ML20236P356

Forwards Response to NRC Request for Addl Info & Clarification of Info Provided in Denial of NOV Re Insp Repts 50-361/97-22 & 50-362/97-22

ML20236P356
Person / Time
Site:	San Onofre
Issue date:	07/14/1998
From:	Nunn D SOUTHERN CALIFORNIA EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20236P359	List: ML20236P356 ML20236P393
References
50-361-97-22, 50-362-97-22, NUDOCS 9807160276
Download: ML20236P356 (22)

Text

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U EDISON 022;""

An LDISON INTE.RAA TIONAL" Compumy July 14,1998 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

Docket Nos. 50-361 and 50-362 l

Maintenance Rule Violation Supplemental Information Request San Onofre Nuclear Generating Station, Units 2 and 3

Reference:

Letter, Mr. A. T. Howell 111 (USNRC) to Mr. Harold B. Ray (SCE),

dated May 11,1998 The referenced letter provided the Nuclear Regulatory Commission's (NRC's) response to Southern California Edison's (SCE's) April 9,1998 denial of a notice of violation (9722-01).

This violation involved the failure to include the turbine building nonradioactive sumps system under the 10 CFR 50.65 monitoring program scope. In the NRC's response to SCE's denial of the violation, the NRC requested additional information, and clarification of l

information provided in the denial. We appreciate the opportunity to provide additional l

information which may help clarify the issue.

The enclosure to this letter provides SCE's response to the NRC request. SCE continues to believe that we were in compliance with the Maintenance Rule on this issue, and believe i

the violation should be withdrawn. As discussed with Mr. Dale Powers on June 5,1998, the due date for this response was extended due to the Institute of Nuclear Power Organization (INPO) site evaluation conducted during this period.

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If you have any further questions, please contact me.

Sincerel,

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9907160276 990714 PDR ADOCK 05000361 G

PDR Enclosures cc:

E. W. Merschoff, Regional Administrator, NRC Region IV J. A. Sloan, NRC Senior Resident inspector, San Onofre Units 2 and 3 J. W. Clifford, NRR Project Manager, San Onofre Units 2 and 3 J. Lieberman, Director, Office of Enforcement, NRC l' O. Ihn 128 San Clemente. CA 92b74 0128 714-368-1480

- l'a 714-3b8-1440

ENCLOSURE 1

Mr. A. T. Howell's letter, dated May 11,1998, requested additional information regarding SCE's denial of a Notice of Violation involving the scoping of the nonsafety-related nonradioactive sumps system under the Maintenance Rule. Specifically, the letter requested the following information:

1)

" we would need to understand the basis for concluding that any potential release would not be significant under the design basis conditions."

2)

" we request that you provide us with the degree of total release mitigation the sump system provides."

3)

"SCE further stated that you enhanced the program definition of " significance" and significant value;...We request that you provide details regarding the specific changes made to your program."

4)

"SCE's assertion that the realignment of the sump discharge is not significant to release mitigation refers the Safety Evaluation Report, Sections 2.4.9 and 15.4.12.

We do not understand how they pertain to the issue under review."

5)

"With respect to your violation response, which indicated that the turbine building l

sump system release mitigation function was placed in scope as of July 10,1996, and removed in October 1997,..Please provide us with copies of your records that document these actions."

The following additional information is provided to clarify these questions.

1.

" we would need to understand the basis for concluding that any potential release would not be significant under the design basis conditions."

l The basis for concluding that any potential release would not be significant under design basis conditions is the SONGS UFSAR Chapter 15 accident analysis of a Steam Generator Tube Rupture (SGTR) event (UFSAR 15.10.6.3.2).

Under design basis conditions as described in the accident analysis, there would be no radioactive effluent release to the turbine sumps and drains as a result of a SGTR event.

As a result, the turbine building sumps and drains provide no release mitigation. The UFSAR Chapter 15 analysis model treats all event radioactivity releases as being i

released directly to the atmosphere.

The Chapter 15 analysis of the radiological consequences of a SGTR considers the most severe release of secondary activity for that event. A SGTR event increases 3

contamination of the secondary system due to reactor coolant leakage through the tube,

I l

l ENCLOSURE break. The analysis assumes an instantaneous, double-ended tube rupture and a postulated coincident loss of offsite station power, which causes closure of the turbine bypass valves. The steam generator pressure will increase rapidly, resulting in steam discharge and radiological activity release through the main steam safety valves (MSSVs). The iodine and noble gas activity is released directly to the environment through the MSSVs. No credit is assigned for plateout, retention, or decay.

l Venting from the affected steam generator continues until the secondary steam pressure drops below the main steam safety valve setpoint and the safety valves close. At this time, the affected steam generator is isolated and no steam or activity is assumed to be released.

The SGTR emergency operating instruction (EOI) includes actions that mitigate offsite exposure by reducing this atmospheric release; however, sump operation is not one of the offsite dose mitigation steps. The EOls contain information for contingencies outside of the design basis of the Chapter 15 analysis and SCE considers the manipulation of the turbine building sumps and drains to fall into this contingency category.

The Chapter 15 SGTR analysis does not postulate any liquid releases and therefore does not postulate an exposure occurring from liquid releases (including contamination of surface or ground water). However, UFSAR Sections 15.7.3.3,15.10.7.3.3, and 2.4 consider potential releases due to tank ruptures. The turbine buildin0 floor is concrete and below grade. There is no path for liquid runoff to surface water. Even if leakage passed through the concrete into the soil, there are no users of ground water that could be affected. Therefore, from the accident analysis standpoint, the lack of operability of the sump discharge valve has no impact upon offsite exposures either liquid or gaseous.

The attached white paper analysis (Attachment C, provides additional details of this summary.

2.

" we request that you provide us with the degree of total release mitigation the sump system provides."

l During a design basis SGTR event, there is no flooding in the turbine building. The safety analysis only consider airborne release paths (UFSAR 15.10.6.3.2). As noted above, the SGTR emergency operating instruction (EOl) includes actions outside of the l

Chapter 15 analysis that mitigate offsite exposure by reducing this atmospheric release.

l The SGTR EOl, presumes that an operator can respond to the event prior to the closure j

of the main steam isolation valve. The EOl mitigates offsite exposure by directing the steam from the affected steam generator to the condenser using the steam bypass piping and valves. Bypassing the steam to the condenser will elevate the inventory of radiciodine in the condenser above normally encountered levels. - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.

