ML20087E155

Nonproprietary Rev 1 to BWROG Rept for Increasing MSIV Leakage Rate Limits & Elimination of Leakage Control Sys.

ML20087E155
Person / Time
Site:	Hope Creek
Issue date:	12/31/1991
From:	Green T, Lee L, Robare D GENERAL ELECTRIC CO.
To:
Shared Package
ML20087E153	List: ML20087E150 ML20087E155
References
NEDO-31858, NEDO-31858-R01, NEDO-31858-R1, NUDOCS 9201210050
Download: ML20087E155 (362)

Text

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lh GE NucIcar Energy '

NEDO-31858 Revision 1 Class I ,

DRF B21-00461 December 1991 BWROG Report for Increasing MSIV  ;

Leakage Rate Limits and  :

Elimination of Leakage Control Systems-l l :.

l L.S. Lee l

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GENuclent Energy 175 WnnerAwnve

. San Jdse, CA 95t!S NE00 31858 REVISION 1 Class I DRF B21-00461 December 1991- ;

GE NUCLEAR ENERGY BWR06 REPORT FOR INCREASING MSIY LEAKAGE RATE LIMITS AND ELIMINATION OF LEAKAGE CONTROL SYSTEMS i

Prepared:

L.S. ' Lee, Sr. Licensing Engineer-Plant Licensing Services Verified: **l'7l4I T.A. GreM , Sr. Technical _ Program Manager BWR Owners' Grc.up Programs Approved: k It}IIlII D.J..Robare,-~ Hanager Plant Licensing Services '

-Approved: $T bL[

2 J. Stark, Manager BWR Owners' Group Programs ae" Approved: D8r R. D. Binz-IV, Chairman-BWR Owners' Group MSIv Leakage Closure Committee Work Performed for the BWR Owners' Group MSIV Leakage Closure Committee d

. 1

-NE00-31858 Rev. 1 DISCLAIMER OF RESPONSIBILITY  ;

i This document was prepared by the General Electric. Company (GE). Neither GE nor any of

~ the contributors 1to this document:

• A. - Makes-any-warranty.or representation, express or implied with -respect -to -the accuracy, completeness, or usefulness of .

~

the information contained in this '

document,--or that the-use of.any information disclosed in this document .,

may not infringe privately owned rights; _!

- or. ,

B. ' Assumes any responsibility.for liability

.or damage of any kind -which may result. 4 fromthe.use-of any information ,

' disclosed in this document.1 4

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i NE00-31858 Rev. I PREFACE This BWROG report rivises the earlier report, NEDO-31BS8 "BWROG Report for !=:reasing MS1V Leakage Rate limits and Elimination of Leakage Control Systems," by L.S. Lee, dated July 1991. After the earlier report was issued, several meetings were held between GE, BWR Owners' Groep members, and the Nuclear Regulatory Commission staff.

As a result of these meetings, GE and the BWR Owners' Group revised the radiological analyses to more accurately model MSIV leakage; provided addltfonal justificatton to support deletion of the Leakage Control-Systems in terms of plant safety improvement; and provided guidelines to verify seismic adequacy of plant-specific main steam

fp!'ng and condensers. This revision incorporates the revised radiological analyses, plant safety justification, and seismic guidelines.

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'NED0-31858 Rev. I s

CONTENTS-Page PREFACE-- 111 ABBREVIATIONS AND ACRONYMS- xi ABSTRACT xi'll ,

'l.0 INTRODUCTION. 1 '1 1.1: Purpose. 1-1 ,

1.2 Summary 1 2.0 SYSTEM DESCRIPTIONS 2-1

-2.1 Main Steam System 2 2.2 Leakage. control System 2-4

, 3.0 SYSTEM PERFORMANCE ISSUES 3 3.1 ~MSIV Leakage Issues 3-1 3.2' Leakage. Control System Issues 3-3 4.0 RESOLUTION OF PERFORMANCE ISSUES 4-1 L4.1 Approach .. 4-1 -

4.2 - Technical Justification for Increasing MSIV. Leakage Limits _4-l' 4.3 Justification for =lsolated Condenser- as Method for MSIV 4-4

• Leakage Treatment.

~4.4 -- Technical Justification' for Elimination of Leakage 4-14 Control Systems-L5;0 OPERATIONAL, MAINTENANCE,'AND SAFETY BENEFITS: 5-1J 5.1 Benefits Associated with Increasing MSIV Leakage Limits .5 -5.2 = Benefits Associated stith Elimination of Leakage Control 5-2 Systems 6.0 METHOD:0F 1MPLEMENTATION 6-1

61 ' Plant

Specific Evaluations ' . 6 .'

6.2 Radiological Analysis of. the MSIV Leakage Path' 6-3 6.3 Technical ~ Specification Changes and Exemption Requests 6-4

'6.4 Changes to Update Safety. Analysis Report 6-8 6.5 LImplementation of Emergency Operating Procedures 8 L6.6 _ Implementation of Plant Design Changes or Modifications 6-8 6.7l Guidelines for Plant Specific Verification of Seismic

. Adequacy of-Main Steam Piping and Condenser 6-8

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NEDO-31858 Rev. 1 CONTENTS (Cont'd)

Page

7.0 CONCLUSION

7-1 B.O REFERENCES 8-1 9.0 PARTICIPATING COMPANIES 9-1 k

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Ir ;m j n n-NED0-31858 Rev. l ,

F l l

APPENDICES I

Page

.A. RECENT MSIV LEAKAGE HISTnRY AND COST BENEFIT DATA: A SURVEY A-1 BY'THE BWR-0WNERS GROUr; APRIL-1988

]

B. EXAMPLES OF NRC SUBMITTALS FOR INCREASING MSIV LEAKAGE RATE LIMITS AND ELIMINATION OF LEAKAGE CONTROL S'J'TEMS l

~ Type 1 Increase Allowable MSIV Leakage B1-1 10CFR100 Appendix A Exemption Type 2- Increase Allowable HSIV Leakage B2-1 10CFR100 Appendix A Exemption 10CFR50- Appendix J Exemption Type 3 Increase Allowable MSIV Leakage B3 Eliminate Requirements for LCS 10CFR100 Appendix A Exemption Type 4 Increase Allowable MSIV Leakage 84-1 Eliminate Requirements for LCS 10CFR100_ Appendix A Exemption 10CFR50 Appendix J Exemption C.- " POTENTIAL ^ OPERATOR ACTIONS.TO CONTROL MSIV LEAKAGE," C-1 NE00-30324, SEPTEMBER 1985 r1

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LE00 31868 Rev.'l

si, .

TABLES o

Table- Tit 1e Page l' MSIV-Le:kage Ferformance' Surveys 3-5 -i

<3v2- MSIV Leakage Performance Summary 6 53-3: Leakage Control System Performance Data 3-7

• 344 Leakage Control-System Performance Summary 3-10 x-

y = 41 Contributions of MSIV Leakage to Off-Site.and Control 4-17. t Ronm' Radiological. Doses 15-1: Man-Hours ~and Expot.ure Required to Repair and Retest 54

'_ w L MSIVs Which Failed Initial Local Leak Rate Tests

2. Estimates of Average Repair Time and Dose Exposures 55 For a.Specified Allowable MSIV Leak Rate

, 6-1 ~ Contribution-to the-LOCA Dose-Exposures for a Maximum 13 MSIV Leak Rate of-200'stfh +

-ILLUSTRATIONS

~

Fig'ure _ Title Page 12-11  ; Simp 1ified_ Diagram of the Main Steam System

- 2-8

2-2;

• Simplified Diagram of Leakage Control System Flow Paths :2-9 2-31 . Simplified Diagram of Inboa'rd Leakage Control Systemi 2 e Simplified Diagram of Outb'oard Leakage Control System J 2-4.- 2-11

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' .4 - 1  : Isolated Condenser Flow Path =-: Turbine Bypass Lines 4-18 r4-22  : Isolated Condenser Flow Path - Drain-Lines' 4-19

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4-3J ~ Mechanical: Vacuum Pump Flow Path- '4-20 ti ~

4421 JSteam Jet Air _. Ejector /0ffgas System Flow Fath 4-21

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L NEDO 31858 Rev. 1 ABBREVIATIONS AND ACRONYMS 10CFR50 Code of Federal Regulation,10 Part 50 10CFR100 Code of Federal Regulation,10 Part 100 ALARA As low As Reasonably Achievable ANSI American National Standards Institute ASME_ American Society of Mechanical Engineers BWR Boiling Water Reactor BWROG BWR Owners' Group cfh- cubic feet per hour efm cubic feet per minute EAB Exclusion Area Soundary, as defined in 10CFR100 E0P Emergency Operating Procedures >

EQ Equipment Qualification 4 FSAR Final Safety Analysis Report GDC General Design Criteria of 10CFR50, Appendix A GE General Electric HE6'A High Efficiency Particulate Air IEEE Institute of Electronic and Electrical Engineers La Designed containment leak rate under accident conditions.

ILRT Integrated Leak Rate Test, as definad in 10CFR50, Appendix J. ,

LCO: Limiting Condition for Operation, as defined in Technical Specifications LCS: Leakage Control System (s)

-LER Licensee Event Report LLRT Local Leak Rate Test, as defined in 10CFR50, Appendix J LPZ Low Population Zone, as defined in 10CFR100 LOCA Loss-Of-Coolant Accident, as defined in the " Accident Analyses" section of the Final Safety Analysis Report mhr Man-hours mrem Millirems MSIV Main Steam Isolation Valve 1

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-1

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- NED0 31858 Rev.-I

- ABBREVIATIONSANDACRONYMS'(Continued)

MSL . Main Steam Line NRC. Nuclear Regulatory Conmiission psia.- - Pounds per square inch, absolute g -psig- PoundsLper square-inch, gage _. j rem'- Roentgen' Equivalent Man j

.  : RG: . Regulatory Guide RPS Reactor Protection-System RPV- Reactor. Pressure Vessel ,

scfh- ' Standard cubic feet per hour -

-scfm Standard cubic feet per minute-

. SGTS -  : Standby Gas, Treatment System

- SJAE Steam. Jet Air Ejector-

SRP Standard Review Plan  ;

SRV . Safety-Relief: Valve -

SSE~ Safe Shutdown-Earthquake USAR Updated Safety Analysis Report t

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NE00 31858 Rev. l' ABSTRACT in 1986 the BWR-0wners' Group (BWROG) formed a Main Steam Isolation Valve (MSIV)' Leakage-Closure Connittee to resolve the issue of high

- MSIV leakage. This issue had been previously addressed; and plant i surveys had indicated a -significant reductton 'in the occurrence of. .j high MSIV leakage rates. However, MSIV leakage rates still l

frequently exceeded the unnecessarily low leakage limits. This

}

resulted in excessive maintenance and prolonged plant ~ outages. The

}

' BWROG; proposed the use of the main ' steam piping and condenser as ^ a i method for:MSIV leakage- treatment to reduce the radiological consequences of MSIV leakagei This report provides justification for increasing MSIV leakage rate limits and for eliminating the requirements for leakage Control Systems.

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'NED0J 3 1858 Rev. I

1.0 INTRODUCTION

1,1- PD? POSE

-The EVR Owners' Group (BWROG) has studied the issues of main steam

isolation valve (MSIV) leakage rates and associated excessive maintenance required for the MSIVs and the Leakage Control Systems (LCS). This report provides justification- for increasing the Technical Specification leakage ra'e c limits for the MSIVs and for eliminating the Technical Specification requirements-for the LCS, in a way that does not adversely affect public exposures-to accident consequences._ To facilitate utilities applying for these proposed changes, this' report provides examples of licensing submittals

" Application for License Amendment to the Technical Specifications" and >

" Exemption Requests" to 10CFR50 Appendix J and 10CFR100 Appendix A.

1. 2.

SUMMARY

i

' The BWROG formed a MSIV Leakage Cr wittee in 1982 to identify -and resolve theLcauses of high MSIV leakage rate' The BWROG then formed a follow-on MSIV

Leakage Closure CommitteeLin 198f r address alternate actions to resolve on-going,~ but 1ess Lse' vere MSIV leakage problems and to address the limited' 1

capability.of the_LCS.-

As a resolution to the MSIV leakage and'LCS-performance issues, the' BWROG proposes:to use the main steam piping and condenser as a method for MSIV

. leakage treatment. - This method'provides effective- and reliable fission product attenuation for. reducing the radiological consequences of MSIV leakage. A radiological. methodology, developed to take advantage of the main steam-piping and condenser, shows that MSIV leakage rata up to 200 scfh per main steam line result in an acceptable increase to the-off-site and control

room doses for-Hope Creek. Furthermore, for Hope Creek, the LOCA doses remain zwithin the regulatory guidelines for MSIV leakage rates up-to approximately f 1500 -'sefh . Therefore, the-proposed method provides a substantial margin-of Lsafety'for mitigating MSIV leakage beyond the proposed Technical Specification i leak rate'. limit of 200 scfh.

l 1-1

, 1

NED0 31858 Rev. 1 The BWROG has evaluated the capability of main steam piping and condensers to process MSIV leakage following a design basis accident coincident with a seismic event. Based on this comprehensive evaluation, the BWROG has concluded there is reasonable assurance that the main steam piping and condenser will remain functional following a design basis accident coincident with a scismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage. The following conclusions provide the bases for this assurance:

(1) Probability for which the resulting dose from MSIV leakage is significant is extremely low. This requires a design basis LOCA, a degraded core where ECCS are not functional, and a significant seismic event.

