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', I UNITED STATES ,7 N U C L E A R , R E.G U L A TO R Y COMMISSION i
.W AS HIN G TON. D. C. 205 5 5 l
Docket tios.JTN 50-447 and STN 50-531 -HEC::ygg General Electric Company ~'" "
ATTth Ivan F. Stuart JUI~ ' , w "a Manager, Safety and Licensing ~
175 Curtner Avenue 'W,...
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San Jose, California 95125
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%c * %- M_ S6 i I Gentlemen: -
g* p The enclosure to this letter which will be transmitted to all licensees ard applicants identifies information that we require in order to i
complete our review of Light Water Reactor Emergency Core Cooling Systems performance to verify conformance to 10 CFR 50.46. The information i requested must be filed on the GESSAR-238 and GESSAR-251 dockets either directly or by reference as pert of the Standard Safety Analysis Report. We will require that this information be reviewed and approved by the HRC (PDA).for staffapplications.
these prior to issuance of a Preliminary Design Approval !
Please contact us if you require additional discussion or clarification of the information requested.
Sincerely.
I' gts @
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. / Ndi Voss A. Moore, Assistant Director h ,
Light Water Reactors, Group 2 Divisinn of Reactor Licensing
Enclosure:
Request for Additional Information -
cc: Mr. W. Gilbert, Manager Safety and Standards Mr. L. Gifford, Manager Regulatory Operations Unit General Electric Company General Electric Company 175 Curtner Avenue 4720 Montgomery Lane San Jose, (faiffornia !TS11tw -4etheWh >
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l 8707070690 870623 PDR FOIA THOMAS 87-40 PDR ,
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Attachment 'l 1 REQUIRED INFORMATION 1
- 1. Break Spectrum and Partial Loop Operation _ i The information provided for each plant shall comply with the )
provisions of the attached memorandum entitled, " Minimum Requirements for ECCS Break Spectrum Submittals."
The ECCS syctem in each plant shou'1d be evaluated by the applicant (or licensee) to show that significant changes in chemical concentrations will not occur durirg the long term after a loss-of-coolant accident (LOCA) and these potential changes have been specifically addressed by appropriate operating procedurec. Accordingly, the applicant should review the system capabilities and operating procedures to assure that boron precipitation would not compromise long-term core cooling capability following a LOCA. This review should , consider all aspects of the specific plant design, including component qualification in the LOCA environment in addition to a detailed review of operating procedures. The applicant should examine the vulnerability of the specific plant design to single failures that would result in any significant boron precipitation.
, 3. Single Failure Analysis
' A single failure evaluation of the ECCS should be provided by the applicant (or lice'nsee) for his-specific plant design, as required by Appendix K to 10 CFR 50, Section I.D.1. In performing this evaluation, the effects of a single failure or operator error that causes any manually i controlled, electrically-operated valve to move to a position that could adversely affect the ECCS must be considered. Therefore, if this consid-cration has not been specifically reported in the past, the applicants upcoming submittal must address this consideration. Include a list of all of the ECCS valves that are currently required by the plant Technical Specifications to have power disconnected, and any proposed plant modifications and changes to the Technical Specifications that might be required in order to protect against any loss of safety function caused by this type of failure. A copy of Branch Technical Position EICSB 18 from the U.S. Nuclear Regulatory Commission's Standard Review Plan is attached to provide you with guidance.
The single f ailure evalurtion should include the potential for passive f ailures of fluid systems during long term cooling following a LOCA as well as single failures of active components. For PWR plants, the single failure analysis is to consider the potential boron concentra-problem as an integral part of long term cooling.
- 4. Submerged Valves The applicant should review the specific equipment arrangement with-in his plant to determine if any valve motors within containment will become submerged following a LOCA. The review should include all valve motors that may become submerged, not only those in the safety injection system. Valves in other systems may be needed to limit boric acid con-centration in the reactor vessel during long term cooling or may be l
required for containment isolation.
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. The applicant (or licensee) is to provide the following information, for each plants (1) Whether or not any valve motors will be submerged following a LOCA in the plant being reviewed.
(2) If any valve motors will be flooded in their plant, the applicant (or I licensee) is tos j 1
(a) Identify the valves that will be submerged. _j (b) Evaluate the potential consequences of flooding of the valves j for both the short term and long term ECCS functions and containment isolation. The long term should consider the s potential problem of excessive concentrations of boric acid in PWR's.
(c) Propose a interim solution while necessary modifications are being designed and implemented. (currently operating plants j only).
(d) Propose design changes to solve the potential flooding problem.
- 5. Containment Pressure (PWR's Only)
The containment pressure used to evaluate the performance capability of ,
the ECCS shall be calculated in accordance with the provisions of Branch Technical Position CSB 6-1, which is enclosed.