ENCLOSURE To the extent that the condenser or its discharge piping leaks, a small amount of liquid could reach the turbine building floor and collect in the sumps. The turbine building would see approximately 150 gallons leakage from the condensate system as a result of normal condensate leakage for the 30 minute event period (UFSAR 11.1.7). The turbine i

building sumps (east and west) have a capacity of approximately 2600 gallons; therefore, they would easily contain the 150 gallons of normal leakage and no radiological release would need to be mitigated by the operation of the sumps and pumps.

In order to further validate the conclusion that neither the radiological consequences of the SGTR leakage in the turbine building are significant, nor the function of the sumps is significant to the release mitigation function of the EOl, SCE has performed an additional evaluation using a more conservative leakage scenario. The additional evaluation, which goes beyond the established NRC accepted design basis, assumes a SGTR, combined with a condensate system line break. This scenario conservatively releases up to 100 gallons per minute for 30 minutes to evaluate the hypothetical amount of radiciodine that could propagate to the sumps (prior to operator action to isolate the break). The potential discharge path to the environment is leakage through cracks in the basemat to the ground. The results of the evaluation show:

a) 0.027% of the SGTR event radiciodine in the steam generator liquid could reach the turbine building sump, b)

Liquid not contained in the isolated sump could also collect in the adjacent condensate piping trench up to a combined volume of approximately 35,000 gallons, which is sufficient to confine the entire conservatively postulated leakage amount of 3,000 gallons, c)

Any liquid leakage through any potential concrete basemat cracks to the ground has no groundwater exposure pathway, other than to the ocean, due to the existing groundwater gradient. Considering the horizontal permeability of the site's water bearing strata, 800 minutes would be required for any radioactivity to reach the beach. Thus, this pathway could not contribute to the 0-2 hour Exclusion Area Boundary (EAB) or 0-8 hour Low Population Zone exposures, d)

Even if the spilled liquid were considered to be released directly to the outfall, and if the steam generator liquid radioactivity were at 60uCi/g dose equivalent iodine (considering no dilution of the RCS leakage in the steam generator mass) the i

maximum liquid concentration in the ocean would be 5.5 E-7 uCi/g. This is about l

half the annual limit of 1.0 E-6 uCi/g for liquid concentration of 131 iodine in unrestricted areas under Appendix B to 10CFR20, e)

Resultant airborne exposure would be more than 3000-fold smaller (<.0001mR I l

J

1 l

l ENCLOSURE j

thyroid EAB) than the UFSAR SGTR airborne release through the MSSVs (2.8 l

mrem thyroid EAB).

This conservative, beyond the design basis evaluation, is also documented in the l

attached white paper analysis (Attachment 1).

t In summary, under design basis conditions, any effluent release to the turbine building as a result of a SGTR would be minute, and easily contained within the sumps. As a l

result, the sumps provide no radioactive release mitigation. Beyond the design basis, a postulated condensate line break, concurrent with the SGTR, would result in any liquid release being confined in the sumps and the adjacent pipe trench. Any leakage which escaped the sumps and pipe trench and seeped through any potential cracks in the concrete to the ground would propagate slowly through ground strata directly to the j

ocean and not affect any potential water users. In addition, the airborne release in this scenario is estimated to be 3000-fold smaller (<.0001 mR thyroid EAB) than the UFSAR SGTR airborne release through the MSSVs (2.8 mR thyroid EAB). SCE considers a radioactive release 1/3000 of the overall SGTR release to be of little significance.

3.

"SCE further stated that you enhanced the program definition of " significance" and "significant value;..We request that you provide details regarding the specific changes made to your program."

Since the baseline inspection, SCE has revised its procedure SO123-XIV-5.3.1, " Scoping for the Maintenance Rule," to strengthen the guidance for determining if a non-safety related SSC referenced in an EOl can be excluded from the scope of the Rule.

Previous definition:

I SO123-XIV-5.3.1, Section 6.6.2.1, stated: "The SSC must add value to the mitigation function of the EOl by providing the total or substantial fraction of the total function ability required to mitigate core damage or radioactive release."

Revised definition:

SO123-XIV-5.3.1, Section 6.6.2.1, now states: "The criteria applied for significant l

mitigating value is that the SSC has to perform directly or directly lead the l

operator to take actions to:

" 1.1 Maintain integrity of one or more of the fission product barriers, i.e., fuel cladding, RCPB or containment, "OR _______________A

ENCLOSURE

" 1.2 Prevent the release of radioactive materials to the environment which are evaluated to be >100 mrem, which is the annual exposure limit given in 10CFR20.1301 for members of the public.

" 2. EOl steps which are for convenience, or evaluation, or that represent good practices in assisting operators in returning the plant to a preferred condition after the plant has been secured following the transient are not considered to be mitigating steps."

4.

"SCE's assertion that the realignment of the sump discharge is not significant to release mitigation refers the Safety Evaluation Report, Sections 2.4.9 and 15.4.12. We do not understand how they pertain to the issue under review."

SCE's reference to Safety Evaluation Report (SER) sections 2.4.9 and 15.4.12 pertain to the issue under review because they confirm the SONGS licensing basis for accidental liquid radionuclides releases. SER section 2.4.9 discusses the flow of groundwater under SONGS (and therefore the path of any liquid release into the groundwater). The section indicates that the ground water gradient flows into the ocean and there are no groundwater users downgradient of SONGS. The section also indicates that there is no potential for reversal of the groundwater gradient. In effect, any potential liquid radiological release will drain to the ocean.

SER Section 15.4.12 discusses the postulated radioactive releases due to liquid tank failures, and it applies to all types of ground water releases. This section states that "We

[NRC) determined that there are no ground water users downgradient from potential releases due to liquid tank failures. Therefore, we did not calculate ground water radioactivity concentrations for potential receptors" and "..we conclude that the provisions incorporated in the applicants' design to mitigate the effects of component failures involving contaminated liquids is acceptable."

SCE's safety analyses are consistent with the NRC's 'SER evaluation that accidental liquid release pathways are not significant at SONGS. The worst case postulated radioactive liquid effluent leakage at SONGS is considered acceptable, under 10CFR20 limits. Any leakage occurring at the turbine building, as a result of a SGTR, would be bounded by this analysis.

5.

"With respect to your violation response, which indicated that the turbine building sump system release mitigation function was placed in scope as of July 10,1996, and removed in October 1997,..Please provide us with copies of your records that document these actions."

r l

ENCLOSURE As requested, SCE has attached applicable meeting minutes and Maintenance Rule l

scope listings that identify the turbine building (non-radioactive) sumps as initially within the scope of the Rule, and then subsequently removed from the scope of the Maintenance Rule in October 1997 (Attachments 3 and 4).

