised.
That is, if the ECCS rule were revised so that calculated peak cladding temperatures were lower, then power would simply be increased until calculated peak cladding tempera-tures again approached 2200*F.
This is thought to be undesirable because this would again place the plant very close to regulatory limits, requiring close 4
Enclosure "C"
7 1400 1300
- 200 180 0 1000 it i
1 i
1 i
l C
2
/
/
4 10 LIMIT FOR SURVIVING f--
8 :
THERMAL SHOCK
/
/Z 2
3 6
w
/
g 4
- +
/
t b
LIMIT FOR SURVIVING
/
O.15-J IMPACT AT g
2 300 K
/
~
~
Y
/
103
/
0 b h% OXIDATION l
O 6
/
LIMIT
/
x l
O 4
i F
e
/
4 gj W
L
-LIMIT FOR SURVIVING E
2
/'
O.3-J IMPACT AT F
300 K l I 2
10 l
1 8
l j
6 1477 K LIMIT 4
l 2
Figure 1.
Comparison of the 50.46 criteria of 2200*F and 17% cladding oxidation with estimates of where actual cladding damage l
would occur.
(NUREG-0562) i 5
Enclosure "C" i
T 1
{
surveillance, immediate reanalys:s and int:ri:a p2nalties if error discoveries indicate violation of the 2200*F limit. We have, therefore, looked at how the licensees might make use of this additional margin and how any potential power a
increases would translate into peak cladding temperature during a large break LOCA.
Both B&W and CE claim that other considerations limit their peak powers and that power increases as a result of a revision of the ECCS rule are not anticipated. Westinghouse indicates that they are limited by LOCA considera-tions and could increase peak local power from 12.2 kw/ft to about 14.4 kw/ft which is a limit imposed by DNB considerations.
If the total power of the plant was held constant, but the peak local power increased to the DNB limit of 14.4 kw/ft by increasing the peaking factor, a revised Westinghouse EM would predict a peak cladding temperature of 2042 F.**
Westinghouse also calculatt:d a peak cladding temperature of 2154*F if the peak local power was increased to 14.4 Kw/ft by increasing total power 13%.
This increase in total power is unre-alistically high, however, since the maximum total power " stretch" capabilities
~
of the plants is likely to be about 5%.
It is, therefore, expected that the revised ECCS rule would not result in increased powers sufficient to again hit the 2200 F limit.
If Appendix K were e
revised, LOCA would no longer be limiting and other factors, such as DNB, would limit peak local power.
REFERENCES 1.
Letter Sullivan to Beckner, November 16, 1984.
2.
P. D. Wheatley, " Analysis of a BWR/6-218 During a large Break Loss-of-Coolant Accident Using Evaluation Model Type Boundary Conditions,"
EGG-SAAM-6519, January 1984.
l 3.
Ltr Hochreiter to Beckner, "Results of Westinghouse Calculations Using a l
Modified Version of Appendix K," April 24, 1984.
4.
" Assessment of Proposed Changes to Appendix K on LOCA Limits," B&W Report 77-1150444-00, April 1984.
5.
Ltr Scherer to Beckner, "NRC Proposed Changes to 10 CFR 50, Appendix K,"
l LD-84-021, May 25, 1984.
- This revised Westinghouse EM is not the EM that Westinghouse would be expected to use under the ECCS rule currently being recommended.
6 Enclosure "C"
7 7
6.
Ltr Quirk to Beckner, " Impact of Propos2d Appendix K Rule Changes on General Electric SAFER /GESTR ECCS Results," July 26, 1984 (Proprietary).
~
7.
Memo Wright to Silberberg, "PBF Data on the Autocatalytic Zircalloy-
{
Oxidation Temperature Threshold Under Core-Uncovery Conditions,"
January 16, 1985.
8.
Van Houten, " Fuel Rod Failure as a Consequence of Departure From Nucleate Boiling or Dryout," NUREG-0562, June 1979.
b s
7 Enclosure "C" l
5
- g..
j5 ENCLOSURE "0" REGULATORY ANALYSIS e
J
m r ecs J
t-L REGULATORY ANALYSIS i
INTRODUCTION The effect of the proposed rule revision would be to reduce the peak cladding temperatures during a LOCA that are calculated using evaluation models.
This would allow some plants to increase allowed peak local power by increasing allowed peaking factors and/or increasing total power. This regulatory analy-sis discusses the effect of the rule change in terms of "value" (e.g., public benefitssuchassafety)and" impact"(e.g.,consequencessuchascosds).
The intent of the proposed rule change is to reduce the negative impact of the cur-rent rule, while producing negligible reduction in the value of the rule.
The proposed rule change may also result is some positive value to safety of the public.
While it appears that the proposed rule change will result in a signifi-cant positive impact for some plants, the risk of a negative impact is minimized by making use of this rule optional for existing plantsr Thus, each applicant or licensee can perform its own analysis to determine if use of the revised rule is advanta0eous.
