ML20072K058
| ML20072K058 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 03/23/1983 |
| From: | Irwin D HUNTON & WILLIAMS, LONG ISLAND LIGHTING CO. |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| ISSUANCES-OL, NUDOCS 8303300352 | |
| Download: ML20072K058 (67) | |
Text
'
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- LILCO, March 23, 19'53 <r 7. ~e,
.p UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
'03 ffg Before the Atomic Safety and Licensing Board DO u.Dr.
In the Matter of
)
'yjygg;p
)
%d LONG ISLAND LIGHTING COMPANY
)
Docket No. 50-322 (OL)
)
(Shoreham Nuclear Power Station, )
Unit 1)
)
i LILCO'S RESPONSE TO "SUFFOLK COUNTY MOTION TO STRIKE PORTIONS OF THE NRC STAFF'S PROPOSED OPINION AND FINDINGS OF FACT, LILCO'S PR9 POSED OPINION AND FINDINGS OF FACT, AND LILCO'S REPLY TO THE PROPOSED OPINION AND FINDINGS OF SUFFOLK COUNTY AND THE STAFF" In a motion served by mail upon the parties on March 8, 1983, Suffolk County has moved to strike three of LILCO's 743 initial Proposed Findings of Fact, three of the Staff's 421 Proposed Findings and their related discussion in the Staff's Proposed Opinion, and certain portions of LILCO's Reply to the Staff's and to SC's Proposed Opinion and Findings of Fact.
The motion is based on the premise that each " refers to, and has the Board draw conclusions based upon data that are not in evi-dence on the record of this proceeding."
SC Motion.at 1.
Although SC's motion to strike appears to identify several' distinct items, the findings at issue involve in actuality only five SNRC lettersl/ relating to two contentions (Safety Relief Valves, SC-22; and Mark II, SC-21) and miscellaneous correspon-dence relating to definitional issues between LILCO and the 1/
Only one item that SC. moves to strike, Note 37 in LILCO's Reply, does not involve correspondence between LILCO and the Staff.
Except for Note 6, below, it is not otherwise specifi-cally addressed in this response.
I 8303300352 830323 i
PDR ADOCK 05000322 Q
D
. Staff on Contention 7B.
Each of these items is a public document and was served on the Board and Suffolk County at the time it was transmitted to the Staff.
The following table clarifies the relationships of the submittals discussed in the SC motion:
Discussion Discussion Reference in LILCO Discussion Discussion in LILCO at Initial in SC in Staff Reply Issue Findings Findings Findings Findings
/
SRVs (SC-22) 22:28/20, Vol. 1 at SNRC-812 H-23 22:28/22 &
212,220 Opinion at 29 SNRC-816 H-37 22:28/38, Vol. I at
& Opinion 218 at 30 Mark II (SC-21) 7 Vol. 1 at SNRC-808/
G-14 SNRC 824 196-97 and Note'38 Vol. 1 at SNRC-8312/
[G-10]
203 (Note 39)
SC-7B Vol. 2 at Two Letters Between 36 LILCO & Staff (Note 13) 2/
SC objects to LILCO's reference to the completion of some confirmatory piping analyses which were to be transmitted to the Staff.
At the time of LILCO's Reply, there had been no submittal; thus the SNRC letter was not referenced in LILCO's Reply.
Subsequent to the Reply, these reanalyses were sent to the Staff in SNRC-831.
The analyses, although not the sub-mittal, were also discussed in LILCO Finding G-10, which SC does not move to strike.
r
. 9 As the table illustrates on its face, the correspondence actually at issue is small -- five SNRC letters 3/ and two letters between LILCO and the NRC Sta'ff not referred to in LILCO's findings.4/
Further, it shows that of the items addressed by LILCO (SNRC-812 and -816 on SRVs, SNRC-808, -824 and -831 on Mark II), all were explicitly cited or referred to in LILCO's Proposed Findings.
Since SNRC-824 and -831 were not written until after the initial filings, they could not have been referred to directly until the reply findings; however, as is shown below, their contents were presaged in the initial findings.
SC's motion should be rejected for both procedural and substantive reasons.
It is overreaching,5/ untimely, and pro-cedurally improper.
It is also ill-taken on the merits, in that it presumes the use by LILCO of the cited documents for the truth of their contents, whereas the use actually made by LILCO of the correspondence at issue in its findings is not to 3/
Copies of each of the SNRC letters are attached for the Board's convenience.
4/
LILCO does not here speak to the reference to the letters between LILCO and the Staff referred to in Staff Proposed Findings, Vol. 2 at 36 (Note 13), though LILCO agrees with the Staff's characterization there of the correspondence at issue.
5/
The motion typically asks for the striking of entire find-l ings, when examination of the findings reveals that only one sentence, at most, in each finding is potentially affected by the correspondence at issue.
For the Board's convenience, copies of the pertinent LILCO initial and reply findings, with affected areas indicated, 'are attached.
1
_4 establish substantive propositions for which there is no other support in the record.
Rather, their purpose is to enable the Board to take official notice of the existence of confirmatory documents which reflect analyses not yet finally written up or commitments still under consideration at the time the issues were being litigated.s/
A.
The Motion is Untimely, Improper and Prejudicial
/
SC's motion was filed fully 50 days after LILCO's January 19, 1983 initial Proposed Findings of Fact, in which every document used by LILCO and now being complained of was either cited directly or presaged.
The facial untimeliness of the motion is exacerbated by the fact that SC filed its own pro-posed findings of fact in the meantime, on January 31, 1983, and in them failed to object to any of the items of correspon-dence now complained of.
That responsive filing, or a contemporaneous motion to strike, was the praper vehicle con-templated by the Commission's Rules of Practice, 10 CFR $
l l
2.754, for SC's objection; and had SC used it, LILCO would have been on timely notice of a disagreeme " over the use of a rela-l tively small class of documents and could either have provided i
p/
In the case of SC's objection to footnote 37 on page 195 of LILCO's Reply, the text of the Reply clearly demonstrates l
that qualification of the strengthened vacuum breaker disc was l
not a prerequisite for licensing the plant.
The note was in-l cluded simply to inform the Board and all parties that this ef-l fort was complete and that results would soon be submitted to the Staff.
l
further authentication of them in its reply findings or clarified their use.
By lying in wait until after LILCO had filed its reply findings (and a rather leisurely wait at thst, given that those findings were filed on February 22), SC has not only defaulted without any showing of good cause on its proper opportunity to object to the use of the documents, but has done so in a manner prejudicial to LILCO, since LILCO relied in its reply findings on the apparent absence of dissent over the use of documents referenced in its initial findings.
SC has waived its proper opportunity under the Rules of Practice to object to the use of any of these documents and has not attempted any showing of good cause therefor; its raising of this matter now is prejudicial to LILCO; and the motion should be denied on this ground.
B.
The Letters Are Not Given Improper Substantive Weight Examination of LILCO's use of the documents now complained of in its initia: and reply findings reveals clearly that in each case the document either contained a confirmatory analysis' of matters discussed on the record in hearings on the SRV or Mark II issues, or memorialized a licensing commitment stated at the hearing to be under consideration.
These documents are not cited for the truth of assertions not otherw'se supported i
in the record, but for their existence.
This is exactly the see 10 purpose for which official.-notice provisions exist, CFR S 2.743(i).
SC's motion suggests baldly, without reference
- to any examples,. that the information in them is not suitable for notice.
SC is clearly incorrect:
there is a substantial body of decisional law supporting the proposition that adminis-trative agencies can and should take notice of official corre-spondence with regulatees.7/
It is not necessary to evaluate, much less rely on, the substance of the SNRC letters now at issue to see from LILCO's initial and reply findings that the letters merely confirm the closing-out of confirmatory analyses promised, or confirm a commitment under consideration, when the issues were being tried.
Taking the SRV-related correspondence first, SNRC-812 merely transmits to the Staff confirmatory analyses of pipe
^
stresses in the alternate shutdown mode; the content of the preliminary analyses had already been accepted by the Staff and been the subject of litigation.
See LILCO Initial Finding H-23; Reply Findings Vol.
1, at 212, 220.~
SNRC-816 merely transmits to the Staff a commitment to implement a lowered MSIV setpoint at the first refueling outage -- a commitment for which credit was not taken in LILCO's analyses, and which had been under consideration by LILCO at the time of the hearings.
See LILCO Initial Finding H-37; Reply Findings Vol. I at 218.8/
7/
The U.S.
Court of Appeals for the D.C.
Circuit has held that a regulatory agency can and should take official notice of the reports filed with it by a regulated company.
Wisconsin v.
F.P.C.,
201 F.2d 183, 186 (1952), cert. denied 345 U.S. 934 (1953); see Market Street R.
Co. v.
Commissiener, 324 U.S.
- 548, 562 (1944); see Midwest Television, Inc. v.
F.C.C, 426 F.2d 1222, 1229 (1970).
8/
It is curious that SC objects to the Board's taking official notice of this commitment,,in view of its Proposed Finding 22:48, which proposed that this change be required at Shoreham as a license condition.
I I
. As to Mark II, SNRC-808 and -824 transmit to the Staff re-sponses to the so-called Humphrey concerns which confirm testi-mony of Staff witness Fields at the hearing, and in any event commit to a mode of operation conceded at the hearing to obvi-ate the Humphrey concerns until final resolution of them.
See LILCO Initial Finding G-14; Reply Findings Vol. I at 196-97 and Note 38.
SNRC-831 merely transmits to the Staff the results of a reanalysis of piping systems at three levels on the contain-ment -- a reanalysis which the Staff had clearly identified as confirmatory.