[

Summary As noted in inspection Report 50-361 & 50-362/97-22 (Reference), SCE's basis for some scoping decisions (i.e., turbine building nonradiological sump system) may not have l

been well documented. Nonetheless, SCE believes that excluding the system from the Maintenance Rule scope was appropriate, based on the function of the system. SCE's Expert Panel considered the function of the system when it determined that the turbine building nonradiological sumps system was not required to be included in the Maintenance Rule scope. As is the nature of an Expert Panel, judgment is used when making decisions. The Expert Panel took several steps to revalidate this decision, including quantifying the significance of SGTR radiological release mitigation. Based on our re-review of this issue, and the additional evaluation performed, it is SCE's position that our Expert Panel made the correct decision when removing the system from the scope of the Maintenance Rule. Therefore, SCE continues to believe that the we were in compliance with the Maintenance Rule on this issue, and believe the violation should be withdrawn.

ATTACHMENTS 1.

NEDO-IPRE-98-001, " Source Term-Based Assessment of the Risk Significance of Turbine Building Releases Following Steam Generator Tube Rupture Events" 2.

Excerpt from Site Technical Services procedure SO123-XIV-5.3.1, " Scoping for the Maintenance Rule" (applicable section 6.6) 3.

Maintenance Rule Expert Panel Meeting Minutes from April 11,1996 (Sumps included in the Rule) 4.

Maintenance Ru' ? coping Summary Matrix STS-SO123-2001, Revision 0

5. Maintenance Rule Scoping Summary Matrix STS-SO123-2001, Revision 1 6.

Maintenance Rule Expert Panel Meeting Minutes from October 17,1997 (Sumps removed from the Rule) 7.

San Onofre 2 & 3 UFSAR, Section 15.10.6.3.2 " Steam Generator Tube Rupture with Concurrent Failure of a Single Active Component" -

S ArrggWar i A

Source Term-Based Assessment of the Risk Significance of Turbine Building Sump Releases Following Steam Generator Tube Rupture Events I.

SUMMARY

Source term reduction afforded by the condenser and its hotwell is the significant component of offsite dose reduction achieved by operator actions following a steam generator tube rupture (SGTR). Operator actions to realign the turbine building sump discharges as part of EOl SO23-

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12-4 are taken to control contamination and are not a significant source of offsite dose reduction.

j While existing safety analyses show that the post-SGTR dose pathway is airbome, only, this paper myiews souwe terms for both airbome and liquid releases. A recent NRC Inspection Repon and a Notice of Violation have focused concem about the scoping of the turbine building sumps and their appurtenances needed for system alignment, in pan due to a role they are thought to have in mitigating liquid releases.

His evaluation considers both the amount of the potential radioactivity release to the sumps as well as the potential pathways for release from the sump to the environment and subsequent effects. The potential radioactive release to the sump is shown to be below 10 CFR 20 limits.

He SONGS criteria for determining release significance, with respect to the Maintenance Rule, has been established as 10 CFR 20 limits.

De floor of the turbine building is below grade. There are only two paths for liquid releases to reach the environment: either by transfer to the circulating water system (CWS) or by leakage i

through the floor and contamination of groundwater. The CWS pathway leads to the ocean and provides substantial, additional source term reduction due to dilution. He groundwater gr$dient is towards the ocean. De underlying strata has a sufficiently low permeability such that

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significant offsite exposure is precluded. There is no groundwater release path onshore.

Therefore, any SGTR-related liquid releases that collect in the sump are not significant to the accident exposure.

Accordingly, under the guidance of NUMARC 93-01, it is not necessary to include the sumps or their appurtenances in the SONGS Maintenance Rule Program.

IL INTRODUCTION AND BACKG.ROUND 10 CFR 50.65(b)(2)(1) states that non-safety related structures, systems, or components that are relied upon to mitigate accidents or transients or are used in plant emergency procedures shall be included in the Maintenance Rule monitoring program. NUMARC 93-01, Revision 2 provides guidance that such items are to be included in the Maintenance Rule monitoring program if they contribute significantly to the mitigatiot Nction of the system.

1 l

'Ihe SGTR EOI, SO23-12-4, Step 22, directs the operator to realign the turbine building sumps.

A recent NRC inspection, December 1-5,1997, expressed concem that there was inadequate justification for the turbine building sumps to be excluded from the Maintenance Rule program.

This was discussed in the Exit Interview of December 5,1997, in the SONGS letter to the NRC 4

of January 2,1998, and in a follow-up Telephone Exit Interview on February 4,1998. Further discussion was included in the Inspection Report and Notice of Violation provided by letter on March 2,1998.

This document provides a source term-based assessment of the risk significance of the operator action in the EOI. This is not a quality-affecting calculation. Rather, it is an assessment in which the baseline UFSAR Steam Generator Tube Rupture analysis and the baseline UFSAR Liquid Release Event analysis are compared to the results that might be expected as a result of operator intervention as contemplated by the EOI. Thb assessment is prepared in order to determine the safety benefit of the operator's action.

III.

METHODOLOGY Section IV presents a summary of the UFSAR Chapter 15 SGTR event analysis. To assess the impact of operator actions, not credited in the UFSAR, that would be taken in accordance with the Emergency Operating Instructions, a presumption of system operation was made. The basis for this operation is provided in Section VI.Section VI.B describes the operation of systems and transport of radioactivity associated with this opemtion. Analysis of the EOI-based radiciodine transport yields release fractions relative to the UFSAR Chapter 15 analysis.

Postulated EOI-based doses are calculated by ratioing the UFSAR Chapter 15 SGTR analysis dose by the fraction represented by the EOI radiciodine release divided by the UFSAR radiciodine release. Thus, Regulatory Guide 1.4 assumptions regarding occupancy, bpathing rates, meteorology, and dose conversion are maintained applicable to the EOI-based assessment.

Site features affect radioactivity transport. Hydrogeological bases and data related to the UFSAR analyses of liquid releases are provided in Section V.

IV.

UFSAR SGTR EVENT (UFSAR 15.10.6)

EA Seouence of Events and Systems Ooeration Integrity of the barrier between the RCS and main steam system is significant from a radiological standpoint since a leaking steam generator tube would allow transport of reactor coolant into the main steam system. Radioactivity contained in the reactor coolant would mix with shell-side water in the affected steam generator. During normal plant operations, some of this radioactivity I

would be transported through the turbine to the condenser where the noncondensible radioactive l

materials would be released via the condenser air ejectors.

