ML20215M587

Preliminary Evaluation of Containment Study Transmitted w/860902 Ltr.Licensee Estimates Appear Optimistic Considering Uncertainties Inherent in Failure Rate Data

ML20215M587
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Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	10/24/1986
From:	Office of Nuclear Reactor Regulation
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ML20215M575	List: ML20215M573 ML20215M587
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f 1 ENCLOSURE 1 PRELIMINARY EVALUATION OF CONTAINMENT SAFETY STUDY BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR THE VERMONT YANKEE NUCLEAR POWER STATION VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271

1.0 INTRODUCTION

. By letter dated June 30, 1986 from J. G. Weigand to H. Denton, the licensee, Vermont Yankee Nuclear Power Corporation, comitted to perform a containment safety study. By letter dated September 2,1986 from J. G. Weigard to H. Denton, the licensee transmitted the study for staff review. On September 12, 1986 representatives of the licensee met with the staff, sumarized the study, and answered questions. On September 28 and 29, 1986, staff representatives visited the plant and the neighboring Vernon Hydro Station to review plant features and portions of the study.

The purpose of this evaluation is to provide the staff's preliminary coments on the licensee's study. Final coments will be based on the licensee response to staff's questions and coments. A regulatory position on generic implementation of improvements must await review of backfitting implications under the requirements of 10 CFR 50.109. The following evaluation should not be considered a Nuclear Regulatory Commission position on the issue of Vermont Yankee Nuclear Power Station containment safety.

2.0 EVALUATION The licensee's study basically consists of two parts. The first part is an evaluation of containment failure probabilities in severe accidents. The l

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second part is an evaluation of potential improvements in five areas (combusti-ble gas control, drywell spray capability, containment venting, core debris control, and training and procedures).

m 2.1 Containment Failure Probability Evaluation The licensee evaluated the design and operation of Vermont Yankee Nuclear Power Station (VYNPS) with respect to core melt probabilities and containment system response. The evaluation used detailed assessments for a similar reactor, adjusted for specific VYNPS features. The principal conclusion by the licensee was that a best estimate of the conditional probability of containment failure was .07. That is, if a core melt accident were to occur at VYNS, there is a 7 percent chance that the containment systems will fail and result in a large release of fission products.

The bases for the licensee's estimate include a number of assumptions, evaluations and judgements. One important judgement pertains to the use of results from an industry developed computer code called Modular Accident Analysis Program (MAAP). The staff has not reviewed this code, at present has no plans to 1o so, but will assess the results and conclusions from its applications to VYNS. The accuracy and application of the code is a subject of considerable debate however. Therefore, uncertainty must be attached to the use of the code at this time.

A number of important conclusions have been drawn by the licensee based on the containment safety study. The more important ones are:

A conditional containment failure probability of .07.

A maximum containment system failure pressure of about 135 psia.

The best estimate core melt probability was 3x10-5 per reactor year (one chance in 30,000 per year of reactor operation).

One accident class (loss of cooling capability with the pressure vessel at high pressure) dominates the core melt probability (about40 percent).

Although station blackout and ATWS are significant contributors to the core melt probability at Vermont Yankee, as for other plants, the largest contributor is loss of makeup with the vessel at high pressure.

The staff's experience with core melt and containment failure estimates indicates that large uncertainties exist in these estimates. In addition, a high pressure core melt entails greater uncertainty regarding containment response than prevails with low pressure core melt. Pending receipt and analysis of the response to staff questions, the staff concludes that the licensee's estimates appear optimistic considering the uncertainties inherent in failure rate data, modeling of systems, human responses, accident initiator identification, as well as the physical processes that follow degraded core conditions.

Vennont Yankee has four SRVs which serve as the automatic depressurization system (ADS), compared to eleven SRVs at Peach Bottom, five of which serve as the ADS. Further, at Vermont Yankee the four SRVs are three-stage Target

Rock valves which have relatively high failure rates based on reported failures.

When the the failure experience of SRVs is considered, accident sequences TPUV, TPQUV, T PQX, T PQUV,gT QUV and TPW need to be added to the dominant E E accident sequences identified by the licensee. As a result, the total core damage frequency of 3x10-5/R-yr could be increased by about a factor of 2.