The value of the proposed rule change may have some nega-tive aspects since an increase in power of a plant may increase risk to the public.
However, this negative value is believed to be very small and offset by other positive benefits to safety. While it is the intent of the proposed rule change to provide only positive impacts, there are also a number of pit-falls which must be avoided.
Information used in this regulatory analysis was obtained from a number of sources:
(1) Previous studies sponsored by DOE (1,2)
(2) Formal responses from the major reactor vendors to a request by NRC for input (3,4,5)
(3)
Informal discussions with reactor vendor, utility, national laboratory and NRC staff.
This information shows a wide variation in the impacts which would result from the proposed rule revision.
Some plants are not negatively impacted by 1
Enclosure "D"
n.
L{
the curr:nt rulo, whilo cthers cre negativ21y impact:;d by the current rule cnd could make use of the proposed rule revision to reduce this negative impact.
ECONOMIC IMPACT 7
LOCA considerations resulting from the present rule are restricting the optimum production of nuclear electric power in numerous ways.
These restric-tions can be placed into the following three categories:
(1) Maximum plant operating power, (2) Operational flexibility and operational efficiency of the plant, and (3) Availability of manpower to work on other activities.
Maximum plant operating power at some nuclear facilities is limited by the present Appendix K licensing results.
However, it can be very difficult to clearly separate these LOCA rule restrictions from other licensing issues and limitations.
Therearenumerouslimitsthatcanre{tricttotalplantpower, as well as the ability to maneuver the power over a wide range.
Typically, this limit is associated with either peak cladding temperature (PCT) calculated to occur during LOCA transients, or as a result of departure from nucleate boiling (DNB) restrictions. Additionally, there are limits to plant power because of NRC guidance on total allowable thermal power of 3800 MW and because of physical hardware limits on the balance of the plant (turbine, condenser, pumps and steam generators).
Regarding the second category of operational flexibility, some plants have very little LOCA margin.
Such a limited margin necessitates additional core power surveillance to prevent peaking factor violations.
This may also require
'special supplementary nuclear or safety analyses and restrictive fuel manage-ment schemes, resulting in inefficient fuel burnup and no extended burnup cycles.
The third category concerns periodic reanalyses which are required by the current rule.
If an error is found in an accepted evaluation model, a new LOCA analysis must be performed even if the error correction results in a decrease in PCT.
This was a major problem several years ago, with each reanalysis costing about $150KII) and diverting both licensee and NRC staff from other, more productive activities.
Very of ten, this reanalysis contributed very little 2
Enclosure "0"
Y to plant safaty. Whi10 re:nalys2s hav2 been less of a problem in r cent years, E
the' proposed rule change will ensure that reanalysis will be required in the f
future only for significant errors or changes.
The degree to which the proposed rule change would benefit a particular
.y plant depends on how limited the plant is by these LOCA restrictions.
The Babcock and Wilcox (B&W) and Combustion Engineering (CE) companies have inform-ally indicated that they do not feel that the plants which they design are limited by LOCA and, therefore, B&W and CE plants would not benefit from the first two categories.
General Electric Co. (GE) plants do tend to be lisited in operation by LOCA restrictions and would greatly benefit from relief from LOCA restrictions.
However, this relief is already availabie 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 (4)
Westinghouse (W) plants are the only plants which would appear to directly benefit from relaxa-tion of LOCA limits.
W plants represent the largest number of plants, however, with 35 plants operating or near operation and 20 additional plants being constructed.
W indicates that most of these plants are7 mited by LOCA consi-li derations.
It can, therefore, be estimated that there are at least 30 nuclear plants on line that are limited by LOCA considerations either in total power and/or in flexibility of operation.
Up to 20 additional plants may also eventually come on line which will be limited by LOCA considerations.
Any rule change that produces a PCT decrease of 100 F can be translated into a total plant power increase of approximately 4% - 6% based on LOCA limit considerations.
This number of a 4% - 6% increase in power represents a power increase which is within the capabilities of typical W plants based on existing hardware and is still well below other limits such as DNB limits.
The rule change proposed would provide a reduction in PCT of more than the required 100 F.
The economic impact of this increase in power can be viewed in terms of capital costs avoided. Assuming forty 1000 MWe plants are upgraded by 5% at an upgrade cost of $150/KWe(5), the total capital cost for 2000 MWe of additional capacity would be $300 million.
This is compared to typical construction costs of $2250/Kwe(5) which would require $4.5 billion dollars to build 2000 Mwe of additional capacity.
Thus by " stretching" existing and planned W plants, 2000 MWe of additional capacity could be available at a savings of approximately $4.2 billion.
i 3
Enclosure "D"
1 h
This increase in power c n also be viewed in terms of savings to the rato payer. A 5% increase in power for forty 1000 MWe plants, assuming a 60% capa-10 city factor, would represent an increase in power generation of 1 x 10 KWH/
year.
Typical replacement power costs for nuclear plants range from less than 2C/Kwh to over 4C/Kwh.