2 1

J Following a reactor trip, the main steam system pressure would increase to the point where the turbine bypass valves would open to control the main steam system pressure. If turbine bypass is unavailable, the steam generator safety valves would open to control the main steam system pressure. He operator can isolate the damaged steam generator and cool the NSSS using manual operation of the auxilian feedwater and the atmospheric steam dump valve of the unaffected steam generator any time after reactor trip occurs. In order to maximize the calculated radioactivity in the seconday system, this analysis conservatively assumes that operator action is delayed until 30 minutes after the first indication of the event.

Diagnosis of this accident would be facilitated by radiation monitors in the blowdown sample lines from each steam generator, in the blowdown processing system neutralization sump discharge sea line which processes blowdown from both steam generators, in the condenser air ejector discharge line, and adjacent to the main steam lines.

For leak rates up to the capacity of the charging pumps in the CVCS, reactor coolant inventoy can be maintained and an automatic reactor trip would not occur. During the first 30 minutes of the accident, a reactor trip is not necessary because the safety limits are not approached and thre is no danger of violating dose limits.

Technical Specification 3.4.13 specifies that the plant is to be shutdown within 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> following detection of leaks greater than 1 gal / min. In addition, the plant emergency procedures provide for rapid plant shutdown usireg operator action in the event of a tube failure. The UFSAR assumes a 30-minute operator action interval before the operator takes actions to ramp down the plant or to manually trip the reactor and place the plant in cold shutdown.

IM UFSAR SGTR Source Yransnort Model I

The analysis of the radiological consequences of a steam generator tube mpture considers the most severe release of secondary activity as well as reactor activity leaked from the tube break.

De inventory of iodine and noble gas fiss,jon product activity available for release to the environment is a function of the primary-to-secondary coolant leakage rate, the percentage of defective fuel in the core, and the mass of steam discharged to the environment.

An SGTR increases contamination of the secondary system due to reactor coolant leakage through the tube break. The coincident loss of off:;ite station power causes closure of the turbine bypass valves to protect the condenser. The steam generator pressure will increase rapidly, resulting in steam discharge as well as activity release through the main steam safety valves.

Venting from the affected steam generator continues until the secondary steam pressure drops below the main steam safety valve setpoint and the safety valves close. At this time, the affected steam generator is effectively isolated, and, thereafter, no steam or activity is assumed to be released from the affected steam generator. He remaining unaffected steam generator removes core decay heat by venting steam through the atmospheric dump valve and steam-driven auxiliary feedwater pump exhaust until cooldown can be accomplished with the shutdown cooling system.

3

FSAR t

Based

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SGTR SS ca**aus Source

'70.'.

Term

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Model

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• 47,700 awn J k 1..M far feG nf.

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so7R Sc uou6d De iodine and noble gas activity is released directly to the environment with no credit for plateout, retention, or decay. He results of the analysis are based on the most direct leakage pathway available. Herefore, the resultant radiological consequences represent the most conservative estimate of the potential integrated dose due to the postulated steam generator tube mpture. He principal assumptions u generator tube ruptum are as follobs: sed to evaluate the radblogical consequences o (1)

Reactor coolant equilibrium activities are based on the Technical Specification Limits.

(2) ne limiting case analyzed included a coincident (pre-existing) iodine spike of 60 uci/g dose equivalent iodine.

(3)

Steam generator equilibrium activity for both steam generators is assumed to be equal to the Technical Specification limit.

(4)

Tube mpture of the steam generator is assumed to be a double-ended severance of a single steam generator tube.

(5)

A steam generator radiciodine partition factor (PF) of 0.01 is used.

4

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1 l

I E EOI Radiological Consequences 4

EAB airbome exposures,0 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, were reponed in UFSAR Section 15.10.6 as 2.8 Rem Thyroid, less than 0.1 Rem Beta Skin, and less than 0.1 Rem Whole Body Gamma. The SGTR analysis does not contemplate any liquid releases and therefore does not postulate any exposures occurring from liquid releases (including contamination of surface or ground water).

1 V.

UFSAR Design Basis Liquid Release Event (UFSAR 2.4.13.3) y_,A Groundwater Re plant lies over the San Onofre Valley groundwater basin. The basin is wholly within the i

l boundaries of the United States Marine Corps Camp Pendleton. He Marine Corps has exclusive jurisdiction over the basin. It is Marine Corps policy to maintain the groundwater table throughout Camp Pendleton sufficiently above mean sea level to eliminate the possibility of saline l

water intrusion from the ocean into the freshwater aquifers.

Here is a groundwater gradient towr.rds the ocean of approximately 0.4 to 0.6 percent. He nearest water wells supply the Marine Corps and are located in San Onofre Creek, over one mile ininnd from the plant site. The established minimum pumping level for the San Onofre Creek wells is above the elevation of the water table at the SONGS site. Rus, a seaward gradient will exist even under extreme pumping conditions. In Section 2.4.9 of tne SER, the NRC concluded that there is no potential for reversal of the groundwater gradient at the site. Based upon this gradient, groundwater movement from the site toward any present or projected users will not j

occur.

l Therefore, it was established in the UFSAR and SER that an analysis bf the effects of an l

accidental release of liquid radioactive material to groundwater was net required. There is no l

liquid release exposure pathway for contaminated groundwater, other than to the ocean.

1 y2 Surface Water Section 2.4.12 of the UFSAR notes that there are no credible accidents that can occur which will result in an accidental liquid release of radioactive effluents as related to existing or potential i

l future water users. As discussed in Section 2.4.12, the radioactive waste discharge line to the l

circulating water system outfall is the only release path for radioactive effluent discharges into the surface water in an unrestricted area.

The maximum pumping flow of the radioactive waste discharge line is 140 gal / min. The normal flow rate in the circulating water system is approximately 830,000 gal / min during full power operation. Therefore, an accidental release of radioactive effluent would result in about a 5900:1 dilution within the circulating water system. The initial dilution of circulating water being 5

r

discharged through the outfall diffusers is about ten times the total volume of flow. Derefore, any accidentally discharged radioactive effluent would be diluted about 59,000:1 in the near field zone.

VI.

EOI SGTR EVENT

.V_LA EOI SGTR Secuence of Events and Systems Ooeration The event is expected to proceed as noted for the UFSAR analysis, except for the following item:

Instead of permitting steam to be released through the main steam safety or atmospheric dump valves, it is assumed that the operator retains sufficient controls to operate the steam bypass control system valves and can send the steam to the condenser.