When SRV failures (spurious opening and stuck open relief valves) rela'ted accident sequences are considered, the release type for Class II accidents (i.e., loss of containment heat removal and subsequent loss of makeup) may change. This is due to possible early failure of the RHR capability after loss of NPSH as a result of stuck open relief valves and subsequent pool heatup. As a result, the containment conditional failure probability for Vermont Yankee could exceed 7% for all sequences and result in containment failure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The frequency of Loss Of Offsite Power (LOOP) at Vermont Yankee is estimated to be 0.07/yr. When compared with 0.22/yr for the national average grid, the Vermont Yankee value is deemed too low based on the conventionally used Bayesian estimation technique. The failure rate of the nearby hydropower plant is estimated to be higher than 0.07/yr. In the Bayesian estimate, the plant data are used as inputs modifying the a priori distribution, i.e., the national average LOOP frequency.

Additionally, the estimated 6-8 hour battery capacity at Vermont Yankee was based on 2175 ampere-hours rated capacity discharged at a constant load of 1.75 volts / cell, with the load estimated to be 27 amperes. This does not include the excessive loads likely to occur during the startup of motors, and the abnormally high ambient temperatures likely during the accident. Thus, a more a _. .

likely estimate for the battery capacity is on the order of 3-4 hours when all those factors are considered. This will, in turn, affect the core damage frequency and the conditional containment failure probability of the complete station blackout sequence. If there is additional battery capacity not included in the Vemont Yankee Containment Safety Study, the Licensee should

, state so for clarification. '

Other important areas of considerable uncertainty are the manner in which the vessel may fail (and the amount of resulting steel in the core debris), the effectiveness of the reactor building for fission product attenuation, iodine may not be a good surrogate for risk estimation (refractory products from core concrete interaction may be better), and the containment loadings and response following vessel head failure.

Based on the above, we conclude the CCFP for VYNP is probably less than 50%.

2.2 Current NRC Concerns For U.S. Containments Severe accidents dominate the risk to the public associated with nuclear power plants. A fundamental objective of the Commission's Severe Accident Policy is to take all reasonable steps to reduce the chances of occurrence of a severe accident, and to assure the capability to mitigate the consequences of such an accident should one occur. The Reactor Safety Study report issued in 1975 found that for BWRs the probabilities of accidents resulting in core melt were low, but the containment performance following a severe accident was poor and tended to offset the benefits of low BWR core melt probabilities.

Subsequent actions resulting from the TMI Action Plan have led to several

i plant modifications and required improvements in plant procedures to further reduce the likelihood and consequences of severe accidents. In concert with the Connission's policy to further reduce the chances of occurrence of severe accidents and to mitigate their consequences, an industry initiative is underway to develop a methodology for Individual Plant Evaluations (IPEs) for use in the

. search for risk outliers. The resulting approach will be applied on a" plant specific basis. The staff may find that, while the IPE approach may satisfactorily address system reliability and containment performance for each plant specifically, the process will necessarily be a slow one. The staff has, therefore, identified several potential deterministic containment enhancements which lend themselves to generic implementation and have the potential to significantly mitigate the consequences of most severe accidents.

The generic approach has the advantage of expeditious implementation on all plants and will be responsive to Commission's policy regarding mitigation of the consequences of severe accidents.

Based on the insights gained from PRAs, the staff has identified the following five areas of potential BWR containment enhancements.

1. Provisions should be made for reliable operation of drywell containment sprays for a broad spectrum of accident sequences, including blackout sequences. The reliability of containment sprays should be enhanced by providing independent water and power sources. Backup water sources and pumps, hose connection and use of fire mains should be considered. The provisions to be implemented should minimize i

occupational exposures that could result from manual actuation, and procedures should be explicitly developed and expeditiously implemented as part of the BWR Owners Group development of the 1

Emergency Procedures Guidelines.

l 2. Provisions should be made for symptomatic response and reliable

. actuation of containment wetwell purge and vent valves. They should open and close under accident conditions as a means to assure that the beyond-design-basis events do not lead to ov n pressure failure of the containment and the selected vent paths. They should provide a path for releases which will maximize the use of the suppression i

pool as a condensing and filtering medium.