This translates to savings of $200-400M per year.
Further savings would be possible from removing LOCA restrictions due to more efficient plant operation, irrespective of whether or not the total power of the plant was upgraded.
These improvements include improved fuel utiliza-tion and improved maneuvering capabilities.
Core management and advanced fuel management concepts are complicated subjects and LOCA limits are only one of many factors to consider.(2) Thus precise estimates of potential savings are difficult.
However, savings of 3 to 6 million dollars per plant per year would not be unreasonable.(5) Even if a utility did not increase power or change fuel management, simpler generic reload calculations possible with less restrictive LOCA limits would save $250,000 per plant per year.(1) y SAFETY VALUE The value of the proposed rule change must also be evaluated in terms of the effect on safety.
The proposed rule change would probably result in increased local power within the reactor core and possibly increases in total power.
Power increases on the order of 5-10% will have an insignificant effect on risk.
The principal effect of increased power would be to increase the fission product inventory.
A 5 percent power increase would result in a 5 percent increase in fission products.
Thus, 5 percent more fission products could be released during core melt scenarios and potentially released to the environment during service accidents.
All current estimates of risk, based on PRAs that have considered the LOCA among the various sequences, still have uncertainties in the range of plus or minus orders of 10 to 100.
Therefore, a 5 percent increase in risk is I
negligible compared with the current uncertainty estimates for risk.
There are also safety benefits derivable from alternative fuel management schemes that could be utilized if the proposed changes were implemented.
An important safety benefit could be realized if restrictions on core power peaking 4
Enclosure "D"
o k'p wera less stringent.
High neutron icakages at the cara cuter boundary are used
~
in PWRs to flatten the radial core power profile.
This inefficient fuel manage-mentprocedureisneededtomaintainpeaklngwithintightlimits.
In addition, the resultant high neutron fluer;ce leakage can enhance vessel embrittlement resulting in PTS concerns.
Higher power peaking factors that would be allowed with the revised rule could be used to reduce neutron flux at the vessel.
POTENTIAL PITFALLS In contrast to the above beneficial economic and safety effects, there are some detrimental impacts that might result from the proposed rule change.
Some believe that a rule change might destabilize the present licensing process, which, while perhaps overly conservative, is well known and predictable.
However, others believe that the licensing process will not be fully stable until the rule is revised.
Any disruption of the licensing process may be circumvented by "grandfathering"; that will give utilities the option of adopt-ing the new analysis methods or continuing with the oldpppendix K procedures.
Perhaps the single most significant concern is the potential for reopening ECC rulemaking.
These original hearings lasted for a year and consumed the time and talents of many engineers and attorneys.
The manpower and resources that would be needed to conduct new hearings would cause major problems for the NRC and an industry in which these items are already in short supply.
The vendors feel that such new hearings are unacceptable and prefer not to modify the rule if it required such hearings.
CONCLUSIONS It is believed that no significant negative impact in terms of economics or negative value to safety would result from the proposed rule change as long as the pitfalls described above are avoided.
The potential positive benefits are difficult to estimate because of the fact that they will be plant specific l
and because it is difficult to estimate how each utility would respond to the i
proposed rule change.
However, it is thought that the potential for significant benefits to economics and safety exist if the rule were revised as proposed.
I r
l 5
Enclosure "D"
E
--REFERENCES 1.
" Revision of Loss-Of-Coolant Accident (LOCA) Rule Licensing Requirements -
A Survey of Opinion within the Nuclear Industry," NUS-4221, April 1983.
2.
P. Wei, et al., " Boiling Water Reactor Uranium Utilization Improvement Potential," GEAP-24965, June 1980.
3.
Ltr fm Hochreiter to Beckner, "Results of Westinghouse Calculations Using a Modified Version of Appendix K," April 23, 1984.
4.
Ltr fm Quirk to Beckner, " Impact of Proposed Appendix K Rule Changes on General Electric SAFER /GESTER ECCS Results," July 26, 1984 (Proprietary).
5.
Ltr Rahe to Ross, "LOCA Margin Benefits," February 8, 1985.
6.
"Probabilistic Risk Assessment (PRA) Reference Document," NUREG-1050, September 1984.
l l
6 Enclosure "D" l
l
T~
- 2 T:.
4*,
e
)
6 ENCLOSURE "E" NOTICEOFPROPOSEDRULEMAKING
Y.
t NUCLEAR REGULATORY COMISSION W
- j
[10 CFR Part 50]
r Revisions to Acceptance Criteria for Emergency Core Cooling Systems AGENCY:
Nuclear Regulatory Commission.
ACTION:
Proposed Rule.
SumARY:
The Nuclear Regulatory Commission is considering amending its regulations concerning acceptance criteria for emergency core cooling systems (ECCS).
These amendments would change the methods used to demon-strate that emergency core cooling systems would protect the nuclear reactor core during a loss-of-coolant accident (LOCA).
Research performed since the current rule was written has shown that calculations performed in accordance with the current rule result in estimates of cooling system performance that are significantly worse than that which would be estimated based on the improved knowledge gained from this research.