Plants that have a Westinghouse nuclear steam supply shall perform their_ECC3 analyses utilizing the proper version of the evaluation model, as defined below:
(1) The December 25, 1974 version of the Westinghouse evaluation model, i.e., the version without the modifications described in WCAP-8471 is acceptable for previously analyzed plants for which the peak clad temperature turnaround was identified prior to the reflood rate decreasing below 1.1 inches per second or for which the reflood rate was identified to remain above 1.0 inch per second; conditions for which the December 25, 1974 and March 15, 1975 versions would be equivalent.
(2) The March 15, 1975 version of the Westinghouse evaluation m'odel is an acceptable model to be used for all previously analyzed plants for which the peak clad temperature turnaround was identi-fied to occur after the reflood rate decreased below 1.1 $nches per cecond, and for which steam cooling conditions (reflood rate less than 1 inch per second) exist prior to the time of peak' clad temperature turnaround. The March 15, 1975 version will be used for all future plant analyses.
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O I$ 1975 MINIMUM REQUIREMENTS FOR ECCS BREAK SPECTRUM SUBMITTALS I. INTRODUCTION The following outline shall be used as a guideline in the evaluation of LOCA break spectrum submittals. These guidelines have been formulated for contemporary reactor designs only and must be re-assessed when new reactor l
concepts are submitted.
The current ECCS Acceptance Criteria requires that ECCS cooling performance be calculated in accordance with an acceptable evaluation model and 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. In addition, the calculation is to be conducted with at least three values of a discharge coefficient (CD ) applied to the postulated b.reak area, these values spanning the range from 0.6 to 1.0.
Sections IIA and IIIA define the acceptable break spectrum for most operating plants which have received Safety Orders. Sections IIB and IIIB define the break Spectrum requirements for most CP and OL case work (exceptions noted 3ater). Sections IIC and IIIC provide an outline of the minimum requirements for an acceptabic complete break spectrum. Su'ch a complete break spectrum could be appropriately referenced by some plants. Sections IIID and IIIE
. provide the exceptions to certain plant types noted above.
A plant due to reload a portion of its core will have previously submitted all
.or part of a break spectrum analysis (either by reference or by specific calculations). If it is the intention of the Licensee to replace expended fuel with new fuel of the same design (no mechanical design differences which could affect thermal and hydraulic performance), and if the Licensee intends to operate the reloaded core in compliance with previously approved Technical Specifications, no additional calculations are required. If the reload core design has changed, the Licensee shall adopt either of Sections IIA or IIC, or of Sections IIIA or IIIC of this document, as appropriate to the plant type (BWR or PWR). The criterion for establishing whether paragraph A or C shall be satisfied will be determined on the basis of whether the Licensee can demonstrate that the shape of the PCT versus break size curve has not When been modified as a consequence of changes to the reload core design. !
the reload is supplied by a source other than the NSSS supplier, the break ;
spectrum analyses specified by Sections IIC or IIIC shall be submitted Additional as a sensitivity l minimum (as appropriate to the plant type, BWR or PWR). !
studies may be required to assess the sensitivity of fuel changes in such areas as single failures and reactor coolant pump performance.
II. PRESSURIZED WATER REACTORS A. Operating Reactor Reanalyses (Plants for which Safety Orders were issued)
If calculational changes
- were made to the LBM** to make it wholly in
- Calculational changes /Model changes--those revisions made to calculational techniques or fixed parameters used for the referenced complete spectrum.
- LBM--Large Break Model; SBM--Small Break Model l
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conformance.with 10CFR50, Appendix K, the following minimum number of break sizes should be reanalyzed. . Each ' sensitivity study performed during the i development of the ECCS evaluation model shall be individually verified as -l
~j remaining applicable, or shall be repeated. A plant may reference a. break spectrum analysis conducted on another plant if it is the same configuration and core design.
- 1. .If the largest break size results in'the highest PCT:
a.. Reanalyze the limiting break.
- b. Reanalyze two smaller breaks.in the large break region.
- 2. If the largest break size does not result in the highest PCT:
- a. Reanalyze the limiting break,
- b. ' Reanalyze a break larger.and'a break' smaller than the limiting break. If the limiting break is outside the range of Moody multipliers of 0.6 to 1.0 (i.e., less than 0.6), then the limiting break plus two larger breaks must be analyzed. .
. 'i If calculational changes have been made, to the SBM to make it wholly in conformance with 10CFR50,~ Appendix K, the analysis of the worst small break (SBM) as previously determined from paragraph C below should be repeated.
B. New CP and OL Case Work A complete break spectrum should be provided in accordance with paragraph C below, except for the following:
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- 1. If a new plant is of the same general design as the plant used a,s a- '
basis for a referenced complete spectrum analysis, but operating parameters have changed which would increase PCT or metal-water 0 reaction, or approved calculational changes resulting in more than 20 F 1 change' in PCT have been made to the ECCS model used for the referenced '
complete spectrum, the analyses of paragraph A above should be provided plus a minimum of three small breaks (SBM), one of which is the transition break.* The shape of the break spectrum in the' referenced analysis should be justified as remaining applicable, including the sensitivity studies used as a basis for the ECCS evaluation model.
