ML20023D601

Forwards Comments on Draft Rept Re Reactor Safety Study Methodology Applications Program for Facility.Changes Indicated in Areas of Validity of Sys Criteria & Correctness & Appropiateness of Assumptions in Sys Analysis

ML20023D601
Person / Time
Site:	Oconee, 05000000
Issue date:	09/22/1980
From:	Parker W DUKE POWER CO.
To:	Bernero R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML20023A436	List: ML20023D601 ML20023D604 ML20023D606
References
FOIA-83-123 004022, 4022, NUDOCS 8305240259
Download: ML20023D601 (11)

Text

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• DUKE Powen' COMPANY Powen Den.mwo 42c Sourn Cauwcu Srazzr, CnAutoriz. N. C. une4a wf LLI Ate O. PA R M r.R, J R.

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- sitaw PmOOUCTION 373-4083 i

September 22, 1980

'Mr. Robert Bernero, Director Probabilistic Analysis Staff Office of Nuclear Regulatory Research

. U. S. Nuclear. Regulatory Commission Washington, D. C. 20555-Subj ect: IfotseneTrIdig)-

Draft of RSSMAP Report

Dear Mr. Bernero:

<Your letter of September 3,.1980 enclosed the draft report on the Oconee RSSMAP study, performed by the Probabilistic Analysis Staf f and

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its contractors, for our review and comment. We have been able to conduct only a limited review of this draft because of the short time made'available to us. Nevertheless, we have found a number of areas

in the study and in the report itself where changes would be desirable and perhaps necessary. These areas of concern include the validity of-the system success criteria, correctness and appropriateness of the assumptions made in the systems analysis, and the applicability of the

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. comparison of the results of the Oconee study to those of the RSS reference plant. Our comments are summarized below and elaborated in the Attachment to this letter.

In the area of the system's success criteria, we found that the criteria used for the emergency core cooling were incorrect and that. tlie success criterion used for the emergency feedwater system needs to he modified.

In the systems analysis area, there exist discrepancies between the

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assumed and. actual c, afigurations .of the HPI and LPI systems. ~Also the assumed' values of~ failure probabilities-for manual actions in a number of instances we believe are overly conservative (HPI cooling, steam generator cooling by HRASWS, reestablishing AC power, etc.).

It-is apparent tint in-some instances an attempt was made to utilize

. Oconec specific component reliability (reliability of the turbine-driven N

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Mr. Robert '2rnero.

Page 2 Sept ember , 1980 emergency feedwater pump and the reliability of the 'Keowee units) . Our review of the applicable failure data indicates that the failure and demand data were not properly interpreted, and as such the unavail-abilities for the affected systems were significantly overestimated.

We understand that in the RSSMAP study the RSS methodology was modified in several critical areas (common mode failure contribution, human reliability, etc.). Thus the indicated difference in the risk and severe core melt probabilities between the Oconee and the RSS reference

-plant does not provide a valid comparison.

The problem areas and discrepancies summarized above and elaborated in the Attachment would have a significant impact on the calculated frequencies of each of the dominant sequences, except perhaps the Event-V sequence, and as such warrant appropriate reanalysis. We suggest that, inasmuch as possible, the draft report be revised taking into account the reviewfcomments provided herein.

We recognize that the Oconee RSSMAP study had to rely primarily on FSAR-

.. type information, which is -outdated in several areas, and did not have the benefit of a more complete information base on the actual plant characteristics in several important aspects. A more thorough characterization of the Oconee accident sequer :es and plant risk is expected by way of the Duke /NSAC Oconee PRA program. The Oconee RSSMAP

' study would certainly serve as a valuable" reference and road map for the Oconee PRA program.

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.It should be pointed out that the Oconee RSSMAP study and our carlier auxiliary feedwater system reliability study have uncovered two areas of the Oconec emergency feedwater system where further improvements p . .

are possible and desirable. Design modifications have been initiated to eliminate the AC dependencies in the turbine-driven EFWS train.