See LILCO Initial Finding G-10; Reply Findings Vol. 1 at 202-03.
The Board does not need to evaluate the substance of any analyses forwarded by these letters to give them their proper weight.
In both their uses, the letters are merely confirmatory:
analyses merely confirm matters already fully lit'igated on the record, and licensing commitments obviate what would have otherwise been areas of technical dispute on the l
record.
While no further authentication of them is needed, the af -
fidavit of Jeffrey L.
Smith attesting to the truth and correctness of the SNRC letters at issue is attached out of an abundance of caution.
The Board can permit the noticed use of i~
the SNRC letters at issue without necessarily receiving the af-I fidavit into evidence.
See Carolina Power & Light Company (Shearon Harris Nuclear Power Plant, Units 1, 2, 3 and 4),
LBP-78-2, 7 NRC 83, 84-88 (1978), and discussion in LILCO's Reply Findings Vol. I at 11-13.
l
. If SC thinks that the Board's taking notice of these SNRC letters is so fundamentally important to the issues in the SRV and Mark II contentions that the outcome of the issues would thereby be affected, and can show good cause for having failed to file a timely objection, it can, of course,~ seek to reopen the record, if it can meet the test for reopening the record, Vermont Yankee Nuclear Power Corp. (Vermont Yankee Nuclear Power Station), ALAB-138, 6 AEC 520, 523 (1973); Duke Power Company (William B. McGuire Nuclear Station, Units 1 and 2),
ALAB-669, 15 NRC 453, 465 (1982).
Even then, the Board has discretion in the procedures it will follow in reopening the record.
Carolina Power & Light Company, supra.
CONCLUSION For the reasons stated above, SC's Motion to Strike, dated March'8, 1983, should be denied.
Respectfully submitted,
.- w Donald P.
Irwin One of Counsel for Long Island Lighting Company Hunton & Williams P.
O. Box 1535 Richmond, Virginia 23212 DATED:
March 23, 1983 Attachments:
Excerpts from Findings Affidavit of Jeffrey L.
Smith (and attachments thereto)
LILCO, MSrch 23, 1983 i
CERTIFICATE OF SERVICE In the Matter of LONG ISLAND LIGHTING COMPANY (Shoreham Nuclear Power Station, Unit 1)
Docket No. 50-322 (OL)
I hereby certify that copies of LILCO's Response to Suffolk County Motion to Strike Portions of the NRC Staff's Proposed Opinion and Findings of Fact, LILCO's Proposed Opinion and Findings of Fact, and LILCO's Reply to the Proposed Opinion and Findings of Suffolk County and the Staff were served this date upon the following by first-class mail, postage prepaid.
Lawrence Brenner, Esq.
Secretary of the Commission Administrative Judge U.S.
Nuclear Regulatory Atomic Safety and Licensing Commission
. Board Panel Washington, D.C.
20555 U.S. Nuclear Regulatory Commi'ssion Atomic Safety and Licensing Washington, D.C.
20535 Appeal Board Panel U.S.
Nuclear Regulatory Dr. Peter A.
Morris Commission Administrative Judge Washington, D.C.
20555 Atomic Safety and Licensing Board Panel Atomic Safety and Licensing U.S.
Nuclear Regulatory Board Panel Commission U.S.
Nuclear Regulatory Washington, D.C.
20555 Commission Washington, D.C.
20555 Dr. James H.
Carpenter Administrative Judge Daniel F.
Brown, Esq.
Atomic Safety and Licensing Attorney Board Panel Atomic Safety and Licensing U.S.
Nuclear Regulatory Board Panel Commission U.S.
Nuclear Regulatory Washington, D.C.
20555 Commission Washington, D.C.
20555
~
. )
i Bernard M. Bordenick, Esq.
David J. Gilmartin, Esq.
David A. Repka, Esq.
Attn:
Patricia A. Dempsey, Esq.
U.S. Nuclear Regulatory County Attorney Commission Suffolk County Department of Law Washington, D.C.
20555 Veterans Memorial Highway Hauppauge, New York 11787 Herbert H. Brown, Esq.
Stephen B.
Latham, Esq.
Lawrence Coe Lanpher, Esq.
Twomey, Latham & Shea Karla J.
Letsche, Esq.
33 West Second Street Kirkpatrick, Lockhart, Bill, P. O. Box 393 Christopher & Phillips Riverhead, New York 11901 8th Floor 1900 M Street, N.W.
Ralph Shapiro, Esq.
Washington, D.C.
20036 Cammer and Shapiro, P.C.
9 East 40th Street Mr. Marc W. Goldsmith New York, New York 10016 Energy Research Group 4001 Totten Pond Road James Dougherty, Esq.
Waltham, Massachusetts 02154 3045 Porter Street Washington, D.C.
20008 MHB Technical Associates 1723 Hamilton Avenue Howard L. Blau Suite K 217 Newbridge Road San Jose, California 95125 Hicksville, New York 11801 Mr. Jay Dunkleberger Matthew J. Kelly, Esq.
New York State Energy Office State of New York Agency Building 2 Department.of Public Service Empire State Plaza Three Empire State Plaza Albany, New York 12223 Albany, New York 12223 Donald P.
Irwin Hunton & Williams 707 East Main Street P.O.
Box 1535 Richmond, Virginia 23212 DATED:
March 23, 1983
7 I
LILCO Finding G-10 i
G-10.
Subsequent to issuance of Revision 5 to the t
Design Assessment Report, the Staff also expressed an interest in learning more about the basis for the selection of the 30 piping subsystems that were analyzed to evaluate the effects of the NUREG-0808 loads on the response of the plant's reactor building.
Tr. 9886 (Terno); Suffolk County Ex. 45.
In partic-e ular, the Staff was concerned with three locations on the con-tainment wall.
LILCO has committed to do a 100% reevaluation of all large bore and certain small bore piping systems at these three elevations before fuel load.
Tr. 9888 (Termo).
The Staff regards this reevaluation as confirmatory only, since no piping system stresses or support loads have been shown to
~
i exceed the code allowables.
Tr. 9889 (Termo).
The Humphrey Concerns G-11.
In a letter to the Applicant dated July 8, 1982, the Staff identified 22 concerns raised by a Mr. John l
Humphrey regarding the adequacy of the design margins'of the Mark II containment system; these concerns were potentially applicable to Shoreham.
Suffolk County Ex. 44; Tr. 9849-50 (Fields).
G-12.
In July 1982, the NRC Staff made a' presentation to the ACRS on the Humphrey concerns for all BWRs, and the ACRS concluded that Humphrey's concerns did not appear significant.
Tr. 9855 (Fields).
-149-n-r
p
.._J LILCO Finding G-14 i
G-13.
Based on the ACRS's conclusion, the Staff indi-catam x at it will evaluate the Applicant's respon*os on a con-firmatory basis only.
Accordingly, the Staff did not see a need to delay issuance of an operating license to await the completion of the review.
Tr. 10005-06 (Eltawila).
G-14.
LILCO submitted its preliminary responses on 1
the Humphrey concerns to the Staff on August 25, 1982 and its final' responses in early December.
All but two of the issues were addressed in that report.
The two remaining responses that relate to the RER heat exchanger relief valve discharge lines will be submitted to the Staff in January 1983.
Those responses may involve a commitment by LILCO not to use the RHR steam condensing mode during normal plant operation until it can be demonstrated that the hydrodynamic loads resulting from operation of the RER heat exchanger in this mode are acceptable.
Steam Bypass Testing G-15.
Both preoperational and periodic tests are per-formed to detect any leak paths that could exist b' tween the e
drywell and the wetwell.
Tr. 9864 (Fields).
The leakage rate is then compared with the appropriate acceptance criterion to determine acceptability.
Eltawila et al., ff. Tr. 9741, at 9.
-150-6
w
h I
e LILCO Finding H-23 Specifically, the Staff wondered how lo' ads measured using a
" rams head" discharge pipe configuration at the test facility were translatable to Shoreham, which utilizes a " tee quencher" at the end of the discharge line.
SC Ex. 34, ff. Tr. 8312, at Question 1.
LILCO replied that the loads measured at the test facility would be larger than the loads on the valve internals at Shoreham, since (1) no dynamic mechanical load generated at the tee' quencher is transmitted to the SRV because there is at least one anchor point between the tee quencher and the valve, (2) the first length of piping downstream of the SRV in the test program was conservatively set at twice that at Shoreham to bound the dynamic mechanical load on the valve, and (3)~
backpressure loads were maximized at the test facility through the use of an orifice plate and conservative pipe lengths.
LILCO Response, ff. Tr. 8402, at 2-4.
The Staff reviewed this response and concluded that the test facility impuced greater loads on the valves than those expected at Shoreham.
Tr. 8407-0;8 (Wright).
H-23.
The Staff noted that the test facility did not utilize spring hangers as pipe supports, but that they are used in conjunction with snubbers and rigid suports at Shoreham.
The Staff asked LILCO to describe the supports used at Shoreham and to explain how the loads measured at the test facility may change given the different support configurations at Shoreham.
SC Ex. 34, ff. Tr. 8312, at Question 2.
LILCO responded that
-168-
o
/
4 the location of snubbers and rigid supports at Shoreham demonstrate that the generic test facility was prototypical, LILCO Response, ff. Tr. 8402, at 5.
In addition to the supports, each SRV discharge line has one or two spring hangers, all of which are located in the drywell.
Id.