In addition, the operation of some nuclear reactors is being unnecessarily restricted by the rule, resulting in increased costs of electricity generation.
The amended rule would allow use of the best information and knowledge avail-able to demonstrate that the ECCS would protect the reactor during a LOCA.
Procedural changes would also be made so that certain reanalyses which do not contribute to safety would not be required to be performed immediately.
The amendmeni.s would apply to all applicants for and holders of construction permits or operating licenses for light water reactors.
DATES:
Comment period expires (60 days following publication in the Federal Register).
Comments received after that date will be considered 1
Enclosure E
Y1 if it is practical to do so, but assurance ef consideration cannot be
?-
[
given except as to comments received on or before that date.
ADDRESS:
Written comments or suggestions for consideration in connection with the proposed amendments should be submitted to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention Docketing and Service Branch. Copies of comments received may be examined in the Commission's Public Document Room at 1717 H Street NW.,
Washington, DC 20555.
FOR FURTHER INFORMATION CONTACT:
L. M. Shotkin (301-427-4254), U.S.
Nuclear Regulatory Commission, Washington, DC 20555.
SUPPLEMENTARY INFORMATION:
===.
Background===
Section 50.46 of 10 CFR Part.50 provides " Acceptance Criteria for Emergency Core Cooling Systems (ECCS) in Light Water Nuclear Power Reactors."
This section requires that calculations of loss-of-coolant accidents (LOCA) be performed to show that the ECCS will maintain cladding temperatures, cladding oxidation and hydrogen generation to within certain specified limits.
It also requires that a coolable core geometry be maintained and that long term decay heat removal be provided.
Appendix K to 10 CFR Part 50 sets forth certain required and acceptable features of the models used to perform these calculations.
The criteria of 10 CFR 50.46 and the calculational methods specified in Appendix K were formally issued in January 1974 after extensive rulemaking hearings and are based on the understanding of ECCS performance available at that time.
2 Enclosure E
.m In the ten years following th2 rulemaking, svar $700 aillion has
[
been invested by the Nuclear Regulatory Commission on research investigat-L t
ing ECCS performance. The majority of this LOCA research is complete and 4
l has greatly improved the understanding of ECCS performance during a LOCA.
The methods specified in Appendix K are now known to be highly conserva-tive; that is, the actual temperatures during a LOCA would be much less than the temperatures calculated using Appendix K methods.
The ECCS research has gone beyond showing that Appendix K is conservative, it has allowed quantification of that conservatism. The results of experiments, computer code development, and code assessment now allow more accurate calculations of ECCS performance during a LOCA than is possible using s
Appendix K procedures, along with reasonable estimates of uncertainty.
I i
It is also known that some plants are being, restricted in operating a
flexibility by limits resulting from conservative Appendix K calcula-These restrictions may be preventing optimal operation of some tions.
I plants.
Based on research performed, it is now known that these restric-tions can be relaxed without affecting safety.
On December 6, 1978, the NRC published an advance notice of' proposed rulemaking (43 FR 57157) calling for a two phase approach to the revision of 10 CFR Part 50 and Appendix K.
The first step would have been to make procedural changes and to permit minor model changes which would not have reduced the conservatism contained in Appendix K.
The second phase would I
have made further technical changes based on research results and operat-ing experience.
NRC activity on the ECCS rulemaking was severely curtailed as a result of the high priority efforts required by the TMI-2 accident.
This ECCS 3
Enclosure E i
1 I
~:
'E rulemaking ess:ntially sat dormant until July 1981, when it was brought up r
7 again in the context of simplifying and streamlining the regulatory process.
The NRC has reviewed the comments made by outside organizations on 1
the advance notice of proposed rulemaking, as well as a number of other comments received since that time.
In general, the commenters support a rule change that would permit greater flexibility in meeting the regula-tions and would incorporate the use of presently available research infor-nation.
Many felt that the Phase 1 scope should be expanded to allow the use of additional information available from the ECCS research that has been performed.
Because of the delay in changing the ECCS rule, the NRC has adopted an interim approach described in SECY.-83-472,1 to accommodate requests for improved evaluation models, submitted by reactor, vendors, for the purpose of reducing reactor operating restrictions. This interim approach requires a best estimate calculation with an evaluation of the uncertainty in the calculation to h demonstrate that an adequate conservatism or safety factor exists.
The NRC has decided to proceed with the rulemaking, but in the form of a more comprehensive rule amendment based on (1) the comments received since the publication of the 1978 notice of proposed rulemaking, (2) the additional research conducted and experience gained since the 1978 notice, and (3) recent experience using best estimate calculations and uncertainty evaluations to support licensing decisions.
A report providing the technical basis for revision of the ECCS rule has been prepared.
It summarizes both the extensive ECCS research 2SECY-83-472, " Emergency Core Cooling System Analysis Methods," November 13, 1983, is available for inspection and copying for a fee at the Commission's Public Document Room at 1717 H Street NW., Washington, DC.