+ . As a result of these modifications, flow of cooling water-to the turbine oil. cooler and to the pump cooling jacket will be by ' gravity flow f rom ~

the high pressure service water. system. This modification should

. nignificantly reduce the-unavailability of the EFWS for transients involving loss. of of f site power or loss of ' all AC power. With regard i to the Event-V sequence, Duke is evaluating measures to reduce the risk

- 'of this event and will implement appropriate procedure changes and/or. ,

modifications to . reduce the risk to an acceptable level.

In su= mary, we consider the report to be flawed to'such a degree that correction of certain portions of the report should be considered. prior to publication. To resolve these comments, we suggest that a meeting be held at'a mutually agreeable time.

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Mr. Robert Bernero Page 3 September 22, 1980 Any questions or comments regarding.the matters discussed in this letter and attachment may be directed to P. M. Abraham of our Project Coordination and Licensing Section phone: . (704)373-4520).

Ve truly yours, x

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. William O. Parker,'Jr.

.PMA:vr

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DUKE POWER COMPANY COMMENTS

, ON DRAFT OCONEE 3 RSSMAP REPORT ,,

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DUKE PO'JER COMPAhT COPJENTS ON OCONEE RSSMAP REPORT

. The following comments primarily address the assumptions and analyses in the Appendices, the source of basic information, but are also applicable to per-tinent sections of the main body of the report.

I. Appendix A1. LOCA Event Trees.

2.1.4 The success criteria for emergency core cooling identified on page 14-57 of the FSAR and used in RSSMAP have been revised as the result of a more recent analysis. The currently applicable criteria, based on Appendix K requirements and documented in BAW-10103, Rev. 3A, are listed below:

LOCA Equivalent Diameter Success Criteria Large Break (A) D > 10" 2/2 CFT '

and 1/3 LPI Pumps Sma'll Break (S1 ) 4" < D $ 10" 2/2 CFT and 1/3 HPI Pumps and 1/3 LPI Pumps Very Small Break (S 2) D-5 4" 1/3 HPI Pumps-Therefore, it is not necessary {.o consider a specific break spectrum in the range of 10" < D $ 13" In addition, the suc-cess criteria used are overly conservative and' the emergency core cooling unavailability will be reduced, thereby reducing the estimated frequencies of sequences involving the term D.

f 2.2.1 Per the preceeding discussion, only three break sizes need to be considered, eliminating the 10" < D $ 13" categoryE' -

2.2.2 The RPS is required to operate for S and t S 2 LOCA's, but not for A LOCA's as redefined above.

2.2.8 The success criteria for ECR are listed below:

LOCA , Functional Success J

A 1 of 3 LPRS S 1 of 3~LPRS 4 . S2 1 of 3 LPRS and' 1 of 3 llPRS e O v

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3.0 Per the preceeding discussion, a stuck-open RCS relief valve results in an S LOCA2 rather than S 3LOCA.

Table A-1 should be revised to reflect the revised success criteria discussed -

above.

II. Appendix A2. Transient Event Tree 2.1 Per the redefinition of LOCA categories, RCS integrity is re-quired to prevent a small small (S 2) LOCA.

2.1.2 The RCS heat removal requirements as stated are unduly con-servative, and the following evaluation demonstrates that flow from any one of the three EFW pumps to either steam generator is sufficient.

Using the August, 1979 ANS Decay Heat Standard and assuming an EFW enthalpy of 61 Btu /lb (corresponding to 90F and 1000 psi), the EFW flow requirements following shutdown are listed below:

Time (Sec) Power Function Flowrate (gpm) 60 0.0342 531 80 0.0322 500

. 100 0.0308 479 150 0.0283 440 200 0.0267 415 400 0.02329 361 600 0.0213 331-The demand for EFW flow occurs when the initial steam generator inventory boils off to the minimum Icvel. At this time the capacity of one motor-driven pump (500 gpm) is adequate for RCS heat removal. This will reduce the unavailability of the EFW system, and consequently the frequencies of sequei:es*

involving the term L.

Two-additional means of removing RCS heat were not discussed explicitly. If off-site power is available, one hotwell -

- pump and one condensate booster pump provide sufficient flow if steam generator pressure is reduced using the turbine bypass valves. The original auxiliary service ' water pump is also still available, requiring opening of the manual atmospheric dump valves to depressurize the steam generators'

'to approximately 60 psig.