Since the dynamic loads resulting from liquid discharge during the alternate shutdown cooling mode of operation are significantly
/
lower than those from high pressure steam discharge, sufficient margins should exist in the Shoreham piping system design to adequately offset the increased dead load on the spring hangers in an unpinned condition due to a water-filled condition.
Id.
6-7.
Nevertheless, LILCO committed to provide additional analyses on all pipes for the alternate shutdown mode, Tr. 8421 (Smith).
The Staff accepted LILCO's response subject to the confirmational analyses.
Tr. 8410 (Cherny).
The results of these analyses, submitted in SNRC-812 on December 15, 1982, demonstrate that Shoreham complies with the requirements of NUREG-0737, Item II.D.l.
H-24.
The Staff also questioned whether the flow conditions tested at the generic facility were similar to those anticipated at Shoreham.
SC Ex. 34, ff. Tr. 8312, at Question 4.
LILCO's response indicated that the fluid conditions of the test program matched those expected to occur at Shoreham when the plant was being operated in the alternate shutdown cooling mode.
As for other event sequences, LILCO first identified
-169-
~.
LILCO Finding H-37 events and displayed various of them, from which BWR owners could' select those best adapted to their circumstances.
Id.
However, most other possible modifications listed in Item II.K.3.16 either provided little relative benefit (5% or less) or tended to have adverse side-effects on plant safety.
Boseman et al.,
ff. Tr. 7959, at 11.
H-36.
LILCO evaluated the possible modifications
/
listed in Item II.K.3.16 and studied in the GE SRV Study, and then applied the results of that study to Shoreham.
Tr.
8617-22 (Smith).
LILCO utilized the two principal methods evaluated in the GE SRV Study for reducing SORV events at Shoreham:
(1) the use of the Target Rock two-stage SRV rather than the three-stage SRV used in the benchmark plant, and (2) use of Emergency Procedure Guidelines that permit manual imple-mentation of a low-low set relief.
The combination of the two
~
was shown to reduce the failure of SORV events by approximately an order of magnitude.
Boseman et al., ff. Tr. 7959., at 10-11; Board Findings E-32 to -35.
H-37.
In addition to these features, LILCO has imple-l mented SORV reduction measures not ' evaluated in the GE SRV Study, for which no specific credit has been claimed: a lowered e
recycling set point, worth about 1% improvement, Tr. 8655 (Hayes), and pneumatic control system modification, worth about 2% to 3%, Tr. 8656-57 (Hayes, Smith).
LILCO witnesses also
-177-O
?~
k
-l.
- j indicated that they were continuing to evaluate the advantages and disadvantages of yet further potential modifications, such II
- i as lowering of the MSIV closure setpoint.
Tr. 8629-31 (Smith, Hodges); Tr. 8672-73 (Smith).
LILCO indicated to the Staff recently that LILCO intended to implement the lowered MSIV J!
setpoint at the first refueling outage (Letter, Smith (LILCO)
- l to Denton (NRC), January 7, 1983 (SNRC-816)).
l e
B-38.
SC contended that the SORV frequency reduction factors asserted by LILCO were too large.
Bridenbaugh and Minor (Challenges), ff. Tr. 8709, at 5.
However, the basis for SC's assertion -- a report by Southwest Research Institute -- relies solely on Target Rock three-stage valves for its reliability analysis.
Tr. 8763 (Bridenbaugh); LILCO-Ex. 18, ff. Tr. 9299, at 20, A-9/A-10.
In addition, SC neglected a second 50% reduction factor attributable to reduction in spurious blowdown (see Board Finding H-34).
SC presented no reliability surveys on two-stage SRV performance.
l I
H-39.
The Staff's prefiled direct testimony agreed that LILCO had reduced the frequency of SRV challenges, contrary to the allegations of SC, by a combination of system modifications and procedural techniques.
Hodges, ff. Tr. 7966, at 3.
On cross-examination, Mr. Hodges confirmed that he had been the author of NUREG-0737, Item II.K.3.16 (Tr. 8022, 8488, 8491-92); that the notion there of " reduction in challenges" to
-178-
. - ~
~.
.-,o. ; -
s e-LILCO Reply at 196-197
}
LILCO submitted its responses to the Humphrey con in SNRC-808 (December 9, cerns 1982) and SNRC-824 (January 28, 1983).31/
These responses confirmed Staff witness Fields' views on these concerns, namely, that a vast majority of the concerns were not significant, and that the only area of potentiaf significance may involve the RHR heat exch anger when operated in the steam condensing mode.
See Tr. 9855 (Fields).
As to the RER heat exchanger, LILCO has reported th at its complex analysis of this matter is still in progress, and that, pending its completion, it has decided to proceed in th of the Grand Gulf Station, e manner see Tr. 9855 (Fields), committing not to use the RER heat exchanger in the steam c'ondensi ng mode 33/.
of LILCO's proposed opinion and findings of fact in t u m ssion ceeding.
The information contained in SNRC-824 updated th formation contained in LILCO Finding G-14 s pro-e in-lowing finding: requests that that proposed finding be replaced with thA e fol-G-14.
the Humphrey concerns to the Staff on August 2 and its final responses in early December 1982 1982 (SNRC-808) and late January 1983 (SNRC-824).
Those responses contain a commitment not to use the RER until it can be demonstrated that the hydrodyn loads resulting from ope' ration of the RHR heat i-changer in this mode are acceptable.
=
ex-
-196-I
\\
for all normal plant operations.
SNRC-P 4; see LILCO Finding G-14.
LILCO has made this commitment with the understanding that it may conduct an engineering effort in the future to dem-onstrate the ability of the RER system to withstand loads while operating in the steam condensing mode.
Id.
Thus, since LILCO has implemented the Staff's recommended interim solution to the RHR problem, there is no basis in the record for SC's request for submittal of Humphrey responses, LILCO Finding 21:80, and approval of any further actions should rest with the Staff.
2.
Shoreham Confirmatory Analysis Subsection (d) of SC Contention 21 questions whether the Shoreham Mark II containment has been demonstrated to with-stand loads from simultaneous design basis transient and LOCA events.
LILCO and the Staff agree that LILCO has demonstrated the design adequacy of the Shoreham Containment.
LILCO Finding G-21; Staff Finding 21:23.
SC contends that the Staff's review of LILCO's assessment is not complete and that, therefore, the Staff's testimony should be given little weight.
SC Findings 21:24-36.
In particular, SC alleges that the Staff has failed to complete its review of LILCO's confirmatory piping analyses -
at three elevations on the containment wall and that LILCO has
-197-1
LILCO Reply at 202-203 k
4 Accordingly, LILCO undertook a program to evaluate the significance of the local exceedances.
Id.
This evaluation included structures, piping, and other components.
The results of the evaluation, documented in the DAR, indicated no exceedances of design stress allowables.
Tr. 9974-75 (Malovrh).
Thus, SC's request that the Board require LILCO to explain and justify the differences in the ARS, SC Finding 21:80, finds no support in the record.
c.
Further Piping Subsystem Analyses Finally, SC contends that the Staff believed LILCO's piping assessment was inadequate. and thus required LILCO to complete a 100% evaluation of piping subsystems at three loca-tions on the containment wall.
SC Findings 21:70-71.
SC further argues that completion of this evaluation is a prereq-uisite for resolving this contention.
SC Finding 21:72.
SC's argument misconstrues the testimony in this proceeding and finds no basis in the record.
LILCO committed to perform a 10d% reevaluation of all large bore and certain small bore piping systems in order to allay Staff concerns about whether earlier piping analyses were representative.
LILCO Finding G-10; Tr. 9887-89 (Termo, Malovrh).
The Staff regarded the f
-202-
)
3 u.
reevaluation as confirmatory, since no piping system stresses or support loads have been shown to exceed ASME code allow-ables.
Id.
The confirmatory analysis undertaken by LILCO is sufficiently circumscribed to be left to Staff review.39/
3.
Steam Bypass Testing The focus of subsection (c) of SC Contention 21 is on the ability of LILCO's test procedures to demonstrate an ac-ceptable steam bypass rate of the drywell floor seal and downcomer vacuum breakers.
SC has proposed no findings that suggest that LILCO's test procedures will fail to demonstrate
.the adequacy of the leakage rate.
Instead, SC has reached be-yond the scope of the contention, which addresses only the
~!
testing procedures and not the test results produced from the
. application of those procedures, to suggest that the high pres-sure test results may not be valid since the Staff did not ver-ify the validity of LILCO's testing.
SC Finding 21:39.
Even if the Staff's review were within the scope of the contention, SC's allegation has no factual basis.
As Staff witness Fields i
39/
The reevaluation, which has now been completed and will soon be forwarded to the Staff, confirms that no code allow-ables have been exceeded.
-203-l
LILCO Reply at 212 for LILCO's conclusions.
See LILCo. Findings H-15,
-16, and -22 to -25.
SC did not quarrel at the time with the appropriateness of the meeting, and in fact attended it.
Tr.
8399-8402.
LILCO's responses to the six questions were sub-4 mitted the following morning, LILCO Findi,ng H-4, and were
[
reviewed by the Staff for approximately the same period of time normally spent on similar issues.
Tr. 8608 (Wright, Hodges, Cherny).
LILCO and Staff witnesses were then subjected to ex-tensive questioning by the Board and SC on this materi'al, including questions regarding the Staff's reason for posing each question, t'he form of response expected by the Staff, and the sufficiency of LILCO's response.
See Tr. 8404-42, 8557-8611.
Based on all these procedures the Staff concluded that l
LILCO had complied with the requirements of NUREG-0737, Item II.D.1.