3. Emergency procedures and training should be reviewed and modified as f necessary to assure that operators are able to recognize severe accident conditions and use plant equipment to best advantage under such conditions. Revision 4 of the BWR Emergency Procedures guidelines i

should be implemented promptly following the staff's review and approval.

4. Paths for core debris travel should be evaluated for conditions repre-

, .sentative of a large scale core melt. Where the expected path of debris i travel indicates a substantial likelihood of loss of the suppression pool as a release filtering or debris quenching medium in BWR containments, the l torus room under the suppression pool should contain a 3 foot high barrier to trap water and core debris.

5. Combustible gas control provisions should provide substantial assurance that containment failure due to hydrogen combustion is not likely in the i

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potential severe accident sequences, including blackout sequences. The period of operations while containments are deinerted while at power, particularly during potential preshutdown conditions, should be minimized by reducing the present Technical Specifications pennitted value of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

, The licensee has evaluated the five areas as appropriate to Vermont Yankee and has found that with the exception of combustible gas control and debris

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control enhancements, the proposed enhancements could be beneficial for Vermont l I

Yankee severe accident mitigation capability. The staff recommends that the  !

licensee should reconsider the combustible gas control and debris control issues in the light of the following staff evaluation. The licensee should also complete the further studies proposed in its containment study report, propose specific modifications that it plans to undertake, and provide a j t

tentative schedule. l 2.2.1 Combustible Gas Control Vermont Yankee is allowed by its Technical Specifications to be deinerted for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at startup and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to shutdown. The staff has proposed the approach of minimizing the deinerting time, thus reducing the vulnerability of hydrogen combustion. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> deinerted period may be sufficient. Based on discussions with the license during a September 11, 1986 site visit, the staff understands that it takes about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to deinert the containment.

Therefore, it appears feasible to reduce the deinerting period in the Technical Specifications.

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-9 In the Vermont Yankee containment study the licensee has concluded based on operating experience that deinerted power operation is about 1.1 percent of the total operating time, and concludes that such a low percentage will not significantly impact safety.

Since the licensee was satisfied that its design objectives have been met, any

. impacts associated with shorter deinerting times were not identified. 'This issue should be explored further in order to address the staff's goal relating

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to increased assurance that the presence of a combustible mixture inside

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primary containment is reduced to as low as practicable levels. To this end, the study should include an evaluation of past practice to determine the impacts to plant operation due to a shorter allowable deinerting period.

2.2.2 Containment Sprays The licensee has identified and implemented means to improve the reliability /

availability of drywell sprays through the use o' the following installed features:

a) redundant sources of water to the containment sprays; namely, the torus, the ultimate heat sink, and the cooling tower basin. l b) redundant pumping capability for the containment sprays; namely, the RHR pumps and the SWS pumps.

c) the capability to operate the containment sprays during loss of a.c. power by using d.c. operated valves and a diesel driven i fire pump.

O 9 d) the capability to use the diesel driven fire pump to pump water through the RPV and into the containment during severe accidents that violate the RPV integrity. This mode could be used to cool the core debris and avoid over depressurizing the containment.

Additionally the licensee indicated the possibility of automating

,' the valves in the Reactor Building that are required to line vp the diesel fire pump with the containment and RPV sprays.

The licensee has concluded that it is practical to use the containment sprays to control containment pressure and temperature and cool core debris during a severe accident.

The staf.f concurs with the licensee that using containment sprays after a severe accident can assist in the control of containment pressure and temperature, scrub the containment atmosphere and cool the core debris.

However, in order to accomplish these objectives sufficient spray flow must be maintained. The diesel driven fire pump may have insufficient flow capacity to produce a spray that is effective for pressure and temperature control, scrubbing, and debris cooling. Insufficient flow to the sprays can result in water droplets that are too large to scrub or control pressure and temperature, and spray coverage may be insufficient to cover the core debris. Additional alternate water supplies (and/or spray nozzles) to the contaf ement sprays should be considered since the diesel driven fire pump cannot supply sufficient water to ensure fission product attenuation and heat removal. The IDCOR/BWROG study (Evaluation of proposed BWR Accident Mitigation Capacity Relative to Proposed NRC Change) of the use of sprays is in conformance with the staff objectives. Further, IDCOR/BWROG estimated that only 250 gpm is necessary

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to cool the core debris to prevent it from damaging the containment. Although 250 gpm may be sufficient to remove decay heat from the core debris when the spray is ideally distributed, a larger flow of water from the sprays may be required to establish a good spray distribution for fission product removal and to. remove the latent heat from the melted core and debris.