- 2. If a new plant (configuration and core design) is applicable to all generic studies because it is the same with respect to the generic plant design and parameters used as a basis for a referenced complete spectrum defined in paragraph C, and no calculational changes resulting in more than 20 0F change in PCT were made to the ECCS model used for the referenced complete spectrum, then no new spectrum analyses are required. The new plant may instead reference the applicable analysis. )
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3 C. Finimum Requirements for a Complete B eak Spectrum Since it is expected that applicants will prefer to reference an applicable complete break spectrum previously conducted on another plant, this paragraph defines the minimum number of breaks required for an acceptable complete break spectrum analysis, assuming the cold leg pump discharge is established as the worst break location. The worst single failure and worst-case reactor coolant pump status (running or tripped) shall be established utilizing appropriate sensitivity studies. These studies
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should show that the wor.=t single" failure has been justified as a function of break size. Each sensitivity study published during the development of the ECCS evaluation model shall' be individually justified as remaining applicable, or shall be repeated'. Also, a proposal for partial loop operation shall be supported by identifying and analyzing the worst break site and location (i.e., idle loop versus operating loop). In addition, sufficie~nt justification shall be pro.vided to conclude that the shape of the PCT versus Break Size curve would not be significantly altered by the partial loop configuration. Unless this information is provided, plant !
Technical Specifications shall not permit operation with one or more !
idle reactor coolant pumps.
a It must be demonstrated that the containment design used for the break spectrum analysis is appropriate for the specific plant analyzed. It should be noted that this analysis is to be performed with an approved evaluation model wholly in conformance with the current ECCS Acceptance Criteria. -
- 1. LBM--Cold Leg-Reactor Coolant Pump Discharge
- a. Three guillotine type breaks spanning at least the range of Moody multipliers between 0.6 and 1.0.
- b. One split type break equivalent in size to twice the pipe cross-sectional area.
- c. Two intermediate split type breaks.
- d. The large-break /small-break transition split.
- 2. LBM--Cold Leg-Reactor Coolant Pump Suction Analyze the largest break size from part 1 above. If the analyses in part 1 above should ind' : ate that the worst cold leg break is an intermediate break size, then the largest break in the pump suction should be analyzed with an explanation of why the same trend would not apply.
- 3. LBM--Hot Leg Piping Analyze the largest rupture in the hot leg piping.
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- 4. SBM--Splits Analyze five different small break sizes. One of these breaks must include the transition split break. The CFT line break must be analyzed for B&W plants. This break may also be one of the five small breaks.
III.. BOILING WATER REACTORS The generic model developed by General. Electric for BWRs proposed that split and guillotine type breaks.are equivalent in determining blowdown phenomena.
The staff concluded this was acceptable and that the break area may be considered at the vessel nozzle with a zero loss coefficient using a two' phase critical flow model. Changes in the break area are equivalent to changes in the Moody multiplier.
The minimum number of breaks required f'or a complete break spectrum analysis, assuming a-suction side recirculation line break is the design basis accident
.( DBA) and the worst single failure has been established utilizing appropriate sensitivity studies, are shown in paragraph C' below. Also, a proposal for partial loop operation shall be supported by identifying and analyzing
. the worst break' size and location (i.e., idle loop versus operating loop). In addition, suf ficient justification shall be provided to conclude that the shape of the PCT .versus Break Size curve would not be significantly altered by the partial loop configuration. Unless this information is provided, plant Technical Specifications shall not permit operation with one or more idle reactor coolant pumps. _
A. BWR2, BWR3, and BWR4 Reanalysis (Plants for which Safety Orders were issued)
If the referenced lead plant analysis is in accordance with Section III.
paragraph C below, the following minimum number of break sizes should be reanalyzed. It is to be noted that the lead plant analysis is to be performed with an approved evaluation model wholly in conformance with the current ECCS Acceptance Criteria. A plant may reference a break spectrum analysis conducted on another plant if it is the same configuration and core design.
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Each sensitivity study published during the development of the ECCS 1 evaluation model shall be ir aividually justified as remaining applicable, l or shall be repeated.
- 1. If the largest break results in the highest PCT:
- a. Reanalyze the limiting break with the appropriate referenced sing 2e failure.
- b. Reanalyze the worst small break with the appropriate referenced single failure.
- c. Reanalyze the transition break with the single failure and model ,
that predicts the highest PCT.
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- 2. If the largest break does not result in the highest PCT:
a.- Reanalyze the limiting break, the largest break, and a smaller break.
If calculational changes have been made to the SBM to make it wholly in 4 conformance with 10CFR50, Appendix K, reanalyze the small break (SBM) in accordance with Section IIIC.