LILCO Findings H-15, -16, and -22 to -25; Staff Findings 22/28:19-27.
Accordingly, the Staff's review was nei-ther incomplete nor imprecise despite SC's allegations (SC Proposed Finding 22:18).
The confirmatory piping analysis undertaken by LILCO at that time is sufficiently circumscribed to be left to the Staff's review (SC Proposed Finding 22:19 to the contrary notwithstanding) and has already been submitted to the Staff in SNRC-812 dated December 15, 1982.
LILCO Finding H-23.
-212-
LILCO Reply at 218 II.K.3.16 -- namely, to reduce SORV events.
As the Staff witnesses notad, Item II.K.3.16 was designed to create a meaningful goal:
a goal not defined simply by numbers, but by a desire to have all reasonable modifications identified and implemented.
Staff Finding 22/28:37.
Judged against this standard, LILCO complied with the requirements of Item II.K<3.16.
LILCO Findings H-39 and -40; Staff Finding 22/28:40.
In an effort to assure that all reasonable modifica-tions have been implemented, LILCO has continued to assess the efficacy and practicality of changing the water level set point for MSIV closure.
See Tr. 8628-29 (Smith).
As noted in LILCO's proposed findings, that review is now complete and LILCO~has committed to implement this modification at the first refueling outage.
LILCO Finding H-37; Staff Finding 22/28:38.
Thus, to the extent SC's proposed findings suggest that the change in the MSIV closure set point is an idea that LILCO is not pursuing, and thus, that all reasonable modifications are not being implemented, see SC Findings 22/28:42 and :47, they are simply inaccurate.
SC also appears to disagree with LILCO't and the Staff's description of the performance of 2-stage Target Rock
-218-4 e
LILCO Raply at 220 REPLY TO SC SRV FINDINGS While LILCO's principal reply on SC Contention 22 and SC Contention 28(a)(vi)/ SOC Contention 7A(6) addresses LILCO's primary concerns with SC's proposed findings, it does not at-tempt to provide a findir.g-by-finding refutation of SC's pro-posed findings.
Nor is that purpose of this section, which simply addresses a small group of findings, no.t specifically addressed in the principal LILCO reply, but which neverthelers do not accurately reflect the record.
This appendix is not meant to be all-inclusive, and the fact that any given finding is not addressed here does not imply LILCO's agreement with it.
22:16.
LILCO did not " acknowledge the need to perform stress analyses."
Instead, LILCO committed to perform confirmatory analyses, even though it felt there was sufficient information to support its conclusions.
L'ILCO Finding H-23.
The results of those analyses, submitted in SNRC-812, support LILCO's earlier conclusion.
Id.
22:17.
The majority of this finding ignores the fact that the amount of time the Staff spent reviewing these re-sponses was approximately the same a,s that spent actually re-viewing similar submittals.
See page 212 above.
-220-
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of
)
)
LONG ISLAND LIGHTING COMPANY
)
Docket No. 50-322 (OL)
)
(Shoreham Nuclear Power Station, )
Unit 1)
)
AFFIDAVIT OF JEFFREY L.
SMITH Jeffrey L. Smith, being duly sworn, depohies and says as follows:
l.
My name is Jeffrey L. Smith, Manager of Special l
Projects for S,horeham Nuclear Power Station, Unit 1.
SNRC Letters 808, 812, 816, 824, and 831 (copies of which are attached) were prepared under my supervision and d_irection.
2.
I hereby solemnly swear and affirm that the contents of the SNRC letters referred to in paragraph one (1) above are true and correct to the best of my k_nowledge and belief.
/'1L_
/Dffy L.' Smith STATE OF NEW YORK, COUNTY OF SUFFOLK b
Subscribed to and sworn before me this oY day of March, 1983.
\\
NANCY ). SCHMITT NOTARY PUBUC, State of New York Notary Public k \\
No. 52-8826330. Suffolk County, My Commission Expires:
'hO g Q
. Term Egires March 30. LQ{
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SNRC-808 ggg LONG ISLAND LIGHTING COM PANY
, h.*ctm.WW St10REHAM NUCLEAR POWER STATION
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- 5-P.O. 80% 618, NORTH COUNTRY ROAD e WADING RIVEA, N.Y.11792 December 9, 1982 SNRC-808 Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation l
U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Concerns Regarding the Adequacy of the Design ]largins of the Mark I and II Containment Systems.
Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322 Reference (1):
Robert L.
Tedesco letter to M. S. Pollock dated July 8, 1982
Dear Mr. Denton:
Reference (1) requested that the Long Island Lighting Company provide a proposed program to respond to the subject concerns that were identified as being potentially applicable to Shoreham.
In response to this request, enclosed please find forty (40) copies of a report entitled " Concerns Regarding the Adequacy of the Design Margins of the Mark II Containment Systems at Shoreham Nuclear l
Power Station."
This report addresses all items in reference (1) with the exception of items 3.3 and 3.4.
A submittal will be made on these items by mid-January, 1983.
I l
should you have any questions, please contact this office.
Very truly yours, 1
l J. L.
Smith Manager, Special Projects Shoreham Nuclear Power Station RWG:jm Enclosure c.c.:
J. Higgins All Parties 1
1
(
SNRC-812 j
LONG ISLAND LIGHTING COM PANY n
- . 4,J.em.LCO,f 1nf SHOREHAM NUCLEAR POWER STATION aujegm bisar4+u - _ r M P.O. BOX 618. NORTH COUNTRY ROAD e WADING RIVER. N.Y.11792 Direct Dial Number December 15, 1982 SNRC-812 Mr. Harold R. Denton, Director l
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 SER Issue II.D.1 - SRV Test Program Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Reference:
(1)
Letter NRC (Mr. A. Schwencer) to LILCO (Mr.
M.
S.
Pollock) dated July 8, 1982 (2)
Letter from BWR Owners' Group (T. Dente) to NRC (D. G. Eisenhut) dated 9/25/81
Dear Mr. Denton:
The reference (1) letter forwarded a request for additional information consisting of six (6) questions on the Safety Relief Valve (SRV)
Operability Test Program Results, and their applicability to the Shoreham SRVs.
The SRV test results have been documented in report NEDE-24988-P, Analysis of Generic BWR Safety / Relief Valve Operability Test Results which was forwarded l
to the staff via the Reference 2 letter.
A response was provided for each of these six questions in a submittal filed with the ASLB on July 29, 1982 (refer to ).
It was determined, however, that the staff required supplemental information for question 2,
involving performance of a stress analysis for each SRV discharge line, and question 4, involving a description of the events and anticipated conditions at Shoreham for which the valves are required to l
operate and a comparison of these plant conditions to the conditions in the test program.
As you are aware, question 2 addressed the issue of how the Shoreham unique SRV discharge piping supports may affect conclusions regarding SRV operability derived from the generic test facility.
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SNRC-812 December 15, 1982 Page The generic test facility was designed to be prototypical of BUR plants in terms of discharge piping configuration.
It was concluded in the generic test program that the fluid transient line forces resulting.from the alternate shutdown cooling mode liquid discharge are of substantially lower magnitude than those resulting from the design basis high pressure steam discharge events.
On this basis it could be concluded that rigid pipe supports and snubbers which would carry the direct fluid transient loads are adequate for the liquid discharge event, since they have been designed for the.more severe steam discharge events., Since spring hangers are affected most by static loads (deadweight and thermal loads), the effects of the added weight of water in the lines need to be evaluated for such' supports.
In order to fully evaluate the adequacy of Shoreham's SRV discharge piping supports to assure no' potentially adverse effects on SRV operability, fluid flow transient analyses as well as pipe stress and support analyses have been performed for the alternate shutdown co aling mode liquid discharge in Shoreham.
The dynamic fluid forces calculated for each of the eleven discharge lines exhibited the same general characteristics as observed in the test facility; particularly, magnitudes of forces were found to be lower than those resulting from'high pressure steam discharge by ratios similar to those found'in the test.
The eleven lines in the dry well (each of which has one or two spring hangers, as well as rigid supports and snubbers) were then analyzed using standard techniques to determine the effects of the dynamic fluid forces.
These lines were also analyzed to determine pipe stresses and support loads due to the deadweight of the water in the lines, the concurrent thermal ef,fects, and also for the effects of an assumed concurrent safe-shutdown earthquake.
All piping stresses calculated for the combination of loads i
described above were found to be well within ASME Faulted condition allowables.
Each pipe support was also found to be within design allowables for the same combination of loads.
It is noted that the spring hanger supports, which were potentially of.most concern, had been designed to carry the full weight of water associated with the hydro test condition (during which time the springs are pinned).
For the condition discussed herein, the spring hanger travel distances were also checked and found to be within the working' range of the springs, assuring that they will not bottom out during this event.
None of the eleven Shoreham discharge lines have any spring hangers in the wet well.
Additionally, since the lines are l
anchored at the dry well floor, loads imposed on the lines in y-
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SNRC-812 December 15, 1982 o
Page this area are in no way' transmitted to the SRVs.
However, the wet well line judged to be most likely to be strongly affected (based on support locations) was also analyzed for the same loading conditions.
Again, all pipe stresses and support loads were well within Besign allowables.
Even though it is clear that on this basis, there is no outstanding concern in this area, for completeness the remaining ten lines in the wet well are also being. analyzed for these conditions.
This final verification analysis will be completed prior to fuel load.
Based on the detailed analytical evaluation described above, it is concluded that the Shoreham SRV discharge piping is adequately supported to sustain the effects of a low pressure liquid discharge.