4 Enclosure E
w_..-
that has be:n c:nducted and the cxperi:nce gained using this information r
in the licensing process and is available at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555.
~
4 SlMtARY OF PROPOSED RULE CHANGES E 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors:
Section 50.46(a)(1) would be amended so as to include require 3that the evaluation model include sufficient supporting justification to show that the analytical technique describe the behavior of the reactor system during a LOCA.
Deleted was the requirement that the features of Appen-dix K to 10 CFR Part 50 be used to develop the evaluation model.
Before the evaluation model could be used it would have'to be accepted by the NRC staff.
The staff will expect-that the analytical technique will, to the extent practicable, utilize the best available methods and data.
The an. ended rule would also require that the uncertainty of the calculation be evaluated and considered when comparing the results of the calculation to the temperature limits and other criteria of S 50.46(b).
The rule would not specifically state the analytical methods or uncertainty i
evaluation techniques that shall be used but guidance is being provided 2 to indicate methods that the NRC staff in the form of a Regulatory Guide l-would consider acceptable.
It should be noted as discussed in SECY-83-4721, l
that the NRC has in the past found acceptable a method for estimating the l
l
' Regulatory Guide xx (issued for comment), " Acceptable Methods for Evaluating ECCS Performance During a LOCA," is available (insert boilerplate on Regula-tpru Guides) l 5
Enclosure E i
n uncertainty that tas judged to be at least at tha 95% probability level.
f" Appendix K Section II, " Required Documentation" would remain generally applicable, with only minor revisions made to be consistent with the
~
amended rule.
Sections 50.46(a)(2) and (3) would be totally revised to eliminate those sections concerned with historical implementation of the current rule and will be replaced as follows:
Section 50.46(a)(2) would be revised to allow the features of Appen-dix K to be used in evaluation models as an alternative to performing the uncertainty evaluation specified in the amended S 50.46(a)(1).
This method would remain acceptable because it is believed that the conservatism in the Appendix K features is sufficient and furthermore, is conservative with respect to the new method as proposed in the amended S 50.46(a)(1).
Thiswouldallowcurrentandfutureapplicantsabdlicenseestouseexist-ing evaluation models if they did not need or desire relief from current operating restrictions.
The revised S 50.46(a)(3) would state specific requirements for reanalyses when errors in evaluation models are discovered or changes are made to evaluation models.
This section would require that all errors affecting peak cladding temperature and any changes or combination of changes such that the sum of the absolute magnitudes of the calculated 1
cladding temperature changes exceeds 50*F shall be reported to the NRC.
l If as a result of the error correction or change, the limiting transient calculation exceeds the criteria of S 50.46(b), then immediate steps l
l would have to be taken to achieve compliance.
If the criteria of S 50.46(b) are not exceeded, a reanalysis shall be performed on a schedule either developed using an approved integrated scheduling system, if available, or one established by the NRC after receipt of a proposed schedule.
6 Enclosure E l
1
j o.
Appendix K ECCS Evaluatien Models E
Yg Minor amendments would be made to Appendix K, I.(C)(5)(b) to modify b
post-CHF heat transfer correlations listed as acceptable.
The "McDonough" reference would be replaced with a later paper which is more generally available and which includes additional data.
The "Dougall-Rohsenow" correlation, which has been found to be non-conservative in some situa-tions, would be eliminated as a generally acceptable model.
Appendix K, II(1)(b) would be removed since this requirement would be clarified under the amended S 50.46(a)(3).
Likewise, Appendix K,11(5) will be 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 should not effect any existing approved evaluation models since the changes are either " house keeping" cha,nges or only changes to
" acceptable features," not " required features."
REGULATORY ANALYSIS The Commission has prepared a regulatory analysis for this proposed regulation.
The analysis examines the costs and benefits of the rule as considered by the Commission.
A copy of the regulatory analysis is avail-able for inspection and copying for a fee at the NRC Public Document Room, 1717 H Street NW., Washington, DC.
l PAPERWORK REDUCTION ACT The proposed rule will be submitted to the Office of Management and Budget for clearance of the appl
- cation requirements that may be appro-priate under the Paperwork Rehetion Act (Publ. L.96-511).
The SF-83 7
Enclosure E
y "R; quest fer Clearance," supp:rting Statement, and related d:cumentation b
submitted to 0MB will b2 placsd in th2 NRC Public Document Room at 1717 H. Street NW., Washington, DC 20555.
The material will be available for inspection and copying for a fee.
REGULATORY FLEXIBILITY ACT In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission hereby certifies that this rule will not, if promulgated, have a significant economic impact on a substantial number of small entities.
This proposed 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 o,f "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Adminis-tration in 13 CFR Part 121.
Since these companies are dominant in their service areas, this proposed rule does not fall within the purview of the Act.
