ML20140B830

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Requests Evaluation & Prioritization of Generic Safety Issue Re Failure of HPCI Steam Line W/O Isolation.Bases for Labeling Issue Generic Provided
ML20140B830
Person / Time
Issue date: 10/18/1983
From: Mattson R
Office of Nuclear Reactor Regulation
To: Speis T
Office of Nuclear Reactor Regulation
Shared Package
ML20140B832 List:
References
FOIA-85-772 NUDOCS 8311020209
Download: ML20140B830 (3)
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%.htta a high presses coolant ta,jection (EPCI)le.1solattaa valves.,The steam M::..

..u-YWrfrom*.ths.enisolatable break emeld discharge late perseemel spaces or speca EITM fa11ere to close,of redondant, het untestab f:c b*;. i. r%Enhistag critical enchimary. The Amector Systems Breach unde a pre 19mfae f

F~.".REaraientiam of the concern and concluded that the cancern is valid.*.! Thel:

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lh n Atsmin peep and a heester paw theegh a reducties pasr. valves, high-pressure *, 'p;.. -

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pressura.unta staan itses at a pelat upstream of the unte steem isolaties valves (NEIVs). :

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. e-verified directly Neuce. the primary basis for this concers is the potential y...

i high probability that both of these isolatten valves will fail to close es...~ c

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' < With regard to the consequences e7 am maisolatable break.the EPCI, steam

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line typically traverses the teres room. There are generally compartments ;-

l 1a the torus room ubich house RCIC and ECC pumps and associated ognipment. '.

At many SW/4 plants, the deers to.these cowartesats are left span to l

l prevent floattap of the terws under extreme fleeding condittees. Therefore.

l if the MPCI iso,ation valves failed to close, these systems would be subject to as environment more severe than the design entireement.

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Gee possible solution to this problem is to regefre that the outboard isolatics valve be kept moraally closed. This muld impact the availabillty of BPCI slightly but would reduce the probability of a HPCI stema line break.

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MEMORANDUM FOR:

cDarrell G. Eisenhut, Director i

Division of Licensing FROM:

Themis P. Speis, Director Division of Safety Technology

SUBJECT:

PRELIMINARY REVIEW 0F SHOREHAM PRA STUDY I

r We have completed our preliminary review of the Shoreham probabilistic risk assessment (PRA) study.

The current Shoreham PRA study, performed by Science Applications Inc. (SAI), considers only internal events (including internal flooding, but not including fire) and considers the frequencies of radioactive releases of various magnitudes but does not consider ex plant consequence. The report on ex plant consequence analysis, which has not been submitted, will be based on the results of the work performed by l

i Pickard, Lowe, and Garrick (PLG).' The Reliability and Risk Assessment l

l Branch, with technical assistance from Brookhaven National Laboratory (BNL),

has reviewed the internal event analysis, and the Reactor Systems Branch (RSB) in the Division of Systems Integration (DSI), also with BNL technical assistance, has reviewed the containment failure and radionuclide release analysis.

The Shoreham PRA study includes flooding.in the reactor building initiated by an internal' event. However, fires and external events such as earthquakes are not considered in the PRA study.

l We and our contractors believe that the Shoreham PRA study is a good and i

comprehensive piece of work within its stated scope. The Long Island t

Lighting Company (LILCO) estimate of the total core vulnerable frequency at Shoreham is about 5 x 10 5/ reactor year.

The Shoreham PRA study indicates that loss of coolant makeup following a transient challenge results in about 58% of the total core vulnerable frequency. Loss of containment heat removal following a transient challenge results in about 16% of the total frequency. Anticipated transients without scram (ATWS) sequences with a failure of alternate rod insertion (ARI) j result in about 25% of the total frequency.

Loss of offsite power (LOOP) events result in about 20% of the total frequency. There are about 20 sequences which contribute to 80% of the total core vulnerable frequency.

There appears to be no single risk outlier which, if it is removed, would significantly reduce the, total frequency.

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Based on the BNL requantification, we estimate that the total core vulnerable frequency at Shoreham is about 1 x 10 4/ reactor year. Our review indicates that ATWS events contribute about 40% to the total frequency and g

LOOP events contribute about 23% to the total frequency.

i The comparison between the Shoreham estimates of core vulnerable frequencies and the BNL estimates are given in Table 1 in the BNL report.

Our review does not identify any safety issue that needs immediate action.

We note that ATWS events at Shoreham contribute significantly to the total r*

core vulnerable frequency. However, we believe that the implementation of the ATWS rule requirements would reduce the contribution to the total core

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vulnerable frequency due to ATWS events.

