ML18038A019
| ML18038A019 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/30/1985 |
| From: | Mangan C NIAGARA MOHAWK POWER CORP. |
| To: | Butler W Office of Nuclear Reactor Regulation |
| References | |
| (NMP2L-0420), (NMP2L-420), NUDOCS 8506030534 | |
| Download: ML18038A019 (230) | |
Text
REGUL~MRY INFORMATION OISTRIBUTI v -'SYSTEM (RIDS)
ACCESSION NBR:8506030534 DOC DATE'5/05/30 NOTARIZED; NO , DOCKET FACIL:50-410 Nine Mile Point Nuclear Station< Unit 2g Niagara Moha 05000410 AUTH, NAME AUTHOR AFFILIATION MANGAN C,V. Niagara Mohawk, Power Corp.
RECIP ~ NAME RECIPIENT AFFILIATION BUTLERgH ~ Licensing Branch 2 s~
SUBJECT:
Forwards nfo on lant-specific ECCS analysis, Info addresses concerns on tR Conf irma or y Issues 1~F1~1,Changes to FSAR pages wi l l be included in Amend 20 ~ Info partial ly deleted.
DISTRIBUTION CODE 5 B001D COPIES RECEIVED LTR ENCL SIZE ~
TITLE; I icensing Submittal t PSAR/FSAR Amdts 8, Related Correspondence dspac NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID LTTR ENCL ID CODE/NAME LTTR ENCL CODE/NAME'RR/OL/ADL 1 0 NRR LB2 BC 1 0 NRR LB2 LA 1 0 HAUGHEYpM 01 1 1 INTERNAL: ACRS R1 6 6 ADM/LFMB 1 0 ELO/HDS3 1 0 IE F ILE 1 1 IE/DEPER/EPB 36 1 1 IE/DQAVT/QAB21 1 1 NRR ROEpM ~ L 1 1 NRR/DE/AEAB 1 0 NRR/DE/CEB 11 1 1 NRR/OE/EHEB 1 1 NRR/DE/EQB 13 2 2 NRR/DE/GB 28 2 2 NRR/OE/MEB 18 1 1 NRR/OE/MTEB 17 1 1 NRR/DE/SAB 24 1 1 NRR/OE/SGEB 25 1 1 NRR/DHFS/HFEB40 1* 1 NRR/DHFS/LQB 32 1 1 NRR/OHFS/PSRB 1 1 NRR/OL/SSPB 0 NRR/OSI/AEB 26 1 1 NRR/DSI/AS 8 1 1 NRR/OSI/CPS 10 1 1 NRR/OS I/CS8 09 1 1 NRR/OSI/ICSB 16 1 1 NRR/DSI/METB 12 1 1 NRR/OSI/PSB 19 1 1 NRR/DSI/RAB 22 1 1
'RR/DSI/RSB 23 1" 1 04 1 1 RG~1 AMI/MIB 1 0 EXTERNAL: 24X 1 1 BNL(AMDTS ONLY) 1 1 OMU/VSS (AMDTS) 1 1 LPOR 03 1 1 isRC PDR 02- 1 1 NSIC 05 1 1 P."lL GRUEL p R 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 52 ENCL 44
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NIAGARA MOHAWK POWER CORPORATION/300 ERIE BOULEVARD WESTSYRACUSE, N.Y. 13202/TELEPHONE (315) 474-1511 May 30, 1985 (NMP2L 0420)
Mr. Walter Butler, Chief Licensing Branch No. 2 U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Butler:
Re: Nine Mile Point Unit 2
-Docket No. 50-410 Attached is information on the Nine Mile Point Unit 2 plant specific ECCS analysis. This submittal should address the staff's concerns on confirmatory issues 10, 15 and 16.
Changes to Final Safety Analysis Report pages will be included in Amendment 20.
Very truly yours, C. V. Manga Vice President Nuclear, Engineering 8 Licensing GW/r la Attachment Project File (2)
R. Gramm, Resident Inspector 8506030534 850530 PDR ADOCK 05000410 E PDR
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NRC DISTRIBUTION FOR PART 50 DOCKET v ATERIAL (TEMPO RA R Y FO RM)
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Niagara Mohawk Power Co. DATE OF DOC DATE R EC'D LTR TWX RPT OTHER Syracuse, N. Y.
TO: ORIG CC OTHER SENT NRC PDR John F,Stolz 1 Signed .
UNCLASS PROP INFO INPUT No cI's REc'D DOCKET NO:
L'LASS 50-410 DESCR IPTION: ENCLOSURES:
Re our ltr of 6-25-75, advising will submit ECCS of the re-analysis'n early FSAR Report. ~....
1980, wi,th the filia
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PLANT NAME: Nine Mile Pt, 0 2 FOR ACTION/INFORMATION VCR 8-2 75 BUTLER (L) SCHWENCER (L) Zl EMANN (L) REGAN (E)
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'I July 29, 1975 Mr. John F. Stolz, Chief Light Water Reactors Branch 2-1 Division of Reactor Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Re: Nine Mile Point Unit 2 Docket No. 50-410
Dear Mr. Stolz:
In regards to your June 25, 1975 letter, we wish to inform you that we presently plan to submit an ECCS re-analysis for Nine Mile Point Unit 2 in early 1980, with the filing of the Final Safety Analysis Report.
Very truly yours, NIAGARA MOHAW POWER C P ATION Gerald K. Rhode Vice President-'Engineering 8167
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JUN 2 5 1975 Docket No 50-410 Niagara Mohawk Power Corporation ATTN: Mr . Gerald K. Rhode Vice President Engineering 300 Erie Boulevard West Syracuse, New York 13202 Gentlemen:
The enclosure to this letter identifies information that we require in order for us to complete our review of the ECCS re-analysis (pursuant to 10 CFR-Part 50, 50.46, and Appendix K to Part 50) on your application. You may have provided much of the enclosed information in your previous submittals. However, to expedite our review and to mitigate the need for requests for additional information, you should check your application to make sure it is complete in this area and amend your application to eliminate any deficiencies that may exist.
If you have any questions regarding the enclosure or this matter, please contact us.
Sincerely, n F. Stolz, Chief ght Water Reactor ranch 2-1 ivision of Reactor Licensing
Enclosure:
As stated above.
cc: Arvin E. Upton, Esquire Mr. Richard Goldsmith LeBoeuf, Lamb. Leiby 8 MacRae Syracuse University 1757 N Street, N. W. College of.Law Washington, D. C. 20036 E. I. White Hall Campus Syracuse, New York 13210 ttiss Juanita Kersey, Librarian Oseego City Library bcc: J. R. Buchanan, ORNL 120 E. Second Street T,. B. Abernathy, DTIE Oswego, New York 13126
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JUN 0;5 )975 Letter to Applicant for Additional Information..
'ISTRIBUTION:
NRC PDR Local PDR Docket File LWR 2-1 File Y. moore R. DeYoung ELD F. Williams R. Klecker IE (3) 6'+ M. P< (YawlZJ H..Smith (2)
J. Stolz ACRS (16) bcc: J. R. Buchanan, ORNL Thomas B. Aber'nathy, DTIE
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Attachment 1 REQUIRED INFO% fATION Break S ectrum and Partial Loop Operation The information provided for each plant shall comply with the provisions of the attached memorandum entitled, "Hinimum Requirements for ECCS Break Spectrum Submittals."
The ECCS system in each plant should be evaluated by the applicant (or licensee) to show that significant changes in chemical concentrations will not occur during the long term after a loss-of-coolant accident (LOCA) and these potential changes have been specifically addressed by appropriate operating procedures. 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.
Sin le Failure Anal sis A single failure evaluation of the ECCS should be provided by the applicant (or licensee) for his specific plant design, as required by Appendix K to 10 CFR 50, Section I.D.l. In performing this evaluation, the effects of a single failure or operator error that causes any manually controlled, electrically-operated valve to move to a position that could adversely affect the ECCS must be considered. Therefore, if this consid-eration has not been specifically reported in the past, the applicants upcoming submittal must address this consideration. Include a list of all of the FCCS 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 failure evalu".tion should include the potential for passive failures of fluid systems during long term cooling following a LOCA as well as single failures of active components. For PHR plants, the single failure analysis is to consider the potential boron concentra-problem as an integral part of long term cooling.
Submer ed 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 cool&>g or may be required for containment isolation.
The applicant (or lice'nsee) is to provide the following information, for each plant:
(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 licensee) is to:
(a) Identify the valves that will be submerged.
(b) Evaluate the potential consequences of flooding of the valves for both the short term and long term ECCS functions and containment isolation. The long term should consider the 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 only).
(d) Propose design changes to solve the potential flooding problem.
S. Containment Pteeente (PMR'e~Onl )
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 ECCG 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'nch per second; conditions for which the December 25, 1974 and March 15, 1975 versions would be equivalent.
(2) The Harch 15, 1975 version of the Westinghouse evaluation model is an acceptable model to be used for all previously analyzed plants ror which the peak clad temperature turnaround was identi-fi'd to occur after the reflood rate decreased below 1.1 inches per second, and for which steam cooling conditions (reflood rate less than 1 inch p'er second) exist prior to t.he time of peak clad temperature turnaround. The Harch 15, 1975 version will be used for all future plant analyses.
I ;l MINIMUM RE UIREHENTS FOR FCCS BRFJK 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 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 rovide assurance that the entire s ectrum of ostulated 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 break 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 later).'ections IIC and IIIC provide an outline of the minimum requirements for an acceptable complete break spectrum. Such a complete break spectrum could be appropriately referenced by some plants. Sections IXID and IIXE 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 'o operate the reloaded core in compliance with previously approved Technical Specifications, no'dditional 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 been modified as a consequence of changes to the reload'core design. When the reload is supplied by a source other than the NSSS supplier, the break spectrum analysis specified by Sections IIC or IIIC shall be submitted as a minimum (as appropriate to.the plant type, BWR or PWR).* Additional sensitivity 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. 0 eratin Reactor Regnal ses (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
t 1 conformance with 10CFR50, Appendix K, the following. minimum number of break sizes should he r'analyZed.. Each sensitivity study performed during the development of the ECCS evaluation model shall be individually verified as 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 des'ign.'.
If the lar est break size results in the hi hest PCT:
a.'eanalyze the limiting break.
- b. Reanalyze two smaller breaks. in the large break region.
- 2. If the lar est break size does not result in the hi hest 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.
If calculational changes have been made to the SBM to make it wholly with 10CFR50, App'endix K, the analysis of the worst smal1 break in'onformance (SBM) as previously determined from paragraph-.C below should be repeated.
