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#REDIRECT [[B12600, Submits Info Re ECCS Mods,Per NRC 870707,10 & 15 Telcons. Attachment 1 Provides Random Stress Summary by Data Point for 8-inch Piping Cross Tie Between RHR & Safety Injection Sys.W/One Oversize Drawing]]
| number = ML20235Y356
| issue date = 07/20/1987
| title = Submits Info Re ECCS Mods,Per NRC 870707,10 & 15 Telcons. Attachment 1 Provides Random Stress Summary by Data Point for 8-inch Piping Cross Tie Between RHR & Safety Injection Sys.W/One Oversize Drawing
| author name = Mroczka E
| author affiliation = CONNECTICUT YANKEE ATOMIC POWER CO.
| addressee name =
| addressee affiliation = NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
| docket = 05000213
| license number =
| contact person =
| document report number = B12600, NUDOCS 8707250215
| package number = ML20235Y359
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 7
}}
 
=Text=
{{#Wiki_filter:_ . _ _    - - _ - _ _ - - _ _ - - _ _ - _
CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N, CON N ECTICU T P.o box 270 e HARTFORD, CONNECTICUT 0614M>270 1 ELEPHONE 2 3**
July 20,1987 Docket No. 50-213 B12600 U.S. Nuclear Regulatory Commission Attn:                          Document Control Desk Washington, D.C. 20555
 
==Reference:==
(1) E. 3. Mroczka letter to U.S. Nuclear Regulatory Commission, "ECCS Modifications - Additional Information - Request for Extension of Single Failure Exemption," dated April 1,1987.
Gentlemen:
Haddam Neck Plant ECCS Modifications - Additional Information Request for Extension of S!ngle Failure Exemption As requested by the NRC Staff during telephone conversations on July 8, (0, and 15, 1987, Connecticut Yankee Atomic Power Company hereby submits the following information on the ECCS modifications at the Haddam Neck Plant.
As previously stated in Reference 1, an interlock will be installed in 1989 to prevent SI-MOV-901 and/or SI-MOV-902 from opening while the LPSI pumps are running. Similarly, the LPSI pumps will be prevented from spuriously starting when SI-MOV-901 and/or SI-MOV-902 are open, by placing the LPSI pump control room switch in the trip / pull-out position. This manual action will remove all automatic start signals.
A question was raised by the NRC Staff as to how a certain accident scenario would be mitigated. The scenario postulates a break in one of the four HPSI injection lines between the MOV and the check valve. It also assumes the MOV in this same line spuriously opens. The check valve effectively isolates backflow from the reactor coolant system (RCS) into lower pressure portions of the HPSI.
The probability of a HPSI injection line breaking in this location followed by a spurious opening of an MOV in the same line is very low, approximately 2E-7/yr.
per line. In addition, given sufficient time between the break and the spurious opening of the MOV, a break in such a line could be detected because the RWST would lose inventory and the containment sump' level would rise. This would occur because the RWST is higher than the HPSI discharge piping, thus causing the RWST to gravity drain. If leakage was insufficient to be detected during
)                normal operation with the HPSI system on standby, the break would be detected I                during the monthly test in which this piping is pressurized to approximately 1,400 lbs. to perform HPSI pump testing.
G707250215 870720                                      "
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U.S. Nuclear Regulatory Commission B12600/Page 2 July 20,1987 If this scenario were to occur, procedures for sump recirculation would direct the operator into one of two scenarios which would rnitigate this event:
: 1. Close two isolation valves located in the intact lines, thus resulting m one intact path to the RCS and one broken path. This is within the design basis and has been determined to provide adequate core cooling.
: 2. Close the valve in the broken line and one other valve, thus resulting in two intact paths to the RCS and one broken path. Closure of the valve in the broken -line will isolate the break and terminate the LOCA, therefore allowing normal shutdown systems to be used.
As requested, Attachment 1 provides a random stress summary by. data point for the 8-inch piping cross tie between the Residual Heat Removal and Safety Injection Systems (line No. 8"--SI-601 R-80). ' It should be noted that ' the                                              -
maximum stress does not occur on this piping run.
Attachment 2 provides supporting clarification for Table 1-A contained in Reference 1 as requested by the NRC Staff on July 15,1987.
If you have any additional questions, please contact us.
Very truly yours, CONNECTICUT Y.ANKEE ATOMIC POWER COMPANY Yba)
E.J.Rh'o7y'kaSenior Vide President (j cc:    W. T. Russell, Region 1 Administrator F. M. Akstulewicz, NRC Project Manager, Haddam Neck Plant
: 3. T. Shediosky, Resident Inspector, Haddam Neck Plant L_ __. _ _ _              _ _ - . _ _      _      _          __    _ _ _ .  -- _ _ _ _ _ _ _ _ _ - . _ _ _ _ . _ _ _ - _ _ _ _ _ _ _ _  _ _ _ _ _
 