(2) Main steam piping and condensers are designed to strict industrial standards and building codes; thus, significant design margin exists.

(3) Main steam piping and condensers exhibit substantial seismic ruggedness. Comparisons of piping and condenser design in GE plants with those in the earthquake experience database reveal that the GE plant designs are similar to or more rugged than those that have exhibited good earthquake performance.

(4) Possibility of significant failure in GE BWR main steam piping or condensers in the event of design basis earthquakes is highly unlikely, and any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

(5) A plant-specific verification of seismic adequacy of the main steam piping and condenser will be performed to provide reasonable assurance of the structural integrity of these components.

1-2 l

NED0-31858 Rev. 1 The proposed increase of the MSIV leakage rate limits will avoid unnecessary dose exposures to maintenance personnel, reduce unnecessary repair c-M. n Le outage schedules, and extend the effective service life of the MSIVs. Furthermore, the proposed increase may significantly reduce recurring valve leakages, and minimize the possibility of needless repair that may compromise plant safety. Depending on specific plant design and radiological consequences, MSIV leakage rate limits can be increased from the typical 11.5 scfh to a maximum of 200 scfh per main steam line. The proposed elimination of LCS requirements will similarly avoid unnecessary dose exposures, reduce maintenance costs, and reduce outage durations.

Furthermore, the risk to the public health and safety will be reduced with the implementation of the proposed MSIV leakage treatment method. For plants that do not have a 1.05, implementing the proposed changes will provide plants with a capability to process MSIV leakage. Use of the main steam piping and condenser provides reliable and effective HSIV leakage treatment without exceedir. the off-site and control room dose limits. The implementation will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit, trom a safety perspective, the proposed change results in an increase in protection to the public.

For plants with a LCS, the proposed changes involve a replacement of the existing LCS with the more reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to reduce dose consequences of MSIV leakage over an expanded operating range and will, thereby, resolve the safety concern that the LCS will not function at-MSIV leakage rates higher than the LCS design capacity. Except for the requirament to esta'lish o a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment.

1-3/1-4

NEDO-31858'Rev. 1

-2.0 SYSTEM DESCRIPTION I

y '

2.1 : MAIN STEAM SYSTEM.

The main steam-system transports steam from the RPV to the power conversion system and to various-types of auxiliary equipment. 'In addition

)the mainisteam system provides a turbine bypass mode that. diverts steam directly.to the main condenser, bypassing the turbine. Figure 2-1 shows a j simplified diagram of the main steam system. l Steam exits from the vessel several feet below the reactor vessel flange through ~ four nozzles. Typically, there are four main steam lines. Carbon

-]

steel steam lines are welded to the vessel nozzles and run parallel to the j vertical axis of the vessel, downward to the elevation where they emerge i

, horizontally from the containment. _ Safety / relief valves that provide-over-pressure protection are flange connected to the main steam lines.

Each main steam line has two, air-operated, MSIVs: one inboard and one outboard of the containment penetration.- The MSIVs Isolate.the reactor system in the event of a LOCA, or other events requiring containment or reactor

system isolation. - Only that portion of the main steam system from the RPV up-to:and including-the' outboard MSIV is part of the reactor coolant pressure

boundary.. -For BWR/6 and ~some' BWR/5 plants, a main steam shut-off valve is installed in each-steam line-downstream of the1 outboard MSIVs. i Typical main steam piping consists of four'28-inch diameter lines from the outboard MSIVs to-a main steam equalizing header. From the

equalizing header are:

~(1)lfour, 28-inch diameter, main steam lines. leading to the turbine stop valves;.

L(2) Two; 18-inch diameter, bypass' lines leading to the turbine bypass val vas;-

2-1

NED0-31858 Rev. 1 (3) Branch lines that supply steam to auxiliary equipment, which include the turbint gland seal system, steam jet air ejec+ ors, off-gas pre-heaters, moisture separators and reheaters.

Drain lines are provided at low points where condensed steam is collected. All drains from the same pressure source are headered wherever possible. All drains discharge to the condenser. A main steam drain header from inside containment is used as a source of warming steam for the main steam lines beyond the outboard MSIVs. This header permits equalization of pressure across the MSIVs following a steam line isolation.

The MSIVs, main steam shut-off valves, turbine stop valves, and bypass valves are designed to close against maximum steam flow and differential pressure. The design pressure of the main steam system is typically 1250 psig. Over-pressure protection is provided by the main steam relief valves, which discharge the steam to the suppression pool. Branch lines having a different design pressure are provided with their own relief valves, all of which discharge to the condenser.

The main steam piping from the vessel rozzle to the outboard MSIVs, including the MSIVs, are designed to the classifications of Seismic Category I, ASME Section III, and Quality Group A. The main steam piping downstream of the outboard MSIVs, or downstream of the seismic restrainer, is typically designed to the classification of ANSI-B31.1 and Quality Group D.

2.1.1 Main Steam Isolation Valves Typical MSIVs are globe valves designed to close by diversified sources of spring pressure and pneumatic pressure. In addition, each valve has an air accumulator to assure adequate supply of pneumatic pressure. The MSIV is designed to close on loss of pneumatic pressure or loss of power to the pilot valves. Each valve has an independent position switch initiating a signal into the RPS scram trip circuit when the valve closes.

2-2

NE00 31858 Rev. 1 I'

Typically the MSIVs are signalled to close upon the following conditions: "

(1) low water level in the reactor vessel, (2) high radiation from the main steam line, (3) high flow rate in the main steam line, (4) low pressure at the inlet to the turbine, (5) high ambient or high differential steam line tunnel temperature (outside containment), (6) low condenser vacuum (unless procedurally bypassed), and (7) high turbine building temperature. Once any of the closure signal is initiated, the valves will continue to close and cannot be opened except by manual ncans. Each valve may be operated by independent remote manual switches located in the control room. Lights in the control room indicate valve position.

The safety function of the MSIVs is to isolate the reactor system in the event of a LOCA or other events requiring containment isolation. MSIV closure actions are included in the design basis LOCA analyses postulated for a main steam line break accident outside the containment. In this accident, the analysis assumes that the MSIVs will be fully closed within ten seconds (typical) to terminate the blow-down of reactor steam to the environment.

This LOCA results in a limited release of high energy fluid into the steam tunnel, and the analysis is evaluated against the structure design for poten-tial sub-compartment over-pressurization. For added margin, the Technical Specifications typically specify MSIV closure times between three to five seconds.

For other LOCAs, the MSI'vs isolate the reactor from the environment in sufficient leak-tightness to minimize the release of fission products from the containment. These LOCAs are evaluated against the off-site and control room dose guidelines as defiied in 10CfR100 (Ref.1) and 10CFR50, Appendix A, GDC 19 (Ref. 2), respectively. The worst-case LOCA for radiological evaluation is the recirculation line break accident, where the fission products are assumed to release immediately from the core to the containment, consistent with the assumptions in RG 1.3 (Ref. 7). The LOCA radiological analysis assumes that MSIVs maintain their leak-tightness and that the release of f ssion products is limited to the designed containment leak rate. MSIV leakage rate is included as part of the containment leak rate.

2-3

NED0-31858 Rev. 1 For some BWRs, in addition to the containment leak rate, the LOCA radiological analysis conservatively assumes that MSIVs leak directly to the environment through the closed MSIVs at a maximum rate allowed by the plant Technical Specifications.

For most BWRs, the Technical Specifications sp: cify an allowable leakage rate of 11.5 scfh for each main steam line. Some Technical Specifications have a limit of 25 scfh per line or 100 scfh for a total of four main steam lines. The surveillance in the Technical Specifications requires leakage rate tests for the MSIVs once every 18 months.

2.2 LEAKAGE CONTROL SYSTEM The safety related LCS is designed to minimize the radiological consequences of MSIV leakages. Most BWRs have a negative pressure LCS, where blowers maintain the boundaries between (1) inboard and outboard MSIVs and (2) outboard MSIVs and downstream valves at less than 1/4 inch of water vacuum.

Leakage from the MSIVs is diverted to the reactor building atmosphere, filtered by the SGTS, and released to the environment through the plant release or elevated stack. Other BWRs have a positive pressure LCS, where the boundaries are maintained at a positive pressure. This is accomplished by pressurizing the boundaries at least ten percent higher than the containment pressure. Under this condition, leakage through the HSIVs will be diracted inward to the containment preventing the release of fission products to the environment.

Since most BWRs use the negative pressure LCS, discussions in this report are based on the negative pressure system. The LCS is a redundant, safety-related system; it consists of an inboard and outboard subsystems. The LCS has mostly small bore piping (3/4" to 1") that connects to the main steam system drain piping. The LCS major components are located in the reactor building. Figure 2-2 shows a simplified LCS flow path.

2-4 l

NED0 1!850 Rev. I 2.2.1 Inboard LCS Sub*ystem ,

l Each main steam line is provided with an individual inboard LCS process line. 1~ t inboard LCS maintains the main steam line between the MSIVs at a i negative pressure. An individual controlled depressurization and bleed line is provided for each main steam line, figure 2 3 shows a schematic diagram of the inboard LC3.

Major features of each inboard LCS line are: (1) an inlet valve that i maintains containtant isolation together with the associated inboard MSlV, (2) an electric heater to ev&porate any condensate in the line, (3) a depressurization line for initial pressure reduction, and (4) a bleed line for ,

MSIV leakage control. The individual bleed lines are connected to a common i flow manifold where the leakage is mixed with dilution air and exhausted by a  !

100 cfm blover. The blower discharges near, or into, a suction plenum of the SGTS.

The inboard LCS lines have individual flow measurement devices and logic controls such that the appropriate portior of the inboard LCS will be shut down when the process flow exceeds the LCS designed capacity. The designed capacit) is typically 100 scfh; however, some plants have lower design capacity.

2.2.2 Outboard LCS Subsystem The outboard subsystem lines are connected to the main steam lines downstraam of the outboard MSIVs. The outboard LCS maintains a slight negative pressure in the main steam line between the outboard MSIVs and the downstream valves, figurt 2 4 shows a schematic diagram of the outboard LCS.

The : individual LCS lines are tonnected to a common process line before branching out to the oepressurization and bleed lines. The depressurization lino vents any steam trapped in the main steam lines for initial pressure reduction.- The bleed line maintains MSIV leakage control and the blowers discharge the process flow near the SGTS suction plenum.

2-S

HEDO 31858 Rev. I 2.2.3 System Operation The LCS is normally isolated from the r,ain steam system. The LCS is manually initiated in the control room approximately 20 minutes following a LOCA, as evident by high drywell pressure and/or low reactor water level.

Both the inbotrd and outbcard subsystems start with the opening of valves in the depressurization mode and start of equipment (such as heaters and blowers), provided all system initiation interlocks are satisfied.

for the outboard subsystem, the initiation interlocks are the main steam line and reactor vessel pressure setpoints (both setpoints are typically at 35 psig). Both setpoints must be satisfied or the outboard subsystem remains inactive.

For the inboard subsystem associated with each main steam line, the initiation interlocks t.re: (1) pressure sensor in the reactor vessel. (2) pressure sensor in the associated main steam line, and (3) position switch in the associated inboard MSIV. These interlocks prevent the inboard subsystem from initiating when the main steam pressure is above a specified setpoint and when the associated MSIV is not fully in a closed position. All three interlocks must be satisfied or the inboard subsystem associated with that main steam line remain; inactive.

For the inboard subsystem, when all initiation interlock setpoints are satisfied, the main steam lines between the MSIVs are depressurized through the opening of the bypass valves. About two minutes into the depressurization mode, flow is directed to the blower through the bleed line by simultaneous actions of bypass valve closing and outlet val 'e opening. In addition to the initiation interlocks, W inboard subsystem has the follo.ving interlocks that will prevent the system f rom operational:

.(1) During the depressurize. tion mode, a pressure switch will reclose the subsystem valves if the steam line pressure has not .ecreased to a preset value. This pressure switch assures that MSIVs are fully closed and that MSIVs are not grossly leaking.

2-6 l

1 NfD0 31858 Rev. 1 (2) A pressure switch in the flow manifold will prevent the subsystem volves and heaters from actuation if a partial vacuum is not established, indicating poor blower performance.

(3) A flow sensing switch will actuate the subsystem valves to close on high flow setpoint. The flow switch is normally set at Technical Specification icakage limit for the MSly and has a range up to 100 scfh.

for the outboard subsy em, when all setpoints of the initiation interlocks are satisfied, the valves in the depressurization line open first, thus permitting the main steam lines to depressurize to atmospheric pressure.

A pressure senor provides the logic control for the biced valves to open and depressurization valves to reclose when the main steam line pressure is reduced to atmospheric. The blowers then estabitsh a partial vacuum in the steam lines ud capture leakage that goes throuoh the closed MSlvs. The outocard subsystem has alsrms that annunciate in the control room upon the con'litions of low partial vacuum in the main steam itnes or low vacuum established by the blowers.

The SGTS filters will remove a large fraction of any fission products in particulate from and in gaseous iodine isotopes, but have only a small effect on noble gas isotopes, finally, the LCS is designed to handle leakages only moderately in excess of the Technical Specification limits.