B. New CP and OL Case Pork A complete break spectrum should be provided in accordance with paragraph C below, except for the following:
- l. If a new plant is of the same general design as the plant used as a basis for a referenced complete spectrum analysis, but operating parameters have changed which would increase PCT or metal-water reaction, or approved calculational changes resulting in more than 20 F change in PCT have been made to the ECCS model used for the referenced complete spectrum, the analyses of paragraph' 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'he 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 F 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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a C. minimum Re uirements for a Coin lete Break S ectrum Since it is expected that applicants will prefer to reference an applicable complete break spectrum previously conducted on another plant, this paragraph defines the ~minim m 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 should show that the worst 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 size and location (i.e., idle loop versus operating loop) . In additions sufficient 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.
It must be demonstrated that the containment design used for the break spectrum analysis is appropriate for. the specific plant analyzed. Xt 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 Hoody multipliers between Oe6 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 indicate 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. LBH Hot Leg Piping Analyze the largest rupture in the hot leg piping.
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- 4. SBM Splits Analyze fiv'e diffexent small break sizes. One of these breaks must include the transition split break. The CFT line break must be analyzed for B6W 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 for a ~com lete 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 paxagraph C below. Also, a proposal for partial loop operation shall be supported by identifying and apalyzing the worst break size and location (i.e., idle loop versus operating loop). In addition, sufficient 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 Regnal sis (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 and core design.
if it is the same confiauration Each sensitivity study published during the development of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated..
- l. If the lar est break results in the hi hest PCT:
- a. Reanalyze the limiting break with the appropriate referenced single 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 pxedicts the highest PCT.
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- 2. If the lar est break does not result in the hi hest 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 conformance with 10CFRSO, Appendix K, reanalyze the small break (SBM) in accordance with Section IIXC.
B. New CP and OL Case Work A complete break spectrum should be provided in accordance with Section III, paragraph C below, except for the following'.
- l. If a new plant is of the same general design as the plant used as a basis for the lead plant analysis, but opexating paxameters 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 20 F 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 20 F 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 Re uirements for a Com lete Break S ectrum This paragraph defines the minimum number of breaks requixed for an acceptable complete spectrum analysis. This complete spectrum analysis is required for each 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.
- l. Four recirculation line breaks at the worst location, (pump suction or discharge), using the LBM, covering the range from the transition break (TB) to the DBA, including CD 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 sic}e of worst location)
D. BWR4 with "Modified" ECCS Same as Section IIIC.
E. BVR5 Same as Section IIIC.
F. BWR6 Same as Section IIIC.
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 graph paper of a quality to facilitate calculations.
Peak clad temperature (ruptured and unruptured node)
Reactor vessel pressure Uessel 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 provided on engineering graph paper of a quality to facilitate calculations. The worst single failure and worst-case reactor coolant pump status will have been established utilizing appropriate sensitivit'y studies.
Flooding rate (PWR on3v>
Core flow {inlet and outi~
Core inlet enthalpy (BWR only)
Heat transfer coefficients MAPLHGR versus Exposure (BWR only)
Reactor coolant temperature {PWR only)
Mass released to containment (PWR only)
Energy released to containment (PWR only)
PCT versus Exposure (BWR only)
Containment condensing heat transfer coefficient (PWR only)
Hot spot flow (PWR only)
Quality (hottest assembly) (PWR only)
Hot pin internal pressure Hot spot pellet average temperature Fluid temperature (hottest assembly) (PWR only)
C. A tabulation of peak clad temperature and metal-water r e reac'tion (local and core-vide) shall be provided across the break spectrum.
D S ft aey Analysis a Reports (SARs) filed with the NRC shall identify on h 1 t th r n date version number, and version date oof the computer 7 p model utilized for the LOCA analysis. Should differences ex s in exist version number or version date from the most cu rrent code listings made avaz'1 a bl e to thee NRC staff, then each modification shall be identified with an assessment of impact upon PCT and metal-water er rreaction (local and I
core-vide).
E. A tabulation of times at which significant events occur shall be provided on each plant and for each break analyzed. The following events shall be included as a minimum:
End-of-bypass (PWR only)
Beginning of core recovery (PWR only)
Time of rupture Jet pumps uncovered (BWR only) liCPR (BWR only)
Time of rated spray (BWR only)
Can quench (BWR only)
-<<End-of-blowdown Plane of interest uncovery (BWR only)
Possible grouping of plants for the purpose of performing generic as well as <ndividual plant break spectrum analyses.
CURRENT OOCKETEO APPLICATIONS
0 BABCOCK ANO WILCOX CATEGORY I: 177 FA w Lowered Loo s Arran ement Re-anal sis Safet Order Plants :
Oconee 1, 2, 3 -- IIA 2568 These lants must;;resubmit at Three Mile Island 1 -- IIA east 3 breaks'hey will do 2535 . so by reference to a complete A~kansas Power 1 -- IIA break spectrum reanalysis sub-2563 mitted generically by BQt.)
. Rancho Seco -- IIA 2772 New OLs:
Three Mile Island 2 --IIB(2) Since these plants are the same 2772 Crystal River 3 --IIB(2) design as the above plant, they 2452 may reference the same reanalysis Midland 1, 2 --IIB(2) of the complete spectrum above.
New CPs:
None CATEGORY II: 177 FA w/Raised Loo Arran ement New OL's:
Davis Besse 1 --IIB Complete spectrum required.
New CPs Davis Besse 2, 3 --IIB Complete spectrum required.
CATEGORY I I I: 205-FA Plants New OLs:
None
. New CPs:
Bellefonte 1, 2 IIB I
lete ectrum re uired.
Com s Greenwood 2, 3 -- IIB Plans are for all to reference a complete spectrum submitted WPPSS 1, 4 IIB probably on MPPSS.)
Pebble Springs 1, 2 CATEGORY IV: 145-FA Plants New OLs:
None New CPs:
North Anna 3, 4 Com lete s ectrum re uired.
One will probab y reference Surry 3, 4 the other.)
0 GENERAL ELECTRIC BHR-2 Oyster Creek LP* Com lete s ectrum re uired. (IIIA)**
Nine Nile Point Reference only required. (IIIA)
BMR-3 (}uad Cities,2 LP* Com lete s ectrum re uired. (IIIA)**
2511 Millstone WW IIIA - 3 breaks required 2011 Nonticello IIIA - 3 breaks required 1670 Dresden 2, 3 IIIA 2527 May reference LP guad Cities 1 I I IA 2511 Pilgrim IIIA - 3 breaks required 1998 BMR-4 Without fix Hatch 1 -- LP* Com lete s ectrum re uired. ( IIIA)*+
2436 Peach Bottom 2, 3 IIIA Com lete s ectrum re uired. One 3293 may reference the other.
Browns Ferry 1, 2, 3 -- IIIA 3293 Cooper IIIA 2381 Fitzpatrick IIIA 3 breaks required.
2436 Hatch 1 may serve Duane Arnold IIIA - 3 breaks required 1658 as a reference Hatch 2 -- IIIA for the others.
2436 Brunswick 1
-- IIIA 2436 Shoreham Fermi ->> IIIB Newbold -- I II B
- Lead Plant
- Original break spectrum'not wholly in conformance with 10CFR50, Appendix K.
BWR-4 Miti+jx Brunswick 2436 2 (Lead Plant) -~ IA - Com
~re lete s ectrum uired.*~
Vermont Yankee IIIA - 3 breaks required (Lead Plant can be 1'593 Ferry* 1, 2, referenced, if Browns & 3 appropriate)
Peach Bottom* 2, 3 See preceding page Fitzpatrick*
BWR-5 Lead Plant IIIE - Com lete s ectrum re uired.
Nine Nile Point 2 -- IIIB Complete spectrum required.
LaSalle 1, 2 -- IIIB (Lead Plant can be referenced by other BWR-5 p'sants, if Ba illy -- IIIB appropriate.)
Zimmer- IIIB Susquehanna 1, 2 IIIB BWR-6 Lead Plant -- IIIF - Com lete s ectrum re uired.
Grand Gulf IIIB Black Fox II I8 Barton 1, 2, 3, 4 -- III.B Complete spectrum required. (Lead Perry 1, 2 Plant can be referenced by other BWR-6 plants, if appropriate.)
Clinton 1, 2 II IB Douglas Point II I8 Hanford 2 -- I I I B Skagit 1, 2 IIIB Hartsville IIIB Somerset I I IB River Bend Station -- IIIB
'liens Creek -- 'IIIB"
~ Nay or may not have the LPCI fix
- Original break spectrum not wholly in conformance with 10CFR50, Appendix K.
0' PLANT SPECIF oyster Creek -- IIIA Complete spectrum required.
Nine Nile Point IIIA Limaerick 1, 2 I I I B Hope Creek
-- IIIB Humboldt Bay IIIA Dresden 1 IIIA Big Rock IIIA
COMBUSTION ENGINEERING The following list is grouped according to similarities in design.
,Some of the older, operating plants are fairly unique, as indicated, and don't fall conveniently into any other groups. The list is in approx, chronological order.
1; Palisades (Unique) IIA
- 2. Ft. Calhoun (Unique) -- I/A 3 breaks required,
- 3. Maine Yankee (Unique) IIA
- 4. 2560 NMt Series
- a. Calvert Cliffs Units 1 8 2 -- IIA - 3 breaks required
- b. Millstone Unit 2 -- IIB Complete spectrum requi red.
(One may reference the other.)
- c. St. Lucie 1 IIB
- d. St. Lucie 2 IIB Complete spectrum required
- 5. 3400 NMt Series ( 3410 MMt 217 Fuel Assemblies)
- a. Pilgrim 2 (3470 Mwt) IiB
- b. Forked River -- IIB Complete spectrum required.
1
{One may reference the other.)
- c. San Onofre 2 & 3 -- IIB
- d. Waterford 3 -- IIB
- 6. Arkansas Class ( 2900 MMt 177 Fuel Assemblies)
- a. Russelville 1 IIB Complete spectrum required.
- b. Blue Mills 1 IIB (One may reference the other.)
Haine Yankee is unique in that it has 3 steam generators, 3 hot legs and 3 cold legs. All other CE plants have 2 steam generators, 2 hot legs and 4 cold legs.
All plants shown above listed before St. Lucie 2 are'f the 14x14 fuel design..
All plants after, and including, St. Lucie 2 are 16xl6.
'. S stem 80 Class- CESSAR -- ggg. Complete spectrum required These plants have not all been named yet. The utility and approx.
~
number of plants expected are as follows:
- a. Duke (6)
- b. WUPPS (1) Hay reference complete
- c. Arizona Power and Light (2),
spectrum, if applicable.