LIST OF SUBJECTS IN 10 CFR PART 50*
l l
Nuclear Power Plants and Reactors Accordingly, notice is hereby given that, pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as
- Comparative text has been used to indicate deletions and additions by l
dashing through and underlining.
l 8
Enclosure E i
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amended and Secti:n 553 ef Titla 5 of the United States Ctde, ad:ption of the following amendments to 10 CFR Part 50 is contemplated.
PART 50-DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES 1.
The authority citation for Part 50 reads as follows:
AUTHORITY:
Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2133, 2134, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat.1242,1244,1246, as amended (42 U.S.C. 5841, 5842, 5846), unless otherwise noted.
+
Section 50.7 also issued under Pub. L.95-601, sec.10, 92 Stat.
e 2951 (42 U.S.C. 5851).
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).
Sections 50.100-50.102 also issued under sec. 186, 68 Stat. 955 (42 U.S.C. 2236).
For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C.
2273), SS 50.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54, and 50.80(a) l are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b));
i SS 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 SS 50.55(e), 50.59(b), 50.70, 50.71, 50.72, and 50.78 are issued under sec. 1610, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).
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l 9
Enclosure E
m 2.
In 5 50.46(c)(1), (a)(2), and (a)(3) are amended to read as follows:
$ 50.46 Acceptance criteria for emergency core cooling systems for light water nuclear power l reactors.
w 4
(a)(1) [Except-as provided-in paragraph-(a)(23-and-(33-of-this section-] Each boiling and pressurized light water nuclear power reactor fueled with uranium oxide pellets within cylindrical Zircaloy cladding shall be provided with an emergency core cooling system (ECCS) which shall 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 shall be
+
calculated in accordance with an [acceptabie] evaluation model that has been accepted by the NRC staff and shall be calcylated for a number of postulated loss-of-coolant accidents of different sizes, locations, and i
l other properties sufficient to provide assurance that the entire spectrum i
of postulated loss-of-coolant accidents is covered.
[ Appendix-K--EEES Evaination-Models--sets-forth-certain required-and-acceptable-features of-evaination-modeis-] Except as provided in paragraph (a)(2) of this section, the evaluation model must include sufficient supporting justi-l l
fication to show that the analytical technique describes the behavior l
of the reactor system during a loss-of-coolant accident.
Comparisons to applicable experimental data should be made and uncertainties in the analysis method and inputs should be identified and assessed so that the uncertainty in the calculated results can be estimated.
This uncertainty shall be accounted for when the calculated ECCS cooling performance is compared to the criteria of paragraph (b) of this section.
Appendix K, 10 Enclosure E t
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q Part II, R:quir:d Documentatien, sets forth the documentatien requira-t*
ments for each evaluation model. One acceptable method for developing I
the evaluation model, calculating the ECCS cooling performance and performing the uncertainty analysis is givk in Regulatory Guide XX,
." Conformance with the criteria set forth in paragraph (b) of this section with ECCS cooling performance calculated in accordance with an [ acceptable] evaluation model that has been accepted by the NRC staff, may require that restrictions be imposed on reactor operation.
(2) Alternatively, an ECCS evaluation model may be developed in conformance with the required and acceptable features of Appendix K, ECCS Evaluation Models.
(3)(i) Each applicant for or holder of an operating license or construction permit shall estimate the effect of any change to an evalua-tion model that has been accepted by the NRC staff to determine if the change is significant.
For this purpose, a significant change 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 latest model that was accepted, or is a cumulation l
of changes such that the sum of the absolute magnitudes of the respec-tive temperature changes is greater than 50*F.
(ii) Each significant change to, or any error discovered in, an l
accepted evaluation model which affects the temperature calculation shall be documented and the effect of the change or correction of the error on the limiting ECCS analysis reported to the Director of Nuclear Reactor Regulation within 30 days.
Each applicant or licensee shall also include with the report a proposed schedule for completing actions 11 Enclosure E
h v.
-need:d to comply with applicable requirements.
This schedule may be developed using an integrated scheduling system previously approved for the facility by the NRC.
For those facilities not having an NRC approved integrated scheduling system, a schedule for achieving compliance will be established by the NRC staff within 60 days of receipt of the proposed schedule.
For any change or error correction that results in a calculated ECCS cooling performance that does not conform to the criteria set forth a
in paragraph (b) of this section, the affected applicant or licensee must notify the NRC promptly and propose immediate steps to bring plant design or operation into compliance.
(4) (Boilerplate on Reg Guide incorporation by reference.)
3.
In 10 CFR Part 50 Appendix K, paragrapb I.C.S.b is amended J
to read as follows:
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-Bengali-Rohsenew fiew-film-boiling-correlation-(R--S:-Bengali-and-W:-M:-Rohsenew; " Film Boiling-en-the-inside-of-Verticai-Tabes-with-Wpward-Fiew-ef-the-Finid-at tow-Qaalities;-MIT-Report-Namber-9079-26--Eambridge--Massachusetts-September-19633-] and the Westinghouse Correlation of steady-state l
transition boiling (" Proprietary Redirect / Rebuttal Testimony of Westing-house Electrical Corporation," USNRC Docket RM-50-1 page 25-1, October 26, 1982) are acceptable for use in the post-CHF boiling regimes.