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o- l NEDO 31858 Rev. 1  !

3.0 SYSTEM PERFORMANCE ISSUES f 3.1 MSIV LEAKAGE ISSUES Operating experience indicated that MSIV leakage frequently exceeded the  ;

Technical Specification leak rate limits. Some of these valves repeatedly  ;

failed the local leak rate tests despite frequent disassembles and refurbish-  !

, ment.

1 The NRC conducted a survey (Ref. 3) of MSIV leakage tests performed between 1979 and 1981, and found that 19 of the 25 operating BWRs had MSlVs that failed to meet the le<' rate limits during one or more surveillance tests. The survey indicated that 54% of the total "as found" MSIV leakages exceeded the typical limit _of 11.5 scfh and 33% of the total exceeded 100 scfh. The BWROG conducted a similar survey for MSIV leakage tests through l 1984 and found lower, but still significant failure rates. Table 3 1 summarizes the results of these earlier surveys.  ;

From a safety perspective, the NRC has been concerned with the potential l consequences of_not maintaining the low leakage requirements. These '

. unnecessarily low leakage requirements are derived from the conservative

. assumption that main stet:A piping will not exist downstream of the outboard .

MSIV following a design basis event, and the resulting MSIV leakage will be

released directly.into the environment. Calculations, _ using these standard conservative acumptions for considering off-site consequences of a postulated i design basis-LOCA, confirm that off-site and control room doses will be withir, regulatory. guidelines for the allowable MSly leakage rates. However, if MSIV l 1eakageLis:only moderately higher than the limits, the calculated doses, with -  ;

these conservative e.ssumptions, will exceed the regulatory guidelines.  ;

From;an operational and maintenance perspective. MSIVs have frequently ifalled to meet the leak rate limits. These failures significantly impact _the

maintenance work load and often contribute to critical path and outage '

extensions. There have been many LERs written for MSIVs failing to_ meet the ,

unnecessarily low leak rate limits. '

3-1

, _ _ ,. _ _ , . . , , ., . .. _ ._.. _ . _. _ - ,_ . .- m. _ ,, . . , , , -

. _ . - . ~ . _ . _ . . . , . . . . . . ~ - , . _ , , , . .

NEDO 31858 Rev. 1 The BWROG formed a MSIV Leakage Committee in 1982 to address the issue of high MSIV leakage rates. This connittee generated an extensive database of MSIV leakage performance and determined corrective actions. Based on a thorough evaluation of this database and a review of applicable LERs, the CWROG identified probable causes for MSIV leakage and categorized those causes as either: (1) primary contributors; (2) secondary contributorst or (3) non contributors. The DWROG reconnended corrective actions for minimizing the primary and secondhry contributors that cause excessive MSIV leakage during local leak rate tests. The NRC reviewed and documented these causes, recommended corrective actions in NUREG 1169 (Ref. 4), and concluded that the BWROG findings and recommendations were reasonable. i

.The BWROG conducted a more recent survey of MSIV leakage tests performed between 1984 and 1988 to verify that the implementation of industry and BWROG actions was offective in reducing MSIV leakage rates. This survey showed significant improvementt however, 23% of the total "as foand" MSIV leakages still exceeded 11.5 scfh and 10% exceeded 100 scfh.

Results of this recent survey are also included in Table 3-1 for comparison with previous survey results. This latest survey demonstrates substantial improvement in leakage performance since the last BWROG survey and greatly improved since the earlier NRC survey.- Improvement in the recent leakage performance is primarily contributed to awareness of the leakage problem. the sharing of good maintenance practices and testing methods, and the implementation of the BWROG recommendations for reducing excessive MSIV leakages. Appendix A contains details of the recent BWROG survey. ,.

Despite the recent improvement in leakage performance, MSIV leakage rates still frequently exceed the current Technical Specification limits and the saf ety and maintenance problems related to high MSIV leakage rates .although less severe, remain as a significant issue. -Table 3-2 summarizos the impact on MSIV maintenance.

3-2

V NE00 31958 Rey, 1 '

furthermore, the BWROG has thoroughly investigated other areas, such as ASME requirements, where potential relaxation of code requirements could possibly be mada to reduce HSIV maintenance. None of these investigations has been beneficial. For example, regarding the ASME code criterion for valve seat indications, it was determined that significant relaxation has been incorporated in the past 20 years, and that additional relaxation will be difficult to justify for the hard valve seat surfaces.

3.2 LEAKAGE CONTROL SYSTEM ISSVES in the late 1960s, the Atomic Energy Commission (AEC) became concerned over the potential consequences of excessive tiSIV leakage. To resolve this concern, the AEC issued RG 1.96 (Ref. 5) and required the installation of a safety related LCS as a positive approach for reducing the radiological effects from HSIV leakages. LCS has been installed for all BWRs (with exception of Nine Hile Point Unit 2) with construction permits granted after Harth 1970, BWRs with construction permits issued before March 1970 are not required to install

• LCS.

In the early 1980s, the Nuclear Regulatory Commission (NRC) became concerned that MSIV leakage rates, as determined by conservative local leak rate testing, were too high and that the LCS would not function at moderate MSIV leakage rates. As discussed earlier, the NRC survey in 1981 indicated that 33 percent of the total tests exceeded leakage rates of 100 scfh. Since the process capability of the LCS is typically designed for MSIV leakage rates of no more than 100 scfh, the potential existed for the LCS not to function as analyzed for a postulated design basis LOCA. Consequently, the conservatively calculated dose contribution from MSIV leakage would exceed the regulatory limits for off-site and control room doses. Due to this concern, the NRC ranked Generic issue C-8 ("HSIV Leakage and LCS Failure") as a high priority item in January 1983. '!his Generic issue was closed in 1990, 3-3

,.._g <+rvw-

1 NEDO 31858 Rev. 1 Although the implementation of industry efforts significantly lowered HSIV leakage rates, thesa valves still require excessive maintenance because of the extreeely low leak rate limits. The BWROG has determined that j alternate pathways are available to effectively process MSlv leakage, and that i these alternate pathways can replace the existing LCS. Because of reported high LCS maintenance, the BWROG evaluated recent LCS performance data for six BWRs operating in the United States. Results of this evaluation indicate that the LCS is extremely difficult to maintain. Plant shutdowns and start up delays have occurred at two of the six plants. At one plant, the LCS has become a critical path task during every refueling outage.

A limited amount of LCS performance data are summarized in Tables 3 3 and 3 4. The system has extensive logic and instrumentation, and calibration of the instrumentation (such a pressure transducers and flow elements) is frequently required to meet the Technical Specification requirements. The

-system has been declared inoperable and required LCO entry at least once every year for each plant. At one plant, the LCS typically causes the plant to enter five LCOs every year. Furthermore, replacement of safety related parts, such as blowers, heaters, and flow elements, has become increasingly difficult and lead times have been long, in addition, the flow element in the LCS has caused many plants _to fail the surveillance test. A replacement that meets the system requirements is currently not available and some plants are in the process of redesigning the LCS to accommodate an alternate flow element, a

3-4

NEDO 31858 Rev. I f Table 3 1 M51Y Leakage Performanet Surseys NRC* BWROG BWROG 1979 - 1981 E.te-1161 Ensh1911 Number of Tests 280 586 329 Percentage of Tests 54 37.8 23 Greater than 11.5 scfh Percentage of Tests 33 21.7** 10 Greater than 100 scfh From Information Notice IN 82 23, dated July 16, 1982 This value is actually the percentage of tests greater than 90 scfh. For this comparison, the 21.7% is assumed the same for results greater than 100 scfh.

3-5

i t

NEDO 31858 Rev. 1  !

Table 3 [

M51V Leakage Performance Sumary (Based on Post 1984 BWROG Survey)

?

Dose Exoosure:' Range from as low as 0.04 rem to as high as 40 l rem (whole body) per valve repair. lypical  ;

exposure is betwoon one to two rem per valvg ,

repair. Average is 1.1 rem per valve repair.

Maintenance Cost: Maintenance resources reported from 82 stn hours I to 1240 man hours per valve repair. Avertigo of- '

350 man hours is celculated. Average cost per valve repair is calculated to be $17.500, based on L labor rate of $50 per hour,

. Critical Pathi Occurs frequently. Most plants have extended Maintenance: the outage durt.tien for MSIV maintenance. In one plant, an additional. 60 days were added to the schedule during one refueling outage for i MSIV maintenance. '

Plant Shutdown or Plagt ' Occurs frequently. In one plant, a mid-cycle Startuo Delavs outage was required for MSly lekk rate tests and' ,

n,aintenance. l Maior Problergsi - Recurring valve leakage. MSIVs-that-have been unnecessarily repaired are likely to require

- future repair to meet the low leak rate limits.

?

f P

~

36 -

1 a

m _ _ _ . .s..;._--...__ . . . - .= ,_.-mm.,_ . w ._ _ . . _ _ _ _ _ . ._,_,.,,,A ,,.,,-.-_.m , ; .. , . , . , . .--.

NED0 31858 Rev. 1 Table 3 3 Leakage Control System Performance Data l Declare System i Plant Total Failures Inoperable (LCO) Plant Shutdowns Performance 1

_ No . ._ ,_let.le.ar Per Yeer or Startup Delavs Comment 1 No Data Approximately 1 0 Poor per-mance, critical path during every re-fueling outage.

2 No Data About 80%

LCS Availability 3 No Data No Data 4 Approx. 5 Approximately 5 2 Poor per-formance.

64 surveil- ,

lance test failures and 250 mainton-ante items since installa-tion in 1975 (13 years).

5 Approx. 12 Approximately 1 -1 Poor per-formance; itew plant.

i; 1

i l 37

_ ~ . - ._~ .. - - - _ . - _ . - - - . _ - _ _ _ ._ _ _.

HED0-31858 Rov. 1 Table 3 3 (Continued)

Leakage Control System Performance Data Declare System Plant Yotal Failures Inoperable (LCO) Plant Shutdowns Performance No.. Per Year Per Year pr Startuo Delay 1 Comment 6 Approx. 3 Approximately 1 0 Testing caused high suppression chambers water temperature and caused ESF actuations.

NOTE: Surveillance testings required by Technical Specifications are:

18-Month System functionr? Test 18 Month Valve Leakage lest Quarterly Valve Stroke Test Monthly System Logic Test 38

t' NEDb 31858 Rev. 1 Table 3 3 (Continued)

Leakage contro'l System Performance Data Equi > ment Replacement Plant Relia)ility Part LLRT Inboard Maintenance Dose  ;

. jit .f.r.gbj ems Problem lest Problemi ,Jost/ Year Exposurci 1 No Data No Data No Data $60K $80K No Data 2 Relays, Relay No Problem $50K 100K Approx. -

Heaters, Replacement. (Approx. 10 1 rem per Switches. EQ Problems Han Months) year Spring Packs 3 Flow Trans- Blower No Problem $15K (1987) < 50 milli-mitters Motors (EQ) rem / year flow Elements (Redesign) 4 Flow Trans- Heaters (EQ) No Problem $40K $50K No Data mitters, Heaters flow Elements (Redesign) 5 flow Trans. Flow Ele- Yes High Low mitters ments (Out of- (New Plant)

(Redesign) Service for 68 Hours) 6 Valve Logics No data No data No data No data i

39

NED0-31853 Rey. 1

~

Table 3 4 i q

Leakage Control System Performance Sumary (6BWRs) l Dose Exposure: Approximately I rem per year. '

Maintenance Cost- $15K to $100K per year ,

(nominal $50K per year).  ;

Maintenance Reauirements: (1) System has extensive logic and b instrumentation.

.i (2) High number of surveillance tests 1 (recalibration frequently ,

required). -!

Limitina Condition of Operation: About one per year per plant. l

- Critical Path Maintenanc3: Occurs frequently. ,

. Plant Shutdowns or' Plant '

. Errtuo Delavs: Has occurred at 2 of 6 plants in survey. ,

Maior Eauipment Problems: (1)

Inboard to trip (redesign flow elements-cause in progress . sy) stem .i (2) Frequent recalibration of pressure transducers.

Ma.ior Replacement Part Problem:! Blower motors, heaters,- flow elements r .(problem in purchasing equipment that meet equipment qualification-requirements).  !

E  :

y

}

T j M f

{ , ,

a 3 t t rv -y ,4 y g-- ,-#.a . .-es., -, ., ,.c,mu.k.-a, ,w, ,,,,,w.. ,y ,e ,, . , .-- , - . . - - ,..-n.,. ,,.,,,++c.w,. e p, w..: w ,,w-.,,, ,,-.--,--,.e

l NED0 31858 Rev. 1 4.0 RESOLUTION OF PERFORMANCE ISSUES 4.1 APPROACH The BWROG MSIV Leakage Closure Committee recommends the following changes i to the plant Technical Specifications-i A. Increase the MSIV leak rate from the current specified lirit (typically 11.5 scfh) to a higher limit up to a maximum of 200 scfh per main steam line. Specific proposed limits will be based on- l individual plant design, and 1

i. Eliminate requirements for the LCS for those plants that currently ,

have a LCS installed I 1

The feasibility of the proposed thanges relf es on the justifiable use of  !

the main steam piping and condenser for MSIV leakage treatment. Using this  :

treatment method, radiological evaluation for dope Creek shows that MSIV 1eakage rates up to 200 scfh per steam line result in an acceptable increase 1 to the off-site and control-room doses. Furthermore, Hope Creek calculations ,

- show that M51V. leakage rates up to approximately 500 scfh per line will not

. exceed 4ne regulatory dose limits. Section 4.3 provides the justification for j

.this. treatment: method, its capability to remain functional following a design j basis: accident, and the radiological assessment, j 4.2 TECHNICAL JUSTIFICATION FOR INCREASING MS!Y LEAK RATE LIMITS The current Technical. Specifications MSIV leak rate is extremely limiting and often leads to routine, unnecessary repair and retesting. This significantly-impacts'tho-maintenance work load during plant outages and often  :!

contributes to outage extensions.- BWR outage planners often schedule several 7d ays of critical path contingency to repair and rotest the MSIVs. In addition, the needless' dose exposures to maintenance personnel are inconsistent.with ALARA principles.