As a prudent measure to limit inadvenent contamination, EOI SO23-12-4, Step 22, routes the turbine sumps to radwaste until such time as sampling and surveys confirm sump activity.

.YLH EOI SGTR Source Transoort Model Source transpon under the EOI is expected to produce identical concentrations of radioactivity in the main steam lines as in the UFSAR analysis. He UFSAR analysis limits the radiciodine panition factor to 0.01. This is a factor of 10 greater than the expected minimum panition factor of 0.001, as discussed in UFSAR Section 10.3.5.3. However, with control of the plant and normal electrical power systems, the operator can take steps to limit offsite releases even with this conservatism. Steam bypass will diven steam directly to the condenser. For consistency and conservatism, the panition factor in the condenser is chosen to be the same as the partition factor in the steam generators. His has the effect, in the analysis, of maximizing iodine transpon to the sump.

Although steam will continue to be released to the atmosphere from the turbine-driven auxiliary feedwater pump exhaust, this assessment considers that half of the exhaust steam came from the affected generator and that this exhaust steam is directed to the condenser. This also conservatively overestimates the radiciodine source input to the condenser.

It is assumed that 100 percent of the noble gases and I percent of the radiciodine in the steam will remain volatile and be released by the condenser air ejector to the atmosphere. This air ejector radiciodine release fraction is equivalent to 0.01 percent of the affected steam generator liquid radiciodine concentration. Iodine filtration at the condenser air ejector is available, but has not been specifically considered for this model, for simplification.

Ninety-nine percent of the radiciodine in the steam is assumed to be directed to and diluted in the condenser hotwell. His amounts to a source (Curie) input to the condenser of 0.99 percent of the RCS release of radiciodine during the SGTR. This model neglects dilution within the steam 1

f 6

1 generator and treats the RCS release as if it were an undiluted release, subject only to iodine species partitioning. A rigorous evaluation would apply dilutica and decay through a series of time-and flow-based differential equations. However, in the inter st of simplification, this model treats the condenser hotwell as being subject to an instantaneous, step increase in radioiodine.

As shown in the accompanying EOI-Based SGTR Source Term Model figure, to take into account the dilution provided by the hotwell, the steam release of approximately 47,700 lbm is diluted in the sum of the steam release and the condenser mass of 835,000 lbm. The average equivalent liquid release to the condenser duririg the 30 minute release from the affected steam generator would be 47,700 lbm /(8.35 lbm/ gal) - 57!3 gal / 30 min or approximately 200 gal / min.

The UFSAR SGTR analysis demonstrates that steam is not released before the reactor and turbine

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trip. Therefore, feedwater demand is abruptly reduced. On average, feedwater Dow will be ec,ual the steam flow from this point onward. It is assumed that this feedwater will continue to be routed from the condenser via the condensate polishers. Therefore, the flow rate of potentially contaminated liquid from the condenser to the condensate polishers will be equal to the steaming rate, e.g.,200 gal / min.

Since the condenser hotwell inflow and outflow rates are assumed ioentical, the concentration of any condensate routed to the condensate polishers or other cleanup systems will be, at most,

[47,700 / (47,700+835,000)]* (100 percent) - 5.4 percent of the assumed final (30 minute) inventory of radiciodine in the condenser hotwell. As the hotwell is presumed to contain 0.99 percent of the steam generator liquid radiciodine source term, the total radiciodine source term that is assumed to be directed towards the condensate polishers is 0.054 percent of the steam generator liquid radiciodine source term.

If it were assumed tha850 percent of such transfers were to leak prior to cleanup, a maximum 0.027 percent of the SGTR event cdiciodine in the steam generator liquid could reach the turbine building sump (the radiciodine concentration would be the same as the hotwell). This presents the potential to release from the sump (or the turbine building floor) to the atmosphere, using the

{

4 established 0.01 partition factor,2.7E-4 percent of the steam generator liquid radiciodine activity.

Be balance,0.027 percent. is assumed to be puddled on the floor of the turbine building or in the sump.

He following figure summarizes the model.

l I

7 i

l l

EOl Based A *'a' SGTR Source 5'o ca==

> gg n

Term

. %,.al,e Model AlmoeWiere Condenser Rosessee 47.F00 Run 0 % NaMae 0.90 % lodnee

{

j L7E-4% loense 1 r Hot Well 0.01 PF for tt

... Dilucon

&01 PF

{

RCS o.ase %loem so % Lama =9' Sump N

tort %leense Ese PF soTn sta Liquid 1 f Condenamie i f Polleheng Dem6nermilaars torF %Indene Lieuld Although the radiation monitor that isolates sump discharges upon high radioactivity is scoped within the Maintenance Rule, it is important to assess the overall significance of liquid discharge following an SGTR 'Ihe sump liquid is intended to be transferred to the liquid radwaste system (LRS), once operators can survey the liquid. Ifit were transferred to the LRS when the LRS was inadvenently aligned for ocean discharge (regardless of thdiation monitoring), circulating water system and outfall dilution of 59,000:1 would suggest that, at most, the ocean concentration could be 0.027 / 59000 - 5 E-7 percent of the steam generator liquid activity. As shown in Section VI.C, even this level of discharge meets 10 CFR 20 limits. Hence, regardless of the effectiveness of the radiation monitor and isolation of the sump discharge, any direct SGTR-related transfer of secondary liquids to the ocean from the sump would not be significant.

If no sump transfer takes place, there is a potential to contaminate the groundwater. Due to the groundwater gradient, there is no groundwater exposure pathway, other than to the ocean.

UFSAR Section 2.4.13 notes that the water bearing strata of the San Onofre Valley gmundwater basin is the San Mateo Formation and that this formation exhibits an average horizontal permeability of 0.025 ft/ min (40 ~ min /ft). The turbine Door and sumps are below grade and are at least 20 feet away from the beach. Even if liquid were to readily seep through cracks to bypass the relatively impermeable concrete basemat of the turbine building, it would still require a minimum of 800 minutes before the radioactivity could reach the strip of the tidal beach 8

immediately adjacent to the plant. This indicates that this pathway could not contribute to the 0-2 I

hour Exclusion Area Boundary or 0-8 hour Low Population Zone exposures.