Since all pipe supports are adequate and all pipe stresses are within allowable levels, the loads on the valves
~
will not adversely affect operability of'the Shoreham SRVs.
With regard to question 4, an amplified response, completely responsive to the NRC question, is included as Attachment 2 to this letter.
The information contained herein should be sufficient to allow the staff to completely close this item on the Shoreham docket.
Should you have any questions, please contact this office.
Very truly yours, l
D.'.:.'nel 6 :-~.-;f 1-:v
- f. L. 'Sinith l
Manager, Special Projects j
Shoreham Nuclear Power Station RWG/ law Enclotures cc:
J.
Higgins All Parties I
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(
LILCL, July 29, 190'2 Atta'chment.1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of
)
[ --.
)
LONG ISLAND LIGHTING COMPANY
)
Docket No. 50-322 (OL)
)
(Shorcham Nuclear Power Station,
)
Unit 1)
)
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RESPONSE OF LONG ISLAND LIGHTING
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COMPANY TO NRC REGULATORY STAFF QUESTIONS OF JULY 8, 1982 RELATIVE TO SRV TESTIN
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Long Islan'd Lighting Compan) has received a letter, dated July 8, 1982, from the NRC Regulatory Staff involving six ques-(
tions relating to the testing of Safety Relief Valves for the Shoreham Station.
The covering letter, as amplified by the oral testimony of Regulatory Staff witnesses, indicated that the Staff felt that it needed c. ore information in the area described in the six questions attached to the letter in order to complete its review of SRV testing for Shoreham.
The[ollowingsubmittal
~.
contains LILCO's response to the six questionq.
~
Rispectfully Submitted -
LONG ISLAND ~ LIGHTING COMPANY
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Donald P.
Irwin Hunton & Williams Post Of"ico Box 1535 707 East Main Street Richmond, Virg' inia 23219 3
DATED:
July 29, 1982 e
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Q.
The test program utilized a " rams head" discharge pipe configuration.
Shoreham utilizes a " tee" quencher configuration at the end of the discharge line.
Des-cribe the discharge pipe configuration used at Shoreham and compare the anticipated loads on valve internals in the Shoreham configuration to the measured loads in the test program.
Discuss the impact of any diff-erances in loads on valve, operability.
A.
The safety / relief valve discharge piping configuration at Shoreham utilizes a " tee" quencher at the discharge pipe exit.
The average length of the 11 SRV dis-
', charge lines (SRVDL) is 137' and the submergence length in the suppression pool is approximately 13'.
The SRV test program utilized a.ramshead at the dis-charge. pipe exit, a pipe length of 112' and a sub-s sergence length of approximately 13'.
Loads on valve
')
internals during the test program are larger than loads on~ valve internals in the Shoreham configuration for the following reasons:
1.
No dynamic mechanical load originating at the " tee" quencher is transmitted to the valve in the Shoreham configuration because there is at least one anchor point between the valve and the tee quencher.
2.
The first length of the segment of piping downstream of the SRV in the. test facility was twice that of l
Shoreham piping,' thereby resulting in a bounding l
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dynamic mechanical load on the valve in the test l
program.
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3.
Dynamic hydraulic loads (backpressure) are experienced N'
by the valve internals in the Shoreham configuration.
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P The backpressure loads may'be either (i) transient backpressures oc,currfng during valve actuation, or
'(ii) steady-state backpressures occurring during steady-state flow following valve actuation.
(a)
The key parameters affecting the transient back-pressures are the fluid inertia in the submerged l
SRVDL and the SRVDL air volume.
Transient backpressure increases with line submergence and decreases with air volume.
The transient backpressure in the test program was maximized by utilizing a submergence of 13', not less than Shoreham, and a pipe length of 112' which is less i
)
than Shoreham.
(b)
The steady-state backpressure in the test program was maximized by utilizing an orifice plate in the S.RVDL above the water level and before the ramshead.
The orifice was sized to produce a backpressure greater than that calculated for,
any of the Shoreham SRVDL's.
e The differences in the line configuration between the Shoreham plant and the test program as discussed above result in the loads on the valve internals for the test facility which bound the actual shoreham loads.
An addi-tional consideration in the selection of the ramshead for the test facility was to allow more direct measure-x O
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4-F ment of the thrust load in the final pipe segment.
Utilization of a " tee" quencher in the test program would have required quencher supports that would unnecessarily obscure securate measurement of the pipe' thrust loads.
For the reasons. stated above, differences between the SRVDL configurations in Shoreham and the test facility will not have any adverse effect on SRV operability at shoreham relative to the test facility.
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The test configtration utilized no spring hangers as P pe supports.
Plant specific configurations do use i
, spring hangers in conjunction with snubber and rigid supports.
Describe the. safety relief valve pipe, sup-ports used at Shoreham and compare the anticipated loads on valve internals fortthe Shoreham pipe supports to the measured loads in the test program.
Describe the impact of any differences in loads on valve operability.
A.
The Shoreham safety-relief valve discharge lines (SRVDL's) are supported by a combination of snubbers, rigid sup-Ports, and spring hangers.
The locations of snubbers
' and rigid supports at Shoreham are such that the loca-tion of such supports in the BWR generic test facility is prototypical, i.e., in each case (Shoreham and the test facility) there are supports near en=h change-of direction in the pipe routing.
Additionally, each SRVDL at Shoreham has only one or two spring hangers, all of which are located in the drywell.
The spring hangers, snubbers, and rigid supports were designed to accommodate combinations of loads resulting from piping dead idcight, thermal conditions, seismic and suppression pool hydrodynamic events, and a high pressure steam discharge transient.
The dynamic load effects on the piping and supports of the test facility due to the water discharge event (the alternate shutdown cooling mode) were found to be sig-nificantly lower than corresponding loads resulting from the high pressure steam discharge event.
As stated in
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NEDE-24988-P, this finding is considered generic to all 1
BWR's sin.ca the test facility was designed to be proto-typical of the features pertinent to this saue.
Fur-thermore, analysis of a typical shoreham SRVDL con-figuration has confirmed the applicability of this conclusion to Shoreham.
During the water discharge transient there will be sig-
'nificantly lower dynamic loads acting on the snubbers and rigid supports than during the steam discharge transient. ' This will more than offset the small increase in the dead load on these supports due to the weight l
l of the water.
Therefore, design adequacy of the snub-bers and rigid supports is assured as they are designed for the larger steam discharge transient loads.
This question ' addresses the design adequacy of the spring hangers with respect to the increased dead load due to the weight of the water during the liquid discharge b
tr ansient.
As was discussed with respect to snubbers and rigid supports, the dynamic loads resulting from liquid discharge during the alternate shutdown cooling mode of operation are significantly lower than those from l
the high pressure steam discharge.
It is believed tha't sufficient margin exists in the shoreham piping system design to adequately affset the increased dead 8
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1
- 7 load on the spring hangers in an unpinned condition due to a water filled condition. Nevertheless, stress analyses are being performed to confirm this assumption regarding the increased deadweight loads for all SRVDL spring hangers.
It should be noted that the effect of dead load weight does not affect the ability of SRVs to open to establish the alternate shutdown cooling path since the loads occur only after valve opening.
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Report NEDE-24988-P did not identify any valve functional 2
deficiencies or anomalies encountered during the test progran.
Describe the impact on valve safety function of any valve functional deficiencies or anomalies en-countered during the program.
A.
No functional deficiencies or anomalies of the safety relief or relief valves, not only for Target Rock two-stage valves but also for all other types of valves tested, were experienced during the testing by Wyle Laboratories for compliance with the alternate shutdown i
cooling mode requirement.
All the valves. subjected to.
1 test runs, valid and invalid, opened and closed without loss o,f pressure integrity or damage.
Anomalies en-countered during the test program were all due to failures'of test facility instrumentation, equipment, data acquisition equipment, or deviation from the approved test procedure.
The test specification for each valve required'six valid runs.
Under the test procedure, any ancmaly caused the test run to be judged inva?,id.
In testing for the Target Rock two-stage SRV, only one anomaly of any sort occurred:
on water test run,No. 302, the test system GN2 regulator failed, resulting in a test which did not comply with the procedural test require-ments.
The Wyle Laboratories test log sheet for the Target Rock two-stage valve tests is attached.
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No anomily encountered during the required test program affects any valve safety or operability function,..,,,.
All valid test runs are identified in Table 2.2-1 of e NEDE-24986-P.
The data presented in Table 4.2-1 for each valve were obtained from the Table 2'.'2-1 test runs and were based upon the selection criteria oft (a) Pr'esenting the maximum representative loading in-formation obtained from the steam run data, 1
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(b) Presenting the maximum representative water loading information obtained from the 15C F subecoled we.ter test data, (c) Presenting' the data on the only test run performed for the 500 F subcooled water test condition.
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OPERABILITY TEST REPORT P
FOR k
TARGET ROCK 6X10 SRV 2
FOR LOW PRESSURE WATER TESTS FOR
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i GENERAL ELECTRIC COMPANY C. ,
.i GENERAL @ ELECTRIC NUCLEAR ENERGY BLSNESSGRO(p 0.0 $ H 8-/e1R2 WI NED CATE 3/~P2 9 % I VFF No, llD A 7O
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TaAusmrrituo.
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175 Curtner Avenue San Jose, California s
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TABLE I TEST LOG FOR SRV TR-1 Test Test Load Line Test No.
Media Configuration Date Remarks 301 Steam I
3/17/81 Acceptable 302 Water 1
3/17/81 GN, Regulator failed.