B. New CP and OL Case Werk A complete break spectrum should be provided in accordance with Section III, ,
paragraph C below, except for the following:
- 1. If a new plant is of the same general design as the plant used as a ,
basis for the lead plant analysis, but operating parameters have !
changed which would increase PCT or metal-water reaction, or approved calculational changes have been made to the ECCS model resulting in more than 200F change in PCT, the analyses of Section III, paragraph A above should be provided plus a minimum of three small breaks (SBM),
one of which is the transition break. The shape of the break spectrum in the lead plant analysis should be justified as remaining applicable, including the sensitivity studies used as a basis for the ECCS evaluation model.
- 2. If.a new plant (configuration or core design) is applicable to all generic studies because it is the same with respect to the generic plant design and parameters used as a basis for a referenced complete spectrum defined in paragraph C, and no calculational changes resulting in more than 200F change in PCT were made to the ECCS model used for the referenced complete spectrum, then no new spectrum analyses are required.
The new plant may instead reference the applicable analysis.
C. Minimum Requirements for a Complete Break Spectrum This paragraph defines the minimum number of breaks required for an acceptable complete spectrum analysis. This complete spectrum analysis is.
required for ecch of the lead plants of a given class (BWR2, BWR3, BWR4, {
BWR5, and BWR6). Each sensitivity study published during the development of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated.
- 1. Four recirculation line breaks at the worst location (pump suction or discharge), using the LBM, covering the range from the transition j break (TB) to the DBA, including Cp coefficients of from 0.6 to 1.0 times the DBA.
- 2. Five recirculation line breaks, us'ing the SBM, covering the range from the smallest line break to the TB.
- 3. The following break locations assuming the worst single failure:
- a. largest steamline break
- b. largest feedwater line break
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- c. largest core spray line break
- d. largest. recirculation pump discharge or suction break (opposite side of worst location)
D. BWR4 with " Modified" ECCS-j Same as Section IIIC.
E. BWR5 1 I
Same as Section IIIC.
F. BWR6
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IV. . LOCA PARAMETERS OF INTEREST A. On each plant and for each break analyzed, the following parameters (versus time unless otherwise noted) should be provided on engineering
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graph paper of a quality to facilitate calculations.
--Peak clad temperature (ruptured and unruptured node)
--Reactor vessel pressure
--Vessel and downcomer water level (PWR only)
--Water level inside the shroud (BWR only)
--Thermal power
--Containment pressure (PWR only)
B. For the worst break analyzed, the following additional parameters (versus time unless otherwise noted) should be providedThe on worstengineering single graph paper of a quality to f acilitate calculations.
failure and worst-case reactor coolant pump status will have been p established utilizing appropriate sensitivity studies.
--Flooding rate (PWR on1"'.
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--Core flow (inlet and c.i. -
--Core inlet enthalpy (BWR only)
--Heat transfer coefficients
--MAPLHGR versus Exposure (BWR only)
--Reactor coolant temperature (PWR only)
--Mans released to containment (PWR only)
--Energy released to containment (PWR only)
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--PCT versus Exposure (BWR only)
--Containment condensing heat transfer coefficient (PWR only)
--Hot spot flow (PWR only)
--Quality -(hottest . assembly) (PWR only)
--Hot pinLinternal pressure
--Hot spot pellet average temperature
--Fluid temperature (hottest assembly)'(PWR only)
C. A tabulation of peak clad temperature and metal-water reaction (local-and core-wide) shall be'provided across the break spectrum.
D. - Sa'fety Analysis Reports (SARs) filed with the NRC shall identify on' each plot the run date,. version number, and version date of'the computer model utilized for the LOCA analysis. 'Should differences exist in version number or version date from the most current code listings made "available to the NRC staff, then each modification shall be identified
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with an assessment of impact upon PCT an'd metal-water reaction (local and core-wide).
E.- A tabulation of times at which significant events occur shall be The following provided on each plant.and.for each break analyzed.
events shall be included as a minimum:
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--End-of-bypass (PWR only)
--Beginning of core recovery (PWR only)
--Time of rupture
--Jet pumps uncovered (BWR only)
--MCPR (BWR only) l
--Time of rated spray (BWR only) +
--Can quench (BWR only)
--End-of-blowdown
--Plane of interest uncovery (BWR only)
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l Possible grouping of plants for the purpose of performing generic as well as individual plant break spectrum analyses. l l
CURRENT 00CKETED i
APPLICATIONS i
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BABCOCK AND WILCOX - . i.
CATEGORY I: 177 FA w/ Lowered Loops Arrangement Re-analysis (Safety Order Plants):
1,2,3 -- IIA i These plants must. resubmit at- {
Ocyggg least 3 breaks. (They will do !
Three Mile Island 1 -- IIA so by reference to a complete 2535 f. break spectrum reanalysis sub-
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Arkansas Power 1 -- IIA mitted generically by B&W.) l 2563 l Rancho Seco -- IIA j 2772 - ,
New OLs: - !