In addi-(
tion the transition boiling correlation of McDonough, Milich, and King (J. B. McDonough, W. Milich, E. C. King, [uPartiai-Film-Beiiing-with l
12 Enclosure E l
I t
s.
Water-at-2000 psig-in-a-Roend-Vsrtical-Tobe 2-MSA-R: search-Eerp---Tech-nical-Report-62-(NP-6976h-(19583] "An Experimental Study of Partial
- ~
Film Boiling Region with Water at Elevated Pressures in a Round Vertical b
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 shall..be restricted as follows:
A A
A A
4.
In 10 CFR Part 50 Appendix K, paragraph II.1.b. is deleted.
5.
In 10 CFR Part 50 Appendix K, paragraph II.1.c. is redesignated II.I.b. and is amended to read as follows:
b.
A complete listing of each computer program, in the same form as used in the evaluation model, shall be furnished to the Nuclear Regulatory Commission upon request.
6.
10 CFR Part 50 Appendix K is amended by' revising Section M.S.
to read as follows:
5.
General Standards for Acceptability - Elements of evaluation models reviewed will include technical adequacy of the calculational methods, including compliance with required features of Section I of this Appendix K, if applicable, and provision of a level of safety and
. margin of conservatism comparable to other acceptable evaluation models, taking into account significant differences in the reactors to which they apply.
l Dated at Washington, DC this day of
, 1985.
For the Nuclear Regulatory Commission.
Samuel J. Chilk Secretary of the Commission 13 Enclosure E I
)
h ENCLOSURE "F" ENVIRdNMENTALASSESSMENT a
e, e
e h
l i.
}
e O
i s
I l
w Ey NUCLEAR REGULATORY COMMISSION
[10 C'FR PART 50]
Acceptance cr.iteria for Emergency Core Cooling Systems; Environmental Assessment and Draft Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (The Commission) is considering revisions to $50.46 and Appendix K of 10 CFR Part 50 which specify requirements of emergency core cooling systems (ECCS) for light water reactors.
ENVIRONMENTAL ASSESSMENT IDENTIFICATION OF PROPOSED ACTION:
Section 50.46(a)(1) would be revised to eliminate the requirement to use 4
the features of Appendix K when calculating ECCS performance during a loss-of coolant accident (LOCA).
The amended rule would require that the uncertainty of the calculation be evaluated and considered when comparing the results of the calculation with the temperature limits and other criteria of $50.46(b).
Section 50.46(a)(2) would be revised to allow continued use of the features of i
Appendix K as an alternative to the uncertainty evaluation required by the amended 550.46(a)(1).
Section 50.46(a)(3) would be revised to specify require-ments for reanalyses and reporting which are excluded from consideration in this environmental assessment per $51.22 of 10 CFR Part 51. Appendix K of 10 CFR Part 50 would be revised to make minor technical changes to the accept-able. features of the, calculations.
NEED FOR PROPOSED ACTION:
l The proposed revisions of 10 CFR Part 50 and Appendix K are required in order to permit new knowledge of ECCS performance gained through research to be used in the calculations of ECCS performance.
The improved calculations wouldallowrelaxationofrestrictionswhicharepreventingoptij$aloperation
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03/12/85 1
APPENDIX X ENC F i
- -r n,- m n, e--, o n..
onLJVRBers and are not nec;ss;ry t3 protect the health cnd safety of the w
v 1
- public, 3
t ENVIRONMENTAL IMPACTS OF THE PROPOSED ' ACTION:
The proposed revisions would be likely to redue I
the cladding temperatures that are calculated during a LOCA and allow an increase in the peak loca of the reactor, while still meeting the temperature limits and other criteria wer of $50.46(b).
An increase in the allowed peak local power could be used in either or both of the following manners:
1.
The total maximum allowed power of the reactor would be uncha but the flexibility in the allowed shape of the power used to improve plant efficiency.
More efficient fuel utilization, more flexibility in changing total power, and reduced derating of plants due to fuel limits might be possible.
2.
The total maximum allowed power of the reactor might be increased.
The expected maximum increase in total power for existing and currently reactors is approximately 5% based on practical limits of plant hardware.
Either of these actions would require an amendment to the plant license to change the technical specification limits and, therefore, would result in cn environmental assessment specific to that particular plant and the specific cmendment being considered.
This environmental assessment is a generic evalua-tion considering the typical impact of the rule revision.
A change in the allowed peak local power, without an increase in total power, would produce no significant environmental impact.
The total fission product inventory, routine releases of radioactive materials and thermal releases to the environment would be essentially unchanged.
Fuel cycle changes would be in the direction of improved use of fuel and should not significantly change the environmental impact of the fuel cycle unless major new fuel cycle cethods (e.g., plutonium recycle) were adopted.