- d. TVA (2)
0 Mestin house 0 eratin Reactors Safet Order Plants
- 2-loo 3-loo 4-loo Ginna Surry 1/2 Yankee Rowe Kewaunee Turkey Pt. 3/4 IP2 Pt. Beach 1/2 H. 8'. Robinson 2 D. C. Cook'1 Prairie Island 1/2 Zion 1/2
~ot. i 2-1oo 3-1 oo 4-1oo Beaver Valley 1 - Trojan*
Far ley 1/2 - Salem 1/2*
North Arna I/2 - Diablo Canyon 1/2 IP-3 "O'. C. Cook 2 NcGuire 1/2 Sequoyah I/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).
1 Construction Permit *"
2-1 oo ~
3-1 oo 4-1 oo North Coast . Sharon Harris 1/4 Byr on/Braidwood 1/2 Koshkonong 1/2 Cata>ha 1/2 Summer 1 Float>ng Nuclear 1/8 Beaver Valley 2 Jamesport 1/2 Wisconsin Utilities Seabrook 1/2 SNUPPS 1-5 South Texas 1/2 Comanche Peak 1/2 Matts Bar 1/2 Millstone 3 Vogtle 1/2
- Complete spectrum required. One plant may reference another if applicable (see paragraph,I?B).
1 I BRAHCH TECHNICAL POSITIOH EICSB 18 APPLICATIOH OF THE Sll!GLE FAILURE CRITERJOH TO YAhUALLY-COHTROLLED ELECTRICALLY-OPERATED VALVES A. BACKGROURD I'here a single failure fn an electrical system can result fn loss of capability to perfor '
safety function, the effect on plant safety. must be evaluated. This fs 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 component performing an undesirable nechanical ration.
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 right cause an un-desirable co.:.ponent action. These provisions are based on the assurption that the co..ponert fs then equivalent to a sinilar component that is not designed for electrical operation, e.g., a valve that can be opened or closed only by direct ranual operation of the valve.
They are also based on the assu';ption that no single failure can both restore Pc:: r to the electrical systen and cause rechanical notion of'he corponents served by th elec:rical systen. The validity of these assumptions should be verified when applying this Position.
I I
B. BRANCH TEC!!HICAL POSIT 10.'<
- 1. Failures in both the "fail to function" sense and the "undesirable functio~" sense o components in electrical systems of valves and other fluid syste... coma cents s".o'id be considered in, designing against a single failure, even though the valve or o.'.er f'Iuid system component ray not be called upon to function in a giver. saft .y operational sequence.
- 2. Hhere it, is determined that failure of an electrical system co-.sonant can cause undesired re'chanical rotion of a valve or other fluid system co.-.ponent and'nis motion results fn loss of the systen safety function, ft is acceptable, in lieu o f design changes that also ray be acceptable, to disconnect power to the electric systems of the valve or other fluid system component. The plant technical specifications should include a list of all electrically-operated valves, and the required positions of these valves, to which the requirement for removal of electric power fs applied in order to satisfy the single failure criterion,
- 3. Electrically-operated valves that are classified as "active" valves, f.e., are required to open or close in various safety syster'. operational sequences, but are nanually-controlled, should be operated from the main control room. Such valves may not be included among those valves fron which >over is removed in order to,meet the single failure criterion unless: (i) electrical power can be restored to the valves from the main control room,(b) valve operation is not necessary for at least ten minutes following occurrence of the event requiring such operation, and (c) it is demonstrated 7A-27
that there is reasonable assurance that all necessary operator actions will be per-formed within the time shown to be adequate by the analysis. The plant technical specifications should include a list of the required positions of manually-controlled, electrically-operated valves and should identify those valves to which the require-ment for removal of electric power is applied in order to satisfy the single failure criterion.'.
1'hen 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 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 by an electrical device (e.g., a motor-operated valve or a solenoid-operated valve) and those valves operated indirectly by an electrical device (e.g., an air-operated valve whose air supply is controlled by an electrical solenoid valve).
C. REFERENCES
- 1. Hemorandum to R. C. OeYoung and V. A. Yoore from V. Stello, October 1, 1973.
BRANCH TECHNICAL POSITIOlt CSB 6-1 HINIHUH COttTAltlHENT PRESSURE MODEL FOR PMR ECCS PERFORHAtiCE EYALUATION A. BACKGROUND Paragraph 1.0.2 of Appendix K to 10 CFR Part 50 (Ref. 1) requires that the containment pressure used to evaluate the performance capability of a pressurized water reactor (PMR) 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
~ all installed prcssure-reducing systems and processes. Therefore, the following branch technical position has been developed to provide guidance in the perfonrancc of ninirum containment pressure analysis. The approach described below applies only to the ECCS-related containmcnt pressure evaluation and not to the containment functional capability evaluation for postulated design basis accidents.
B. BRANCH TECMttICAL POSI TIOtt
- l. ~In ut Information for Model
- a. Initial Contain-..ent Internal Conditions The minimum containncnt gas tenpcraturc, rinimur containment pressure, and maxinum humidity that may be encountered under limiting norr.:al operating conditions should bc used.
- b. Initial Outside Contain",.ent Ambient Conditions A reasonably low arbient temperature external to the contain-.cnt shculd bc used.
- c. Containment Volur.c The maximum nct free containment volu.;.c should be used. This raxir un free should be determined fro".. the gross containmcnt volume ninus the volu-:,es
'olume of interrial structures such as walls and floors, structural steel, rajor e"uip..ntt, and piping, 'he individual volurac calculations should reflect the uncertainty in the component volumes.
- 2. Active Meat Sinks
- a. ~S ra and Fan'oolin S stems The operation of all enoineercd safety feature containment heat removal systems operating at maximum heat removal capacity; i.c., with'll contain.-.'.cnt spr'ay trains operating at maximum flow conditions and all emergency fan cooler units operating, should he assumed. In addition, the mininun 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.
6.2.1.5-3
1 k Deviation rom the foregoing will be accepted if i n be shown 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 ECCS models
- b. Cnntainnent Stean> Ni~xin "itn S tiled CCCS Vatee The spillage of subcooled ECCS water into the containment provides an additional 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 containment pressure calculations.
- c. Containment Steam Hixin ".ith Vater from Ice S!elt The water resulting from ice melting'n an ice condenser containment provides an additional heat sink as the subcooled water mixes with the steam while draining from the ice condenser into the lower contain,ent volume. The effect of the steam-water mixing should be considered in the containment pressure calculations.
- 3. hssive Heat Sinks
- a. 'Identification The passive heat sinks that should be included in the containment evaluation model should be established by identifying those structures and co..ponents within the containment that could influence the pressure response. The kinds of struc-tures and components that should be included are listed in Table 1.
Data on passive heat sinks have been compiled from previo's revie's and have been used as a basis for the simplified model outlined below. This model is acceptable for minimum containment pressure analyses for construction pes-~it applications, and until such time (i.e., at the operating license revie") tnat a complete identification of available heat, siraks can be made. Tnis si;.plified approach has also been folio:s d for o-crating plants by licensees co-plying witn Section 50.46 {a)(2) of 10 CFR Part 50. For such cases, and for construction permit reviews, where a detailed listing of teat sinks within the contain-ent e
often cannot be provided, the following procedure may be used to model tne pass;ve heat sinks within the contain...cnt:
(1) Use the surface area and thickness of the primary containment steel shell cr steel liner and associated anchors and concrete, as appropriate.
(2) Estimate the exposed surface area of other steel heat sinks in accordance with Figure 1 and assume an average thickness of 3/8 inch.
{3) 5'lodel the internal concrete structures as a slab with a thickness of 1 foot and exposed surface of 160,000 ft .
The heat sink thermophysical properties that would be acceptable are shown in Table 2.
6.2.1.5-4
I E ~ .\
At the crating license stage, applicants shou ovide,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 exposed 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-increase in the condensing heat transfer coefficient from h.initial=8 Btu/hr-ft2 - 0 F, at t 0 0, to a peak value four times greater than the maximum calculated condensirg heat trans-nax where h
't fer coefficient at the end of blowdown, using the Tagami correlation (Ref. 2),
max 0.62
= maxir,.um
~
heat transfer coefficient, Btu/hr-ft primary coolant energy, Btu
-'F net free containment volume, ft3 n
V ~
tp time interval to end of blowdown, sec.
(2) During the long-tern post-blowdown phase of the accident, characterized by low turbulence in the containment atrosphere, assir.".e condensing heat transfer coefficients 1.2 times greater than those predicted by the Uchida data (Ref. 3) and given in Table 3 ~
Q (3) During the transition phase of the accident, between the end of lowdo0n and the long-tern post-blowdown phase, a reasonably conservative exponser.tial transition in the condensing heat transfer coefficient should be ass.".ed
( See Figure 2).
The calculated condensing heat transfer coefficients based on the a".ove re'.nod be applied to al) exposed passive heat sirksboth metal and ccrcre:e, ard 'hould for both .painted and unpainted surfaces.
Heat transfer between adjoining raterials in" passive heat sinks should b based on the assumption of no resistance to, heat flow at the material interfaces. An example of this is the containment liner to concrete'nterface.
C. REFERERCES
- l. lg CFR lSD.46. "Acceptance Crlterla for Emergency Core Cooltng Systems for Light Mater Nuclear Power Reactors," and 10 CFR Part 50, Appendix K, "ECCS Evaluation Hodels."
- 2. T. Tagami, "interim Report on Safety Assessments and Facilities Establish.-.ent Project in Japan for Period Ending Junc 1965 (No. 1)," prepared for the National React Testing Station, February 28, 1966 (unpublished work).
Q 6.2.1.5-5
O.
~ ll h
H. Uchida, A. Oy, and Y, Toga, "Evaluation of Post-Inci Cooling Systems of Light-ktater Power Reactors," Proc. Third International Conference on the Peaceful Uses of Atomic Energy, Volume 13, Session 3.9, United Nations, Geneva (1964).
6.2.1.5-6
I TABLE 1 Q IOENTIF ICATION OF COl<TAI!!HENT HEAT 5 I f<KS 1, Containment Building (e.g., liner plate and external concrete walls, floor, and sump, and liner anchors).
- 2. Containment Internal Structures (e.g., internal separation walls and floors, refueling pool and fuel transfer pit walls, and shielding walls).
3, Supports (e.g., reactor vessel, steam generator, pumps, tanks, ma>or components, pipe supports, and storage racks).
- 4. Uninsulated Systems and Components (e.g., cold water systems, heating, ventilation, and air conditioning systems, pumps, motors, fan coolers, recombiners, and tanks).