41

. ..... ...__.~..._i...... ~  ;._,i._.....,.,_2.,_-.m.. . _ , . .. _ , _ , , .,,,.,,m_, , . . . , . , , , -

NED0 3185S Rev. 1 The current Technical Specification leakage limit is established by excessively conservative LOCA radiological analysis. The valve's physical size and operability characteristics (large size and fast-acting) and existing turbino building treatment equipment were not considered at the time the leakage limit was established.

Based on an extensive evaluation of valve leakage data, the BWROG concludes that MSIV leakage rates exceeding 500 scfh usually indicate that significant maintenance or valve modification is required; whereas, leakage rates less than 500 scfh can generally be eliminated by proper routine maintenance and testing procedures. This supports the BWROG position that MSIV leakage rates up to 500 scfh are not indicative of substantial defects that would challenge the isolation capability of the MSIVs. This conclusion is verified by LERs in which plant-maintenance personnel routinely do not find specific defects when the HSIVs have been disasseeled for repairs. This finding is also consistent with the observation of MSIV leakage performance as documented in NUREG-ll69 (Ref. 4). In addition, valve manufacturers have stated that these large and fast-acting MSIVs are expected to have no substantial defects at leakage rates up to 200 scfh. Therefore, it can be conservatively concluded that the proposed increate of the HSIV leakage limit up to 200 scfh will not inhibit the isolation capability of the valve.

The BWROG has also found that disassembling and refurbishing MSIVs to meet the unnecessarily low leakage limits frequently contributes to repeated failures. Examples of the these maintenance induced defects include machining-induced seat cracking, nachining of guide ribs, excessive pilot valve seat machining, and mechanical defects induced by assembly and

~ disassembly. By not having to disassemble the valves and refurbish them for minor leakage, the utility may avoid introducing one of the root causes of recurring leakage. Industrial experience suggests that, by attempting to correct non existing or minimal defects in the valves, it is likely that some actual _ defects may be introduced that lead to later leak test failures.

l I

42

NED0 31858 Rev. 1 Two recent incidents occurred in which the safety function of the MSIV was compromised by attempting to repair the valve to meet the low leakage limit. In one incident, lapping of the valve damaged the teat such that 56 days of critical path time were required to resolve the problem, in the other incident, maintenance personnel over tightened the stem packing to reduce leakage and, as a result, all eight MSIVs failed to close fully under spring force.

Another HSIV incident (not included in Table 5 1) that occurred in late 1990 resulted in a significant exposure of 6.8 man rems to repair MSIVs that barely failed the LLRT. Work required to support this repair inclued:

Install scaffolding and work platform; Stage equipment; Remove insulation; Perform valve repair (machine as required);

LLR1; Quality inspections; Supervision and engineering inspection; Area clean up and decontamination; Scaffolding and equipment removal; Install insulation; and Radiation protection support.

In summary, the proposed increase in the MSIV leakage limit will significantly reduce recurring valve leakages and will minimize the possibility of needless valve repair, which can comp omise plant safety.

Based on the recent survey of MSIV leakage performance, the proposed increase in the allowable leakage rate up to 200 sefh will increase the chance for successful local leak rate test to greater than 90%, up from the 77% success rate at the current typical limit of 11.5 scfh.

i 4-3

NE00 31858 Rev. 1 1

4.3 JUST!F1 CATION FOR ISOLATED CONDENSER AS METHOD FOR M51V LEAKAGE TREATMENT  :

i 4.3.1 Description of Treatment Equipment '

The BWROG evaluated the effectiveness and availability of leakage j treatment methods that utilize existing plant equipment and that can be used for reducing the radiological consequences of MSIV leakage. These methods are I

- the isolated condenser, mechanical vacuum pumps, steam jet air ejector (SJAE)/offgassystem,andisolatedsteamlines, t

The BWROG recommends the isolated condenser method, which utilizes the main steam drain piping to convey MSIV leakage to the condenser, as the _

3 preferred method for MSIV leakage treatment. This method provides effective  !

fission product attenuation in the condenser such that the consequences of MSIV leakage can be significantly reduced. This method usually requires  ;

operator action to open the specific drain valves to mitigate the MSly leakage. Minor design changes may be required to facilitate opening of the valves if off site power is unavailable. -

Other leakage treatment methods, such as the mechanical vacuum pumps and SJAE/offgas system, are very effective for reducing the radiological consequences!but require ~ substantial operator actions to initiate certain non-safety related systems. Therefore, these methods are not recommended as  ;

a design basis-for MSly treatment but can be used as a backup to enhance MSIV.

leakage treatment capability. Each of these methods is-briefly discussed in

• the following paragraphs.

L4.3.1.1 holated Condenser This leakage treatment method takes advantage _ of the large volume in the l main condenser to provide h'old up and plate-out of fission productr that may leak-from closed MSIVs. The condenser is isolated from the turbine building

- and the outside environment. MSIV leakage that enters the condenser is.

subsequently released to the turbine butiding from the low pressure-turbine seals.-

44' 7 m'4W w.r wWT-m- i $> 4-r*8'-'1 *--w v+ w 1 md*-$ t - W W- w-c9T-----MF-NFr*9cw-M'T?=W---r*w'.rme-pT% w t- gzN sm w_ w r_- -e m 3supat w'de *-e y 4 Frt**.. .W=.i-*N.F- WErpv'WW WP. te m

l NED0-31858 Rev. 1  !

There are two alternate pathways to convey leakage to the condenser. The first path (Figure 4 1) utilizes the turbine bypass lines to direct the leakage to the condenser. This path requires operator action to open the turbine bypass valves, which are normally closed and remain closed following an accident. These valves may be opened by initiating the turbine electro hydraulic system, but this is usually not possible if off-site power is not available. Some BWRs have interlocks to prevent the valves from opening following an accident. l Thesecondpath(Figure 42)usesthemainsteamdrainlinestodirect i the leakage to the condenser. This path usually requires operator actions to l open the specific drain valves and may require minor modification to assure valve opening after a loss of off site power.

The BWROG recommends the drain lines over the turbine bypass lines to  !

direct MSIV leakage to the condenser because it is much more difficult to assure the turbine bypass valves can open be opened following a loss of off-site power.

4.3.1.2 Mechanical Vacuum Pumo.1 In series with the isolated main condenser, the condenser mechanical vacuum pumps or the gland seal / exhaust system can be used to maintain the condanser at'a lower pressure than the surrounding environment and prevent leakage from the turbine seals. Figure 4 3 shows a schematic diagram of the mechanical vacuum pump flow pathway. Assuming the main steam drains are open.

MSIV leakage wil' inion to thro 7h the main steam lines to the condenser. The mechanical vac,yr camps or glana seal exhausters will then discharge the leakage-to-the ph nt 0,acks (elevated release point).

-This method is more effective than the isolated condenser method in terms

,of dose reductico because the elevated release provides more dispersion of the fission products. Howevet Pts method requires substantial operation actions

.and availability of certun po. safety-related systems.

45

NED0 31858 Rev. 1 4.3.1.3 S_ team Jet Air Eiector - Offons System in series with the isolated condenser method, the steam jet air ejectors i (SJAEs), steam seal system, and offgas system can be used to collect and process MSIV leakage where it is filtered, treated, and delayed prior to discharge to the plant stack (elevated release point). Figure 4 4 shows a schematic diagram of this method.

This method is more effective than the mechanical vacuum pump method because of the filtering effect by the offgas system. Ilowever, this method requires substantial operator actions and availability of SJAEs and the offgas system.

4.3.1.4 Isolated Steam Lines In this method, the main steam lines are isolated from the condenser, turbine. and environment. The turbine stop, control, and bypass valves allow some minor leakage that will eventually migrate to the high pressure turbine and the condenser.

This method is not as effective as the isolated condenser method because the high pressure turbine has a smaller hold up volume for fission product attenuation. The iodine removal capability in the high pressure turbine is not adoquate to reduce the radiological consequences of significant MSIV leakage. Therefore, this method is not recommended.

4.3.2 Seismic Evaluation According to 10CFR100 Appendix A. those systems, components, or structures which are designed to assure the capability of mitigating the consequences of design basis accidents must be Seismic Category I. Main steam lines downstream of the outboard MSIVs and the condenser are not usually designed to Seismic Category I requirements.

1 6

e- .+1 wr w-- += s--e ums-.. A- --e-

NED0 31058 Rev. 1 The BWROG has evaluated the availability of the main steam system piping and condenser for HSIV leakage treatment. The BWROG has reviewed the potential combinations of LOCAs and seismic events of interest:

(1) t0CA WITHOUTE&LLOINCIDENT sElsMIC tytH1, for this occurrence the pressure in the piping system downr.tream of the MSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the flow path through main steam system piping to the condenser is assured.

~

(2) ifl1MIC EVENTJ11tt0VT NEAR COINCIDENT LOCA. Without A LOCA and the potential associated core degradation, the radioacti/ity transported with MSIV leakage is of no radiological significance.

(3) LOCA WITH NEAP M INCIDENT SEISMIC EVENT. For this occurrence (also assuming significant core damage), the consegurnces are of interest because a seismic induced failure in the main steam or condenser system could allow HSIV leakage to bypass th>; treatment pathway, it has been previously well documented that the probability of a near coincidentLOCAandseismiceventisextremelysmall(designbasis earthquake probability approximately 0.001 per reactor per year; cere melt probability is plant specific and typically ranges from 0.00001 to 0.0001 per reactor per year). It is also noted that a LOCA does not induce a seismic event, and that a seismic event has very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10-7 per reactor per year).

Considering that the probability of a near coincident LOCA and seismic event is much smaller than other plant safety risks (less than 1 x 10*7 per reactor per year for coincident events, less than 5 x 10 7 per reactor per year for seismic. induced LOCA), the concern for main steam piping or condenser L

l l

47

NED0 31858 Rev. I damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design basis seismic events.

ANSI B31.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin.

In addition, specific seismic design provisions have been incorporated into some newer BWR main steam and condenser systems.

To further justify the capability of the main steam system piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non seismically designed piping and condensers (in past earthquakes). The study summarizes data on the performance of main steam piping and condensers in past strong motion earthquakes and compares these piping and condensers with those in typical U.S. GE Mark 1. !!, and 111 nuclear plants. This limited earthquake experience data and similarity comparisons are then used to further strengthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in a design basis earthquake in conjunction with a LOCA occurring just before or after the seismic event.

The earthquake experience data are derived from an extensive database on the performance of power plants and industrial facilities and over 100 power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world ; rom 1934 to the present.

The study concluded that piping and condensers in the earthquake-experience database exhibited substantial seismic ruggedness, even when they were not designed to resist earthquakes. This is a common conclusion in stujies of this type on other plant items such as welded steel piping in general, anchored equipment such as motor control centers, pumps, valves, structures, and so forth. That is, with limited exceptions, normal industrial construction and equipment typically have substantial inherert seismic 4-8

--

• g y *<w + +r, e -, w m--+-rw ---2,m- e

NED0 31858 Rev. I ruggedness, even when they are not dcMgned for earthquakes. No failures of

• In steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.

Comparisons of piping and condenser design in example GE Mark I, 11, and )

111 plants with those in the earthquake experience database reveal the GE plant designs are similar to or more rgged than those that exhibited good earthquake performance. The BWROG concluded that (1) the possibility of significant failure in GE BWR main steam piping or condensers in the event of an eastern U.S. design basis earthquake is highly unlikely and that (2) any such failure would also be contrary to a large body of historical earthquake expeelence data, and thus unprecedented.

Earthqucke experience data methods have been applied in seismic equipment qualification issues associated with Unresolved Safety issue A-46 (Seismic Qualification of Equipment in Operating Plants). Piping performance data have been presented in NUREG 1061, a report from the NRC Piping Review Committee, which proposes changes to criteria that are directed toward the recognition of the superior performance of piping in earthquakes and establishing more realistic seismic criteria for piping qualification. The NRC has published NUREG 1030 and NUREG 1211 " Seismic Qualification of Equipment in Operating Nuclear Power Plants," which conclude that the seismic experience data approach provides the most reasonable and preferred alternative to other current equipment qualification methods.