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We note that the Shoreham PRA study has been used to address two issues, 1

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namely, flooding in the reactor building and reactor water level measurement i

system. These issues as well as the associated actions are discussed in I and in our previous memorandasea, h

Since the reactor building at Shoreham is an open annulus, a break in the high pressure coolant injection (HPCI) or reactor core isolation cooling (RCIC) steam lines in the reactor building with a subsequent failure to i

isolate the break may have damaging effects on the safety equipment in the reactor building. This issue is still under study and will be addressed in i

our final report.

With respect to the Shoreham containment response and radionuclide release analyses, BNL has completed their preliminary review and submitted to the RSB in the DSI. The evaluation from the RS8 will be included in our final t

report.

Our review of the ATWS events at Shoreham indicates that there is a large discrepancy between the determinstic analyses regarding the magnitude of the reactor power when the reactor water level is maintained at the top of the r

active fuel (TAF). We believe that the times available for operators to i

f take critical actions are dependent on the magnitude of the reactor power.

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We request the RSS in the OSI to provide us with feedback on this issue.

In addition, we request that our evaluation be sent to LILCO for comments.

We request that all feedback and comments from LILC0 as well as other NRR divisions be forwarded to us in three weeks to allow us sufficient time for consideration in our final evaluation.

' contains a summary of our preliminary findings and discussions of areas that may need further resolution. contains the preliminary report from BNL.

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l With this evaluation the Phase I (preliminary review) work on the Shoreham PRA study is complete.

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Themis P. Speis, Director Division of Safety Technology l

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Enclosure:

' l 1.

Preliminary Review of Shoreham PRA Study 2.

"A Review of the Shoreham Nuclear Power

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Station Probabilistic Risk Assessment,"

BNL, November 1984 5

cc:

H. Denton E. Case l

R. Vollmer R. Bernero H. Thompson A. Schwencer G. Burdick I

F. Rosa

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B. Sheron i

D. Ziemann G. Thomas M. Wigdor D. Silver R. Caruso M. Campagnone 4

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Contact:

492-4727 i

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FEB 19 mg NOTE FOR:

Themis P. Speis FROM:

Arthur Busiik

SUBJECT:

INTERIM REPORT ON HPCI/RCIC/RWCU LINE BREAK IN SHOREHAM REACTO BUILDING This note gives a summary of the status of work related to HPCI/RCIC/RWCU line breaks in Shoreham reactor building, and gives our plan for addressing We understand that Harold Denton may wish to be briefed on this this issue.

The concern is that the isolation valves in these lines may progress.

not be able to close under blowdown conditions. Generally speaking, they have not been qualified for these conditions.

Current Knowledga (1)

We estimate that the core-vulnerable frequency due to HPCI/RCIC (i) line breaks is about 2x10 7/ reactor year, even if the inboard isolation valve fails to isolate. The contribution of the HPCI line break dominateJ; the RCIC line is of only 3" diameter; the operator has more time to depressurize the reactor and recovery is more likely.

(ii) The outboard isolation valves are normally closed at Shoreham, while in most BWRs the outboard isolation valves are open. Moreover, the piping between the two isolation valves at Shoreham is of " break-exclusion" type, and is assumed to have an order of magnitude lower failure probability The estimate of the core-vulnerable frequency at than other primary piping.

Shoreham due to a HPCI line break takes into account these two considerations.

(iii) We note that the BNL analysis of the HPCI line break gives credit for use of the condensate system. The condensate system is estimated to have a 20% chance of failure for these sequences.

The possibility of the break causing failure of the. ability to use the condensate system was investigated.

The only identified dependency was the effect of the steam environment on the motor control centers (MCCs) for valves in the feedwater line; these MCCs are located in the reactor building annulus. However, it was found that these MCCs would very likely not be affected.

They are at a higher elevation than the HPCI line; they are on two opposite sides of the containment; and they are in enclosed cubicles, and are protected from the environmental conditions in the reactor building, according to information obtained by BNL and verified in an informal conference call between the staff and LILC0 on January 31, 1985.

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Contact:

Ed Chow, RRAB 49-24727

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. FES 131935 (iv)

Ex' cept for MSIVs, qualification of an isolation valve to determine its ability to close under blowdown conditions is generally not done by test, as far as we know.