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i ECCS MODir1CATIOtG                                    ,
STRE3S StiMARY 8"-SI-601R-80 CONDITION ANALYZED
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PRESSURE &                                            l DEAD 7EIGHT        THEPf1AL &        DBE SEISMIC &
DATA POINT STRESS (psi)      SAD STRESS (psi)  7PA STRESS (psi) 530          1520              7906                3287 553          1295              ?155                2203 4
575          1462              1520                3004 605          1109              1927                1139 619          1351              901                1505 655          1118              1946                1243        j 675          1298              9415                8919 3000          3186              4924                6730 3026          3651              3255                4415 3040          5370              3912                5001 l
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                  ' July 1987--                    ,1
 
N                                                                                                                                          )
l Attachment 2 I
The follouing information provides supporting clarification for Table 1-A contained b Peference 1.
I Table 1-A daicribed acceptable ECCS mitigation for the full spectrum of design'                                                          l oesis LOCAs fer both the injection and recirculation phases of ECCS operation.
The bases cf injection phase acceptability are the Westinghouse analyses (References 7 through 10) which demonstrated acceptable F.CCS performance in accordance with the Interim Acceptance' Criteria (Reference li) for the full spectrum of breaks which includes the small- and_large-break LOCAs. The result                                                            ,
of the Westinghouse analyses remain valid and unaffected by the Haddam Neck                                                              !
ECCS Long-Terrn Modifications.
1 Results of the analyses performed by NU using the NULAP5 met'hodology                                                                    !
(Reference 12), currently under NRC review, . vere also included in Table .1-A as                                                        I confirmatory evidence of acceptability of the long-term cooling perbrmance                                                                l using state-of-the-art ; methods. Specifically, the small-break LOCA p%k clad                                                            i temperatures of 13950 and ll810F were computed by the NULAP5 transient                                                                    j thermal hydraulic blowdowa code.                                                                                                          '
l For the recirculation phase of the LOCA, Table 1-A presented data which demonstrates the acceptab:!ity of ECCS performance without the need fer -                                                                j additional supporting ECCS licensing anc!ysis. The recirculation column of Table 1 A provides core boil-off rates with corresponding ECCS delivery flows ic,r the intended ECCS recirculation aligr. ment of two-out-of-ioub injection lines isolated. (In the Table 1-A recirculation case wherein the break occurs in the-                                                          'j j  HPSI line, onfy one irdection path provides delivery flow to the RCS.) Recircula-tion acceptability is demonstrated since the RCS boil-off rate is exceeded by l  ECCS delivery for all cases. Therefore, no additional analyses are required since                                                        i acceptable ECCS performance is assrd during the recirculation or long-term                                                                !
l  phase following a LUCA.
I
 
==References:==
 
l
: 1. E. 3, Mroczka letter to U.S. Nue: ear Regulatory Commission Document                                                                ;
Control Dask, "BCCS Modifications-Additional Information--Request for                                                                l Extension of Singte Failure Exemption," dated April 1,1984.                                                                        1
: 2. WCAP-8213, Effects of Fuel Densification on the Connecticut Yankee                                                                i
{      Jteactor, Octo'ber 1973.
I
: 3. Description and Safety, Including the Effects of Fuel Densification on_ the                                                        L
        , Connecticut Yankee Reactor, Cycle V, CYAPCO, November 1973.                                                                        I
: 4. D. C. Switzer letter to Assistant Director for Operating Reactors, USAEC, dated December 5,1972.
: 5. D. C. Switzer letter to A. Schwencer, dated May 2,1977.
(  6. D. C. Switzer letter to A. Schwencer, dated October 31,1977.
i
  ?. W. G. Counsil letter to D. M. Crutchfield, dated December 14, 1982.
L          m                  -                - . _ . - _ . _ - - . ---_--____--_____.__.--_-_______._______-____-__._____.___.-___-___4
 
B12600 Attachment 2                                                                        i Page 2
: 8. W. G. Ceunsil letter to D. M. Crutchfield, " Proposed Revisions to Technical 5 specifications-Coastdown at End of Cycle 12 Full-Power Life," dated March 30,1984.
: 9. 3. F. Opeka letter to C, I. Grimes, " Proposed Revision to Technical Specifi-cations Cycle 14 Reload," dated December 11,1985.
: 10. D. C. Switzer letter to D. 3. Skovolt, dated May 19,1972.
: 11. Interira Acceptance CrJteria; Federel Register 12247, dated June 29,1971.
: 12. NU Topical Repor t, "NULAF3, 2 FORTRAN IV Digital Cornputer Program for NSSS Blowdown and Fuel Rod Heat Up Anahses," April 1983.                    .
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2 OVERSIZE DOCUMENT PAGE PULLED SEE APERTURE CARDS NUMBEM OF OVERSIZE PAGES FILMED ON APERTURE CARDS ~
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APERTURE CARD /HARD COPY AVAILABLE FROM RECORD SERVlCES BRANCH,TfDC FTS 492
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