Data not acceptable.
303 Water i
3/17/81 Acceptable 304 Steam I
3/17/81 Acceptable 305 Water i
3/18/81 Acceptable 306 Steam I
3/18/81 Acceptable 307 Water 1
3/18/81 Acceptable 308 Water 1
3/18/81 Special test at elevated
' C, ~'g temperature and low pres-sure requested by G.E.
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Q.
The purpose of the test program was to determine valve i
performance under conditions anticipated to be encountered i
I in the plants.
Describe the events and anticipated con-ditions at Shoreham for which the valves are required to operate and compare these plant conditions to the condi-tions in the test program.
Describe the plant features assumed in the event evaluations used to scope the test program and compare them to plant features at Shoreham.
For example, describe high level trips to prevent water from entering the steam lines under high pressure operating conditions as assumed in the~ test event and compare them to trips used at Shoreham.
A.
The purpose of the test program was to determine valve
, performance, under conditions anticipated to be.en-countered in the plants, which could result in liquid or two phase flow through the va.lves.
The alternate shutdown cooling mode is the only anticipated event which is expected to result in liquid at the valve inlet.
)
Consequently, this was the event simulated in the SRV test program.
This conclusion and the test results applicable to Shoreham are discussed below.
The alter-nate shutdown cooling mode has been described in the response to NRC question 5.
The SRV inlet fluid conditions tasted in the BWR Owners' Group SRV test program, as documented in NEDE-24988-P, are representative of the fluid conditions expected to occur in the alternate shutdown cooling mode of. opera-tion at Shoreham.
These fluid conditions at the SRV 0
l inlet are 15 F to 500 F subcooled liquid at 20 psig to 250 psig.
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- 11 The BWR Owners' Group, in their enclosure to the
, September 17, 1980 letter from D. B. Waters to R. H.
Vollmer, identified thirteen events which could result in liquid or two phase SRV inlet flow.
These events were identified by evaluating the initiating events described in Reg. Guide 1.70, Rev. 2, with and without the additional conservatism of a single active component
, failure or operator error postulated with the event sequence.
Of these thirteen events, only.eight are applicable to the Shoreham plant because of its design
.and specific plant configuration.
For these eight events, the Shoreham specific features, such as trip logic,
(
power supplies, instrument line configuration, alarms and operator actions, have been compared to the base case analysis presented in the BWR' Owners' Group September 17, 1980 submittal and subsequent discussions with the NRC Staff.
This comparison has demonstrated that in each case, the base case analysis is applicable t'o Shoreham in that the base case assumptions are applicable.
For example, the base case analysis for the reactor level 8 failure /HPCI overfill event included a level 8 trip schtme with two out of two logic, two variable in-strument legs and one power supply inputting to one HPCI 1
turbine trip mechanism with one turbine stop valve.
This
.(
scheme is the same as the Shoreham design.
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,1 As discussed above, the Shoreham plant features are represented in the base case analysis performed in the BWR Owners' Group evaluation.
This evaluation concluded that the alternate shutdown cooling mode i's the only expected operating event involving liquid or two phase flow and therefore requires testing.
The alter-nate shutdown cooling mode fluid conditions tested in the BWR Owners' Group test program accurately bound
' the Shoreham plant specific fluid conditions expected for this event.
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Q.
The valves are likely to be ext'ensively cycled in a con-
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trolled depressurization mode in a plant specific appli-cation.
Was this mode simulated in the test program?
What is the effect of this valve cycling on valve per-formance and probability of the valve to fail open or to fail closed?
A.
The BWR safety / relief valve (SRV) operability test program was designed to simulate the alternate shutdown cooling mode, which is the only expected liquid dis-charge event for Shoreham.
The sequence of events
' leading to the alternate shutdown cooling mode is given below.
Following normal reactor shutdown, the reactor operator depressurizes the reactor vessel by opening the
()
turbine bypass valves and removing heat through the main condenser.
If the main condenser is unavailable, f
the operator could depressurize the reactor vessel by using the SRV's to discharge steam to the suppression pool.
If SRV operation is required, the operator cycles the valves in order to assure that the cooldown rate is maintained within I
the technical specification limit of 1000 F per hour.
This would require on the order of 1-10 cycles of the SRV.
When the vesse'l is depressurized, the operator
~
initiates. normal shutdown cooling by use of the RHR system.
If that system is unavailable because the valve on the RHR shutdown cooling suction line fails to open, the operator initiates the alternate shutdown cooling mode.
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j For alternate shutdown cooling, the operator opens one SRV and initiates either an RHR or core spray pump utiliz-ing the suppression pool as the suction source.
The re-actor vessel is filled such that water ~is' allowed to flow into the main steam lines and out of the SRV and back to the suppression pool.
Cooling of the sys-tem is provided by use of an RHR heat exchanger.
As a
, result, an alternate cooling mode is maintained.
In order to assure continuous long term heat removal, the SRV is,kept open and no cycling of the valve is performed.
In order to control the reactor vessel cooldown rate, the operator is instructed to throttle l
ms the injection valve into the vessel.
Consequently, no cycling of the SRV is required for the alternate shutdown cooling mode, and no cycling of the SRV was
~
performed for 'the generic BWR SRV operability test program.
The ability of the Shoreham SRV to be extensively cycled for steam discharge conditions has been confirmed during steam discharge qualification testing of the valve by the valve vendor.
This qualification testing for the Target Rock two-stage valve used in Shoreham l
has been previously identified in the Shoreham response to NRC question 212.51.
Based on the qualification t
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T testing of the SRV's, the cycling of the valves in a controlled depressurization mode for g Nam discharge conditions will not adversely affect u hig. performance and the probability of the valve to fail open or closed is extremely low.
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D scriba how tha valuco of valva C 'o in rcpart
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NEDE-24988-P will be used at Shoreham.
Show that i
the methodology used in the test program to determine the valve C will be consistent with the application y
at Shoreham.
The flow coefficient, C, for the Target, Rock 6 x 10 A.
y two-stage pilot operated safety relief valve (SRV) utilized in Shoreham was determined in the generic SRV test program (NEDE-24988-P).
The average flow coefficient calculated from the test results for the
, Target Rock two-stage vaive, model 7567F, is reported in Table 5.2-1 of NEDE-24988-P.
This test value has been used by LILCO to confirm that the liquid discharge
. flow capacity of the Shoreham SRVs will be sufficient to remove core decay heat when injecting into the
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reactor pressure vessel (RPV) in the alternate shutdown cooling mode.
The Cy value determined in the SRV test l
l demonstrates that the Shoreham SRVs are capable of return-(
ing.the flow injected by the RER or CS pump to the suppression pool.
If the operator were to place the Shoreham plant in the alternate shutdown cooling mode, he would assure that adequate core cooling was being provided by monitoring the following parameters:
RER or CS flow rate, reactor vessel pressure and reactor vessel tc=perature.
The flow coefficient for the Target Rock two-stage valve reported in NEDE-24988-? was determined from the SRV s
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flow rate when the valve inlet was pressurized to
'ap roximately 250 psig The valve flow rate was measured p
with the supply line flow venturi upstream of the steam chest.
The Cy for the valve was calculated using the nominal measured pressure differential between the valve inlet (steam chest) and 3' downstream of the valve and the corresponding measured flowrate.
Furthermore, the test conditions and test configuration were repre-sentative of Shoreham plant conditions for the alternate shutdown cooling mode, e.o. pressure upstream of the valve,' fluid temperature, friction losses and liquid flowrate.
Therefore the reported Cy values are appropri-i~
ate for application to the Shoreham plant.
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Attcchment 2-NRC QUESTION '4 The purpose of the test program was to determine valve performance under conditions anticipated to be encountered in the plants.
Describe.the events and anticipated conditions at Shoreham for which the valves are required to operate and compare these plant conditions to the conditions in the test program.
Describe the plant features assumed in the event evaluations used to scope the test program and compare thcm to plant features at Shoreham.
For example, describe high level trips to prevent water from entering the steam lines under high pressure operating conditions as assumed in the test event and compare them to trips used at Shoreham.
RESPONSE TO NRC QUESTION 4 The purpose of the S/RV test program was to demonstrate that the Safety Relief Valve (S/RV) will open and reclose under all expected flow conditions.
The expected valve operating conditions were determined through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2.
Single failures were applied to these analyses so that the dynamic forces on the safety and relief valves would be maximized.
Test pressures were the highest predicted by conventional safety analysis procedures.
The BWR Owners Group, in their enclosure to the September 17, 1980 letter from D. B. Waters to R. H. Vollmer, identified 13 events which may result in liquid or two-phase S/RV inlet flow that would maximize the dynamic forces on the safety and relief valve.
These events were identified by evaluating the initiating events described in Regulatory Guide 1.70, Revision 2, with and j
without the additional conservatism of a single active component failure or operator error postulated in the event sequence.
It was concluded from this evaluation that the alternate shutdown cooling mode is the only expected event which will result in liquid at the valve inlet.
Consequently, this was the event simulated in the S/RV test program.
This conclusion and the test results applicable to Shoreham are discussed below.
The alternate shutdown cooling mode of operation has been described j
in the response to NRC Question 5.
l The S/RV inlet fluid conditions tested in the BWR Owners Group l
S/RV test program, as documented in NEDE-24988-P, are 15*F to 50*F subcooled liquid at 20 psig to 250 psig.
These fluid conditions envelope the conditions expected to occur at Shoreham in the alternate shutdown cooling mode of operation.