In response to ACRS questions on valve qualification, Hope Creek recently submitted a data sheet from a valve closure test conducted by W'le Laborator'ies. The test'd'a'ta indicated that the valve was y

capable of closing in about 2 seconds against a differential pressure of 1370 psig. The test demonstrated that the valve differential pressure was never as large as the initial upstream pressure during the closing cycle.

This introduced further margin.

However, no tests or analysis under blowdown conditions have been performed for these valves.

For Limerick, the valves were shown by analysis to be capable of closing during blowdown i

conditions.

For Shoreham, we have obtained the test data for closing the HPCI and RCIC isolation valves.

The tests were performed by the valve vendor, Velan Engineering Company.

The test data for the RCIC isolation valves indicated that these valves closed in about 16 seconds under a differential pressure of 1135 psig. The test data for the HPCI isolation valves indicated that, under a differential pressure of 1135 psig, the HPCI inboard isolation valve closed in about 17 seconds and the HPCI outboard isolation valve closed in 44 seconds.

LILCO is still trying to retrieve test data on RWCU isolation valves and the procedures for testing HPCI/RCIC/RWCU isolation valves from Velan. We have received no schedule from LILC0 as to when they will submit this test data and procedures to us. We will ask Equipment Qualification Branch (EQB) to review this information in order to determine if the valves can close under blowdown conditions.

(2) Planned Future Efforts The preliminary BNL review of the Shoreham PRA study.did not explicitly address the RWCU line break because the RWCU line is 6" in diameter and is much smaller than the HPCI line which is 10" in diameter.

If a RWCU line break occurs, there is more time for the operator to take recovery action.

However, the RWCU line is always open, so that the advantage of a closed outboard isolation valve, as is the case with the HPCI line, is lost.

BNL is pursuing this issue.

Furthermore, BNL is examining other line breaks in the reactor building in addition to breaks in HPCI, RCIC and RWCU. We believe that in general the other lines are less than 4" in diameter and there would be greater time for the operator to take recovery action. We have expanded the BNL contract effort to devote more manpower and resources to this issue.

BNL will perform sensitivity analysis and provide estimates of core-vulnerable frequencies due to these breaks, assuming that the isolation valves fail to isolate during blowdown conditions (the probability of the valve failing to close is assumed to be 1).

We expect BNL to complete this effort by March 8, 1985.

In addition to the sensitivity studies, our program of ef fort includes assessing whether the valves can close under blowdown conditions.

The schedule for this

FEB 1b ng depends on when LILCO submits the test data and procedures; the length of 4

time needed by EQB to perform the review will depend on the nature of the material submitted.

t 4

Arthur Busiik, Section Leader Reliability and Risk Assessment Branch cc:

D. Eisenhut R. Bernero H. Thompson R. Vollmer A. Schwencer B. Sheron M. Hodges G. Thomas M

R. Wright, EQB R. Silver 4

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,l} }l BROOKHAVEN NATIONAL LABORATORY Q(ll ASSOCIATED UNIVERSITIES, INC.

Upton. Long Island. New York 11973 (516) 282' 2147 Department of Nuclear Energy FTS 666' March 1, 1985 Mr. George Thomas Reactor Systems Branch Division of Systems Integration U.S. Nuclear Regulatory Commission Mail Stop P-1132 Washington, D. C.

20555

Dear George,

I have enclosed the revised list of information items for the containment re-sponse review which you requested (Attachment 1).

I have included a schematic of the secondary containment (reactor building) indicating the expected fail-ure location (see Appendix M of the Shoreham PRA) and the water relocation for gross overpressure failures (Attachment 2).

The previously transmittd BNL decontamination factors are included as Attach-ment 3.

These are tirae-averaged pool scrubbing factors based on SPARC (NUREC/CR-3317) and do not take credit for in-vessel retention and primary containment hold-up. Note that for the Class-I sequences the suppression pool is subcooled and that the BNL decontamination factors cre high and in substan-tial agreement with the Shoreham PRA results.

However, for Class-IV se-quences, the pool is heated to saturation before core melt and the BNL decon-camination factors are much lower.

For the Class-IV ATWS with failure at the basemat (y"), it is assumed that the pool is relocated to the annular region of the reactor building which surrounds the primary containment (see Attach-ment 2).

Thus, the in-vessel release through the SRV's see the same scrubbing as the ex-vessel release through the vent pipes. The slight difference in the decontamination factors (9 compared to 14) for the two releases depend on gas blowing rates at the tiime the scrubbing occurs.

If you have any questions, please call.