The rapidly growing use of the seismic experience data approach is further illustrated by the fact that this method of analysis is now referenced in:

r A. Draft RG 1.100, Revision 2 " Seismic Qualification of Electrical and Mechanical Equipment in Nucicar Power Plants" B. Recent approved revision of IEEE Standard 344-1987, " Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations" 4-9

NE00 31858 Rey, 1 i l

C. Draft report of ASME Standard " Recommended Practice for Seismic j Performance Qualification of Mechanical Equipment Used in Nuclear Power Plants "

The earthquake experience database includes a large number and variety of piping systems, in fact, piping is probably the strongest area in this regard j (compared to areas like electrical or mechanical equipment, cable trays, i etc.), it has been concluded that the earthquake experience data on piping, l and in particular data on main steam piping, are applicable to main steam piping in BWRs.

In both nucicar and conventional power pla e ,s the condenser is designed to reduce the low pressure turbine outlet pressure (thereby increasing turbine efficiency) and to co.idense the steam. The nuclear environment does not impose additional significant design considerations on the condenser. With the exception of hotwell size, a conveniunal plant and nuclear plant with similar performance parameters have similar condensers.

None of the condensers within the seismic experience database has seismic design criteria. However, in view of the performance of the condensers within the database, it is concluded that the condensers have an inherent seismic ruggedness and that the earthqueke experience date on condensers are applicable to condensers in BWRs.

Another recent study to develop, by c;ata collection and statistical analysis, updated estimates of pipe breaks in commercial U.S. nuclear power plants was completed in 1987 (Ref 6). This study evaluates both LOCA sensitive systems and non-LOCA sensitive systems. For BWR non LOCA sensitive systems, ten pipe failures have occurred over 313 years of operating experience. None of these failures occurred in the main steam piping. Based on the observed failure rates, this study estimated the failure rate for the main steam system piping to be 0.0007 failure / year /BWR with an upper bound of 0.0096 failures / year /BWR. These results are consistent with the conclusion i

! 4-10 l

NED0 31858 Rev. 1 '

from the earthquake databases and NUREG ll69: BWR main steam piping is designed to_ withstand severe plant transients such as turbine trips and is expected to remain intact following accidents as severe as a design basis  :

LOCA. Thus, the non-seismically designed main steam piping and the main  !

condenser can be used to mitigate the consequences of MSIV leakage. '

1 i

'In addition.to the comprehensive earthquake experience database evaluation, a plant specific verification of main steam piping and condenser l seismic adequacy will be performed to provide reasonabic assurance of the integrity of these systems and components. Section 6.7 provides guidelines for this verification.

4.3.3 Radiological

Evaluation The BWROG has develo;)ed a radiological dose methodology using the isolated condenser method for MSly leakage treatment. ,

I The BWROG analysis assumes LOCA conditions in which the core is degraded with source term releases consistent with those assumed in RG 1.3 (Ref. 7).

Noble gases and iodines in an atrount _of 100% and 50% of the core inventory, respectively, are immediately released to the containment air space, and in

.accordance with RG 1.3, 50% of the iodine released to the containment is ,

considered to be_ removed by the containment. lodine composition are 91%

-elemental, 5% particulate, and 4% organic. The following summarizes the scenarios and assumptions for MSIV leakage calculations:

(1). As a result.of H51V leakage, a fraction of the containment activity -

enters.the main steam lines. In accordance with the single active failure requirements, one HSIV in_one line remains open after containment. isolation.

(2) Leakage from closed MSIVs enters the following pathways:

(A) Main steam drain lines transport leaktge into the condenser -

This path is preferred for MSly leakage treatment. When the selected drain valves (See Sectim 6.1.1 for requirements) are h

4-11

)

an , , w, .. - -- - - - -- - -

-A

NED0-31858 Rev. 1 open, a large fraction of the MSly leakage (approximately 99%)

rnters this pathway. MSIV leakage that enters the condenser is

. subsequently released to the turbine building through the low-pressure turbine seal.

(0) Tra.n , ort through the closed turbine tJ ess valves into the condenser - A small fraction of MblV leakage ent irs this path when the dr:iin valves are 9 pen. Dose reduction from this path is essentially equivalent to the drain line pathway because -

MSIV leakage frcm either pathway is precessed via the condenser, (C) Leakage through the closed turbine stop valves and control valves into the high-pressure turbine. A small fraction of MSIV leakage (typically less than 1%) enters this path when the drain valves are open. This path is not desirable because the mitigating capability in the high-pressure turbine is not as effective as the condenser (especially in the first few hours when temperatures are elevated in the high-pressure trubine).

-MSIV leakage that enters the high-pressure turbine is subsequently released to the turbine building through the high-pressure 5urbino seals.

The consequences of MSIV leskre c.3 reduced by iodine deposition on inside surfaces of the main steam s.ipir; nd the condenser, and by the delay time for the activity transport. Iodine kposition is a function of the inside surface temperatures of the piping and condenser, surface areas, and residence time of the activite transport. The time delay is_ based on a conservative plug-flow model. In addition, a fraction of the elemental iodine that has been deposited is calculated to be resuspended and converted to-organic iodine.

4-12

_. -__._. _ _ _ . .~_ __

NEDO-31858 Rev. 1 The analysis calculates MSIV contribution doses at the control room, exclusion area boundary (EAB), and low population zone (LPZ). For most BWRs, the revised LOCA doses are the sum of the previous LOCA doses (as identified in the USAR) and the calculated specific MSIV doses at the proposed leak rate limit. This method is very conservative since the previous dose calculation already included dose contribution from MSIV at the current Technical Specification leak rate limit. Some BWRs may elect to replace the previous LOCA MSIV contribution doses with the new doses using the proposed treatment method.

Table 4-1 summarizes MS!V dose contributions for Hope (., reek fcr MSIV leakage rates of 100 and "00 s:fh per steam line. MSly leakage rntes that can be tolerated for aev given site depend on plant-specific parameters. The total dose frce MS!V leakage includes specific contributions from noble gases, inorganic iodines, organic iodines from RG 1.3 sources, and organic iodine from resusp.. ion / conversion from the deposited iodines. The analysis also delineates de * ;ntributions from the drain line pathway and from the high-pre, cra tur.iine pathway.

For 110 scfn ,>er steam line, the total dose contribution at the off-site from MSIV ie &..e is 22.1 reni '7.3% of 300 rem limit) to the thyroid. Only 0.05 percent of this dose is frou the high-pressure turbine pathway. 96% of the off-site dose is from organic iodine. The total dose at the control room is 1.05 rem (3.5% of 30 rem limit) to the thyroid. 96% of the control room dose is from organic 'odine. The whole body doses are insignificant.

For 200 scfk 'eam line, the total dose contribution at the off-site from MSIV leakage sa .5 rem (19% of 300 rem limit) to the thyroid. Only 0.05% of this dos ;s froc the high-pressure turbine pathway. 87% of the off-site dose is from organic iodine. The total dose at the control room is 2.7i rem (9.0% of the 30 rem limit) to the thyroid. 87% of the control room dose is from organic iodire. The whole body dose remains insignificant.

4-13

. , _ . . . . . . . . . "- i.i ' .  : ' ; - - - - ' - - - - - - - - - ' - - - - - - - - - - - - - - - - - -

l NED0 31958 Rev. 1 The 2-hour EAB off-site doses are not shown in Table 4-1. For 200 scfh per steam line, the Hope Creek EAB doses are 0.10 rem whole body and 2.6 rem thyrold.

From results of the BWROG analysis, the MSIV contribution doses are dominated by the organic iodine because of the ultraconservative source term assumptions used in this analysis. For 200 scfh per steam line, more than 85%

of the off-site and control room doses are from organic iodine, based on the assumptions from RG 1.3 source terms and from organic iodine converted from the elemental iodine deposited in main deam piping systems. If a realistic iodine composition from the fuel release (cesium iodide) is used in the calculations, essentially all of this organic iodine dose would be eliminated.

4.4 TECHNICAL JUSTIFICATION FOR ELIMINATION OF LEAKAGE CONTROL SYSTEM As discussed in Section 2.2, at MSIV leakage rates greater than 100 scfh, the safety-related LCS will not function following a postulated design basis LOCA. In lieu of the LCS, BWRs have inherent MSIV leakage treatment capability. The main steam lines and condenser are significantly more reliable than the dedicated safety-related LCS for treatment of MSIV leakage.

Furthermore, the risk to the public health and safety is reduced with the implementation of the proposed MSIV leakage treatment method.

For plants that do not have a LCS, implementing the BWROG proposed changes will provide plants with a capability to treat MSIV leakage. Use of the main steam piping and condenser provides reliable and effective leakage treatment without exceeding the 10CFR100 off-site dose limits following a Design Basis Accident. The implementation also will provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to the public.

4-14

NE00-31858 Rev. I For plants with a LCS, the BWROG proposed changes involve a replacement of the existing LCS with the more reliable and effective main steam piping and condenser for MSIV leakage treatment. Except for the requirement to establish a proper flow path from the main steam drain line to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tignt barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment.  !

l The existing leakage control systems have limitations for mitigating MSIV leakage. For negative-pressure LCS, operation of the system induces highe-MSIV leakage by increasing the differential pressure across the valves, and is inconsistent with the philosophy of multiple barriers for-limiting fission product releases (does not take advantage of the outboard MSIVs). For positive-pressure LCS, operation of the system increases the containment pressure and thereby increases the containment leakage. The LCS requires multiple logic controls, interlocks, timers, containment isolation valves, and other equipment to ensure containment integrity and protection from the high

-pressure main steam lines. Based on plant operating experience, leakage control systems installed in most BWRs do not provide a high degree of reliability. Extensive control logic and interlocks in the LCS design frequently require calibration and maintenance; and spare parts.are increasing difficulty to- find. Also the LCS has limited capacity and does not function at moderate MSIV leakage rates above 100 scfh.

Even though the resulting off-r.te doses may be slightly higher for low MSIV leakage rates, the effectiveness of the proposed method even for leakage rates greater than the proposed increased allowable limits, ensures off-site l dose limits to the public are not exceeded. Overall, the proposed treatment method can handle MSIV leakage over an expanded operating range. Thus, a l margin of safety exists. Furthermore, it is clearly a safety improvement to f replace a system with known limitations with tne alternate main steam piping and condenser treatment pathway, which has been shown to har excellent reliability.

4-15

NEDO 31858 Rev. 1 The safety significance of the LCS in terms of public risk was addressed in NUREG CR-4330 (Ref. 8). This NUREG evaluated the possibility of regulatory modification concerning elimination of the LCS requirements and disabling the systems currently installed at BWRs. The NUREG concluded that streamlining the regulatory requirements would have little impact on public risk, and that substantial savings in operating costs may be realized. The NUREG also concluded that the increased overall public risk is less than one percent, but the overall savings are estimated over several million of dollars for the remaining life of the plants (sum of all plants).

s 4-16

NED0-31858 Rev. 1 Table 4-1

- Contributions of MSIV Leakage to Offsite and Control Room Radiological Doses CONTROL ROOM OFF-SITE Whole Whole MSIV Leakage at Body Thyroid BETA Body Thyroid 100 scfh/line liL (30) (30-75) 121L. (300)

1. Noble Gas (DL) 0.04 0.00 0.59 0.10 0.00

2. Inorganic I (DL) 0.00 0.04 0.00 0.00 0.81

3. - Organic I (DL) 0.00 0.60 0.00 0.01- 12.73

4. Noble Gas -(HPT) 0.00 0.00 0.00 0.00 0.00

5. Inorganic I (HPT 0.00 0.00 0.00 0.00 0.00

6. ' Organic I (HPT) ) 0.00 0.00 0.00 0.00 0.01

7. Organic I via 0.00 0.41 0.00 0.00 8.60 Re-suspension /

Conversion (DL)

TOTAL (100 scfh) 0.04 1.05 0.59 0.11 22.12 Whole Whole MSIV Leakage at Body Thyroid BETA Body Thyroid 290 scfh/line _ (5) (30) (30-75) (25) (300)

1. Noble Gas (DL) 0.10 0.00 1.64 0.31 0.00

2. Inorcinic I-(DL) 0.00 0.36 0.00 0.D1 7.58

3. Organic I (DL) 0.00 1.53 0.00 0.02 32.48

4. Noble Gas (HPT) 0.00 0.00 0.00 0.00 0.00

5. Inorganic I (HPT) 0.00 0.00 0.00- 0.00 0.00

6. Organic I (HPT) 0 30 0.00 0.00 0.00 0.03 .

7. Organic I via 0.00 0.82 0.00 0.00 17.38 Re-suspension /

Conversion (DL)

TOTAL (200 scfh) 0.10 2.71 1,64 0.34 57.47 I - Iodine DL - Drain Line Path HPT - High-Pressure Turbine Path 4-17

g i i j NEDO.31858 Rev, 1 Plant Vent Stack k

Secondary Turbine Primary Containment Building .-

Containment r- m MSIV MSIV e

................n...-......n..................n..........j I

SV E  :

[T*Q K SRV E BV Vacuum E MO CV g Breaker I \ /

l E Mo ..[ -

i ongas RPV condenser u

u_ ,-

i .l SJAE Mechanical Vacuum Pump -

• MAIM STEAM SHUT-OFF VALVE (BWR/6 AND SOME BWR/5)

Figure 4-1. Isolated Condenser Flow Path - Turbine Bypass Lines 4-18

NEDO 31858 Rev. 1 Plant Vent Stack 5

Secondary Turbine Primary Containmen; Building :l Containment

r m usiv usty .

....=....n.. ....x. ....n ,

,y g .