The BWR Owners Group identified 13 events by evaluating the initiating events described in Regulatory Guide 1.70, Revision 2, with and without the additional conservatism of a single active component failure or operator error postulated in the events sequence.
These events and the plant-specific features that mitigate these events, are summarized in Table 1.
Of these 13 events, only 8 are applicable to the Shoreham plant because of i
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its design and specific plant configur& tion.
Five events, namely 2, 5, 6, 10, and 13 are not applicable'to the Shoreham plant for the reasons listed below:
a.
Event 2 will only result in steam conditions at the S/RV inlet because the Shoreham plant.has safety relief valves located higher than the MSIVs.
b.
Events 5 and 10 require initiation of the HPCS system.
This system is not present in the Shoreham design, c.
Event 6 requires initiation of the RCIC system with head sprays.
The Shoreham plant desigm dog; not include head sprays.
d.
. Event 13 is not applicable for the Shoreham plant because t)ere are no procedures or specific design features that lead to break isolation in the event of a large breck accident.
For these 8 remaining events, the Shoreham specific features, such as trip logic, power supplies, instrument line configuration, alarms and operator actions, have been compared to the base case analysis presented in the BWR Owners Group submittal of September 17, 1980.
This comparison has demonstrated that in each case, the base case analysis is applicable to Shoreham because the base case analysis does not include any plant features which are not already present in the Shoreham design.
For events, 1, 3, 4, 8, 9, 11, and 12; Table 1 demonstrates that the Shoreham specific features are included in the base case analyses presented in the BWR Owners Group submittal of September 17, 1980.
It is seen from Table 1, that all plant features assumed in the event evaluation are also existing features in the Shoreham plant.
For example, the base case analysis for Event 3, the reactor Level 8 failure /HPCI overfill event, included a Level 8 trip scheme with 2 out of 2 logic, 2 variable instrument legs and 1 power supply' inputting to 1 HPCI turbine trip mechanism with a turbine stop valve.
All features included in this base case analysis are similar to plant features in the Shoreham design.
Furthermore, the time available for operator action, is expected to be longer in the Shoreham plant than in the base case analysis for each case where operator action is required.
Event 7, the alternate shutdown cooling mode of operation, is the only expected event which will result in liquid or two-phase fluid at the S/RV inlet.
Consequently, this event was simulated in the BWR S/RV test program.
In Shoreham, this event involves flow of subcooled' water (approximately 20'F subcooled) at a pressure of approximately 50 psig.
The test conditions clearly envelope these plant conditions.
As discussed above, the BWR Owners Group evaluated transients including single active failures that would maximize the dynamic
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forces on the safety relief valves.
As a result of this evaluation, the alternate shutdown cooling mode is the only expected event involving liquid or two-phase flow.
Consequently this event was tested in the BWR S/RV test program.
The fluid conditions and flow conditions tested in the BWR Owners Group test program conservatively envelope the Shoreham plant-specific fluid conditions expected for the alternate shutdown cooling mode of operation, s
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SURC-816 LONG ISLAND LIGHTING COM PANY
- '/ SNCO d'mvamw SHOREHAM NUCLEAR POWER STATION s, s we.
_ u... :. sm.w P.O. SOE Sts. NORTH COUNTRY ROAD e WADING river. N.Y.11782 January 7, 1983 SNRC-816 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 SER Open Issue II.K.3.16 MSIV Setpoint Changes Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Reference:
(1)
LILCO letter, SNRC-563, (J. P. Novarro) to the NRC (H. R. Denton), dated 5/15/82
Dear Mr. Denton:
r As a member of the BWR Owners Group, LILCO participated in the evaluation of NUREG-0737, Item II.K.3.16 " Reduction of Challenges to the SRVs" (Reference 1).
The Shoreham Nuclear Power Station incorporated several of the recommended modifications to reduce the number of challenges to the safety / relief valves.
1 1
These modifications, in conjunction with the use of the two stage Target Rock SRVs, result in an estimated stuck open relief valve (SORV) frequency which is approximately an order of magnitude less than the BWR 4 benchmark plant.
The present Shoreham design there-fore meets the intent of NUREG-0737, II.K.3.16.
During the ASL3 hearing litigation on SC Contention 2Ba (vi), LILCO indicated that it was investigating the feasibility cf another modification which would further reduce the number of SRV challenges; i.e., changing the MSIV isolation setpoint from the present low reactor water level (L2) to the low-low level (L1).
This investi-gation, which is complete, has determined that it is feasible for Shoreham to modify the setpoint so that isolation occurs at level 1, without ir.pacting the safe operation of the unit.
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N Jcnuary 7, 1983 Mr. Denton Page 2 l
As previously stated, Shoreh'am has already incorporated several design changes which have reduced the SORV frequency by an order of magnitude and, as shown in reference 1, the incremental benefit of the proposed MSIV isolation setpoint chcnge will result in a minor improvement to the total SORV event frequency.
This small incremental benefit is due to the fact that the previous modifica-tions have already remedied common SORV initiators that were I
factored into the estimate of the benefits of the proposed mod-l ification taken singly.
The benefits are not additive; each l
subsequent modification results in a smaller incremental improve-ment.to the predicted SORV event frequency.
Not withstanding this small incremental reduction in total SORV event frequency resulting from the incorporation of this change, LILCO commits to its implementation.
The aforementioned investigation, determined that several actions are necessary to support the actual physical incorporation of the hardware to implement the MSIV logic setpoin,t revision.
Although the feasibility investigation determined that the FSAR Chapter 15 bounding analyses remain unaffected, two of the less
, critical transients will require revision.
Specifically, the loss of offsite power and the loss of.feedwater transients must be reanalyzed to reflect the lowered isolation setpoint.
.The engineering revisions' involve several organizations.
The NSSS vendor must revise all related documentation such as elementary and one line drawings, logic diagrams, and the master parts list and provide formal permission to revise the MSIV isolation logic setpoint.
Upon receipt of the formal approval and the technical details of the revision, LILCO's architect / engineer will revise the applic-able plant drawings and authorize the change.
The Start-Up or ~
the Plant Maintenance group'will then implement the actual hard-ware modification.
To support the modified MSIV isolation set-point logic, the Plant Staff-must revise the applicable station Procedures.
In addition, the operators must be trained using these procedures to ensure familiarity with the modified plant parameters resulting from the change.
In view of the efforts involved in this modification, the implementation.of the MSIV logic setpoint revision can not support the start-up testing schedule.
In summary, the Shoreham design already meets the intent of NUREG-0737.
Previous modifications have reduced the predicted SORV frequency by an order of magnitude vis-a-vis the benchmark BWR-4 plant.
The modificatica cf the isolation logic requires hardware
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.I January 7, 1982 Mr. Denton Page 3 modifications to systems that will not be an extended shutdown when access is permissible and the MSIVs are not required to perform a safety function.
In light of the above, and since the proposed MSIV isolation setpoint revision does not provide a significant contribution to the overall plant safety, LILCO intends to incorporate this plant modification during the first refueling outage.
In the course of the ASLB hearing litigation, LILCO committed to complete a feasibility evaluation for this modification.
LILCO believes the information stated herein and the commitment to modify the 15IV isolation setpoint logic adequately fulfills
,this commitment.
- - Very truly yours, e.-~
gnp @!"' p J. L. Smith Manager, Special Projects Shoreham Nuclear Power Station RJT/jm cc:
J.
Higgins All Parties l
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A SURC-824' peu - w.
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LONG ISLAN.D LIGHTIWG COPv1PANY-5 /' f.G C O
- C SHOREHAM NUCLEAR POWER'OTATION y+
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- WADING RIVER, N.Y.11792 m--.-
Direct Dial Number i,
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s' January 28, 1983 SNRC-824 Mr. Harold R. Denton, Director
^
Office of Nuclear Reactor Regulation U.S. Nuclear Regulat.ory Commission 1
Washington, D.C. 20555 x
x.
7, Humphrey Concerns 3.3 and 3.4 Shoreham Nuclear power Station - Unit 1 Docket No. 50-322
References:
(1)
Robert L. Tedesco letter to.[
M.S. Pollock, dated July 8, 1982 (2)
J.L.
Smith letter to 1
H.R. Denton, SNRC-808, dated December 9, 1982 s
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Dear Mr. Denton:
N Reference (1) requested that Long Island. Lighting Company pr6 vide-a proposed program to respond to the' subject concerns that were identified as being potentially applicable to Shoreham'.
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In Reference (2), LILCO transmitted J.ts responses to those items, with the exception of. items 3.3 and 3.4. -Re ferenA'a (2) committbd to a submittal in January of 1983 relative to these tuo items.
1 Since the time that Reference (1) was transmitted, continued analyti-cal investigation has been conducted by Stone & Webster Engineeiing Corporation into the submerged structure loads and piping / pipe support loads associated with RER heat exchanger relief valve dis-charge events during the Steam Condensing mode (SCM) of RHR system operation.
This analysis is complex d'ue to: a) the various poten--
tial modes of system failure, and b) the response characteristics of system components.
To date, we have been unable to demonstrate that the existing piping, supports, and structures can withstand the present conservatively predicted leadings which 6ay be imposed by this event.
.l FC.19 35.1 sw
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I SNRC-824 Jcnuary 28, 1983 Page 2 As a result, LILCO has elected to delete the SCM as an operating mcde of RHR, for all normal plant operations.
This would allow LILCO to utilize available engineering resources more efficiently for the purpose of completing construction and supporting the current schedule for fuel load.
Accordingly, a standing order will be issued to operating personnel prohibiting the use of SCM during normal plant operations.