As reported in BAW-1610, the maximum RCS pressure expected for an event involving failure of the RPS is 3600 psig,.

rather that 4000 psig.

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' . x 2.2.3 The unavailability estimate of the PCS during T2 transients is unrealistically high for Oconee. To obtain the necessary main _feedwater flow prior to SG dryout, only one. train of the

-hot-well - condensate booster - main feedwater. pumps combination

• is_necessary. In the event the PCS recovery is unsuc'cessful prior to SG dryout, the hotwell-booster pump combination can be utilized since the:resulting SG pressure is low, and further the turbine bypass valves can be controlled to maintain SG pres-sure within the HWP-CBP flow capability. The Oconee turbine

, driven main feedwater pumps can be supplied with motive steam

• via the station auxiliary steam header.

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. A survey of recent Oconee operating experience for 1979 and 1980 identified nine reactor trips involving feedwater tran-sients. In all nine occurences,-the main feedwater. system was either available throughout the transient or was recovered within 30 minutes of.the. reactor trip. This operating ex-perience suppo,rts the contention that a PCS non-recoverability probability of 1.0 grossly exagerates the unavailability lof j- the PCS during transients other than T3 transients.

Based on the above considerations, it is inappropriate.to  !

assume a value of I for the PCS unavailability during T2 events. Rather, a value of 10 2, consistent'with the operat-ing experience, would be appropriate. .This change would

,, reduce the. presently calculated probability _of sequences involving the term M.

2.2.4 The discussion should_ identify hFW capabilities'as two 100%

motor-driven pumps and one 200% turbine driven pump.

2.2.7 reseat af ter being The frequency challenged of the PZR is assigned relief value the same valves(10 to_2) for the PORV and the safety values. We believe that the value_for the safety. valves is much smaller (of the order of 10 4).fWith respect to the PORV, operator action to close the PORY block

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value should be considered in light"of the recent' changes

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in operator training and procedures and thetimplemenation of direct PORV position indications. Considering the smaller frequency of safety value failure ~and the changestin PORV is ' -.

olation~capabilit'y, it-is suggested that_a much smaller value for. Q, perhaps . in .the range of -10 4 - 10 3 would:be appropriate.

.This change #'ould reduce the currently calculated frequencies

. of sequences involving Q by a factor of 10 to 100.

TheLunava11 ability of feed and bleed cooling _during transient.

? -sequences-involving failure of the PCS, EFS, and HASWS is overestimated. .HPI cooling in the: absence of SGLcooling is

[-- explicity' required by_ the emergency procedures. Because of-the RCS' saturation alarm and the obvious ~ indications of inadequate.SG~ cooling, and considering that?a~ time' inter-

, Lval :offgreatcr : than 20 ' minutes -is available to initiate thi,s -

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function, we believe that the unavailability of this fun-ction is of the same order as that of the ECC recirculation I function. Furthermore, continued boil-off of the RCS in-ventory through the PZR relief values would eventually ac-tuate the ESF on high RB pressure.

3.0 The stuck' open relief safety valves r esult in S 2 1.0CA's rather than S3 LOCA's.

III. Appendix Bl. Emergency Power System 2.1.1 At least one of the two Lee combustion turbines is started and energizes the dedicated 100 KV line to Oconee whenever.

one of the Keowee units is unavailable. In the event-the

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Keowee outage is an extended one, a load from Oconee (nor-mally a 4160 volt bus) is placed on the Lee turbine for operat-ing reliability.

5.2 The unavailability calculated for the Keowce units based on the number of tests and failures appears to be too high. It was apparently assumed that only a monthly test of each Keowee unit was made, yielding 168 tests. However, seven annual tests have been performed in addition to system demands upon the Keowee units to supply power to the grid. Properly accounting for Keowee startup demands from.all sources indicates that ireater than 2500' demands have occurred. . Twelve instances have been identified-wherein a unit failed to deliver power upon demand, regardless of the source of the demand. Of the twelve instances, seven involved -a unique failure of a 3 articular component over a five month period. The problem is belived to have been rec-'

tified. This type of failure has not reoccurred in the suc-ceeding two years and may justifiably be counted as a single failure. Thus, the Keowee failure per demand ratio can be conservatively calculated as 6/2000.