Y1C EOl Radiological Consequences Noble gas-related radiological consequences will be essentially unchanged from the UFSAR presentation. A specific determination of airbome offsite exposure resulting from iodine evolution from the sump, or turbine building floor, has not been calculated inasmuch as the amount of such evolution is a minute fraction of the steam source term. Dispersion and transit time from the release point to the offsite dose receptor is expected to be much greater than that described in the UFSAR. However, the ratio of percentage of radiciodine released, 1:0.00027, indicates that there will be more than a 3000-fold smaller impact in airbome offsite exposure from that given in the UFSAR (2.8 Rem Thyroid). His is certainly a small portion of the annual exposure limit of 100 mrem given in 10 CFR 20.1301 for members of the public.

Herefore, airbome evolution of radiciodine related to sump operation or non-operation is not risk-significant.

He region near the circulating water outfall is not normally occupied. The tremendous release dilution at the outfall in conjunction with the low,0.027 percent, portion of the steam generator liquid activity assumed directed to this pathway provides an exposure at any likely dose receptors so far below the airbome dose already assumed in the accident analysis that any liquid pathway contribution would be meaningless.

Validation of the magnitude of this exposure can be seen by evaluating the magnitude of the concentration at the outfall, alone, without crediting occupancy reduction factors. As the sump has an iodine activity concentration equal to about 0.054% of the SG liquid iodine activity I

concen: ration, the factor of 59,000 dilution reduces the CWS release to a level of 9.2E-7 percent of the SG liquid iodine activity concentration. Even if the steam generator liquid'radiciodine I

activity were at 60 uCi/g dose equivalent 131-iodine, that is, assuming no credit at all for dilution o

of the RCS leakage in the steam generator mass, this would create a maximum liquid concentration in the ocean of (9.2E-9)(60) - 5.5 E-7 uCi/g. His is approximately half the annual limit of 1.0 E-6 uCi/g for liquid concentration of 131-iodine in unrestricted areas under Appendix B to 10 CFR 20. When occupancy factors are considered along with the radioactive decay and dispersion that would substantially reduce the concentration prior to reaching a dose receptor, it is clear that this pathway dose is not significant. Herefore, direct liquid release thmugh the radwaste system is not risk-significant.

In Section 2.1.2 of the SER, the NRC concluded that SONGS has the authority to determine all activities within the ex:lusion area as required by 10 CFR Part 100. Control of the strip of tidal beach has been adjudicated in a Commission proceeding (see ALAB-432) and has been declared "de minimis" on the basis of its occasional use, together with analyses which indicate that any radiation exposure to individuals in this zone will be within the guideline values of 10 CFR Pan 100 in the event of emergency. The 800 minutes needed for contaminated groundwater to l

l 9

1 pem1eate to the tidal beach is in excess of the time needed to exclude the public from the beach area. Herefore, the EAB exposure pathway dose would be zero and the p,remise for the ALAB Decision remains valid.

I Residual seepage into the ocean would be readily diluted and subject to decay before it could reach other dose receptors at the low population zone boundary. Therefore, the groundwater release pathway is not risk significant.

His finding is consistent with the NRC Safety Evaluation Report which stated: "We determined that there are no ground water users down gradient from potential liquid releases due to liquid tank failures. Therefore, we did not calculate ground water radioactivity concentrations for potential receptors."

VIL CONCLUSION He SGTR EOI release estimates include substantial conservatism - especially regarding the potential condensate leakage estimate of 50 percent. Even so, the analysis demonstrates that sump alignment in the EOIis not a significant safety action. Any alignment of the sumps, whether intentional or caused by equipment failure, produces offsite exposure within the annual limits of 10 CFR 20 following an SGTR. Herefore, it is not necessary to include the sumps or their appunenances in the SONGS Maintenance Rule Program.

A sensitivity study was prepared to assess variation in offsite exposure as a function of secondary leakage and iodine partition factor. Secondary leakage would be expected to be at normal levels, not the high leakage rate considered in the evaluation. He partition factor would be expected to be consistent with that used for normal releases,0.001, as described in UFSAR Section 11. rather than the bounding limit of 0.01 established in UFSAR Chapter 15.

he study demonstrates that for either normal or accident-based partition factors over a range of secondary leakage from normal (5 GPM) to 100 percent of the steam bypass control system flow, offsite exposures remain well below the reference limits of either the UFSAR SGTR accident analysis or 10 CFR Part 20.1301 for annual exposures. ne offsite 0-2 hour EAB exposure described in this assessment is indicated by the 0.01 partition factor,50 percent leakage case shown in the sensitivity analysis.

Hence, sump or turbine building floor airbome evolution of radiciodine is not of regulatory concem and sump operability does not significantly affect offsite airbome exposures. The following figure presents the results of the sensitivity study.

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VIII. APPENDIX VIII.A Design. Inputs

l. Radiological consequence assumptions of Section III.B: UFSAR Section 15.10.6.3 and Table 15.10.6.3.2-3, Amendment 13.

2. Primary steam release (47,700 lbm): Ref ABB-CE design analysis 1370-TS-109 as used in Design Input 4.19 of Rev 2 of SONGS Calc N-4075-004 for 43,703 lbm released fmm affected steam generator. Per Assumption 3.22 of Rev 2 to N-4075-004, turbine-driven auxiliary feedwater pump receives steam equally from each steam generator. Per Design Input 4.20 of the calculation,8071 lbm is vented from turbine-driven auxiliary feedwater pump. Hence, the event release from the affected generator is 43,703 + (8071/

2) - 43,703 + 4035 - 47,738 - approximately 47,700 lbm.

3. UFSAR Doses (2.8 R 'Ihyroid, >0.1 R Beta Skin and Whole Body Gamma): UFSAR

\\

Table 15.10.6.3.2-4, Amendment 13.

4. Gmundwater Gradient (0.4 to 0.6 percent): UFSAR Section 2.4.13.3, Amendment 13.

5. LRS discharge line flow (140 GPM): UFSAR Section 2.4.12, Amendment 13.

6. Circulating water flow rate (830,000 GPM): UFSAR Section 2.4.12, Amendment 13.

7. Ten-fold dilution at outfall diffusers: UFSAR Section 2.4.12, Amendment 13.

8. Condenser inventory (835,000 lbm): Per SD-SO23-240, normal hotwell volume is 96,600 gallons. Considering inventory in piping and normal fluctuations, round to 8

100,000 gallons. I gallon - 8.35 lbm [(1 g/ce) * (1000 cc/ liter) * (3.7853 liters / gallon) *-

(1 lb / 453.6 g)). 100,000 gallons - 835,000 lbm.

i l

i I

12

A STn c p e r 2_

{

v. rEna,ce:sOm-vi-1.0.