Addi-tionally, manual icolation valves which isolate RHR system pressure control valves lEll*PCVOO7A, B,
leve] and pressure controllers lEll*
LTOO2A, B and lEll*PTOO3A, B respectively, and upstream manual isola-tion valves which isolate RHR system pressure control valves lEll*PCVOO3A, B will be maintained normally closed and administratively controlled to preclude inadvertent operation in the SCM.
The Station Test Pro-cedure'for SCM operation will not be conducted until an analytical justification has been developed.
A review of the Shoreham licensing application was conducted and it has been determined that there are no design basis transients or accidents which, for the purpose of mitigation, require the use of the SCM of the RHR system.
However, for non-design basis events involving multiple failures (e.g.
Station Blackout), the SCM flow path of the RHR system will be retained as an available option for use by operating personnel when all other means of core / containment cooling have been exhausted.
For such an x
event, reactor pressure would be less than the relief setpoint of 450 psig and the system would only be operated in a local manual mode.
I.t is our belief that, given the above commitments, this issue is removed as a licensing concern, with the understanding that LILCO na'y conduct an engineering effort in the future to demonstrate the
-m ability of the RHR system to withstand the potential loads while operating in the SCM.
1 If you should have any questions concerning this matter, feel free to contact this office.
Very truly yours, Orig nal slanan %
J. L.
Smith
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Manager, Special Projects Shoreham Nuclear Power Station DWD/jpb cc:
N11 Parties J. Higgins, Site NRC e
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SNRC-831 i
LONG ISLAN D LIG HTIN G UUlv1FAINY '
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SHOREHAM NUCLEAR POWER STATION raw P.O. BOX 618, NORTH C'OUNTRY ROAD e WADING RIVER. N.Y.11792 a.wus
-u Dhet Dial Number 1
February 18, 1983 SNRC-831 Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washin,gton, D.C.
20555 Mark II Hydrodynamic Loads Confirmatory Program Shoreham Nuclear Power Station - Unit 1 Docket No.
50-322 ReferGnce 1)
SSER for Seismic and Dynamic ~ Qualification of Safety Related Electrical And Mechanical Equipment, dated December 27, 1982 2)
Plant Design Assessment for SRV and LOCA Loads Revision 5, December 1981 Shoreham Nuclear Power Station - Unit 1
Dear Mr. Denton:
In response to reference 1, item (3),' LILCO hereby submits its report entitled, " Mark II Hydrodynamic Loads Confirmatory Program, Pipe Mounted Equipment Evaluation, Phase I."
This report presents the current qualification levels for all motor operated valves (MOV's), on the 30 piping subsysteis discussed in Reference 2, and the acceleration levels calculated for the Generic Long Term Program (LTP) confirmatory loads.
P As stated in this report, all MOV's on the 30 piping subsystems have been evaluated and found to be adequately designed to accomodate the final generic (LTP) hydrodynamic leads.
l
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FC.4335.1
s, Mr. Harold R.
Denton SNRC-831 Page 2 LILCO believes this information sufficient to constitute closure of Phase I, Pipe Mounted Equipment concerns.
Should you have any further questions regarding this matter, please feel free to contact this office.
Very truly yours, Original pgne8 by
~
J.L. Smifh Manager, Special Pro ects j
Shoreham Nuclear Power Station DWD:bc Attachment cc:
J. Higgins All Parties Listed in Attachment 1 l
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. as ATTACHMENT 1
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I,awrence Brenner, Esq.
.Berbert H. Brown, Esq.
Administrative Judge
..,.. Lawrence Coe Lanpher, Esq.
Atomic Safety and Licensing
~ Earla J. Letsche, Esq.
Board Panel
'Kirkpatrick, Lockhart, Hill' U.S. Nuclear Regulatory Commission Christopher & Phillips
- Washington, D.C.
20555 Sth Floor 1900 M Street, N.W.
- Washington, D.C. --20036 s
Dr. Peter A. Morris Administrative Judge Atomic Safety and Licensing Mr. Marc W. Goldsmith Board Panel Energy Research Group U.S. Nuclear Regulatory Co H ssion 4001 Totten Pond Road l
Washington, D.C.
'20555 Waltham, Massachusetts 02154
/
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... ~...... :-p:.;..:......i ', 2s, -
- ~- -
Dr. James R. Carpenter
- HEB Technical Associates Administrative Judge i;,;.1723 Hamilton Avenue Atomic Safety and Licensing Suite K x-Board Panel
. San Jose, California 95125 U.S. Nuclear Regulatory Commission Washington, D.C.
20555 3.~ p.,.-.nStephen B. Latham, Esq.
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- ' '. Twomey, Latham & Shea Daniel F. Brown, Esq.
~ 1.F 33 West Second Street Attorney P.O. Box 393 g
Atomic Safaty and Licensing-Board Panel
,.. Riverhead, New York 11901 U.S. Nuclear Regulatory Commission Washington, D.C.
20555 r.-
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....;. Ralph Shapiro, Esq.
~
.y.' Cammer and Shapiro, P.C.
". N-9 East 40th Street
- l-Bernard M. Bordenick, Esq.
Rew York, New York 10016 David A.
Repka, Esq.
~
U.S. Nuclear Regulatory Commission 3
.L, -
Washington, D.C.
20555
~
.." Matthew J. Kelly, Esq.
State of New York
- Department of Public Service Three Empire State Plaza
., Albany, New York 12223 s
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MARK II HYDRODYNAMIC LOADS C0hTIRMATORY PROGRAM PIPE MOUNTED EQUIPMEhTEVALUATION - PHASE I SHOREHAM NUCLEAR POWER STATION - UNIT 1 LONG-ISLAND LIGHTING COMPANY The objective of this report is to present additional information on Shoreham equipment evaluation results as a supplement to the Shoreham Design Assessment Report (DAR) Revision 5 (Reference 1) Appendix L " Mark II l
Hydrodynamic Loads Confirmatory Program."
As stated in the DAR, the Shoreham Mark II hydrodynamic loads confirmatory program has evaluated the plant with respect to the final generic Long-Term Program (LTP) hydrodynamic loads.
The LTP hydrodynamic loads, the scope I
and procedure of the confirmatory program, and the evaluation results have been discussed in Reference 1. The evaluation concluded that the Shoreham reactor building structures, piping, and equipment had been adequately designed to accommodate the final generic.LTP hydrodynamic loads, with the exception of a number of motor-operated valves (MOVs).
These MOVs are generally the same valvas that had acceleration values due _to the design bases hydrodynamic loads which exceeded original qualification levels.
i l
l Since the evaluation results beca:e available, analytical efforts as well as a requalification test program have been completed te demonstrate that the integrity and operability of the valves can be assured.
This report presents the current qualification levels for all MOVs on the 30 piping subsystems discussed in the DAR and the acceleration levels calculsted for the LTP confirmatory loads.
As indicated in Reference 1,
the Shoreham reactor building structural dynamic analysis results had clearly shown that the most significant final generic load is the CO-basic load.
The load definition is a direct application of the 4TCO test data on the Shoreham pool boundary with a conservative spatial distribution.
NUREG-0808 (Reference 2) has acknowledged the conservative nature of the load definition.and allows credit to be taken for the pool size effect and pool temperature range (Reference 3).
Shoreham has performed a plant unique assessment and concluded that a CD-basic load reduction ratio of 0.7 can be applied for i
the pool size effect.
Shoreham has conservatively elected not to take l
credit for the pool temperature range effect for results reported to date.
As it was discussed throughout the DAR, the structural analyses have generally employed simplifying assumptions that are conservative in nature.
An example is the treatment of axisymmetric hydrodynamic loads such as the CO-basic load definition.
The support excitation to a piping subsystem that is attached to the containment wall is a one-directional radial excitation.
The design analyses generally employed have been conservatively performed with the full amplitude of radial excitation applied in two perpendicular horizontal directions.
This substantial conservatism in the CO-basic load analysis has been removed in the piping analyses for the 30 piping subsystem evaluated herein.
1 i
1
Th2 ccc31arctica vclu23 ct th2 NOVs h;va bien calculetzd with thi Shercham pool size effect taken into account for the CO load definition and the tangential component of excitation removed for axisymmetric loads.
The results are summarized in Table 1 for the 60 MOVs on the 30 piping subsystems evaluated. Also shown are the qualification levels for both the
.valv6 operators and assemblies.
The operator qualification levels were arrived at by test.
The assembly qualification levels reflect the acceleration values used in the original qualifict.-ion stress analyses ratioed up by the faulted condition factor of safety.
For the c3 valves evaluated, all calculated accelerations were found to be acceptable.
For one valve, 1E11*MOV-039A, the calculated horizontal acceleration exceeds the valve assembly qualification level, but the combined effect with a lower vertical acceleration results in acceptable calculated stresses.
Based on the information presented herein, it is concluded that all Shoreham equipment will be proven to be within qualificat; ion levels. Phase II of this program is underway to address all valves on the remaining piping subsystems attached to the primary containment at locations of high amplitude ARS.
It is expected that the program will provide positive confirmation that the qualification of Shoreham MOVs conform with the requirements of the Mark II LTP hydrodynamic load definitions.
l l
2
References:
1.
Shoreham Nuclear Power Station Unit 1 Plant Design Assessment Report for SRV and LOCA Loads, Revision 5, December 1981 2.
Mark II Containment Program Load Evaluation and Acceptance Criteria, NUREG-0808, August 1981 3.
Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria, NUREG-0487, Supplement 2, February 1981
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