Such changes are beyond the scope of this rule revision and are not considered.
03/12/85 2
APPENDIX K ENC F
r For th2 case of a small (i.e., 5%) incrcase in total power, there would 4p.
be a carrespondingly small increase in fissicn prtduct inv3nt0ry, routine y
releases of radioactivity and fuel use.
However, maximum allowed releases of radioactivity during both accident situations and during routine operation are a
specified by technical specifications and other sections of 10 CFR Part 50 which are unchanged by this revision.
It is not expected that the small increase in total power that could result from this revision would result in difficulty meeting the existing release limits.
An increase in total power would increase the thermal discharge to the environment by an amount approximately propor-tional to the increase in power.
The environmental impact of such an increase in thermal discharge would be site specific and a site specific environmental assessment would have to be performed prior to the license amendment that would be required for a change in total allowed power.
It is believed, however, that the small increase in thermal discharge (e.g., less than 5%) that might be possible with this rule revision should not significantly change the environ-mental impact of the thermal discharge for typical plant sites, a
ALTERNATIVES TO THE PROPOSED ACTION:
The staff has considered a number of alternatives to revise 650.46 and Appendix K.
However, all the alternatives considered would allow similar increases in local or total power, with the exception of the alternative of making no changes.
Since the environmental impact of the proposed revision is considered to be not significant and the revision is required to reduce unwar-ranted restriction on the operation of some reactors, the Commission has decided to proceed with the proposed revision.
AGENCIES AND PERSONS CONSULTED:
^
~
The NRC staff consulted U.S. manufacturers of nuclear power plants to determine the maximum increase in local *or total power that might result from application of the proposed rule revisions.
The staff did not consult other agencies or persons.
e 03/12/85 3
APPENDIX K ENC F
_,,,,,,r,
.y,--
DRAFT FINDING OF NO SIGNIFICANT IMPACT 4
?p-f-
The Commission has determined not to prepare an Environmental Impact Statement for the proposed action.
Theforegoing$nvironmentalassessmentof this action has concluded that the proposed action would not significantly effect the quality of the human environment.
Written comments or suggestions for consideration in connection with this Draft Finding of No Significant Impact should be submitted to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention Docketing and Service Branch.
Copies of comments received qpy be examined in the Commission's Public Document Room at 1717 H Street NW, Washing-ton, D.C. 20555.
The comment period expires (60 days following publication in the Federal Register).
Comments received after that date will be considered if it is practical to oo so, but ass,urance of consideration cannot be given except as to comments received on or before that date.
a Dated at Silver Spring, Maryland this day of '
1985.
For The Nuclear Regulatory Commission R.B. Minogue Director, Office of Nuclear Regulatory Research i.
03/12/85 4
APPENDIX K ENC F
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?
USER OFFICE CONCURRENCE
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4 UNITED STATES i
E' "
7.
NUCLEAR REGULATORY COMMISSION y/l!
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7 y
E
- ,E 4wAsHWGTON. D. C. 20555
,/
APR 3 as MEMORANDUM FOR:
Robert M. Bernero, Director Division of Systems Integra n,-
Office of Nuclear Reactor (gv1'ation FROM:
- 0. E. Bassett, Director Division of Accident Evaluation Office of Nuclear Regulatory Research
SUBJECT:
DRAFT RECOMMENDATIONS TO EDO CONCERNING WHETHER AND HOW TO CONTINUE WITH ONGOING RULEMAKING SPONSORED BY RES,
" ACCEPTANCE CRITERIA FOR EMERGENCY CORE COOLING SYSTEMS (ECCS) FOR LIGHT WATER NUCLEAR POWER PLANTS" Enclosed for your consideration are draft recommendations supported by a draft office review concerning whether and how to continue with an ongoing rulemaking sponsored by RES for which your office is identified as the user office.
This memorandum constitutes my concurrence in the enclosed draft reconnendation.
? $ ^ ^ M " " ' '- - - 9 " ' - - - ^ ~ - ' ^ _.. _ '
"t:
ith or 6 60 on the draft reconsnendations as indicated below.
Under present ED0 guidelines, the EDO will not consider or transmit to the Commission any rulemaking which has not been subjected to an independent staff review to determine whether the rulemaking action should proceed. Therefore, once I have your agreement that the rulemaking should proceed, I plan to expedite dispatch of the enclosed package for RES independent review to avoid any unnecessary delay in processing of the rulemaking package now in preparation.
> >hh
. E. Bassett, Director Division of Accident Evaluation Office of Nuclear Regulatory Research
Enclosures:
1.
Draft Recommendations 2.
Rulemaking As Currently Proposed 3.
Task Leader Evaluation 4.
NRC Regulatory Agenda Entry 5.
Supporting Documentation Receipt Acknowledged:
See Page 2
A G.
j'?
Robert M. Bernero '+
6 Receipt acknowledged.
Comments are as follows.
1 I
/s Robert M. Bernero, Director Division of Systems Integration Office of Nuclear Reactor Regulation 4
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