Very truly yours, enneth R. Perkins a

KRP:tr Attachments cc:

W. Y. Kato (w/ attachments)

J. Rosenthal (w/ attachments)

R. A. Bari M. Wohl W. T. Pratt F. Eltawila c.

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ATTACHMENT 1 REQUEST FOR INFORMATION 1.

Table II of Appendix M gives different pressure limits for the longitudi-nal reinforcement bars at the base of the containment and in the wetwell region. However, the longitudinal bars appear to be continuous and should therefore have the same stress. Please explain the basis for the dif-ferent results.

2.

Table II of Appendix M indicates that the shear bars at the base and dry-well head have the low'st pressure holding capability (121 psi and 120 e

psi, respectfully) but the discussion indicates that the additional rein-forcement will preclude this failure mode. Since the containment failure mode is a key ingredient of the release estimates, please provide a quan-titative estimate of the additional shear strength provided by the non-shear reinforcement bars.

3.

If shear failure is precluded as discussed in Section 3.2 of Appendix M, "it appears that the ultimate capacity is controlled by the yield of the longitudinal and the hoop bars at about 123 psi."

These two failure modes appear to be very important to subsequent fission product release (particularly for Class IV ATWS) since they will both occur in the wetwell region. Please provide an estimate of the size, location and direction (vertical or horizontal) containment failures for each of the three possi-ble failure modes.

4.

Section 3.6 of the PRA takes credit for containment leakage which will prevent gross containment failure for all pressurization rates except the very rapid pressurization associated with large breaks. However, the structural analysis by Stone and Webster (Appendix M) did not identify any significant source of leakage. The basis for the expected leakage source and the leakage rate as a function of pressure should be provided.

5.

The basis for the partitioning between release category 10 and 11 (no pool bypass vs. partial pool bypass) should be provided. The phenomeno-logical basis for the estimate of only 10% bypass should be provided.

Preliminary results from IDCOR indicate that for some BWR sequences the

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vessel will fail with only 20% of the core molten. Presumably 80% of the melt release would bypass the SRV's and be released into the drywell.

6.

The basis for binning into release categories is poorly described and the transfer from Tables H.4-8 etc. into the 16 release categories is diff t-cult to interpret. A table listing the specific sequences which are binned into each category should be provided.

7.

The lack cf R5 sequences in the release categories makes it apparent that these releases have been binned " downward" into the lesser release cate-gory Rg.

The basis for this " downward" binning and,any other sequences that are moved to less severe categories should be provided.

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8.

Table H.4-25 appears to be incomplete in that it does not include se-quences 06 and 08. The completed table should be provided.

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REQUEST FOR INFORMATION (Cont.)

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9.

The source escape fractions used for end state screening (Table 3.6-10) appears to be quite arbitrary yet it greatly influences the importance ranking.

In particular: the use of I as the surrogate for melt release ignores the fact that there are noble gases in the melt release which will not be scrubbed at all; the use of a large scrubbing factor (500) for Cg transients is inappropriate since most of the melt release will be released directly to a failed containment; the reduction facter of 0.01 for y" failures is indefensible since the event tree precludes everything but large ruptures where the pool will be blown out into the reactor building at high pressures, i

Table 3.6-10 should be replaced by a table with defensible reduction fac-4 tors. As a minimum the table should include a separate category for Cg transients, uhich recognizes the defined sequence of events (containment failure before core melt).

In addition, a detailed justification for i

J each reduction factor should be provided along with the numerical results of the ranking process. This revised table will provide the basis for our independent importance ranking based on revised estimates of accident frequency and reduction factors.

10. Sheet 1 of Figure H.4.2 has been reduced so that it is illegible. A full-size legible copy should be provided.

4 i

11. Appendix L provides a detailed discussion of the disposition of the i

corium (90% is expected to go down the vent pipes) based on the revised i

reactor pedestal geometry illustrated in Figure L.3-2.

However, this I

figure is inconsistent with other descriptions of the geometry (e.g.,

i, Figure 2.3-2) and provides inadequate information for an independent j,

assessment of the corium disposition. Please provide detailed (as built) drawings of the vent pipes and their covers within and external to the reactor pedestal region.

Include a description of whether the air ducts and manways in the reactor support wall will be blocked during operation.

r

12. Provide the estimate of the fraction of the molten corium which is ex-pected to spread out of the pedestal area through the open manways and air ducts in the reactor suoport wall, i

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' ATTACHMENT 3 i

Table 1 Comparison of Suppression Pool Decontamination Factors for Core Malt Accidents in Shoreham 1

Decontamination Factors i

Shoreham 8NL

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Sequence SRV Vents SRV Vents a

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