Three Mile Island 2 --IIB (2)l since these plants are the same design as the above plant, they Crystal. River 3 --IIB (2) v. ; may reference the same reanalysis 4 2452 of the complete spectrum above.
Midland 1, 2 --IIB (2).f New cps-i a
l None ,y CATEGORY II: 177 FA w/ Raised Loop Arrangement New OLs:
Davis Besse 1 --IIB Complete spectrum required.
New cps Davis Besse 2, 3 --IIB Complete spectrum required.
CATEGORY III: 205-FA Plants New OLs:
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s New CP=s:
Bellefonte 1, 2 -- IIB )
Complete spectrum required _.
Greenwood 2, 3 -- IIB (Plans are for all to reference a complete spectrum submitted
( Probably on WPPSS.)
i' WPPSS 1, 4 -- IIB f Pebble Springs 1, 2 -- IIB j CATEGORY IV: 145-FA Plants New OLs:
None i
L New cps: 1 North Anna 3, 4 -- IIB Complete spectrum required _. 1
}P (One will probably reference Surry 3, 4 -- IIB J theother.)
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. GENERAL ELECTRIC i
i B"R-2_' Oyster Creek -- LP* Complete spectrum require'd. (IIIA)**
. Nine Mile Point -- Reference only required. (IIIA)-
BWR-3 Quad Cities 2 -- LP* Complete spectrum required. (IIIA)**
2511 Millstone- -- IIIA - 3 breaks required 2011 ,
Monticello -- IIIA - 3 breaks required 1 i
1670 '
Dresden 2, 3 --IIiA B 2527 L May reference LP Quad Cities 1 -- IIIA ) -
i 2511' Pilgrim -- IIIA - 3 breaks required 1998 BWR-4 Without fix Hatch 1 -- LP* Complete spectrum required. ( III A)** l 2436 a
Peach Bottom 2 -_3 -- III A ].Completespectrumrequired. One may reference the other, Browns Ferry 1, 2, 3 -- IIIA 3293 '
i Cooper -- IIIA
> l 2381 f3 breaks required. !
Fitzpatrick -- IIIA [' '
2436 ) Hatch I may serve Duane Arnold -- IIIA - 3 breaks required as a reference 1658 Hatch 2' -- IIIA
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for the others. '
2436 _ _:
Brunswick 1 -- IIIA ,
2436 '
Shoreham -- IIIB Fermi -- IIIB :
1 Newbold -- IIIB.
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- Lead Plant 1
- Original break spectrum not wholly in conformance with 10CFR50, Appendix K. ,
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1 BWR-4 dithfixBrunswick2(LeadPlant) -- IIIA - Complete spectrum l 2436 required.**
Vermont Yankee -- IIIA - 3 breaks required (Lead Plant can be 1593 referenced, if Browns Ferry
- 1, 2, & 3 S appropriate)
Peach Bottom
- 2, 3 > See preceding page Fitzpatrick* / ;
BWR-5 Lead Plant -- IIIE - Complete spectrum required. l 5 1 l
Nine Mile Point 2 -- IIIB l i Complete spectrum required. j LaSalle 1, 2
. -- IIIB $ (Lead Plant con be referenced ,
f by other BWR-5 plants, if Bailly -- IIIB appropriate.) l i
i Zimmer -- IIIB .
k Susquehanna 1, 2 --IIIB} q l
BWR-6 Lead plant -- IIIF - Complete spectrum required. '
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" 3 Grand Gulf -- IIIB 1 Black Fox
-- IIIB Barton 1, 2, 3, 4 -- III.B Complete spectrum required. (Lead j Perry 1, 2 -- IIIB ' Plant can be referenced by other i BWR-6 plants, if appropriate.) l Clinton 1, 2 -- IIIB ]
Douglas Point -- IIIB l Hanford 2 -- IIIB i
Skagit 1, 2 -- IIIB j 1
Hartsville -- IIIB Somerset -- IIIB River Bend Station -- IIIB Allens Creek -- IIIB
- May or may not have the LPCI fix q
- Original break spectrum not wholly in conformance with 10CFR50, Appendix K.
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PLANT SPECIFIC, Oyster Creek -- IIIA Complete spectrum required. l' Nine Mile Point. -- IIIA " "
Limerick 1, 2 -- IIIB Hope Creek -- IIIB "
Humboldt-Bay -- IIIA Dresden 1 --
IIIA ,
a " i Big Rock -- IIIA ,
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- COMBUSTION ENGINEERING The following list is grouped according to similarities in design. !
Some of'the older, operating plants are fairly unique, as indicated, i and don't fall conveniently into any other groups. The list is in approx, chronological order.
- 1. Palisades (Unique) -- IIA S'
- 2. Ft. Calhoun (Unique) -- IIA 3 breaks required
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- 3. Maine Yankee (Unique) -- IIA J l
- 4. 2560 MWt Series
- a. Calvert Cliffs Units 1 & 2 -- IIA - 3 breaks required Millstone Unit 2 Complete spectrum required.