. The licensee's proposed usage of the containment sprays is a reliable and cost effective means for post severe accident control of containment pressure

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and temperature, scrubbing of the containment atmosphere, and cooling the core debris, and is consistent with the staff's views. However, in order to be assured of sufficient flow for severe accidents further study is recommended to better understand spray performance at derated flow conditions.

2.2.3 Venting With regard to the venting issue, the licensee discussed two of the most significant accident scenarios; station blackout and ATWS. The report concluded that containment venting should only be considered as a last resort.

The staff agrees that venting the noble gas activity is a procedure of last resort. The licensee also concluded that venting may be of little value in preventing core damage for an ATWS. In the case of station blackout, the licensee states that opening a wetwell vent is difficult and the potential benefits are limited. Therefore, the licensee concluded that containment venting is not considered to be practical.

Even though the report concluded unfavorably with regard to venting, it identified six possible vent paths from the existing piping arrangements.

After evaluating each path, a "12 inch vent path" was proposed as being the i

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most desirable. This proposed path bypasses the Standby Gas Treatment System filters. The modification would require that approximately 15 feet of 12-inch pipe be added and that one new motor-operated isolation value he added to the line. The proposed three-inch Atmospheric Control System line includes the Standby Gas Treatment System. For this pathway, additional vendor analysis

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and testing would be required for the reliability of the valves. Both paths would utilize the vent stack. The final selection was left open pending further study, which will include consideration of competing safety requirements.

None of the potential vent paths identified by the licensee has been shown capable of opening and closing at the pressures, temperatures, radiation, and steam environments associated with severe accidents. The issues of a.c.

independent power sources for the needed valves have not been evaluated by the licensee.

The staff agrees that further study should be performed. In the study, consideration should be given to more recent investigations such as the Evaluation of proposed BWR Accident Mitigation Capacity Relative to Proposed NRC Change, submitted by IDCOR/BWROG, dated August 1986.

In particular, the availability of adequate independent power under accident condition should be considered (station blackout). The Vermont Yankee study did not consider independent power sources because an earlier study (Harrington, R.M. and Hodge, S.A.; " Containment Venting as a Severe Accident Mitigation Technique for BWR Plants with Mark I Containment"; June 26, .

1986) concluded that wetwell venting during Station Blackout has limited beneficial potential for BWR Mark I containment plants. This limited benefit may be significantly understated because of key assumptions in their analysis

e 0 and, in addition, it should be weighed against the IDCOR finding that station blackout is a significant contributor to core melt probability. The study should also include an engineering evaluation that compares venting requirements with venting capabilities. The report stated that the need for substantial venting capability may arise during an ATWS scenario in which the energy in

. the form of steam that must be removed from the containment would be in the range of approximately 10% to 40% of rated reactor power. However, it is not clear if the vent paths noted earlier have this capability. The IDCOR study indicated that vent sizes for power levels of 20 to 30 percent may require vents of 26 to 34 inches in diameter, whereas the sizes required to remove decay heat after 10 minutes decay for other scenarios would range from 4 to 6 inches.

Another aspect of venting that can have a significant effect on accident consequences (onsite as well as offsite), and that needs further consideration, is the consequences of elevated releases.

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If the vent path includes the plant stack, the elevated release during venting can prevent the contamination of onsite buildings and equipment. As noted in the licensee's study, some vent paths have the potential for making portions of the plant inaccessible for accic'ent control and mitigation functions. Specifi- l cally, some of the vent paths have the potential fu releases into the Reactor and Contrcl Buildings. One advantage of an elevated release through the plant stack is that it maintains operator accessibility to vital plant areas. Elevated releases can also reduce substantially offsite dose consequences in the vicinity of the plant. Reduction factors of 10 or more are conceivable within two or three miles. Hence, the licensee should evaluate and quantify the specific l

benefits that can be achieved with respect to offsite dose consequences by venting through the plant stack. . Specifically, the licensee should consider site specific meteorology and topography in the evaluation of venting through the plant stack. The offsite atmosphere transport should extend to distances where the effects of stack height become negligible.