E sRv m av y.,yy, l  ;.

!! E MO cv r smaker 5  :

\/

5 m uo I"" ]E -

-.3 .

i..............+

-- +

RPV L._ ong.

u s condenser 7 ,

f I I ME y,,,,,,,,

vacuum Pump

• MAIN STEAM SHUT-0FF VALVE (BWR/6 AND SOME BWR/5)

Figure 4-2. -Isolated Condenser Flow Path - Orain Lines 4

1 NED0 31858 Rev.-.1-  :)

l l

1

)

Plant Vent Stack 4-Seconda7 Turbine

_ Primary : Containment Building Containment r m-MS!V WSlV e

_.. ....n........n.............n .

7r * '! E EV

-sv z u ggy -. Vacuum g

II MO CV g Breaker MO; '

i... ... ......... L' 4

RPV'.' cmgu p consenwr 7

c- a _ _ _ ,

I I;;

&JAE Wechanical '-

Vacuum Purg -

L t

• MAIN STEAM SHUT-0FF VALVE-

'(BWR/6ANDSOMEBWR/5)~

Figure:4-3. Mechanical Vacuum Pump Flow Path

4-20 l

7 NED0 31858 Rey, 1 Plant Vent Stack '

A, i

Seconda7 Turbine ,,

Primary Containment Building  :.

Containment r m .

o usiv - Msiv . .

.. . . x . _ . . x . . . . . . . . .. x __

Tq ,

[T'il -

E SRV i e av 8V E Yzuum

{  :'

x w wo: cv E 'er aer  :.

b uo -- I'~  !

.!. N -

y.

I.......

. . ... ... A 'I RPV L_. -

ongu p 1

e u Condenter . su.

a 1 I SJAE u cnancel Vacuum Pump i

i

• MAIN STEAM SHUT-OFF VALVE l

__(BWR/6 AND SOME_BWR/5)

Figure 4-4. Steam Jet- Air Ejector /0ffgas System flow Path

{

l I

4-21/4-22 4

NE00-31858 Rev. 1 5.0 OPERATIONAL, MAINTENANCE, AND SAFETY BENEFITS 5.1 BENEFITS ASSOCIATED WITH INCREASING MSlY LEAK RATE LIMITS By setting the leakage limit for MSIVs at a realistic value (up to 200 scfh), significant safety and plant operational benefits can be realized.

Recent test data show improved leakage performance, but MSIV leakage rates in excess of the current Technical Specification limits still frequently occur. Although the MSIVs can be repaired or refurbished to meet the current limits, the proposed increase in the allowable MSIV leak rate up to 200 scfh  !

will provide a more realistic leakage limit, which will reduce MSIV repair and refurbishment costs, reduce dose exposures to maintenance personnel, reduce outage durations, and extend the effective service life of the MSIVs. l To quantify some of these benefits, the BWROG participants have provided data on repair time and dose exposures for MSIVs that failed the initial local leak rate tests. This data is included in Appendix A.

Of the initial local leak rate tests, 45 out of the 62 tests that failed

.to meet the current leak rate limit included data on repair time and/or dose exposures. Review of the data showed that there is no correlation between MSIV leakage rates and repair times to the specific HSIV design or to inboard versus outboard configuration. The dose exposures to maintenance personnel are inherently higher for the inboard than the outboard MSIVs.

Table 5-1 summarizes the data. Key conclusions supported by this data are summarized below:

a. An average of about 350 man-hours is required to repair or refurbish a typical MSIV that fails the initial local leak rate test.

. Maintenance times.as short as 82 man-hours and as long as 1240 man-hours are reported.

l 5-1

NED0-31858 Rev. I

b. For BWRs with a long operating history, dose exposures to personnel for each MSIV repair are one to two rem whole body. Since there is no inherent difference in valve failure rates between the inboard and outboard MSIVs, an averaged dose exposure value based on all inboard and outboard data is applicable for estimating plant exposures.

c. Table 5-2 contains an estimate of average cost and exposures for BWRs with a long operating history. For a typical BWR with 11.5 scfh allowable leak rate, an average labor cost to repair / refurbish MSIVs is $31,500 per test interval and the average whole body dose exposure is approximately 2.0 rem per test interval,

d. For a proposed allowable MSIV leak rate up to 200 scfh, labor cost savings of greater than $18,000 per test interval can be realized.

The corresponding reduction in dose exposures to maintenance personnel is greater than 1.1 rem per test interval (see Table 6; estimates conservatively based on comparison of leak rate test failures at 11.5 and 100 scfh),

c. MSIV repair often results in extended outage lengths, thereby reJuc-ing plant availability. One failed valve, listed in Appendix A, resulted in a 48-hour increase in outage duration, which cost the utility approximately $1 million in replacement energy costs,

f. Excess machine lapping of MSIV seating surfa m @ mes seat thickness and increases m,ir +cm.Eu rMbirements in future years.

5.2 BENEFITS ASSOCIATED WITH DELETING LEAKAGE CONTROL SYSTEMS The LCS has blowers, isolation valves, and extensive instrumentation that must be tested in accordance with the Technical Specifications. Six BWROG participants have provided cost and dose exposures associated with maintaining the LCS. Other data provided include information on equipment reliability and replacement parts.

5-2 l

NED0-31858 Rev. 1 An evaluation-of the performance data presented in Section 3 shows that:

a. An average cost of about $50,000 per year is required to maintain the system. Annual maintenance costs as low as $15,000 and as high as

-$150,000 are reported,

b. Dose exposure to maintenance personnel is about one man rem per year per plant.

c. Because the LCS consists of extensive instrumentation, the system is frequently declared inoperable. At one plant, there have been 64 surveillance test failures and about 250 maintenance items since installation. At another plant, the LCS has become critical path during'every refueling outage,

d. A ct:rrent probler with the flow element has a high potential for plant operational impact. A solution to the problem is in progress at this time; however, repair costs are projacted to be high.

Irrespective of the direct labor cost to repair the MSIV and maintain the

'LCS, the major costo are. associated with extended outage durations and critical paths. These costs can be very high,-in the' order of millions, because of the high cost of energy replacement. Finally, it is becoming more difficult-to obtain appropriate environmentally qualified replacement

! components for -the safety-related LCS.

Furthermore, as previously discussed in Section 4.4, the risk to the public health and safety is reduced with the implementation of the proposed MSIV leakage treatment method. The proposed method can handle realistic MSIV leakage rates up to 200-scfh. Thus, a margin of safety exists. -It is clearly a' safety improvement to replace a system with known limitations with the

. alternate' main steam piping and condenser treatment pathway that has been shown to have excellent reliability.

5-3

NEDO-31858 Rev, 1

. Table 5 1 Man Hours and Exposure Required to Repair and Retest NSIVs that Failed Initial-Local Leak Rate Test Avg - Exposure Exposure . Average No, of

' Pl ant - Man- -Inboard Outbeard Repairs:to Pass LLRT

-No. lig.ttti frem) (rem) After Initial failure Comment 1- 165 1,03 -0.36 1.7 2 1,55 0,77

2. 530 - ---

-$60K Cost J3'- -275 9,00 1.25 ---

4; 1020= ~0.05- 0,05 1.3 Newer - Plant

- 5 ': 240 0'.04 0,03 1.0 Newer Plant l6' '350 2.25 1.37 --- $20K Cost 71 300 -- --- ---

Newer Plant NOTE:t Data is'11sted in order of quality of input-(i.e., _ plants at the . top-of the-listing maintained better. records; data shown at the bottom ' -1 are: estimates provided by responsible maintenance personnel).

. 5-4

u _ , . _ . .- _ _ . _ . , _ . _ _ . _ _ _ . . . . _ _ . . . . . . _ . .

.)

fj NED0-31858 Rev.-l- j i

Table 5 2--_

Estimates of Average Repair Time and Daso Exposures For a A Specified MSIV Leak Rate Limit '

H51V Leak Rate limits 11.5 scfh -Ja,0_irf).  !

1 ,

x

.~

(1) Probability =that initial- 77% 90%

test' meets the limit-  ;

l(2)lNo. of MSIVs that meet test limit 6.2 7.2

. (3)1 No.fof: MSIVs = that ' fail ' test limit - 1. 8 - 0.8- >

(4)-Average total man-hours 630 mhr 280 mhr

.requiredito repair / refurbish

~

-failed MSIV (#3_x 350 mhr per MSIV)

,(5) Average: total labor cost $31,500 $14,000- a w

(#4 x:550) -i (6) Whole body. dose exposure to 2.0 rem 0.9 rem. ,

maintenance personnel

- (#3 x.l.1: rem /MSIV)-

9 n

a _ BASIS:

1 . Average man-hour.per'MSIV = 350 A: l Average _ labor, cost = .550'per hour ,

- Average dose-exposure per valve = 1.1 rem Tutel numbers of MSIVs - 8 i:

5-5/5-6

w . . .- . - - - ~. - - - - .

Nr00-31858 Rev, 1:-

t 6.0; METHOD OF IMP CMENTATION LThe BWROG_MSIV-Leakage Closure Committee recommends-tho following proposed changes to'the plant Technical Specifications:

(1) Increase the MSIV leak rate from the current specified limit to a

' plant. specific higher limit, and (2) For.cpplicable plants, eliminate the requirements for the LCS.

-The-current MSIV leak rate is typically limited to 11.5 scfh per main steam line. DependingTon plant-specific design and radiological consequences, the' limit'can;be increased up to a maximum of 200 scfh per steam line. The y BWROG and valve manufacturers have-concluded that MSIV leak rate up to 200 scfh per_-valve will'_not inhibit the isolation performance of the valve, For plants where the. radiological consequences approach the regulatory

~

guidelines, MSIV leak rate: limits can be increased on per line and total c leakage basis. 'For example, not to exceed -100 scfh per main steam line while not exceeding 200 scfh total,

!!mplementation of the proposed changes involves the following actions by

the utility:;

(1)' _ Pl' ant:Lspecific' evaluations, (2) Radiological: analysis-for the specified MSIV leakage pathway;

-(3) Amendment to the Technical' Specifications and Exemption Requests,

'(4)' Changes _to the Updated Safety Analysis Report,'

[(5) Implementation.of.EmergencyOperatingProcedures, (6) Implementation of plant-design changes or modifications. _l

,1 6-1  :

i

NED0-31858 Rev. 1 The utility will complete the plant-specific evaluations, radiological analysis, and submit the licensing change request for NRC review and approval.

Upon NRC approval, the utility will then complete the USAR update, verification of seismic adequacy, E0P implementation, and re,uired design changes.

6.1 PLANT-SPECIFIC EVALUATIONS 6.1.1 Isolated Condenser and Main Steam Drain Lines The feasibility of increasino MSIV leakage rate limits rely on use of the main steam piping-and the condenser for fission product attenuation. Section 4.3. identifies this leakage treatment pathway as the " isolated condenser" method. The utility will evaluate the plant capability to divert MSIV leakage from the main steam drain line to the main steam condenser following a design basis LOCA. The evaluation will consider the following requirements.

(1) The main steam drain line downstream of the outboard MSIV should be used to convey MSIV leakage to the condenser.

(2) The internal cross-sectional area of the drain line should be approximately 2.88 square inches or larger. This area corresponds to a nominal diameter of two inches. This drain line flow araa is needed to minimize the fraction of the MSIV leakage which leaks through the turbine stop and control valves 'into the high pressure turbine. Depending on plant specific parameters a somewhat lower flow area may be acceptable.

(3) In each drain line, there may be two or more drain valves that must be capable of being opened even if off-site power is not available.

Therefore, power must be supplied from the essential power busses.

Since these drain valves are usually powered from non-essential power sources, minor design changes may'be required.

6-2

NED0 31358 Rev. 1 (4) A plant-specific verification of seismic adequacy of the main steam pi $ g and condenser must be performed to assure the integrity of these components. Section 6.7 provides guidelines for this verification.

S.I.2 Backup MSIV Leakage Treatments Section 4.3, NUREG-ll'69, and NEDO-30324 provide discussion of several methods for HSIV leakage treatment, which may be used as a backup or enhancement to the " isolated condenser" method. The utility may evaluate the ,

feasibility of using these backup methods for diversity and enhancament in treating MSIV leakages. Plant procedures should be modified accordingly.

6.1.3 Leakage Control System For plants with the LCS, the utility may completely remove or disable the entire LCS, or disable the inboard part of the LCS and use the outboard part as a backup for MSIV leakage treatment.

6.1.4 Plant Specific MSIV and LCS Issues Sect;ons 3 and 5 provide discussion of MSIV and LCS performance issues and benefits based on the BWROG surveys of some BWRs. The utility ehould evaluate its plant-specific MSIV and LCS performance issues and benefits when applying for changes to the Technical Specifications.

6.2 RADIOLOGICAL ANALYSIS OF THE MSIV LEAKAGE PATH GE has developed, for the BWROG, a radiological dose methodology using the isolated condenser method for MSIV leakage treatment. A computer code calculates the radiological effects of the proposed MSIV ieak rate in terms of off-site and control room doses.