- 5. Niscellaneous Equipment (e.g., ladders, gratings, electrical cable trays, and cranes).
6.2.1.5-7
TABLE 2 HEAT SINK THERHOPHYSICAL PROPERTIES Specif(c Thernal Densi)y Heat Conductivity Fteteriel ~lb ft utullh-'F ~8tu hr-ft-'F Concrete 145 0.156 Oe92 Steel 490 0. 12 27.0 6.2.1.5-8
UCHIDA HEAT TRANSFER COEFFICIENTS f
Hass Ratio lb air lb stean Heat Transfer Coeffici~nt Hass Ratio Hea t Trans er
"'"ficipt Coef
" "l.
50 2 3 29 20 8 2.3 37 18 9 1.8 46 14 10 1.3 63 10 14 0.8 98 7 17 0.5 140 5 21 0.1 280 4 24 6.2.1.5-9
IV Figure 1 Area of Steel lkeat Sinks Inside Containment 2'
Containment Free Volume, x 10't 6 3 Revised 12/74 D
J Figure 2 Condensing Heat Transfer Coefficients for Static Heat. Sinks C
O CJ 0 = 4. x?L
'U iO max Tagami linear iJT I
W 4l h=h stag +(h max -h stag ) e'p
.025(t-t )
Q b0 C
h stag
~ 1.2xh Uchida
~
C Ol o h =8.
'V 0 I t Time I P I
blowdown I reflood I
I I
~ ~
(4 ~ ~, W r,oc'fret le. 50"610 APR 18 1975 1'iapara I'ohavk '3over Corporation
&TV.: l!r. CeraM K- Rhode Vice President Pngineering 3OO Erie Boulevard rr'est Syracuse, Fess Pork 13202-Centlecen:
pursuant to 5 50.N('f) of 10 CK part: 50, the nuclear plepulatory Co~visaion (bRC) staff requires that certain inforuation related to the desipp of the containr.".ant for your facility be subr,itted promptly to 1'RC for its'reviev.
This requirer..ent pesults from'ecent developments associated 9ith the $
larpe-scale B>lm Farl: XD testing being conducted by the General Electric Co 3pany. These tests indicate that suppression pool hydrodynarQic loads
, during a loss-ot-coolant'ccident (XGCA) should be considered in the detailed desiFn, of co. ponents and structures of t>M Nark XX1 contain.ent.
Zn "ddition$ there appears t'o be a potential for the occurrence of similar dynamic loads on plants vith a 1;artN; II type of gontaintlent.
Therefore, ve require that you provide the inkorration specified in Fnclo ure 1 concartsinp; the potential aaonitudp of these -t>ydrodynmic loads, and the effects of these loads, in co'bination !;ith other desi(n loads, on the design of your containr..ent structures.
Enclosure 1 specifies the inf'orr;.ation required to ce>piete our review L
of the effect of pool dynamic loads on the design of your contains!ent structures. $I@closure 2 contains baclzpround inforc~ation on the status of efforts directed at de'temininp pool dynavic 1'oads; for general inforvation$ ae have also provided in Enclosure 3 a description of the various phenomena during a. postulated LOCA which result in pvssible hydrodynamic loads. Please note that certain key phrases in .Lnclosure 3 have been underlined. These phrases (1} identify those specific, hydro-dynsr ic loads vhich2 as a ~ini~m, should be considered in your revise of containn!ont design, and (2) establish the standard nq"..enclature by which phenomena should be discussed or referenced in ypur docur!enkation.
Your propra~ and schedule for prm~pt resolution of the proble;:.s asso-ciated mith LOCA-caused suppression pool hydrodynrvic loads should be filed vith the Comission not later than 30 days after r'eceipt of this lett.er. 7ne schedulinF of vort on this r-;.atter should par llel relcte$ d efforts on other containment. design/operational control aspectsr,i.e. $
relief valv'e vent clearing and stoaFA quenching; vibration phenomena.
OFFICE>
SURNAMEW DATE+
Form ABC-318 (Rev. 9.53'ECM 0240 O' $ $ COVCRNHRH$ 'RlHTINC ORRlCRl l$ 7$ ~$ $ 2'l$
pi+5, P
I I'
,Niagara Mohawk Power Corporation - 2 Please contact us if you desire additional di'scussion or clarification of the information requested.
Sincerely, Original Signed by Q. M~~49 Q~
John F. Stolz, Chief Light Mater Reactors Branch"2-1 Division of Reactor Licensing
Enclosures:
- 1. Required Information
- 2. Background
- 3. Description of Potential Pool Dynamic Pheno"ena This request for Generic Inforr,.ation twas approved by GAO under a No. B-180225 (R0072);'his clearance expires July 31, 1977.
blanket'learance cc~ LeBoeuf, Lamb, Leiby h l.acRae Carmine J. Clemente, Esq.
ATTN: Hr. Arvin E. Upton, l'sq. l,ew Yorl: State Atomic 1737 N Street, N.$ ,. l.nergy Council Uashinpton 9 D. C. 20036 L'epartnent of COImnerce 99 l"ashinpton Avenue J. Bruce liacDonald Esq., Deputy Albany, l; w York 12210 Commissioner and Counsel New 'fork State Department of Anthony Z. Roisman, Esq.
Comnerce L'erliu, Roisman and Kessler 99 Washington Avenue 1712 U Street, h.il.
Albany, New York 12210 Uashington, D. C. 20036 liiss Suzanne Weber 78 Uest Senaca Street D ISTR I BUT ION:
NRC PDR TR Branches Oswego, New York 13126 Local PDR 'LWR 2 File BC's'CRS Hr. Richard Goldsmith Docket (14)
Syracuse University R. Klecker J. Norris College of Law F. Williams E. I. White Hall, Campus F. Schroeder A. Kenneke Syracuse, New York 13210 V. t1oore R. DeYoung ELD IE (3) bcc: J. R. Buchanan, ORNL W. Kane Thomas B. Abernath DTIE H. SNith OFFICE~ L:LW 2-1 L:LWR2-1
~
SURNAME+ e .i JStolz +gg OAT~ 4//$ '/75 4JgS J75 Form AECDl8 (Rev. 9-53) AECM 0240 'A V.D OOVERKHEKT TRIKTIKO OTTDCE( IDTD~DD 959
4,
.~ a) s pr( fx.,i a'aQ
~ 1 I
1
'l I
5
'I II ~
rZCLnS<iRt:. 1 RE UIRI!D IYFORibKTION Provide large size plan and section drawings of the suppression chamber which illustrate the structures, equipment, and piping in and above the suppression pool. These drawings should be in sufficient detail to describe all equipment and structural surfaces which could be subjected to suppression pool hydrodynamic loadings.
(2) Provide a chronology of all potential pool dynamic, loads during a LOCA which identifies the source of the load (see enclosure 3),
the time interval over which the load is active, and the structures which are affected. (For an example, see GESSAR, Response 3.82.)
II For each structure or group of structures identified in paragraph (2) above, provide the anticipated load as a function of time due to each of the pool dynamic loads which could be imparted to the structure.
0 For each structure or group of structures identified in paragraph (2) above, provide the total load as a function of time due to the sum pool dynamic loads. o'nticipated Describe the manner in which the pool dynamic load characteristic shown in (4) above is integrated into the structural design of each structure.
Specify the relative magnitude of the pool dynamic load compared to other design basis loads for the structure.
Describe the manner by which potential asymmetric loads were considered in the containment design. Characterize the type and magnitude of possible as@metric loads and the capabilities of the affected structures to withstand such a loading profile. include consideration of seismically induced pool motion which could lead, to locally deeper submergenc s for certain horizontal vent stacks.
Provide justification for each of the load histories given in (3) above by the use of appropriate experimental data and/or analyses.
References to test data should indicate the specific test runs and data points and the manner by which they were converted to loads.
As an inter m measure, use of available experimental data may be acceptable; however, if it appears necessary, additional tests directl>
applicable to the LOCA pool dynamic load phenomenon and its analyses will be required.
Provide a description of the structural analysis methods, and a smeary of the results of your structural design evaluation which either demonst ate that the containment design can withstand the pool dynamic loads imposed upon the structure within adequate margins, or the design modifications required to meet allowable design limits.
(.
~
ENCLOSURE 2 BAC!(GROUND Pool Dynamics The need to consider suppression pool hydrodynamic loads in the design of certain parts of the Mark III containmcnt developed during the carly phases of the large-scale Mark III test program being conducted by the General Electric Company. A series of air tests were performed in March 1974 to scope the rang and magnitude of pool dynamic
'loads. It was recognized that more definitive tests >>ere required and therefore comprehensive tests in 1/3 scale were initiated in the summer of 1974 and are currently still in progress. Parallel efforts to develop analytical models for the various pool dynamic phenomena have been implemented by the General Electric Company, several architect/engineers, and by NRC consultants.
The HRC staff has maintained contact with GE regarding the planning and progress of the pool dynamics testing and associated analyses. Due to the commonality of the >>ater prcssure suppression feature in Mark I, II, and III type containments it was apparent that pool dynamic loads could also be a consideration for Mark I and II plants. GE, in fact, is in the process of planning a series of tests for ASEA/Atom of Sweden.
The purpose of tnese tests would be to determine pool dynamic loads for a structure located immediately above the suppression pool for. a,containment with vertical vent pipes. The basis for applying this data to specific Mark I and II designs has not yet been established.
Wc
~ 'NCLOSURE 3 DESCRIPTION OF POTENTIAL POOL DYNAMIC PHENOMENA, Following a design basis loss-of-coolant accident in the drywell, the dry-well atmosphere will be rapidly compressed due to bio>>down mass and energy addition to the drywell vo'lume. This compression would be transmitted in the form of a com ressive >>'ave and propogate through the vent system into the suppression pool. The pool response to this effect could include a load on the suppression chamber walls.
1('ith pressurization of the drywell, the water in the do>>ncomers will be
'epressed and forced out through-the vent system into the suppression pool. This movement of pool water can result in a water et im in ement load on the suppression chamber.
Following clearing of the vents an air/steam/>>ater mixture will flo>> from the drywell through the vents and be injected into the suopression pool
'elow the surface. Depending on the characteristics of the suppression system (i.e., the vent area comspared to the drywell volume and break flow area) dry>>ell overexpansions could occur. Overexpansion of the drywell results when the initial vent flow, following vent clearing, evacuates the drywell more rapidly than the volume is replenished by blowdown mass and energy input. If the drywell volume is relatively small compared to the area of the vents, then there is insufficient capacity to abso b 'he transitipn in venting rates and loads due to ~i~a.l~overe.>>ans'on peril lati.ons can occur on the suppression chamber and vent system.