- b. -- IIB
)> (One may reference the other.)
- c. St. Lucie 1 -- IIB s
- d. St. Lucie 2 -- IIB . Complete spectrum required 5 '. 3400 MWt Series ( 3410 MWt 217 Fuel Assemblies) 1
- a. Pilgrim 2 ,(3470 Mwt) --IIB) .
Complete spectrum required.
b.. Forked River 1 -- IIB ht (One may reference the other.)
- c. San Onofre 2 & 3 -- IIB
- d. Waterfoid 3 -- IIB j l
- 6. Arkansas Class ( 2900 MWt 177 Fuel Assemblies)
- a. Russelville 1 -- IIB '
Complete spectrum required.
(One may reference the other.)
- b. Blue Hills 1 -- IIB s
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- j Maine Yankee is unique in that it has 3 steam generators, 3 hot legs and j
3 cold legs. All other CE plants have 2 steam generators, 2 hot legs and 4 cold legs. l 1 '
+ *All plants shown above listed before St. Lucie 2 are of the 14x14 fuel design..
All plants after, and including St. Lucie 2 are 16x16.
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- 7. System 80 Class-(CESSAR) -- IIB. Complete spectrum required These plants have not all been named yet. The utility and approx.
1 number of plants expected are as follows:
- a. Duke (6) \
- b. WUPPS (1) . May reference complete I - spectrum, if applicable.
- c. Arizona Power and Light (2),
- d. TVA(2) ) - l l
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e Westinghouse Ope'ratina Reactors (Safety Order Plants)*
2-loop 3-loop 4-loop
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Ginna Surry 1/2 Yankee Rowe Kewaunee Turkey Pt. 3/4 IP2 Pt. Beach 1/2 H. B'. Robinson 2 D. C. Cook 1 Prairie Island 1/2 Zion 1/2 Operating License **
a 2-loop 3-1000 4-loop i
Beaver Valley 1 - Trojan * /
Farley 1/2 - Salem 1/2*
North Anna 1/2 - Diablo Canyon 1/2*
IP-3 D. C. Cook 2 McGuire 1/2 Sequoyah 1/2
- 3 breaks required (IIA). One plant may reference another if applicable.
- Complete spectrum required. One plant may reference another if applicable (see paragraph IIB).
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1 Construction Permit **
1 2-1$op 3-loop 4-1000
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Sharon Harris 1/4 Byron /Braidwood 1/2 l
. North Coast. 1 Koshkonong 1/2 Cataea 1/2 {
Summer 1 Floating Nuclear 1/8 l .
Beaver Valley 2 Jamesport 1/2 Wisconsin Utilities Seabrook 1/2 SNUPPS 1-5 South Texas 1/2 Comanche Peak 1/2 j
Watts Bar 1/2 J
Millstone 3 Vogtle 1/2
- Complete spectrum required. One if applicable (see paragraph IIB) plant may reference anot I
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l BRANCH TECHNICAL POSITION E!C5818 APPLICATION OF THE SINGLE FAILURE CRITERIO4 TO mat.vALLY CONTROLLED ELECTRICALLY-OPERATLD VALVES A. BACKGROUND Where a single failure in an electrical system can result in loss of capability to perforr.
a safety function, the effect on plant safety.must be evaluated. 1his is necessary regard-less of whether the loss of safety function is caused by a co:ponent failing to ;crform a ]
requisite mechanical motion, or by a coeponent performing an undesirable a4chanical ration. {
l This position establishes the acceptability of disconnecting power to electrical components of a fluid system as one means of designing against a single failure that eight cause an un- {
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desirrhle co-ponent action. These provisions,are based on the assurption that.the co".ponent [
is the equivalent to a similar corponent that is not designed for electrical operation, e.g., a valve that can be opened or closed only by direct canual operation of the valve.
They are also led on the assu*rption that no single failure can both restore PC'..'er to the electrical systam and cause mechanical motion of the corponents served by the electrical d system. .The validity of these assumptions should be verified when applying this rcsition.
[ B. BRANCH TEC CICAL POSITION
- 1. Failures in both the " fail to functfon" 5(nse and the " undesirable functicC se .se of cor:ponents in electrical systens of valves and other fluid syste* co.-Xre".tn sto ld ;
be considered in, designing against a sini.le failure, even though the valve or ct'.er fluid syster.i component may not be called upon to fMetion in a given safcty e?crational sequence.
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- 2. Where it is determined that failure of an electrical systen co.Donent ca*. CaJse undesired ee'chanical r:otion of a valve or other fluid system cc-ponent and tds motion results in loss of the systen scfety function, it is acecptable, in lies of design changes that also may be acceptable, to disconnect power to the electric systers of the valve or other fluid system component. The plant technical specifications should i include a list of all electrically-operated valves, and'the recuired positions of these valves, to which the requirement for removal of electric power is applied in order to satisfy the single failure criterion.