For most BWRs, the calculated MSIV doses at the proposed leak rate limit can simply be added to current LOCA doses (identified in the USAR) to yield the resultant LOCA doses. Hope Creek uses this approach and Table 6-1 6-3

-l

'NEDO 31858 Rev..'l

'contains a. summary of the results. This_ approach is very conservative since

'the' current-USAR LOCA doses already include contribution from MSIV leakage at the maximum' rate-permitted in the Technical-Specifications. Some BWRs may elect to replace:the current MSIV doses with the new HSIV doses' calculated by the BWROG code.-

o

-6.3- TECHNICAL SPECIFICATION CHANGES AND EXEMPTION REQUESTS 6.3.1 License Amendment' Per 10CFR50.90, the proposed Technical Specification changes require an

-application- for license. amendment. Plant-specific changes _ vary depending on

-the implementation method for the 10CFR50 Appendix J requirements, and whether -

a LCS has been -installed.

-6.3.2 Exemption Requests 6.3.2.1 Exemotion Reouest to 10CFR100. Anoendix A 10CFR100, Appendix A requires equipment thEt-assure the capability to

mitigate .the-10CFR100 radiological consequences!for off-site. doses.be designed to remain functional. following.a seismic event coincident.with a design basis

accident.- The engineering-. method used to assure that the required safety functions are maintained following a SSE; involves the use of either dynamic

analysis or a suitable qualificati.on test to demonstrate that components can withstand the seismic and'other concurrent loads. >

For. all BWRs, an application for. exemption to 10CFR100, Appendix A_ is

~

required Lto' allow the use of the existing' main _ steam piping and condenser-as a-idesign basis for MSIV;1eakage treatment.. This exemption is only for_MSIV~

cleakage~ evaluation against the off-site' dose guidelines in 10CFR100 for the.-

idurat_ ion-of a postulated LOCA. Although there is:no specific provision for

_ exemption'in'10CFR100, the BWROG_ realizes that the NRC has the authority to evaluate:and approve such. exemption. Justifications for the exempti_on request are discussed in Section 4.3;2 and are summarized below:-

6-4 n >

~- "-

,; . , , _ . _ , _ . . . . , _ . m . ._. , ,

NED0-31858 Rev. 1 (1) Probability for which the resulting dose from MSly leakage is significant is extremely low. This requires a design basis LOCA, a degraded core where ECCS are not functional, and a significant seismic event.

(2) The main steam piping and condenser are designed to strict industrial standards and building codes; thus, significant design margin exists.

(3) The piping and condenser exhibit substantial seismic ruggedness.

Comparisons of piping and condenser design in GE plants with those in the earthquake experience data base reveal the GE plant designs are similar to or more rugged than those that have exhibited good earthquake performance.

(4) Possibility of significant failure in GE BWR main steam piping or condensers in the event of design basis earthquake is highly unlikely, and any such failure also would be contrary to a large body of historical earthquake experience data, and thus unprecedented.

(5) A plant-specific verification of seismic adequacy of the main steam piping and condenser will be performed to provide reasonable assurance of the structural integrity of these components.

6.3.2.2 Exemption Reauest to-10CFR50 Anoendix J For some BWRs, exemption requests to the Type A and/or Type C test acceptance criteria of 10CFR50 Appendix J are necessary for increasing the

, - allowable MSIV leak rate. Appendix e test requirements ensure that containment leakage following a design basis LOCA will be less than the designed containment leakage (La) assumed in the safety analyses for radiological consequences. Type A test requires that an Integrated Leak Rate Test (ILRT) of the containment be periodically tested for leakage and that the test result be less ~ than or equal to 0.75 La. The ILRT usually incluoes leakage through the closed MSIVs. Also, Appendix J Type C test requires that a Local Leak Rate Test (LLRT) of each containment isolation valve be l

6-5

NE00 31858 Rev. 1  !

I periodically tested for leakage and that the combined test result of all penetrations and valves be less than or equal to 0.601 a. The LLRT acceptance criteria somet.mes include MSIV leakage.

If the ILRT and LLRT leakage limits include MSIV leakage, then the proposed increase in the MSIV leak rate limit will ex:eed the Appendix J acceptance test criteria. For example, a typical BWR/6 plant has a design leak rate of 417 scfh (t4ased on a containment free air volume of 1 x 106 ft3 and a design containmeni leak rate of 1% per day), Increasing the MSIV leak rate limit from 11.5 scfh to 200 scfh per steam line will exceed the Appendix J Type A and Type C test criteria. Therefore, MSIV leakages must be exemptsd from inclusion into the Type A and Type C acceptance test criteria. Some BWRs currently have exemption to Type C criteria for MSIV leakage because this leakage bypasses Standby Gas Treatment System (SGTS) and, therefore, has a different radiological impact from the containment leakages that are filtered by the SGTS.

Justifications for the exemption to 100FR50 Appendix J are:

(1) The radiological consequences of MSIV leakage (release directly to turbine building, bypasses the SGTS) are calculated separately from the radiological consequences of containment leakage.

(2) MSIV leakage rates are periodically measured as part of the LLRT to ensure the leakage rates will not exceed the proposed r..aximum rate assumed in the MSIV radiological analysis. Therefore, the intent of Appendix J is maintained.

6.3.3 Example of License Amendment and Exemption Requests To facilitate the utilities applying for the changes, GE has drafted four different generic examples of License Amendment and Exemption Requests for NRC submittals. These example submittals are included as Appendix B. These example submittals include the following applicable License Amendment and Exemption Requests:

6-6

e .- -

NEDO 31858 Rev. 1 (1) The Type 1 example submittal applies to BWRs that do not have a LCS but have exemptions from 10CFR50 Appendix J Type A and Type C for MSIV leakage. The submittal includes a license amendment for inct easing the MSIV allowable leakage rate and exemption requests to 10CFR100 Appendix A. For BWRs that currently have an exemption from 100FR50 Appendix J Type C only, the Type 2 example submittal can be revised to incorporate exemption for Appendix J Type A criteria only.

(2) The Type 2 example submittal applies to BWRs that do not have a LCS.

The example submittal includes a license amendment for increasing the MSIV leak rate limit, exemption request to 10CFR100 Appendix A, and exemption request to 10CFR50 Appendix J Type A and Type C.

(3) The Type 3 example submittal applies to BWRs that have a LCS and have exemptions to 10CFR50 Appendix J Type A and C. The submittal includes: (A) a license amendment for increasing the MSIV leak rate limit and for eliminating the requirements for the LCS; and (B) exemption request for 10CFR100 Appendix A. For BWRs that currently have exemption from 10CFR50 Appendix J Type C only, the Type 4 example submittal can tw revised to incorporate the exemption from 10CFR50 Appendix J Type A testing criteria. "

(4) The Type 4 example submittal applies to BWRs that have a LCS. This submittal includes: (A) license amendment for increasing the MSIV

. leak rate limit and for eliminating the requirements for the LCS; (B) exemption request to 10CFR100 Appendix A; and (C) exemption request to 10CFR50 Appendix J Type A and Type C.  :

For BWRs that do not have MSlv leak rate limits specified in the Technical Specifications, utilities usually specify administrative leak rate limits for the MSIVs similar to those for other containment isolation valves.

The leakage rate limits for these isolation valves, including MSIVs, are

-inherent in 10CFR50 Appendix J acceptance criteria. For these BWRs, the BWROG recommended changes are not applicable unless the utility proposes: (1) to add a MSIV leak rate limit in the Technical Specifications; and (2) to exempt MSIV leakage rate from the acceptance test criteria of 10CFR50 Appendix J.

6-7

NEDO 31858 Rev. 1 6.4 CHANGES TO UPDATE SAFETY ANALYSIS REPORT The utility should revise the USAR to address LOCA radiological analysis and control room habitability that are affected by the revised LOCA doses, With the deletion of.LCS,-the USAR section on the MSIV Leakage Control System (typically Section 6.7) will be affected by system deletion.

6.S INPLENENTATION OF ENERGENCY-OPERATING PROCEDURES Chtnges to the E0Ps may be required to identify necessary operator actions-to mitigate MSIV leakage consequences. An earlier BWROG study (Appendix C) astessed potential operator actions to limit radioactive gas ,

release through the_MSIVs and-provided typical changes to the E0Ps.

6.6- INPLEMENTATION OF-PLANT-DESIGN CHANGES OR MODIFICATIONS Upon NRC approval of the Technical Specification amendment, plant design changes may be necessary to remove or disable the LCS, and to install essential-power supply to the specified main steam drain valves. In addition, the normal plant' operation and maintenance procedures will need to be reviewed and updated as required.

- Also,- the Utility will institute into the MSIV maintenance and test program, the requirement that if any MSIV exceeds the proposed leak rate limit

(e.g., 200 scfh), that MSIV will be repaired and retested to meet a leak rate

-limit _ at the current Technical Specification limit (e.g.,11.5 scfh). This will assure' continuation of-high quality repair and refurbishment efforts to improve the overall performance and reliability of the MSIVs.

~6.7' GUIDELINES FOR PLANT-SPECIFIC VERIFICATION OF SEISMIC ADEQUACY OF MAIN STEAM PIPING AND CONDENSER

'The primary compeaents-to be relied upon for pressure boundary integrity c in resolution of the BWR MSIV leakage issue are:-(1) the main turbine condenser, -(2) the main steam lines from the MSIVs to the turbine stop and 1

6-8 l

NED0 31858 Rev. 1 bypass valves,.and (3) the main steam turbine bypass and/or drain line piping to the condenser. Plant-specific reviews and evaluations should be performed to provide reasonable assurance of the integrity of these systems and components.

6.7.1 Seismic Failure Modes In general, design conditions which may result in seismic induced pressure boundary failure and inventory release of main steam and drain piping include support failure and falling of non-seismically designed plant features (II/I), proximity impact, and differential seismic anchor motion on piping systems.

Seismic (II/I) proximity interaction _is the impact of adjacent equipment or structures on components due to their relative motion during seismic excitation or support failure and falling of non-seismically designed components. Relative motion can be the result of the vibration and movement of the pipina and eqitipment or failure and movement of adjacent equipment or

-structures.

Seismic proximity and impact issues have been extensively studied and reported on in the post earthquake reconnaissance literature. Substantial work has been done by the Seismic Qualification Utility. Group as well as the structural engineering and research communities in the identification of proximity as a potential source of damage to a limited number of fragile components. Welded steel piping has demonstrated ruggedness and few instances

-of interaction damage are known.

The BWROG seismic experience study has concluded that possibility of significant-failure in GE BWR main steam piping or condensers is highly

- unlikely and that any such failure also would be contrary to a large body of historical earthquake experience. data, and thus unprecedented.

6.7.1.1 Failure and Proximity Imoact Equipment and piping can be damaged and lose pressure boundary integrity 69

NED0-31858 Rev. I equipment, systems, or structures. Non-seismically designed and installed equipment can fail by either loss of structural integrity or support anchorage failure. Seismic experience data has demonstrated that structural failure of anchored equipment leading to collapse is generally not credible with few exceptions which are readily identifiable by plant walk-down.

Assurance that detrimental falling hazards and proximity interactions df not exist may be provided by review of existing non-safety design programs, plant walk-throughs, and/or analysis. Potential seismic interactions should be evaluated for piping components such as valve operators, vents, drains, instrumentation, and fragile appurtenances.

Existing non-safety design programs may be reviewed to identify structures and components which have been evaluated for position retention integrity under earthquake loadirigs. These could include original seismic and wind design bases for the turbine building structure, responses to Generic Issue 80-11 on masonry walls and design basis programs to comply with RG 1.29 for II/I type issues.

Earthquake experience data also may be used to assess component structural integrity in direct field walk-down assessment for II/I type effects. Walk-downs can be used to focus only on realistic hazards and verify design attributes important to seismic performance and to identify non-typical commodities with uncertain seismic capacity.

6.7.1.2 Differential Seismic Anchor Motions Piping which crosses independently founded structures or similar conditions which impose differential anchor motions should be evaluated to ensure that adequate piping flexibility exists to precluded failure.

Piping should be reviewed for seismic anchor movements imposed by the following three conditions:

6-10

NE'10-31858 Rev.1 (1) Excessive Movement of Terminal End EAViDERD1 Piping performance can 3 be insured by verifying that adequate anchorages are provided to the terminal end equipment such as pumps, tankt, aat exchangers, etc.

(2) Differential Movement Between Pine SuDDorts in Ad_iacent. Uncoupled fLuildinas - Differential displacement concerns can be identified by 7 evaluating potential relative displacements and assuring that the piping has adequate flexibility.

(3) Excessive Moyements Imposed on Small Branch Lines by Flexiole Headers

- For rigidly attached branch piping, the effects of the movements of a flexible header should be evaluated.

6.7.2 Scope of Review Specifically, the seismic effects discussed above will be reviewed for main steam piping, drain line piping, condenser components, and their supports as described below:

6.7.2.1 Main Turbine Condenser The condenser design will be reviewed to demonstrate that it falls within the bounds of design characteristics found-in selected conventional power plant condensers. These conventional power plant condensers will have demonstrated good seismic performance end will be equivalent and applicable to nuclear power plant condenser designs. This should include a review of "as-built" design documents and/or walk-c)wn to verify that the condenser has adequate ancharage.

6.7.2.2 Main Steam and Drain Line/Bvoass Pinina Portions of main steam and e ain line/ bypass piping designs that have not been seismically analyzed will be reviewed to demonstrate that piping and supports fall within the bounds of design characteristics found in selected conventional power plant steam piping. These conventional power plant steam piping designs will have demonstrated good seismic performance and will be 6-11 l

. .- .m._._._ _ . . _ - _ - _ . , _ _ _ . , - _ _ . - .. _ . _ . . _ _ _ _ . _ .