2.2.4 Core Debris Barriers The use of core debris barriers to prevent molten core debris from penetrating the steel containment shell or torus vent pipes has been investigated by the licensee. During initial construction the bottom of the steel drywell at Vermont Yankee was backfilled with concrete to El. 238'. Should core debris melt through the reactor it will fall on this concrete in a 17' -2" diameter enclosed area (231.5 ft 2) known as the Subpile Room. The only paths available for core debris to exit the Subpile Room are through its only doorway, or i

through either the three-inch steel floor drain or the three-inch steel equip-ment drain. The Subpile Room floor and equipment drain lines are imbedded in l

the drywell floor concrete and are routed to sumps inside the drywell. There is.a minimum of approximately 1.5 feet of concrete underneath the sumps and piping before reaching the steel drywell shell. The Subpile Room doorway is a curbless opening in the portion of the biological shield wall which supports

( the reactor skirt.

The licensee states that should core debris melt through the reactor onto the circular Subpile Room floor and stay molten with enough vertical head available to cause it to flow, it could pass out of the Subpile Room though

its only doorway, or through either the three-inch steel floor drain or the three-inch steel equipment drain. The Subpile Room floor and equipment drain lines are imbedded in the drywell floor concrete and are routed to sumps inside the drywell. If the entire volume core debris is available for spreading, a maximum layer thickness of approximately 1.1 inches would result. Since the

. vent line is approximately 1 foot above the drywell floor, it is not expected that the molten core debris could enter the torus through the vent pipes.

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The licensee concludes that Vermont Yankee analysis indicates that a scenario in which sufficient core debris melts and propagates to the contain-ment boundary does not appear feasible. Vermont Yankee has a similar size drywell and reactor vessel as the Peach Bottom plant analyzed in WASH-1400, yet has less than half the number of fuel assemblies and control rods.

Significant uncertainty exists in core and steel debris volume and transport analyses. The Vermont Yankee analysis indicates that core debris will not reach the drywell shell. Vennont Yankee already has diverse capability to spray water into the containment. The licensee states that additional physical barriers may be counter-productive as they may prevent containment spray from cooling the debris while it is confined in the Subpile Room. The Vermont Yankee analysis shows core debris control is closely coupled to contain-ment spray capability.

Our preliminary assessment is that full core debris barriers as considered by the licensee may not be cost effective for Vermont Y6nkee. However, the licensee should examine localized barriers protecting the downcomer opposite

the doorway of the Subpile room and two adjacent downcomers (one on each side).

Such a modification may not be costly, and will provide added assurance that core debris travel will not cause suppression pool failure or bypass.

2.2.5 Severe Accident Procedures Vermont Yankee's present symptom-oriented Emergency Operating Procedures (EPGs) are based upon the latest approved version of the BWROG EPG's (Revision 3). However, containment venting and reactor power control using water level for ATWS conditions were not included in the Vermont Yankee Emergency Procedures.

The licensee has indicated a willingness to consider updating the emergency oper.iting procedures to include the EPG Rev. 4 guidance on venting and ATWS power control. The licensee should implement the operating procedures currently being considered by industry in EPG Rev. 4, following NRC staff reviews. Any future addenda which may result from the current program of enhancing the containments capability to mitigate the consequences of severe accidents.

2:2.6 Other Improvements Other potentici containment performance improvements should also be considered. For example, the licensee should consider supplemental power supplies for improved autoenatic depressurization system operation for station blackout response. This is particularly important since loss of makeup at high pressure appears to be the largest contributor to core melt.

3.0 CONCLUSION

S The staff concludes that the Vermont Yankee containment safety study has provided evidence that the containment is more capable of performing its function during severe accidents than previous assessments of Mark I type containments would indicate. That is, given a core melt accidents, a prelimi-nary evaluation indicates a less than 50 percent likelihood of containment failure. The staff also concludes that the likelihood and consequences of

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such a failure may be reduced substantially with modest improvements as discussed above. The staff, therefore, recommends that the licensee perform the feasibility studies indicated by the study, and those proposed in this evaluation.