On a qualitative basis, the Oconee EPS reliability is" jqual to ar better than that of many other nuclear stations. Alternate power sources are available to an individual Oconee unit from the other nuclear units at the station, from the system grid, from the near site Keowee units, and from the' combustion tur- ~

bines at the Lee' Steam Station. The hyd,roelectric generators at the Keowee facility have inherently simpler design and operating characteristics relative to diesel generators and thus. represent a more reliable backup power source. Further-more,-the Keowee' units are frequently called upon toisupply power to'the system grid -and therefore problems are more likely to be detected and corrected prior to emergency.use. Based upon these considerations, the AC power non-recovery probability

-(within toi3 hours) of 0.1 - 0.5 used in the report 'is un-realistic.

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IV. . Appendix B3. Reactor Protection System 5.2 Q(RPS) = 2.6 x 10~5 /RY, not 2.6 x 10 5 /RY Table B3-1 Two reactor trip setpoints are incorrectly listed.

The correct setpoints are identified below:

Over Power 105.5% of rated power RC Pressure 2300 psig - High V. Appendix B6. Low Pressure Injection System 2.1 During normal operation, motor-operated valves LP-21 and LP-22 in the LPIS suction lines from the BWST are left open. This will reduce somewhat the unavailability of the LPIS.

Although it is not important to the results, the notation for the LPIS pumps and coolers refer to Oconee-1 components. The correct notati'on for Oconee 3 is LP-P3A, LP-C3A, etc.

5.1 The correct success criterion for LPIS for both A and S LOCA's t

is one out of two pump trains.

5.2.1 Since the LPIS success criterion is the same for LOCA's A and S2 , only one Boolean equation _is required, i.e., Equation B6-1.

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5.2.2 The LPIS unavailability for failure of 2 of 2 trains is re-calculated for the case when MOV's LP-21 and LP-22 are normally open. g for LP-21 plugging 1x10[4 LP-22 operator error 3 x 10 4 (Consistent with LP-28)

Q2 total 4 x 10 4 E. B: LP-22 + LP-30 4 x 10~4 + 10~4 = 5 x 104 C: LP-21 + LP-29 - 5 x 10~4 [ -.

. The unavailability of ~the LPIS is therefore reduced fr'om 2.6 x 10~3 to 2.1 x 10 3, so the change is not very significant.

Table B6-4 should be deJeted.

VI. Appendix .B7. - Low Pres,sure Recirculation System 5.1 Successful ECR requires one LPRS train for A and S LOCA's.

1 5.2.1 The Boolean equation for LPRS failure excludes some terms in-cluded in LPIS since success of LPRS is important only given success of LPIS. Therefore, the terms from the LPIS equation which are -included here do not have the same unavailabilites, and should be' lower.

Also, although the discussion in 2.1 includes the third LPIE

, pump, no credit is-taken forLits availability in ,the analysis.

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VII. Appendix B8. High Pressure Injection System 2.1  :

The~ electric power for each of the three HPIS pumps is sup-plied from a different 4160 volt bus. It appears that this discussion should refer to the digital ES channels which -

actuate the three pumps.

2.2 The injection valve in the B train (HP-27) is normally left open with power to its motor operator removed. This should reduce the unavailability of this train somewhat. A cross connection is available between all three injection lines, with isolation by normally closed motor-operated valves HP-409 and HP-410. The injection points are downstream from the normal isolation valves, HP-26 and HP-27, allowing operator action to assure two trains of HPIS flow in the event of failure of the injection' valves or an HPI pump. This also should reduce the unavailability of the HPIS.

5.2.1 - The HPIS unava'ilability due to' pump testing is incorrectly calculated. At Oconee, the A and B HPI pumps are used alter-nately to . supply normal makeup sea 1 ' injection. Test procedures for the C pump 'specify that it be used to supply normal- makeup by' closing.HP-27 (refer to Figure B8-1) and opening thel cross-tie valves from the B injection line, HP-116 and 117 (not labeled in the. figure). Thus, this pump is available should

,, an HPI actuation signal occur- during testing of that pump.