- 3. Electrically-operated valves that are classified as " active" valves, i.e., are required to open or close in various safety syster operational sequences, but are ranually-controlled, should be operated from the, r.aln control room. Such valves may not be included among those valves from which ,ower is removed in order to freet the single failure criterion unless: (t) electrical power can be restored to the valves from the main control room,(b) valve operation is not necessary for at least ton minutes k following occurrence of the event requiring such operation, and(c) it is demonstrated 7A 27
- - a - _ . _ . _ - _ . . . _ _ - . _ _ _ _ . - _ - - - - .
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- * 'st that there is reasonable assurance that all necessary operator actions will be per- ,, s/
formed within the time shown to be adequate by the analysis. The plant technical specifications should incluoe a list of the required positions of manually-controlled, electrically-operated valves and should identify those valves to which the require-
- l ment for removal of electric power is applied in order to satisfy the single failure criterion.
- 4. When the single failure criterion is satisfied by removal of electrical power from valves described in(2) and (3), above, these valves should have redundant position i indication in the main control room and the position indication system should.itself.
meet the single failure criterion.
- 5. The phrase " electrically operated valves" includes both valves operated directly bj en electrical device (e.g., a motor-operated valve or a solenoid operated valve) and those valves operated indirectly by an ciectrical device (e.g., an air-operated valve whose air supply is controlled by an electrical solenoid valve).
C. REFERENCES
-4. Hemorandum to R. C. DeYoung and V. A. Moore from V. Ste11o. October 1. 1973.
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I BRANCH TECHNICAL POSITION CSB 6-1 .
MINIMUM CONTAINMENT PRESSURE MODEL FOR PWR ECCS PERFORMANCE EVALUATION A. BACKGROUND
^
j Paragraph I.D.2 of Appendix K to 10 CFR Part 50 (Ref. 1) requires that the containment i
' pressure used to evaluate the perfont.ance capability of a pressurized water reactor (PWR) emergency core cooling system (ECCS) not exceed a pressure calculated conservatively for that purpose. It further requires that the calculation include the effects of operation of i
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all installed pressure-reducing systems and processes. Therefore, the following branch technical position has been developed ato provide guidance in the perfonr.ance of mininum containment pressure analysis. The approach described belos applies only to the ECCS-related containrr.cnt pressure evaluation and no( to the containment functional capability evaluation for postulated design basis accidents.
B. BRANCH TECHNICAL POSITION a 1. _!_nput Information for Mudel
- a. Initial Contain ent Internal Conditions The minicum contairmnt gas temperature, minimum containt ent pressure, and maximum humidity that may be encountered under lin.iting nornal operating
. conditions shnuld be used,
- b. Initial Outside Contain ent Arbirnt Conditions A reasonably low ambient temperature external to the contain .cnt sbc.ld be used.
- c. Containment Volure )
The maximum not free co'itain. ment volu c sbeuld be used. This taxi.un free l volume should be deten11ned fro : the gross containment volu e minus the voluves l of interrial structures such as walls and floors, structural steel, rajor ecuip .ent, and piping. The individual volume calculations should reflect the uncertainty in the component volumes.
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- 2. Active Heat Sinks
- 4. Spray and Fan Coolino Systems The operation of all engineered safety feature containment heat removal systems l operating at maximum heat rernoval capacity; i.e., with all containment spray trains operating at maximum flow conditions and all emergency fan cooler units operating, should be assumcd. In addition, the minimum temperature of the stored water for the spray cooling system and the cooling water supplied to the fan coolers, based on technical specification limits, should be assumed.
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Deviations from thc' foregoing will be accepted if it can be shewn that the worst conditions regarding a single active failure, stored water temperature, and cooling water temperature have been selected from the standpoint of the overall (CCS model,
- b. Contain..cnt stea g ixing With Spilled ECCS Water The spillage of subcooled ECCS water into' the containment provides an additional
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-heat sink as the subcooled ECCS water mixes with the steam in the containment.
. The effect of the steam-water mixing should be considered in the contain. ment l
--pressure calculations.
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- c. Containment Steam Mixinc 1.'ith Water from lee Melt The water resulting from ice melting-in an ice condenser contaira,cnt provides an ,
additional heat sink as the sub,codied water mixes with the steam while draining g from the ice condenser into the lower containment volume. The effcet uf the l :! steam water mixing should be considered in the containment pressure calculations.