TEMPORARY CHANGE NOTICE Page 1 of h (PERMANENT WHEN FINAL APPROVED)

1) g the Docurnent is QA Program Affecting,2130 a Tectinical Specification /Lloonese Controlled Specification Violation will occur if NOTES:

CFDM final oval is not obtekted teethin 14 days from TCN date of leauences

2) y only Ectt Corrections are required, Hig0 form PF(123) 111 should be used (refer to SO123 Vl 1.0.11 Part A - ForCont Use Only:Qg}g lasuance Date Single Use TCN Cancels On TCN No.

04 Copy forwarded to Nuclear Training Divason AlWS Coordinator. E-50,if Part 8, Line 7 la marked YES.

Part 8

1. Docurnent No.

SO123-XIV-5.11 Revision No.

00 SingleUseTCN: YesO No Document Title Scoolna for the Maintenance Rule Document Author / Originator __

Dave Schafer 86906 03/27/98_ Site Technical Services PRitif CA TYP6 heeME PA.E DATE ORGANi2AfiON

2. TCN Devastion Approval (R for TCN numbers over 5): APPROVED BY:

N/A CFDM ter emelyant ssQhAfuRE/ IP Bf TELECON, PRedl NAMI ANo so sf ATE DArE/TJWE

3. Check appropriate box:

Entire Document Attactied O AffectedPage(s) Attached Supersededancorporated TCN(s)/EC(s) 0-3 Peen #er&PM (NoteMSe TCNs)

NutSER OP NUNE,30 sTA TE)

' " ~ '

4. This Charige Carirlot waI uritil the tieXt revWn Cf the Documerit arid is retMed:

a O To impiement face desian chanoe (oca,M,.uM.ete >

liar 311998 APR 021998 r

Facirny design chen0e identAer

/

esoCAiE DCP, MMP PFC.1F4 ETC.

imsememanon of tne inca,dwgn has been eete,mi,ied.

COPY SITE 1%"00PY Yes O No O St No, then a TCN cannot tie approved uned the aumen change has boon t

Other (e.g CAR, NRC Comirwments) Speerse Reason:

Resoonse to NOV datod OT02/98. from D. A. Powers to R B. Ray.

Maintenance Rule (MR) hrono-Turbine Rmria norwar5=+ve sumos.

Description of Change (s) Use Reverse Sade if Hogu6scf Provide stronandance for 11 dr6 Mac if a nonaafety related SSC referenced in the EOl can be ewtded from the scoce of tie MR. and 2) c6 Mao the definrtion of siontficant value to the functon of the eof _

/

5. Could implementation of this change pose adrerne envoonmental effects of any type directly or indirecsty?

YesO nog tif Yea, then a TCN is not aulnanzod unW a renew nom Erwirarwnental Protection a otnamed. Reier to SO123-V8-1.3)

6. Revleur requested from other organtrations/ disciplines? (s Yes PF(tra t soA, e esaiesse '

YesONo[

en,nor tio enes.su

7. Is training requirect?

If Yes, prrit name of contact for trainin0 coordlnatiorr YesO No M PAX Port C

]

1. Is the document being TCN'd QA Pro 0rarn Affectino or t.evel 1 OA Program Affecting7 YesLW NoO Answer No any d document es clasedied as Net QA Prograrn Aneceng. This is escoted on the Tatzie of Conserits page of the document.

FYes. compes6e eine eachort rienproceed to Part D. # No. Pienproceed to Part E. tSee &

• below for irdlaal approval. If time permits, otitaan nsaal and smal approvals.)

la the document to be changed an Emergency Operating instruction?

Yes,C No a.

Of the answer to 1.s above is Yes,!!!gD a TCN is not authertzed. A revistoros raouired; see SO123.Vl.0.9 and sot:3-VI 1.)

Part D

1. 10 CFR 50.50 Consideratloc Has the proposed change alrevy been evaluated for 50.59 consideration or a 7 question Safety Evaluation prepared using an approved process?

tEsarrpes at approved grocesses are: MP!FCNMMP/TFMNCR. Tecri Spec Amoruenent impiamentaton. LCS change.-*-8 procedure Safery Evaluation p NOTE; Soth YES and NO rnay be checked,if applicable, auestonsl, etc.)

j Yes O Enter k$entifier and samm"=ned no.

scrE: e VEL pie pneenes aunyt mue as amassess No !!B Attach PF(123) 109-1 (Refer to SO123 VI-1.3.)

ei se no.as - -__

- seener ymiermee.

l 5tDicKTthr. rm, mur, a u. nun, ws, i em e. n., muust area up.

Y.s O No }

& la em silent er the engmal document enered? tCheck Yes il Part il of PF(123) 109-1is checked'YES* full 50.59 Safety Evaluason cor@esied.

w enaar otipecove or procacLee cmanged.)

• 3 Yes, tad oeteln CFDM pnel approval prior to '- -

precedure change (Tech speec 05.&2 & LCS 54.1011.2).

M&s 2.

O h $

2'2f',,%.

PLANT MANAGEMENT 5TArF-UNIT 1

• DATE TIME PEANT MANAGEMENT ST AFF-UNITS 2&3 DATE TIMG d gCN affegctsome il represent a change to a Yes'O No CoalINe TCN etes e to a YeSe Q No 7 0b"K

$1cICA MS(.

a.

7 3.

9 YM /W) pH.-UNiI l i DATE TIME SRO-Urtr35 2&3 y

'OATE TIME 9 s Leum i QA Peeram asecamg a Nel QA Papara Aescarg then sessen esprewet irem em Capuaant sneenemortal as em =e===st ure(eHays Plerd Managumee slefiImeqw) and eds oF se sROrCFM tries prior se snaindlel kiCoM. DAssener overesgre appno,el may be seested for Lowel 9 QA Peepwe Anedrig TC#es s QA Propam Anecemqb tuon alonevel ehet to by ye noneer el pe Ploes ^^

, : Stelt em erm SROCFM Lanneed en she esd y utste ellacent Po TCN appewel, nespees er og Piers "

^ Stas eso eWees sia 9e esparwisar eitnesse as the anilt. er as doesywded a esesq by to CFD64 eessansg vespermammy si the spaare sees anil watral addressed by ee trier 48J

• s yet even em she sweerdwidererShe sueenweer smet promes the remmed s40CFH ano.cmet 9en9 Fanel A IbCHra 4-.w.Nzd APPnOVEDbO :f 4hIn an ll rwH cocNIZAN1 fMNCTIONAL Dr5ibOS MANAGER tcFDhn care NUCLEAR OVERSCHT ~AJ/A t.