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ENCLOSURE 2 VERMONT YANKEE CONTAINMENT STUDY REQUEST FOR ADDITIONAL INFORMATION

. 1. What is your estimate of the overall uncertainty of conditional

containment failure probability and its basis?

2. What is the affect on core damage frequency when accident sequences TPUV, TPQUV, T E PQX, E T PQUV, Tg QUV, TPW are included in the dominant accident sequences based on reduced battery life, and the number and type of SRVs compared to Peach Bottom, and on the CCFP?

3. Given that the national average value for frequency of loss of offsite power is on the order of 0.22/yr justify, on the basis of the Bayesian estimate that the frequency of loss of offsite power at Vermont Yankee is 0.07/yr.

4. Verify that the total battery capacity availsble at Vermont Yankee is greater than 2175 ampere-hours, and that it could be maintained at a voltage greater than 1.75 volts / cell in high ambient temperature during the accident for 6-8 hours.

5. How often are the RHR/RHRSW interconnecting valves actuated to assure that the valves work properly?

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6. How often are the interconnecting valves between the RHRSW and the fire protection system (fire pumps) actuated to assure that the valve works properly?

7. How readily can the MSIVs be reopened following closure at operating conditions? What interlocks must be bypassed and how complicated are the procedures (e.g., must the differential pressure across the MSIVs be reduced for the valves to be re-opened)?

8. It is not clear how the CCFPs given or page 74 of the report were obtained. Please explain.

9. What SLCS modifications are proposed for V.Y.? Page 86 discusses the advantages of two different possible modifications, but gives no connitment to either.

10. Identify the testing and maintenance requirements you use for the diesel driven fire pump. Do these requirements conform to those contained in the flational Fire Codes? Also identify any reliability informatior. fer the >

system such as outages and failurer to start on demand. What outage time limitations do you use for the system while at power?

11. Identify the scope of modifications required to the spray system, or increases in the pumping capacity, to assure a uniformly distributed spray with proper droplet size (as opposed to a dribble) if the diesel fire pump were used in a core melt event. Approximately what would the costs be of such modifications?

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12. Can portable AC generators be used effectively to power vital valves and/or small pumps for station blackout accidents? If so, what modifications would be required, and what would be their approximate costs?

13. In section 2.2.1 you conclude that the containment can be " expected to withstand pressures approximately two times design prior to failure."

Provide the bases for your conclusion.

14. In section 2.2.10.3, isn't the water supply from at least a portion of the

- . cooling tcwers also available?

15. The use of the Vernon Hydroelectric Station is referenced in Section 2.2.11.1, and discussed in more detail in Section 4.4.2.3. Reliability estimates are presented on page 62. Please provide the basis for the reliability estimates with reference to both the historical operation of Vernoa Hydro, and the transmission line and substations to Vermont Yankee.

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16. The Nitrogen Containment Atmosphere Dilution (NCAD) system is referenced on pages 25 and 115. What maintenance and surveillance procedures are used to ensure operabilit'y?

. 17. In Section 4.1.4 MARCH /RMA and MAAP code package for Vermont Yankee is referenced. Were calculations made for Vermont Yankee, or were the results of computation for other reactors evaluated for the Vermont Yankee design? What calculations were made?

18. Invessel and exvessel steam explosions were not considered credible (pg.

55, 1st para.) based upon research. Identify the research that forms the basis for this conclusion. -

19. As we understand tables 4.7 through 4.10, the designations E, L, and NCL refer to early (E) or late (L) containment failure estimates, and NCL refers to no containment failure. The second designators H, M and L refer to high, medium and low releases, respectively. To what extent can mitigation through manual actions in the time available, and in the temperature and radiation environments associated with such accident i

types be expected to be successful for early failures; for late failures? 1 Specifically, for the combustible gas control, spray, and venting evalta- 1 i

tions, what do you judge the effectiveness of the existing plant and procedures to be versus the possible improvements for early and late sequences?