During vent flow the steam component of the'flow mixture >>ill condense in the pool >>hile the air, being noncondensible, >>ill be released to the pool as high pressure air bubbles. Initial air bubble loads would be experienced by all pool retaining structures and could be of an osc'llatory mode due to overexpansion'nd recompression of the bubbles.
The continued addition and expansion of air within the pool causes .the pool volume to swell and therefore an acceleration of the surface vertically upward. This response of the pool is referred to as bulk pool s>>ell since the air is confined beneath the pool and is driving a solid 1'gament of
'water. Bulk ool s>>ell air bubble and flow dray. loads are imparted to the suppression chamber walls and to structures, components, etc., which may be located at low elevations. above the normal pool surface. Bulk vool swell Due to the ef ect of buoyancy, air bubbles >>ill rise faster than the pool water mass and will eventually break through the s>>ollen surface and relieve the driving force beneath the pool. This breakup of the water ligament
'eads to the upward expulsion of a 2-phase mix ure of air and water and is referred to as pool swell in the, froth mode. Struciures>>hich are located at higher clovaticns above "hc initial pool su..i'ace could experience a pool swell froth im inqemcnt.load due to impact of 2-phase flow.
0 2-Froth flow will continue until the fluid kinetic energy has been expended, followed by fallback of the water to the initial suppression pool level.
Structures located above the pool could be subject to water fallback loads.
~
Following the initial pool swell event the suppression system will settle
.,into a generally coherent phase during which significant vent flow rates are maintained from th'e dry>>ell to the pool. A resultant effect is the occurrence of hi h vent flow. steam condensation loads, which can be of an ~ 1~" v . on pool retaining structures. As the reactor coolant system inventory of mass and eneigy is depleted, near the end of blowdown, venting rates to the suppression pool diminish allowing water to reenter the downcomers. During phases of low vent mass flux the suppression system behaves in an oscillatory manner, referred to as chugging, whereby periodic clearing and subsequent recovery of vents occurs since the vent flow cannot. sustain bulk steam condensation at the vent exit. The resultant local fluctuations in pressure and water levels generate fichu Lggin
, predominantly on the vent system.
It should be further noted that the magnitude and range of any of the hydro-of the suppression system, either in the circumferential or radial direction.
One possible initiator of such response >>ould be seismicall induced ool motion >>hich could lead to locally deeper submergences for certain downcomers and therfore larger pool swell loads. Full account of this potentiality should be made in establishing hydrodynamic load capabil'ities for the suppression chamber structures design.
SER CI-10 The number of ADS SRVs is determined by a LOCA analysis. NMP2 will require seven SRVs to achieve a rapid depressurization during a small break LOCA. The applicant will confirm this by a plant-specific ECCS analysis.
RESPONSE
In a small-break ECCS analysis, a particular ADS flow rate is required to bring the vessel pr essure down in a prescribed time to allow the operation of low pressure core cooling following the postulated failure of HPCS. This ADS flow rate determines the number of ADS valves. Before the completion of NNP2 plant specific ECCS analysis, seven SRVs were designated for ADS function based on generic BWR/5 calculations including that of WNP-2 (Hanford) .
Subsequently, the NHP2 plant specific ECCS analysis in compliance with the criteria of 10CFR50, Appendix K, has confirmed the adequacy of 7-ADS-valve design by using only six ADS valves in the small b~eak analysis. See FSAR Table 6.3-1.
J II II 0
a I'fl, I
('k
~ 4
SER CI-15 NMPC must submit the results of a plant specific LOCA analysis.
~Res ense See revised FSAR Section 6.3
r,~f ~
~ > )I <<*
SER Cl-16 The applicant will pe~form the plant-specific ECCS analysis and will provide calculated maximum total hydrogen generation from the chemical reaction of the cladding with water or steam for the most limiting LOCA case.
RESPONSE
The NMP2 plant specific ECCS analysis shows that the most limiting core-wide metal (cladding) -water reaction is 0.07K of the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding around the plenum volume, were to react with core coolant. This is well within the 10CFR50.46 limit of 1X. See FSAR Table 6.3-4.
Il H
t
~
y I'
H ll l I h
II o ~
'I '"g R 0 ~ f
Nine Nile Point Unit 2 FSAR Criterion 4 Coolable Geometr "Calculated changes in core geometry shall be such that the core remains amenable to cooling." As described in NEDE-20566-P',Section III.A. Conformance to Criterion 4 )
is demonstrated by conformance to Criteria 1 and 2.
Criterion 5 Lon -Term Coolin "After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core." Conformance to Criterion 5 is demonstrated generically for GE BWRs in NEDE-20566-P',Section III.A. Briefly summarized, the core remains covered to at least the jet pump suction elevation and 'the uncovered region is cooled by spray cooling.
6.3.3 ' Single-Failure Considerations The functional consequences of potential operator errors and single failures (including, those that might cause any manually controlled electrically operated valve in the ECCS to move to a position that could adversely affect the ECCS),
and the potential for submergence of valve motors in the ECCS are discussed in Section 6.3.1. No potential single failures are more severe than the single failures identified in Table 6.3-3. It is, therefore, only necessary to consider each of these single fai lures in the ECCS performance analyses.
For large breaks, failure of one of the standby diesel generators is in general the most severe failure. For small breaks, the failure of the HPCS is the most severe failure; neither failure results in unacceptable consequences.
A single failure in the ADS (one ADS valve) has no effect on large breaks. Therefore, as a matter of calculational convenience, it is assumed in all calculations that one ADS valve fails to operate in addition to the identified single 20 failure. This assumption reduces the number of calculations required in the performance analysis and bounds the effects of one ADS valve failure and HPCS failure by itself. The only effect of the assumed ADS valve failure on the calculations is a small increase (approximately 100 F) in the calculated temperatures following small breaks.
Amendment 20 6.3-23 July 1985
Nine Mile Point Unit 2 FSAR 6.3.3.4 System Performance During the Accident In general, the system response to an accident can be described as:
Receiving an initiation signal,
- 2. A small lag time (to open all valves and have the pumps up to rated speed), and
- 3. ECCS flow entering the vessel.
Amendment 20 6.3-23a July 1985
Nine Mile Point Unit 2 FSAR THIS PAGE INTENTIONALLY BLANK Amendment 20 6.3-23b July 1985
Nine Mile Point Unit 2 FSAR Key ECCS actuation set points and time delays are provided in Table 6.3-1. The delay from the receipt of signal until the ECCS pumps have reached rated speed is subject to the physical limitations of accelerating the diesel generators and pumps. The delay time due to valve motion in the case of the high-pressure system provides a suitably conservative allowance for valves available for this application. In the case of the low-pressure system, the time delay for valve motion is such that the pumps are at rated speed prior to the time the vessel pressure reaches the pump shutoff pressure.
The flow delivery rates analyzed in Section 6.3.3 can be determined from the head-flow curves on Figures 6.3-3A, 6.3-4A, and 6.3-5A and the pressure versus time plots discussed in Section 6.3.3.7. Simplified piping and instrumentation and functional control diagrams for the ECCS are provided in Section 6.3.2. The operational sequence of the ECCS for the DBA is shown in Table 6.3-2.
Operator action is not required, except as a monitoring function, during the short-term cooling period following the LOCA. During the long-term cooling period, the operator takes action as specified in Section 6.2.2.2 to place the containment cooling system into operation.
Using the standard, approved licensing models and an assumed flow control valve (FCV) closure rate of 11 percent per second, generic BWR/5-6 analyses showed an increase of the peak cladding temperature (PCT) of <45 F. The ECCS calculations applicable to Unit 2 yield a PCT that can accommodate this increase without violating the 2,200 F limit of 10CFR50.46.
Provisions for limiting the FCV closure rate to are discussed below.
ll percent ECCS analysis for the double-ended recirculation pipe break design basis accident takes credit for a core flow coastdown resulting from recirculation pump coastdown in the unbroken loop. A hypothesized closure of the FCV in the unbroken loop during the first few seconds of a LOCA gives rise to the question of whether degraded flow coastdown could lead to increased values of peak clad temperature. Actual recirculation flow control valve closure, leading to degraded flow coastdown, is not an expected consequence of a LOCA event.
Amendment 20 6.3-24 July 1985
Nine Mile Point Unit 2 FSAR the entire system. The maximum allowable out-of-service time is a function of the level of redundancy and the specified test intervals (Section 15A).
6.3 '.7 ECCS Analyses for LOCA 6 '.3.7.1 LOCA Analysis Procedures and Input Variables The procedures approved for LOCA analysis conformance calculations are described in detail in Section S.2.5.2 GESTAR II. These procedures calculations discussed in Section 6.3.3.
were used in the The significant input variables used by the LOCA codes are listed in Table 6.3-1 and Figure 6.3-9.
6.3.3.7.2 Accident Description A detailed description of the LOCA calculation is provided in Section S.2.5.2 GESTAR II.
6.3.3.7.3 Break Spectrum Calculations zo A complete spectrum of postulated break sizes and locations is considered in the evaluation of ECCS performance. For ease of reference, a summary of all figures and tables presented in Section 6.3.3 is shown in Table 6.3-6.
A summary of the results of the break spectrum calculations is shown in Table 6.3-5 and graphically on Figure 6.3-8. 2 0 Conformance to the acceptance criteria (PCT 52,200 F, local oxidation 517 percent, and core wide metal-water reaction Sl percent) is demonstrated. Details of calculations for specific breaks are included in subsequent paragraphs.
6.3.3.7.4 Large Recirculation Line Break Calculations The characteristics that determine which is the most limiting large break are:
- 1. The calculated hot node reflooding time.
- 2. The calculated hot node uncovery time.
Amendment 20 6.3-25 July 1985
Nine Mile Point Unit 2 FSAR
- 7. Fuel rod convective heat transfer coefficient as a function of time.
- 8. Peak cladding temperature (PCT) as a function of time .
- 9. Average fuel temperature as a function of time.
- 10. PCT rod internal pressure as a function of time.
The maximum average planar linear heat generation rate >0 (MAPLHGR), maximum local oxidation, and PCT as a function of
~
exposure from the DBA analysis are shown in Table 6.3-4.
6.3.3.7a5 Transition Recirculation Line Break Calculations Important variables from the analysis of the transition (1.0 sq ft) break are shown on Figures 6.3-21 through 6.3-32. These variables are:
Core average pressure (large break methods) as a function of time.
- 2. Core flow (large break methods) as a function of time.