1 oArE Part G - For NPQ Use Orsly' i

1. Is Nuclear Oversignt Review! Approval Required? NOTE; Une t e NDMS or Nudear Overaget Review Rockered t.mt to respond.

Yes O No*

n No. emer =A on the Nuclear ov.r.gm Re 4poroven i. ei pari F.

2. Has a 50.59 Safety Evaluation (7 Questions) been attached?

Yes 0 No i

a Y.s. tarwaro e ao the PFt123 1094and so so Safety Evaustion to Nuclear t.icenseg. as amiscable trelet to sot 23Mi.31 PERFORMED BY:

d~

"N scE PF023 sto REV 9 8the

/

NUCU;AH FHUU:.UUHE5 UHOUP (NPU)

DA i t; Site File Copy

NUCLEAR ORGANIZATION SITE TECHNICAL SERVICES PROCEDURE S0123-XIV-5.3.1 l

UNITS 1. 2 AND 3 REVISION O PAGE 4 0F 6 TCN 0-4 6.3.2 Nosystemhasbeenlistedmorethanonc$duetocodingor nomenclature peculiarities.

6.4 Determine whether the SSCs are Safety Related 6.4.1 A search of the MOSAIC database shall be performed to detemine which systems are classified as Safety Related.

6.4.2 Each SSC that is Safety Related shall be identified in l

Reference 2.3.2.

6.5 Determine whether the SSCs are required to mitigate accidents and j

transients.

t 6.5.1 Accidents and Transients are defined for each unit to be the aggregate of:

1

.1 DesignBasisEvents(DBEs)asdefinedinNEDODesignBasis Documents.

.2 Plant specific Initiator Events identified within the units' Probabilistic Risk Assessment (PRA).

l 6.5.2 The systems required to mitigate accidents or transier.ts i

are:

I

.1 Systems identified in each Final Safety Analysis Report (FSAR) as required to mitigate the consequences of the I

listed DBE.

NOTE:

These systems had beer previously identified using the approved Nuclear Regulatory Comission

(?!RC) methodology for developing the FSARs and the design bases of the units.

.2 Systems identified within the PRA which are required to prevent, control, mitigate or teminate an accident or transient 6.5.3 Each SSC which is required to mitigate accidents or transients shall be identiffeu in Ref. 2.3.2.

6.6 Determine whether any non-safety related SSCs are referenced and used l

withintheunits'EmergencyOperatingInstructions(E01s).

l 6.6.1 E0Is are those procedures which guide the operation of a l

unit following a plant trip or safety system actuation.

1 NUCLEAR ORGANIZATION SITE TECHNICAL SERVICES PROCEDURE S0123-XIV-5.3.1 UNITS 1, 2 AND 3 REVISION O PAGE 5 0F 6 TCN 0-4 l

6.6.2 SSCs are selected if they add significant value to the mitigation function of the E01 by providing the total or l

substantial fraction of the total functional ability required to mitigate core damage or radioactive release.

.1 The criteria applied for significant mitigating value is that the SSC has to perfom directly or directly lead the operator to take actions to:

l

.1.1 Maintain integrity of one or more of the fission product i

barriers, i.e., fuel cladding, RCPB or containment, l

OR

.1.2 Prevent the release of radioactive materials to the en-vironment which are evaluated to be >100 mrem, which is the annual exposure limit given in 10CFR20.1301 for members of the public.

i

.2 E0I steps which are for convenience, or evaluation, or l

that represent good practices in assisting operators in i

returning the plant to a preferred condition after the plant has been secured following the transient are not considered to be mitigating steps.

6.6.3 Each SSC which is required by an E0I shall be identified in Reference 2.3.2.

6,7 Determine whether the SSCs can fail and prevent a safety-related system from fulfilling its safety-related function.

6.7.1 Each SSC which can fail and prevent safety-related systems from fulfilling their safety-related functions shall be identified in Reference 2.3.2.

6.8 Determine whether the SSCs can fail and cause a reactor scram or safety system actuation. Forced plant outages and forced outage j

extensions shall be considered the equivalent of a reactor scram.

6.8.1 SSCs shall be included if data from DBDs or the PRA indicate that failure of the SSC would cause a reactor scram or safety system actuation.

6.8.2 The evaluation shall include the review of Licensee Event Reports (LER), NPRDS, and other applicable documents to determine the following for the SSCs:

.1 That failure of the SSC caused a reactor scram or safety l

system actuation, and

.2 That the scram or safety system actuation did not result from an operator error while operating the SSC.

l

AITha(MM S

[

Maintenance Rule Expert Panel Meeting 4/11/196 Performance Criteria The purpose of this meeting is to begin the review / approval of the purposed performance criteria for risk significant and non-risk significant SSCs. Attached to this agenda are a number of attachments that are being provided to help facilitate this process. The following steps are recommended for prforming this review:

1.

Review the list of risk significant SSCs in attachment 1, i.e., the list reviewed and approved by the expen panel on March 19,1996. This is refresh the panel recollection of what SSCs are in risk significant space.

2.

Review attachment 2. " Risk Ranking of Risk Significant SSCs". The PRA representative to explain the ranking methodology and results.

3.

Review attachment 3,"Long Range Schedule Availability Discount', STS will discuss how this information was extracted from the Long Range Schedule 1995/1996 Rev.l.

4.

Review attachment 4," Maintenance Rule Availability Data", STS will lead a discussion on the availability data contained therein along with proposed availability performance criteria.

4.

Review attachment 5," Maintenance Rule and Selection of Performance Criteria", NOD will discuss the data contained within this attachment which represents the unavailability used in the PRA model. They will also explain the PRA definition of unavailability.

STS will provide the definition used for unavailability in maintenance rule space. NOD will discuss how the proposed allowed unavailabilities affect the results of the PRA.

5.

STS will provide a brief overview of what came out of the March 26,27, and 28, Maintenance Rule Industry Workshop regarding the industry approach to setting performance criteria.

6.

The expert panel will now review, on a case by case basis each of the risk significant SSC and provide their collective opinion of the proposed criteria using the information provided. Should the panel as a group believe that all or any individual criteria need to be changed, they will propose the change and provide a basis fcr the change. This information will be collected and will be used as a historical perspective during future reviews of performance criteria and proposed changes thereto. The results of the review of performance criteria will be captured and reported on Form 1.1 " Performance Criteria" and attached to the appropriate procedures.

7.

STS will discuss the development of the plant level performance criteria contained in attachment 1.

l