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20. NPSH during spraying is identified as a concern on page 117. To what extent will further investigation be undertaken to determine whether NPSH is an issue? Verify that procedures exist for the operator to lineup ECCS water sources outside the containment in the event NPSH requirements are not met. If analysis indicates it is an issue, what do you propo'se be done to eliminate or reduce the level of concern?

21. Venting is considered for the station blackout sequences only. Please discuss your rationale for not considering other events when venting may be beneficial.

22. Since there is a substantial difference between the heights of the VYNPS plant stack (318 feet) and the reference plant stack (500 feet), indicate how this was taken into account in the comparison of the two plants.  :

23. Evaluate the differences in offsite dose consequences due to venting at ground level versus through the stack. Using site specific meteorology and topography, provide an estimate of the offsite dose differences between the two types of release as a function of distance from the site.

24. On page 125, rapid containment depressurization which could fail the drywell is offered as an uncertainty relative to containment venting.

What analysis and/or tests are being conducted to reduce this uncertainty?

If no analysis or tests are contemplated, what actions are proposed to minimize the uncertainty?

25. Remote manual valve operation is discussed in Section 5.3.5.1.1.,

primarily with respect to station blackout. To what extent can the remote vent valve and any spray valve alignment be counted on for the other classes of sequences you assessed? That is, if remote manual operation is not available, would the local environment the operators would en' counter allow successful local operation?

26. Severe accident venting discussed in Section 5.4 does not include an evaluation of the reliability of the ADS system. Given the types of severe accident challenges you have described, provide your estimates of the reliability of ADS valves; i.e., their potential as a suppression pool bypass path. Can battery packs or portable generators be used to assure high reliability? If so, at what approximate cost?

27. What is the approximate cost for improving the valving for the diesel fire pump?

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28. For the improvement options you have evaluated, what maintenance and surveillance guidelines would you propose to use?

29. To what extent do you consider the option of drywell flooding to be effective? If effective, would you include the option in future revisions to your emergency operating procedures.

30. It is estimated that the maximum debris layer thickness on the drywell floor would be approximately 1.1 inches (page 136). Provide the bases for such a conclusion.

- 31. What is the thickness of the vent duct between torus and drywellf~ (Fig 1, Page F-6).

32. It is stated that " additional physical barriers are believed to be counter-productive as they may prevent containment spray from cooling the debris while it is confined in the Subpile Room." Please elaborate how such a barrier would prevent the spray from effectively cooling the debris. (p 137).

33. It is indicated that, with the exception of the one-inch nitrogen CAD line and the six-inch nitrogen purge line, the pipe lines associated with four potential vent paths are likely to fail. Provide background information which led to this conclusion (p 131).

34. It is implied that a layer of debris (1.1 inch thick) would not penetrate the drywell steel shell and enter the torus (p 136). If such is the ,

conclusion, please discuss why such a burn through is unlikely while core debris is attacking the dry well floor. There is a gap between drywell steel shell and concrete shield. If the molten core were to burn through

O the steel shell at the indicated corium elevation, what would prevent the fission gas from entering the reactor building since the concrete shield outside the drywell shell is not designed as a pressure boundary?

. 35. It is stated that 135 psia is a reasonable value for the VY Conta'inment failure pressure (page F-5) a) What is the uncertainty range associated with this value?

b) What would be a change in core melt and conditional containment failure probabilities associated with the uncertainty?

c) Provide references for the Ames and Sandia calculations mentioned in the Appendix F. (page F-1)

36. Your evaluation of deinerting indicates a relatively few hours of power operation while the containment is deinerted (i.e., about 1% in the RunMode). For each such instance, please identify the following:

a) the number of hours deinerted;

.b) the purpose for the deinerted condition, and whether it was successful; before shutdown was required by your technical specification, and c) the power level and its correspondirig reactor pressure at which containment entry and exit were made.

Please indicate the impacts you would expect for a deinerting technical specification to either 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

o 's o 9_

37. Please provide your estimates of pressure and temperature as a function of time for the accident sequences you analyzed for CCFP estimates.

38. It is not clear from your evaluation why the probability estimates of early failures with higher releases are lower than for Class IV e' vents.

Please explain if venting of ATWS sequences before core melting was assumed?

6

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