- 3. Core inlet enthalpy (large break methods) as a function of time.
- 4. MCPR (large break methods) as a function of time.
- 5. Water level (large break methods) as a function of time.
- 6. Pressure (large break methods) as a function of time.
- 7. Fuel rod convective heat transfer coefficient (large break methods) as a function of time.
- 8. PCT (large break methods) as a function of time.
- 9. Water level (small break methods) as a function of time.
- 10. Pressure (small break methods) as a function of time.
ll. Fuel rod convective heat transfer coefficients (small break methods) as a function of time.
Amendment 20 6.3-27 July 1985
Nine Mile Point Unit 2 FSAR
- 12. PCT (small break methods) as a function of time.
6.3.3.7.6 Small Recir'culation Line Break Calculations Important variables from the analysis of the small break yielding the highest cladding temperature are shown on Figures 6.3-33 through 6.3-36. These variables are:
- 1. Water level as a function of time.
- 2. Pressure as a function of time.
- 3. Convective heat transfer coefficients as a function of time.
- 4. PCT as a function of time.
The same variables resulting from the analysis of a less limiting small break are shown on Figures 6.3-37 through 6.3-40
'.3.3.7.7 Calculations for Other Break Locations Reactor water level, vessel pressure, fuel rod convective
! heat transfer coefficients, and PCT are shown on Figures 6.3-41 through 6.3-44 for the HPCS line break, Figures 6.3-45 through 6.3-48 for the feedwater line break, 20
) and Figures 6.3-49 through 6.3-52 for the main steam line break inside the containment.
An analysis was also done for the main steam line break outside the containment. Reactor water level, vessel pressure, fuel rod convective heat transfer coefficients, and PCT are shown on Figures 6.3-53 through 6.3-56.
6.3.3.8 LOCA Analysis Conclusions Having shown compliance with the applicable acceptance criteria of Section 6.3.3.2, it is concluded that the ECCS will perform its function in an acceptable manner operation and meet all of the 10CFR50.46 acceptance criteria, given at or below the MAPLHGRs in Table 6.3-4.
6.3.4 Tests and Inspections 6.3.4. 1 ECCS Performance Tests All systems of the ECCS are tested fo operational ECCS function during the preoperational and/or startup test program. Each component is tested for power source, range, Amendment 20 6.3-28 July 1985
Nine Mile Point Unit 2 FSAR TABLE 6.3-1 SIGNIFICANT INPUT VARIABLES USED IN THE LOSS-OF-COOKANT ACCIDENT ANAKYSIS A. Plant Parameters Core thermal power 3, 461 MWt ao Vessel steam output 15.0 x 10~ ibm/hr Corresponding percent 105 of rated steam flow Vessel steam dome pressure 1,055 psia Maximum area of recircula- 3.1 ft~
tion line break B. ECCS Parameters B. 1 KPCI system 20 Vessel pressure at 5225 psid (vessel I which flow may commence to containment)
Minimum rated flow at 21,200 gpm io vessel pressure 20 psid (vessel to containment)
Initiating signals:
Low water level (Ll) 1.0 ft above or top of active fuel High drywell pressure 2.0 psig Maximum allowable time 27 sec delay from initiating signal to pumps at rated speed Pressure at which inject- 275 psig~ >>
ion valve may open Injection valve fully Greater of: 20 open a)40 sec after DBA b)20 sec after pressure permissive Amendment,20 1 of 3 July 1985
Nine Mile Point Unit 2 FSAR TABLE 6.3-1 (Cont)
B.2 LPCS system Vessel pressure at which 5289 psid (vessel flow may commence to containment) t rated flow 6,250 2'inimum gpm at vessel pressure 122 psid (vessel to containment)
Initiating Signals:
Low water level (Ll) 1.0 ft above top or of active fuel High drywell pressure 2.0 psig 20 Minimum runout flow 7,214 gpm Maximum allowed delay time 27 sec from initiating signal to pump at, rated speed Pressure at which in- 335 jection valve may open psig'2'reater 20 Injection valve fully of:
open a)40 sec after DBA b)20 sec after pressure permissive B. 3 HPCS Vessel pressure at which 1,160 psid (vessel flow may commence to sources)
Minimum rated flow avail- 516 gpm 9 1, 160 psid able at vessel pressure 1, 550 gpm 6,250 gpm 9 200 psid I 1, 130 psid 20 (vessel to pump suction)
Initiating signals:
Low water level (L2) or 8.6 of ft above top active fuel 20 High drywell pressure 2.0 psig 1I Minimum runout flow 6,250 20 gpm I Maximum allowed delay time 27 sec from initiating signal to rated flow available and injection valve wide open Amendment 20 2 of 3 July 1985
Nine Mile Point Unit 2 FSAR TABLE 6.3-1 (Cont)
B.4 ADS Total number of valves installed Number of valves used in I 20 analysis Minimum flow capacity of 4.8 x 106 ibm/hr 20 six valves at pressure 1, 125 psig Initiating signals:
Low water level (Ll) 1.0 ft above top of I 20 and active fuel Signal that at least LPCS(145) psig 20 one LPCS or LPCI pump LPCI 125 psig is running (pump dis-charge pressure)
Delay time from all initi- 120 sec I 20 ating signals completed to the time valves are open C. Fuel Parameters Fuel type Initial core Fuel bundle geometry 8 x 8 Lattice Number of fueled 62 rods/assembly Peak technical 13.4 kw/ft specification LHGR Initial minimum critical 1.2 0 20 power ratio Design axial peaking factor
' 'Corresponds to injection valve differential pressure of ao 62 psid.
'Corresponds to injection valve differential pressure of 20 35 psid.
Amendment 20 3 of 3 July 1985
Nine Mile Point Unit' FSAR
~ TABLE 6.3-2 ACCIDENT'i'0 OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEMS FOR DESIGN BASIS Time
~sec Event Design basis LOCA assumed to start; normal auxiliary power assumed to be lost.
.w 0 Drywell high pressure'2'nd reactor low water /
20 level reached. All diesel generators signaled to start; scram; HPCS, LPCS, LPCI signaled to start on high drywell pressure.
20 Reactor low-low water level reached. HPCS receives second signal to start.
Reactor low-low-low water level reached. Second 20 signal to start LPCI and LPCS; auto-depressurization sequence begins; MSIVs signaled to close.
HPCS diesel generators ready to load; energize HPCS pump motor; open HPCS injection valve.
20
< 16 Divisions 1 and 2 diesel generators ready to load.
~ 30 HPCS injection valve open and pump at design flow, which completes HPCS startup.
Pressure permissive for LPCS injection valve reached.
% 33 Pressure permissive for LPCI injection valve reached. LPCI and LPCS pumps at rated speed. 20 LPCS pump at rated flow, LPCS injection valve open, which completes the LPCS startups' 53 LPCI pump at rated flow and LPCI injection valve open which completes LPCI startup.
Amendment 20 1 of 2 July 1985
Nine Mile Point Unit 2 FSAR TABLE 6.3-2 (Cont)
Time
~sec Event See Fig. Core effectively reflooded assuming worst single 6.3-15 failure; heatup terminated.
210 Operator shifts to suppression pool cooling.
min
For the purpose of all but the next to last entry on this table, all ECCS equipment is assumed to function as designed. Performance analysis calculations consider the effects of single equipment failures (Sections 6.3.2.5 and 6.3.3.3).
'~'No credit is taken in the LOCA analysis for ECCS 20 start on the high drywell pressure signal.
'LPCS considered failed for the limiting case.
Amendment 20 2 of 2 July 1985
Nine Mile Point, Unit 2 FSAR TABLE 6.3-3 SINGLE ACTIVE FAILURES CONSIDERED IN THE ECCS PERFORMANCE EVALUATION~ii Suction Break Assumed Failure S stems Remainin Division II standby 6 ADS, HPCS, LPCS, 1 LPCI 20 diesel generator 20 Division I standby 6 ADS, HPCS, 2 LPCI I diesel generator 20 HPCS 6 ADS, LPCS, 3 LPCI
Other postulated failures are not specially considered because they all result in at least as much ECCS capacity as one of the above designed failures.
'2'Systems remaining, as identified in this table, are applicable to all non-ECCS line breaks. For a LOCA from an ECCS line break, the systems remaining are those listed, less the ECCS in which the break is assumed.
'~'Analysis performed with six of the seven installed ADS valves (see Section 6.3.3.3).
Amendment 20 1 of 1 July 1985
Nine Mile Point Unit 2 FSAR TABLE 6.3-4 MAPLHGR, MAXIMUM LOCAL OXIDATION, AND PEAK CLAD TEMPERATURE VERSUS EXPOSURE (DBA LOCA)
Average Planar Exposure MAPLICGR PCT Oxid'
~MWD T ~kW ft ~oF ac
'fr Fuel T e P8CRB233 200 11 9 F 1904 0.009 1,000 12 ' 1898 0.008 5,000 12.1 1865 0.007 10,000 12.2 1859 0.007 15,000 12.2 1874 0.007 20,000 12.2 1878 0.008 25,000 11.7 1828 0.006 30,000 11.3 1765 0.005 35,000 10.7 1695 0.004 40,000 10 F 1 1637 0.003 45,000 9.5 1579 0.002 Fuel T e P8CRB183 2 0 200 12.0 1901 0.009 1,000 12. 1 1901 0.009 5,000 12. 7 1903 0.008 10,000 12. 8 1905 0.008 15,000 12.9 1921 0.009 20,000 12.8 1913 0.008 25,000 11.8 1806 0.006 30,000 10.9 1697 0.004 35,000 10 F 1 1608 0.003 40,000 9.4 1549 0.002 45,000 8.8 1487 0.001 Fuel T e P8CRB071 200 11.5 1802 0.006 1,000 11.4 1770 0.005 5,000 11.4 1728 0.004 10,000 11.5 1724 0.004 Amendment 20 1 of 2 July 1985
Nine Mile Point Unit 2 FSAR TABLE 6.3-4 (Cont)
Average Planar >>
~MWD T ~kW ft ~F ~Fr ac Fuel T e P8CRB071 (Cont) 15, 000 11.5 1723 0.004 20, 000 11.1 1686 0.004 25, 000 10.4 1633 0.003 30, 000 9.8 1571 0.002 35,000 9.1 1499 0.002 40,000 8.5 1425 0.001 45,000 7.8 1355 0.001 20
' 'The corewide metal-water reaction for the subject plant has been calculated using Method 1 described in Reference 2. The value is 0. 07%.