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I l ,l 3. Rissive Heat Sinks l a. Identification ,
The passive heat sinks that should be included in the containment evaluation a model should bc established by identifying those structures and components wittin the containment that could influpice the pressure response. The kinds of strue-tures and components that should be included are listed in Table 1. l sm
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Data on passive heat sinks haveJecn compiled frora previo s revie.<5 and have -
been used as a basis for the sir:iplified model outlined below. This rodel is i acceptabic for mini.tum containment pressure analyses for construction permit applications, and until such time (i.e., at the o;,erating license revic d inat a '
complete identification of available heat sinks can be made. This sirplified approach has also been followed.for o; crating plants by licensees co' plying wits ,
Section $0.46 (a)(2) of 10 CFR Part $3. For such cases, and for constra tion permit reviews, where a detailed listing of heat sinks within the centai--en', '
. of ten cannot be provided, the following procedure ray be used to model tae passhe heat sinks within the containment:
(1) Use the surface area and thickness of the primary containment steel shell er ,
steel liner and associated anenors and concrete, as appropriate, ,
I (2) Estimate the exposed surface area of other steel heat sinks in acecrdance with Figure 1 and assume an average thickness of 3/8 inch.
(3) Model the internal concrete structures as a slab with a thickness of I foot j 2
and exposed surface of 160,000 ft . f 1
i The heat sink thermophysical proporties that would be acceptable are shown in l
- Table 2. a
) 6.2.1.5-4 .
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' 4 At the operating license stage, applicants should provide.a detailed list of passive heat sinks. with appropriate dimensions and properties.
- b. Heat Transfer Coefficients The following conservative condensing heat transfer coefficients for heat transfer ,
to the c.xposed passive heat sinks during the blowdown and post-blowdown phases of the loss-of-coolant accident should be used (See Figure 2):
(1) During the blowdown phase, assume a linear ircrease in the conder. sing heat
- F. at t = 0, to a peak i
transfer coefficient from hinitial=8 Btu /hr-ft
'value four times greater than the maxinum calculated condensirg heat trans-fer coefficient at the end of blowdown, using the Tagani correlation (hef.2). 0.62 .
h max a 72.5 .yf
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where hg = ma irtum heat transfer coefficient. Stu/hr-ft 'F Q = primary coolant energy. Stu V = net free contaireen't volune, f t t = time interval to end of blowdown, sec.
p (2) During the long-term post-blowdown phase of the accident, characterized by j d low turbulence in the containment atrtsphere, assume condensing heat transfer I coefficients 1.2 times g-eater than those predicted by the Uchida data (Ref. 3) and given in Tabic 3. '
(3) During the transition phase of the accident. between the end of blow % n aad the long tem oost blowdown phase, a reasonably conservative exponer tial transition in the condensing heat transfer coefficient should be assu cd ,
. (see Figure 2).
The calculated condensing heat transfer coefficients based on the se:ve r etud should be applied to all exposed passive heat sir.ks, both tretal ard ccrcrete, and for both. painted and unpainted surfaces.
Heat transfer between adjoining materials in passive heat sinks s*.culd 'e c based on the assumption of no resistance to heat flow at the raterial interfaces. An example of this is the containment liner to concrete' interface.
C. R E F E R E NC,E_S,
- 1. 10 CFR 150.46. " Acceptance Criteria for Emergency Core Cooling Systems for Light L'ater Nuclear Power Reactors." and 10 CFR Part 50, Appendix K. "ECCS Evaluation Models."
- 2. T. Tagami, " Interim Report on Safety Assessments and Facilities Establisi-ent Project in Japan for Period Ending June 1965 (No. 1)." prepared for the National React ~ Testing Station. February 28.1966(unpublishedwork).
6.2.1.55 I
- 3. H.' Uchida. A. 0yama, and Y. Toga. ." Evaluation of Post-Incident Cooling Systems of Light.
Water Power Reactors." Proc. Tntrd International Conference on the Peaceful Uses of AtomicEnergy. Volume 13. Session 3.9. Unite'dNations. Geneva (1964).
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-. TABLE 1 10ENTIFICAT10NOFCONTAINMENTHEATStts i
- 1. Containment. Building (e.g..linerplateandexternalconcretewalls, floor,andsump,and lineranchors).
l l 2. Containment In:ernal Structures (e.g., internal seoaration walls and floors, refueling pool and fuel transfer pit walls, and shielding walls).
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- 3. Supports (e.g., reactor vessel, steam generator, pumps, tanks.. major compor.cnts, pipe l supports, and storage racks).
4 Uninsulated Systems and Components (e.g., cold water systems, heating, ventilation, and air conditioning systems, puvs. motors, fan coolers, recombiners, and tants).
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- 5. Miscellaneous Equipment (e.g.. ladders, gratir.gs,.clectrical cable trays, and cranes).
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TABLE 2 HEAT $1NK THERM 0 PHYSICAL PROP [RT!ES Specific Thermal Densi3y Heat Conductivity
- Material _lb/ft Stu/lb 'F. Btu /hr-ft 'F Concrete 145 0.156 0.92 L
. Steel ,
490 0.12 27.0 e
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M DA HfAT TRANSFER C0 Err 1CIENTS Mass Heat Transfer Mass Heat Transfer Ratic . Coefficient Ratio (1b air /lb steam) (Stu/hr ft 'F) (lbair/lbstead (8tu/hrftgF1 Coefficient
. 50 2 3 29 20 8 2.3 37 18 9 1.8 46 e :
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