Therefore, HPI pump testing does not contribute to the HPIS unavailability.

D VIII. Appendix B9. High Pressure Recirculation System -

2.1 The discussion concerning electric power contains the same error identified above, i.e., three separate buses supply power-to the three respective pumps.

IX. Appendix B10. Engineered Safeguards Protection System I'~.

2.2 In response to NUREG-0578, reactor building isolation is act-uated by channels 1 and 2 to provide isolation on either low reactor coolant pressure or high containment pressure. -

X. Appendix B11. Containment Spray Injection' System -

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Valves LP-21 and LP-22 are normally open and are therefore not required to change positions on ES 7 and 8 actuation. This should reduce the unavailability of the CSIS.

5.0 The dominant failure contributor of the CSIS during transients and small' break LOCA events was treated to be the failure of the operator.to start the system whenever the CSIS is manually.

bypassed. Resetting ~the applicable ESFAS channels'would'still-maintain,the system in the safety mode. Deliberate operator

- action is required to-bypass ~an automatic safety function and

- is not-permitted unless it'is confirmed that the plant mode does~not require that function. Even under bypass conditions, T-1 - I a g - . -- - - - ' ' ii - -' m

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the high reactor building pressure alarm would alert the C - operator to activate the sprays. Therefore, the assignment of a higher probability for the CSIS failure during transients and small break LOCA events does not seem to be necessary.

XI. Appendix B13. Emergency Feedwater System and.High Head Auxiliary Service

~ Water System

1. Since 1973, 158 tests of the turbine driven emergency feedwater pumps have been performed. A review of the test results iden-tified four incidents wherein the pump failed to start. This data supports a pump unavailability probability of 2.5 x 10 2 which is more than three times lower than the value used in the report.

2. The auxiliary service water system cross-ties frons the other units are still available, in addition to the new high head system. Thus, two additional backup systems are available.

3. In addition, as discussed in the comments for Appendix A2, the

-correct success criterion for EFW is one of the three pumps, or a flowrate of 500 gpm.

4. The unavailability of the HMASWS is estfrated at 0.1 based on human error. This seems unrealistically high considering the

~ fact that it is a dedicated system and further that the operators will be dedicated to the SSF. It is expected to yield fairly high availability, on the order of that for the 'rps, D

XII. Appendix BIS. Reactor Building Cooling Systems 5.2.2 The failure of'the RBCS was estimated for cases with and with-out?AC power initially available, although these cases were not investigated for other systems.

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Dun e POWER COM PANY Pownu Dun.oixo 422 Sourn Cut ncis Si uner. CnAnt.oTrz, N. C. ::nm W 8 L LI A *4 Q. PAR et C R, J R. ,

Tg 6 t >=o = CAnta 704 V*Cr Pers.or-t sic.- e-coves o= November 3, 1981 Mr. Thomas M. Novak Assistant Director .for Operating React ors Division of Licensing U. S. Nuclear Regulatory Cor.. mission Washington, D. C. 20555 Subj ect: QPcuegdig .

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Dear Mr. Novak:

s The following information is submitted in response to your 1ctter of September'8, 1981 concerning the Reactor Safety Study Methodology Application Program (RSSMAP) study of Oconee Unit 3. Duke Power Company previously (by our letter of September 22, 1980) provided comments on an earlier draft report of this study.

We note with appreciation that some of our comments were considered in the preparation of the final version of the report. Specific comments on the con-clusions of the study reported in Section 6.3.1 and some discussion on changes in plant systems and procedures implemented at Oconce subsequent to the RSSMAP study with potential positive impact on the RSSMAP estimated probabilitics and consequences of accidents are provided ,in the following paragraphs.

With regard to the frequency of an interf acing system LOCA event. we have instituted a program for periodic leakage testing of the check valves of inter-est and further have ceased the stroke testing of the normally closed MOV's-LP-17 6 -18 at operating conditions. (Stroke testing of these values is done

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only at cold conditions.) Since the Icak-leak failures are essentially elimi-nated by the periodic Icak testing program and by climination of the MOV stroke _

testing at operating conditions and since the Icak-rupture and rupture-rupture

. f ailures are significantly small with the normally closed configuration of the MOV's, the event V is'now believed to'be a non-significant risk. contributor.