Amendment 20 2 of 2 July 198S
Nine Mile Point Unit 2 FSAR TABLE 6.3-5
SUMMARY
OF RESULTS OF LOCA ANALYSIS Peak Local Single PCT Oxidation Break Size Location Failure ~OF O 3.1 sq ft Recirc.
suction LPCS diesel 1,921' o.ss I (DBA) break generator failure 1.0 sq ft Recirc. HPCS 1,>Set >> O.e6 (large break suction diesel methods) break generator failure 1.0 sq ft Recirc. HPCS 1,20m< >> <0.2 (small break suction diesel methods) break generator failure 0.09 sq ft Recirc.
suction HPCS diesel 1,522'~~ <0.2 break generator failure
Large break methods
'Small break methods Amendment 20 1 of 1 July 1985
Nine Nile Poi t 2 FSA4 TABLE 6 3"5 KFT TO PIGVRE W'NBERS+
Transition Recircula- Saall Reqirculation tion Line Brea',s Liner Breaks Large 1 0 ft* 1 ' ft~ 0 09 fthm . 0.7 ft~ Other Break Locations Recirculation Large Snail Highest Additional Core Feed- Main Stean Nain Stean Line Breaks Break Break Teap Snail. Snail Spray vater Line Inside Line Outside DBA Method Method Break Brea'k Line Line Containnent Containnent Core average pressure 21 Core average inlet flov 12 22 Core inlet enthalpy 23 N. ninua critical pover ratio 14 24 Rater level inside shroud 15 25 29 33 37 41 45 53 20 Reactor vessel pressure 16 26 30 38 46 50 Convective heat transfer 20 coefficient 17 27 35 39 43 47 51 55 I Peak cladding tenperature (PCT) 18 28 32 36 40 52 Average fuel temper at ure 19 PCT rod internal pressure 20 Nornalized pover vs. tine Anendaent 20 1 of 2 July 1985
Nine Mile Obit 2 FSAR TA BLE 6. 3-6 (Cont)
Transition Recircula- Snail Recirculation tion Line Breaks Line Breaks Large 1.0 fthm 1.0 fthm 0. 09 f tR 0.7 ft2 Other Break Locations ti Rec irc ula on Line Breaks Large Snail Highest hddit ional Core Feed- Main Stean Main Steam Break Break Te np Snail Snail Spray water Line Inside Line Outside BBA Method Method Break I Break Line Lin e Co nta in ne nt Cont ai n aent Peak claddinq teapera't ure and peak local oxidation versus break area vo
/
Total tire for which hi qhest powered node renains uncovered versus l reak 10 area
+All figure nunbers refer to figures in Section 6.3.
laendnent 20 2 of 2 July laSS
5000 C
O 4000 K
0 0
C x 3OOO 9
2000 g
1000 1000 2000 3000 4000 5000 6000 7000 8000 9000 F LOW (gpm)
F IG U RE 6.3-3A HEAD VERSUS HIGH PRESSURE CORE SPRAY FLOW USED IN LOCA ANALYSIS NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
800 700 600 C) cr 500 C) 400 300 200 100 3 4 FLOW (1000 GPH)
FIGURE 6.3-4A HEAD VERSUS LOW PRESSURE CORE SPRAY FLOW USED IN LOCA ANALYSIS NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
I 600 500 400 C) 300
.200 a.
100 3 4 FLOW (1000 GPH)
FIGURE 6.3-5A HEAD VERSUS LOW PRESSURE COOLANT INJECTION FLOW USED IN LOCA ANALYSIS NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
4 SUCTION BREAK LPCI 0/G FAILURE SUCTION BREAK LPCS 0/G FAILURE SUCTION BREAK HPCS FAILURE.
ta.
O 200 I HAX. CSLN BREAK LPCS 0/G FAILURE 4J I
CB C)
C) llAX~ STHO BREAK ED CI liPCS FAILURE C7 8
IOOO t b HAX. FOMR BREAK HPCS fAILURE HAX. STHL BREAK LPCI 0/G FAILURE Sl ALL BREAK HET1100 LARGE BREAK HETHOO 20.
IOR SUCT BREAK FOR SUCT BREAK 10.
0.
0.01 O. I 1.0 BRCAK AREA (SO. CT.)
FIGURE 6.3-8 PEAK CLADDING TEMPERATURE AND PEAK LOCAL OXIDATIONVERSUS BREAK AREA NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
200 180 o 160 o
m'40 E
120 o
o 1 ~ 0 FT2 BREAK cr: 100 I
g ep 60 40 I
o I
20 10 20 30 40 50 60 20 ep 90 100 BREAK AREA If OF DBA)
FIGURE 6.3-10 TOTAL TIME FOR WHICH HIGHEST POWERED NODE REMAINS UNCOVERED VERSUS BREAK AREA, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
1 200 cr 800 W
cQ
~C 4l 400 0.
0 8. 12. 16.
TIRE AFTER BREAK (SEC.)
FIGURE 6.3-11 CORE AVERAGE PRESSURE FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE IVIILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
~ >~ <a, j4
JET PBNP URCOTERT
$ 0.2 LGHER PLEHVN FLASHING
-0,$
- 8. 12.
TINE AFTER BREAK (SEC.)
FIGURE 6.3-12 NORMALIZED CORE AVERAGE INLET FLOW FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
560.
590.
W 5Ll 520.
500.
0 8. 12. 16.
TINE AFTER BREAK (SEC.)
FIGURE 6.3-13 CORE INLET ENTHALPY FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
1.4 1.3 1.2 CI 1.1 JET PINP UNCOVERY 1
0.9 0.8 0.7 ROTE 1 0.8 0.5 0.4 0.3 CPR 1.0 at SPACER 2 0.2 NOTE 1 0.1 TIRE AFTER BREAK {SEC.)
FIGURE 6.3-14 MINIMUMCRITICAL POWER RATIO FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
di, eo g AO Ial I
lal TAF CI
~ II Ial lal 20 I BAF 0
0 100. 200. 300. 400.
TIHE AFTER BREAK tSEC.)
FIGURE 6.3-15 WATER LEVEL INSIDE SHROUD FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
1200 800 le vs 400 CC CI I
LJ 0.
0 ~ 100. 200. 300. TI00.
TINE AFTER BREAK (SEC.)
FIGURE 6.3-16 REACTOR VESSEL PRESSURE FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
kl
'I
10 JET PUNP UNCOYERY I
I 4.
2 I
lal IO't Ll I
IO ONSET OF LONER PLENUH FLASHING TO ZERO 0.1 0 10. 20. 30. 40.
TINE AFTER BREAK (SEC.)
FIGURE 6.3-17 FUEL ROD CONVECTIVE HEAT TRANSFER COEFFICIENT DURING BLOWDOWN AT THE HIGH POWER AXIALNODE FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
3000 HIGHEST PONERED AXIAL PLANE HIGHEST PDNERED AXIAL PLANE TO EXPERIENCE CPR
~ 1.0 PRIOR TO JET PUHP UNCOVERY HIGH PONER AXIAL PLANE REFLOODED 2000 W
I ri1 BEGIHHIHG OF SPRAY COOLIHG /
CI 1000 r
p l HIGH PDHER AXIAL PLANE UNCOVERED ONSET OF BOILING TRANSITION 0.
0.1 10. 100. 1000.
TIHE AFTER BREAK ISEC.)
FIGURE 6.3-18 PEAK CLADDING TEMPERATURE FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MlLE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
3000 2000 III I
CI 1000 0.
0.1 10. 100. 1000.
TltK AFTER BREAK (SEC.)
FIGURE 6.3-19 AVERAGE FUEL TEMPERATURE FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
P>>
150 '
100.
50.
0.
0.1 10. 100. 1000.
TllK AFTER BREAK (SEC.)
FIGURE 6.3-20 PCT ROD INTERNAL PRESSURE FOLLOWING A DESIGN BASIS ACCIDENT RECIRCULATION SUCTION BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER COR PORAT ION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
1200 800 4J la>
4I 5LI F00 0.
0 10. 20. 30. 40.
TlHE AFTER BREAK {SEC.)
FIGURE 6.3-21 CORE AVERAGE PRESSURE FOLLOWING A 1.0 SQ FT BREAK (LBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
>'j T-HIHDOH 0.2 LGHER PLEHUH FLASHING
-0. 2
- 0. 10. 20. 30. LIO.
TIHE AfTER BREAK (SEC.)
FIGURE 6.3-22 NORMALIZED CORE AVERAGE INLET FLOW FOLLOWING A 1.0 SQ Ft BREAK (LBM)
RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
560.
500.
520.
- 10. 20. 30. 40.
TIE AFTER BREAK (SEC.)
FIGURE 6.3-23 CORE INLET ENTHALPY FOLLOWING A 1.0 SQ FT BREAK (LBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
'l tl 2.8 2.6 2A 22 DN 2
I 1 8 1.6 1.4 1.2 0.8 0.6 0.4 0.2 8 10 12 14 16 18 TIHE AFTER BREAK (SEC.)
FIGURE 6.3-24 MINIMUMCRITICAL POWER RATIO FOLLOWING A 1.0 SQ FT BREAK (LBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
h 60 CI o 40 4I W TAF
~ W 20 BAF 0.
0 100. 200. 300. 400.
TJNE AFTER BREAK tSEC ~ )
FIGURE 6.3-25
. WATER LEVEL INSIDE SHROUD FOLLOWING A 1.0 SQ FT BREAK (LBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
1200 800 5
I tJ loo 0.
0, 100. 200. 300. 000.
TIHE AFTER BREAK (SEC.)
FIGURE 6.3-26 REACTOR VESSEL PRESSURE FOLLOWING A 1 0 SQ FT BREAK (LBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATIO NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
h:
0
I U ONSET OF BOILING TRANSITION I
~g 10 I
la, HIGH PURER AXIAL PLANE UNCOVERED 10 8
I 10 W
I ONSET OF LONER PLENUH FLASHING Cl Ll TO ZERO 0.1
- 10. 20. 30. 40. 50. 60. 70. 80.
TINE AFTER BREAK (SEC.)
FIGURE 6.3-27 FUEL ROD CONVECTIVE HEAT TRANSFER COEFFICIENT DURING SLOWDOWN AT THE HIGH POWER AXIALNODE FOLLOWING A 1.0 SQ FT BREAK (LBM) RECIRCULATION SUCTION BREAK, HPC8 FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
<~ ~ 'Ps "I '
3000 a
2000 HIGH PONER AXIAL PLANE REFLOOOEO 5
I C5 CD LP BEGINNING OF SPRAY COOLING 1000 HIGH PONER AXIAL PLANE UNCOVERED ONSET OF BOILIPS TRANSITION
- 0. 1000.
O.I 10. 100.
TIME AFTER BREAK tSEC.)
FIGURE 6.3-28 PEAK CLADDING TEMPERATURE FOLLOWING A 1.0 SQ FT BREAK (LBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
60 CI o 40 TAF OC 20 BAF
- 0. il00.
0 100. 200. 300.
TINE AFTER BREAK (SEC.)
FIGURE 6.3-29 WATER LEVEL INSIDE SHROUD FOLLOWING A 1.0 SQ FT BREAK (SBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
1200 800 4I lal CC C7 I
400 0.
- 0. 100. 200. 300. 400.
TIME AFTER BREAK (SEC.)
FIGURE 6.3-30 REACTOR VESSEL PRESSURE FOLLOWING A 1.0 SQ FT BREAK (SBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
200. 300. 1IOO.
TIRE AFTER BREAK (SEC.)
FIGURE 6.3-31 FUEL ROD CONVECTIVE HEAT TRANSFER COEFFICIENT FOLLOWING A 1.0 SQ FT BREAK (SBM) RECIRCULATION SUCTION BREAK. HPCS FAILURE NIAGARA MOHAWK POWER CORPORATIO NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
3000 2000 I
5al
~
I CI El W
1000 0.
- 0. 100. 200. 300. IIOO.
TINE AFTER BREAK (SEC.)
FIGURE 6.3-32 PEAK CLADDINGTEMPERATURE FOLLOWING A 1.0 SQ. FT BREAK (SBM) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
9 60 CI o~ 40 4I I
UI
à 20 BAF 0.
0 200. GOO. GOO. 800.
TIHE AFTER BREAK tSEC.)
FIGURE 6.3-33 WATER LEVEL INSIDE SHROUD FOLLOWING A 0.09 SQ FT BREAK (HIGHEST TEMPERATURE SMALL BREAK) RECIRCULATION SUCTION BREAK. HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
8~
1200 400 0.
- 0. 200. 1IOO. Gon. 800, TIKE AFTER BREAK (SEC.)
FIGURE 6.3-34 REACTOR VESSEL PRESSURE FOLLOWING A 0.09 SQ FT BREAK (HIGHESY TEMPERATURE SMALL BREAK) RECIRCULATION SUCTION BREAK. HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MlLE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
106 I
104 L
@II 10 I
V 5LJ 1
- 0. 200. 1I00. 600. 800, TIIK AFTER BREAK tSEC.)
FIGURE 6.3-35 FUEL ROD CONVECTIVE HEAT TRANSFER COEFFICIENT FOLLOWING A 0.09 SQ FT BREAK (HIGHEST TEMPERATURE SMALL BREAK)
RECIRC. SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
3000 4.
OOOO 1000 0.
- 0. 200. 1IOO. GOO. 800.
TINE AFTER BREAK (SEC.)
FIGLIRE 6.3-36 PEAK CLADDING TEMPERATURE FOLLOWING A 0.09 SQ FT BREAK (HIGHEST TEMPERATURE SMALL BREAK) RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
e 4tp
60 io TAF CL'0 BAF
- 0. 400,
- 0. 100. 200. 300.
TIKE AFTER BREAK tSEC.)
FIGURE 6.3-37 WATER LEVEL INSIDE SHROUD FOLLOWING A 0.7 SQ FT BREAK (ADDITIONALSMALL BREAK)
RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
o ~, r 1200 BOO 4J i@I OC 4l Ot.'
400 CL'.
- 0. 100. 200. 300. IIOO.
TIME AFTER BREAK (SEC.)
FIGURE 6.3-38 REACTOR VESSEL PRESSURE FOLLOWING A 0.7 SQ FT BREAK (ADDITIONALSMALL BREAK)
RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
4>
v'4
~ ~
' ~ ~ 0 0
~ o o ~
~ ~ ~ ~
~ a 0 ~
0
~
0 0
~ 0
3000 g 2000 5
I Ch 4l, 1000 0.
0 100. 200. 300. ((00.
TJME AF1ER BREAK {SEC.)
FIGURE 6.3-40 PEAK CLADDING TEMPERATURE FOLLOWING A 0.7 SQ FT BREAK (ADDITIONALSMALL BREAK),
RECIRCULATION SUCTION BREAK, HPCS FAILURE NIAGARA MOMAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
PW W~~ C 4
I4
60 gCI 40
à 20 I BAF 0.
- 0. 200, 400. 600, 800.
TINE AFTER BREAK (SEC ~ )
FIGURE 6.3-41 WATER LEVEL INSIDE SHROUD FOLLOWING A MAXIMUMHPCS LINE BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MlLE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
0 1200 B00 W
0 i@I OC C7 400 0.
0 200. 400. 600. 800.
TIME AFTER BREAK (SEC.)
FIGURE 6.3-42 REACTOR VESSEL PRESSURE FOLLOWING A MAXIMUMHPCS LINE BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POlNT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
a Ja s 10 4
I
~ V I
104 I
h
%J 10 I
5 I
LP 5
- 0. 200. 400. GOO. 800.
TIKE AFTER BREAK ISEC.)
FIGURE 6.3-43 FUEL ROD CONVECTIVE HEAT TRANSFER COEFFICIENT FOLLOWING A MAXIMUMHPCS LINE BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
>o'll ~
)
3000 2000 P
H 1000 Ll 8
0.
0 200. IIOO. 600, 800.
TIME AFTER BREAK (SEC.)
FIGURE 6.3-44 PEAK CLADDING TEMPERATURE FOLLOWING A MAXIMUMHPCS LINE BREAK, LPCS DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
h 1 60 Cl 40 Vl W
I W
0 200. 900. 600. 800.
TIHE AFTER BREAK tSEC.)
FIGURE 6.3-45 WATER LEVEL INSIDE SHROUD FOLLOWING A MAXIMUMFEEDWATER LINE BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
'+1% I 1200 800 4J OC 400 DC I
EJ lal
- 0. 600. 800.
0 1IOO.
T!KE AFTER BREAK (SEC.)
FIGURE 6.3-46 REACTOR VESSEL PRESSURE FOLLOWING A MAXIMUMFEEDWATER LINE BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
C.
106 10
- 0. 200. EIOO. GOO. 800.
TIHE AFTER BREAK (SEC.)
FIGURE 6.3-47 FUEL ROD CONVECTIVE HEAT TRANSFER COEFFICIENT FOLLOWING A MAXIMUM FEEDWATER LINE BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
1&tlAI"-
3000 2000 P
W I~
1000 0.
- 0. 200. 1IOO. 600. 800.
TIHE AFTER BREAK (SEC.)
FIGURE 6.3-48 PEAK CLADDING TEMPERATURE FOLLOWING A MAXIMUMFEEDWATER LINE BREAK, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
60 40 Q
g W TAF W
CI ZQ BAF 0.
- 0. 200. 400. 600. 800.
TtHE AFTER BREAK tSECo)
FIGURE 6.3-49 WATER LEVEL INSIDE SHROUD FOLLOWING A MAXIMUMMAIN STEAM LINE BREAK INSIDE CONTAINMENT LPCI DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
O4"
~
N~
l
1200 400 5
I 0.
- 0. 200. 400. 600. 800.
FIGURE 6.3-50 REACTOR VESSEL PRESSURE FOLLOWING A MAXIMUMMAIN STEAM LIN'E BREAK INSIDE CONTAINMENT, LPCI DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATIO NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
C>
4
000. 600. 800.
>ted mEa aacav (SEC.)
FIGURE 6.3-51 FUEL ROD CONVECTIVE HEAT TRANSFER COEFFICIENT FOLLOMING A MAXIMUM HAIN STEAM LINE BREAK INSIDE CONTAINMENT, LPCI DI ES EL GENERATOR FAI LURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
3000 3000 g
5I u 1000 0.
0 200. 400. 600. 800.
TIHE AFTER BREAK {SEC.)
FIGURE 6.3-52 PEAK CLADDING TEMPERATURE FOLLOWING A MAXIMUM MAIN STEAM LINE BREAK INSIDE CONTAINMENT LPCI DIESEL GENERATOR FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
'f'yl I 40
~ al 20 BAF CL'al I
- 0. 1200 0 300 600 900 TINE AFTER BREAK {SEC. I FIGURE 6.3-53 WATER LEVEL INSIDE SHROUD FOLLOWING A MAXIMUMMAIN STEAM LINE BREAK OUTSIDE CONTAINMENT, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
1200 800 le i@I IJI 5
4oo 0.
0 300 600 900 1200 T))]E AFTER BREAK (SEC.)
FIGURE 6.3-54 REACTOR VESSEL PRESSURE FOLLOWING A MAXIMUMMAIN STEAM LINE BREAK OUTSIDE CONTAINMENT, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
10 CJI U
I DC I
CD 10 CD lJI IJ I
I lJI IJI 10 I
LC R
CD 1
- 0. 300 600 900 1200 TIKE AFTER BREAK (SEC.I FIGURE 6.3-55 FUEL ROD CONVECTIVE HEAT TRANSFER COEFFICIENT FOLLOWING A MAXIMUMMAIN STEAM LINE BREAK OUTSIDE CONTAINMENT, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
Mi 3000 2000 I
1000 0.
0 ~ 300 600 900 1200 TIHE AFTER BREAK (SEC.)
FIGURE 6.3-56 PEAK CLADDINGTEMPERATURE FOLLOWING A MAXIMUMMAIN STEAM LINE BREAK OUTSIDE CONTAINMENT, HPCS FAILURE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
Nine Mile Point Unit, 2 FSAR
'TABLE 15.6-5 SEQUENCE OF EVENTS FOR STEAM LINE BREAK OUTSIDE CONTAINMENT Time
{ serac Event 0 Guillotine break of one main steam line outside primary containment.
0.5 High steam line flow signal initiates closure of MSIVs.
(1.0 Reactor begins scram.
S5.5 MSIVs fully closed.
~30 RCIC and HPCS would initiate on low-low water ao level (RCIC considered unavailable; HPCS assumed single failure and therefore not available).
~70 SRVs open on high vessel pressure. The valves open and close to maintain vessel pressure at approximately 1,100 psi.
Reactor water level begins to drop slowly due to loss of steam through the SRVs; reactor pressure still at approximately 1,100 psi.
~720 ADS would signal to initiate on low-low-low water level.
~840 ADS initited. Vessel depressurizes rapidly.
~l, 040 Low pressure ECCS initiated with reactor fuel 20 partially uncovered.
~1, 135* Core effectively reflooded and clad temperature heatup terminated; no fuel rod failure.
~See Section'.3.3.
Amendment 20 1 of 1 July 1985