Plant codification has been completed to eliminate the AC dependency of the turbine driven emergency feedwater pump. With this modification, the avail-ability of the emergency feedwater system during accidents involving loss of offsite power or loss of all AC power has improved. Consequently, the frequency of core melt accidents initiated by.or involving these events would be less ,

than that estimated in.the RSSMAP~ study.. 1

' TVo. changes which are outside -the scope of 'the RSSMAP study and whidh came to

- our attention in conjunction with the : ongoing NSAC-Duke Oconce PRA program are being implemented'now. . One is a change in the emergency procedures to deal with a situation in which the LPI pumps could be running at shutoff head for ar. cxtended period of . tin . Such a ~ situation is postulated to occur during 8 -

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Mr. Thomas M. Novak November 3,1981 Page 2 certainsmallbreakLOCAandseveresteamlinebreakeventsiftheRbSdoes not depressurize sufficiently below the LPI actuation setpoint. Although the operators are aware of the need to secure the LP1 pumps within reasonable time under this situation, the existing emergency procedures do not include this requirement. A change in the applicable procedures is now being imple-mented to include the necessary guidance. The other change pertains to two ICS simulator relays. A postulated spurious energization of these relays could lead to a feedwater transient (resulting in a reactor trip) and the turbine bypass valves failing closed. A modification of the system to de-activate this circuitry is being impicmented. The interim results of the Ocor.e5 PRA are being monitored to assure early identification of any important risk outliers. At this point, no additional items have been identified which merit consideration in the near term.

A number of the post-TMI changes to plant 1 systems and procedures have con-tributed to improved safety both with. respect to probabilities and. consequences of accidents. Among the measures contributing to reduced probabilities of accidents are: ~ -

(a) Modification to the turbine EHC System to reduce the frequency ,

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of turbine-reactor trips.

(b) Modifications to the main feedwater system to minimize tLa occurrence of feedwater transients. -

(c) Changes in the control system power supply to minimize the occurrence of power supply failure induced transients and to better cope with such events.

(d) Modification of the emergency feedwater system initiation, control, -

and indication functions for better reliability and performance.

The following other post-TMI efforts, which are in various stages of impicmenta-

- tion, are considered to effect further reduction in the probabilities and cons'equences of accidents.

'(a) Renewed vigilance and scarching reviews now being conducted on operational occurrences through the operating experience evaluation program. ,

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(b) Control room design review and incorporation of safety pargpeter display system.

(c) Improved operator training, development and-impicmentation of improved 4

-procedures and the utilization of shift technical _ advisors.

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Mr. Thomas M. Novak November 3, 1981 Page 3 *

(d) Implementation of RCS high point vents, post-accident samp5ing panel, and dedicated hydrogen penetrations.

(e) Implementation of PORV/PSV position indicator and RCS subcooling monitor.

(f) Impicmentation of accident monitors and expanded emergency planning programs and facilities.

We are not certain whether the conclusion reached.in the Oconee RSSMAP study rega-Jing hydrogen burning and the associated impact 1on containment is valid.

There is reason to ,believe that the Oconee containment failure pressure is much higher than.that assumed in the RSSMAP study (183 psia versus 133 psia).

Furthentore, the MARCH code treatment of hydrogen in regard to its generation in the core, accumulation in the containment, and degree of burn- in the con-tainment is generally recognized to be very conservative, particularly for small break and transient induced core melt events. A more realistic analysis of the containment accident; process is expected in the NSAC-Duke Oconee PRA analysis, It is our impression that the Oconce RSSMAP study has been a very worthwhile undertaking. Although there are some limitations in this-study, it still

.provides some useful insights into the dominant accident sequences and their contributing factors.

Ver 'truly yours,

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William O. Parker, Jr. ,

PMA/php cc: Mr. Robert. F. Bernero, Director Division of Risk Analysis

Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, D. C. 20555 ,

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