I'd h g .NEDO-31858 Rev..-1 equivalentand applicable to nuclear plant steam piping designs'. This_should

include.a review of- design codes and standards used to insure adequate dead lo'ad support margin and ductile support behavior where subject to lateral loads, and a walk-down to verify that small diameter piping and j instrumentation is free of impact 1 interactions from falling:and prcximity or idifferential' motion hazards. ,

-6.7.2.3 'Interconnt.qted Systems Process steam lines #whichLinterconnect with the main steam line between 1the_MSIVs-and turbine stop-valves:also will -

tiewed to provide reasonable 4.

assurance of- seismic performance - These systems vary from plant to plant' but maylinclude-RFP turbine, hydrogen recombiners and other steam extraction

systems. -The'se systems should be reviewed to insure piping system design, freedom from adverse--interactions,=and adequate andorage up to major system equipment.-  ;

In the extremely, unlikely-failure of other balance-of-plant piping that

interconnects with.the main ' condenser, the radiological consequences are significantly smallerithan_forfpiping which interconnects directly with the mainisteam_ piping.--Leakage from the condenser. effects only the-decay of

=

(organic iodine ((remainingjelemental iodine entering condenser is rapidly

removed)

and the typicalf removal factor forf organic iodine 131: (I-131) entering {thecondenserislets'than.afactoroftwo.

r 4

6-12 1  !

NED0 31858 Rev. 1 Table 6-1 Contribution-to the LOCA Dose Exposures for A Naximum NSIV Leak Rate of 200 scfh Hope Creek Generating Station Whole Body Thyroid Beta Ir.tal frem) frem)

Exclusion Area A) 10CFR 100 Limit 25 300

• Boundary (2-Hour)

B)- Previous Calculated 0.6 76.7 Doses **

C) Contribution From 0.1 2.6 MSIVs at 200 scfh D) New Calculated Doses 0.7 79.3 Low Population A) 10CFR 100 Limit 25 300

• Zone (30-Day)

B) Previous Calculated 0.08 7.7 Doses **

C) Contribution Frem 0.34 57.5 MSIV at 200 scfh D) New Calculated Doses 0.42 65.2

-Control Room - A) GDC-19 Limit 5 30 30/75***

(30-Day)

B) Previuus Calculated 0.04 0.26 .0.91 Doses **

C) Contribution From 0.10 2.71 1,64 MSIVs at 200 scfh D) New Calculated Doses , 0.14 2.97 2.55 No 'imit specified.

USAR Sections 15.6.5.5 and 6.4 (includes MSIV leak rate at a total of 46 scfh for the first 20 minutes; control room dose assumes 100% per day i reactor building inleakage).  !

75 if prior commitment has been made to use protective clothing, i 6-13/6-14 l

NEDO 31858 Rev. 1 7.0 C'dCLUSIONS Based on a thorough and comprehensive analysis, the BWROG proposes to increast the Tnchnical Specification MSIV lt d rate limit from 11.5 to a maximum of 20 scfh per steam line, and to eliminate the requirements for the LCS. The pr osed increase in the MtlV leak rate limit will sigrificantly reduce rem eing valve leakage problems, and will minimize needless valve repair that can compromisa plant safety. The proposed deletion of the safety-related LCS, and the proposed alternate method (main steam piping and condenser) for HSIV leakage treatment, will reduce the risk to the public heelth and safety.

Fr plants that do not have a LCS, impicmenting the proposed changes will provide plants with a copability to process MSIV leakage. Use of the main ste e "oing and condenser provides reliable and etfective HSIV leakage treatment without exceeding the off site and control room dose limits. The implementation also will provide a uniform ba.is for estabitshing a .

plant specifim MSiv leakage rate i mit. From a safety perspective, the proposed change results in an increase in protection to the public.

For plants with a LCS, the proposed changes involve a replacenient of the existing LCS with the more reliable and effective main steam piping and cordenser for MSIV leakage treatment. Although the LOLA doses may be slip tly

^

higher for low HSIV leak 196 rates, the proposed treatment method cen effectively treat tiSIV leakage over an expended operating range. Thus, a margin of safety exists. Furthermore, it is clearly a safety improvement to replace a system with known limitations with the alternate main steam piping and condenser treatment patnway, that has been shown to have excellent reliability.

The BWROG has concluded that increasing MSIV leakage rates up to 200 scfh per main steam line will not inhibit the safety function of the valve, and will improve the chance for a successful local leak rate test to greater than 90%, up from the 77% success rate at the current typical limit of 11.5 scfh, 7-1

, . , . , . . . . . i. . -

NED0 31850 Rev. I l

The BWROG has evaluated the effectiveness and availability of several '

leakage treatment methods that use existing plant equipment and that can be used for reducing the radiological consequences of MS!V leakage. The BWROG recommends the isolated condenser method, that uses the main steam drain lines to convey MFIV leak n e to the condenser, as the preferred method for MSIV leakage treatment for BWRs. This method may require minimum plant modifications to permit opening of the selected drain valves upon a loss of l

off site power. j The BWROG has evaluated the capability of main steam piping and condenser to process MSIV leakage following a design basis accident coincident with a i l

seismic event. Based on the comprehensive evaluation, the BWROG has concluded there is reasonable assurance that the main steam piping and condenser will remain functional following a design baeis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage. The following conclusions provide the bases for this assurance:

(1) Probability for which the resulting dose from MSIV leakage is significant is extremely low. This requires a design basis LOCA, a degraded core where ECCS are not functional, and a significant

-seismic event.

(2) The main steem piping and condenser are designed to strict industrial standards and building codest thus, significant design margin exists.

(3) The piping and condenser exhibit substantial seismic ruggedness.

Comparisons of piping and condenser design in GE plants with those in the earthquake experience data base reveal the GE plant designs are similar to or more rugged than those that have exhibited good earthquake performance.

(4) Possibility of significant failure in GE BWR main steam piping or condensers in the event of design basis carthquakt is highly unlikely, and any such failure also would be contrary to a large body of historinal earthquake experience data, and thus unprecedented.

12

. . . - - - - -- . ~ . - - . . -.. . . - . . - -

NEDO 31858 Rev. 1 (5) A plant specific verification of seismic adequacy of the main steam piping and cor, denser will be performed to provide reasonable assurance of the structural integrity of these components.

The BWROG has developed a radiological dose methodology that conforms to 3

regulatory guidance to calculate the dose contributions from M51V leakage using the isolated condenser method for MSIV icakage treatment. Analysis for 40pc Creek demanstrates that the increase in the LOCA doses is insignificant for HS:V leakage rates up to 200 scfh per steam line. Furthermore, for Hope

reek, the LOCA doses remain within the regulatory guidelines for MSIV leakage

'es ur approximately 500 scfh. Therefore, the proposed method provides a

{'

st.'stantial o :aargin of safety for mitigating MSIV leakage beyond the proposed Tcchnical Specification leak rate limit.

7-3/7-4

t NED0 31858 Rev.1

8.0 REFERENCES

Il 1.- 10 Code of Federal Regulation, Part 100 - Reactor Site Criteria. '

2. 10 Code of Federal Regulation, Part 50, Appendix A, General Design Criteria 19 - Control Room.

l l

i

- 3. Information Net.ics IN 82+23. " Main Steam Isolation Valve (MSIV) Leakage",

July 16,' 1982.

l

-4.- .NUREG1169,1"NcsolutiontoGenericissueC-8,AnEvaluationofBoiling l

Water Reactor Main Steam Isolation Valve Leakage and the Effectiveness of

)

Leakage Treatment Methods", August 1986.  !

t

5. Regulatory Guide 1.96, Rr'ision 1, " Design of Main Steam isolation Valve  !

Leakago Control Systems Ft ' Boiling Water Reactor Nuclear Power Plants", t

-June 1976.- '

~

6.- NUREG/CR4407.'(EGG 2421),"PipeBreakFrequencyiEstimationforNuclear  !

Power Plants",' idahoL National - Engineering, Laboratory", May 1987.

i

7. Regulatory. Guide 1.3, Revision 2 " Assumptions Used for Evaluating the-  !

Potential Radiological Consequences of a loss of-Coolant Accident for

-Boiling-Water Reactors", April.:1974.

8.1 . NUREG CR 4330 (PNL-5809),;'" Assessment of Selected Regulatory Requirements F that May Have Marginal Important to Risk",-June-1986.

l l-.

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n 8-1/8-2 -

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NEDO 31858 Rev.)

9.0 PARTICIPANTING CONPANIES f

Ut111tv Etantis1 Boston Edison Company Pilgrim I i Ctrclina Power-and Light Company Brunswick 1, 2 I Cleveland Electric 1110minating Company Perry 1, 2 Commonwealth (dison Company Dresden 2, 3:

La Salle 1, 2; Quad Cities 1, 2 Detroit Edison Company- '

Fermi 2 Entergy Operations Incorporated Grand Gulf 1  !

Georgia Power Company Hatch 1, 2 GPU Nuclear Corporation Oyster Creek  :

i Illinois Power Clinton j Iowa Electric Light & Power Company Duane Arnold Niagara-Mohawk Power Corporation Nine Mile Point 1, 2 f

Northern States Power. Monticello 1

Pennsylvania-Power and. Light Susquehanna l',2 Philadelphia Electric Company Peach Bottom 2, 3';

Limerick 1, 2:

Public Service Electric & Gas: Hope Creek-1 1

-Tennessee Valley Authority Browns Ferry 1, 2, 3 i Washington Public Power Supply -Hanford 2 l b

N 9

u 9-1/9-2 1 y

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NEDO.31858 Rey, 1 APPENDlX A RECLNT HSIV LEAKAGE HISTORY AND COST BENEFIT DATA (REVISION 2)

COMPILED BY GE NUCLEAR ENERGY APRIL I?88

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NED0 31858 Rev. I try!$10N 1 PAGI .t or .1L gyrca ntiv flart c!Ptitt tiotim2 ggj;322,,jggIY tht ACf Mf ff0PY A@ Tft 1&Efft BATA MllV K M TACT \ft,tm n*rWil . &

WUT I TOTAL INCREA!!)I 1ktoARD I Artta m>tBIA ovtAct HP,5 CULACI

(* INITIAL PitAIR CT 1ptitAL DIPAlt hitA!k flv.m! _

!!21311 IIAR DNAP,I .12L. .EZL fid74) .E21. A lJ211 ,,U5L, ]M ll411 $1*f1Vl1_

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2. L t ,,nk., t 1a t .,,, n

s. s .hk. o As _ __

6 1 .Ak, 1 1.L., p

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W

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7. f .Ak, O ,,.id.,, , _

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13. l tl., o f1 ,,.1,1. o? .,,,,

36 1 .,31, 1 LL _

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26. $ .41. f 1a _

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' o L1,

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16 1

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IVIAL PJPA!R$ 9 sen tutas icLi Lt.,

bt W 11.5 $CrH  ?? 1 S! W 100 $CFH 11,,,,, 1 WJtt$t 3.  !

• Inboar18, O e outboard

1. Il repair is required

3. V e fot, N

• No. U

• Unknown 4 3 e DVRG3 modific.ations incorporated. W e SWROG sodtlications not incorporated. WP e new plant data S. ** 1aek rate not knowns bowever, talte was returned to acceptable leak rates With portal maintenance pr.actates. (beta not included in 4.ntstal loaa. age everage.)

A4

i NED0+31858 Rev. 1 i I

i i

~l.

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Metr I tutAl, tuotAsto l -

INDOAR!d AfTER Wl#StR UU1Mt HR$ OL'IME ,

INITIAL kpAIR OF IN!T1AL kEPA!R ht?Alk timM OR b lEEN 3348 $53QARD . XZL. .112L IDER$1 .1GDL. JtL. RIB 1 .12E h.4 $l.lL3 Gietm . ,

1. i 1 _f LL '

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s. 1 -- 1 1 ** .mi. _

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1

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e.

• 1- e 14.

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8. If repair is reptred t 3.. Y e Yes, N

• me. U

• Unknam

~6 t e DWm00 endificattens arcertorated W

• DWh0G andtf tcations not incorporated, kt a few plant data j

5. ** 1aak rate not knows however. Selve was returned to acceptable leak totes with normal estatenance

- practicos. (!sta not tacluded 1A thittal leakage everage.)

'A5

{ y rga wigp pr.m ,mmnM ~ q - -M-.4- --p g 34 c9 m mg,yqw mqw yp em 3- wygg. y; 4-Ma m ie r -se. y an + q -7g.g--p g

NE00 31858 Rev. 1 ItfYl810N 3 7 AGE ,1 OF J ),,,

B*'MO M!IV IJJJACf t'11'fUF1 Otstd!TTII Pl* twt M.!!Y if>JACf MittJ .tY mfd (Y'ff ID'f7ff bATA l

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3 Ntxf 3 TUrAL 1Nett.Astr INboAAD 1

AFTER ff)St1 (VIACI HR$ OU1 AGE 1 OR IN!!!AL f.11 A!R OF INIf1AL REPA!R RFPA!R 114G"nf _

lE'.C3 1162 PNAPJ .EDL .EDL F M II .EDL. 5L ImC11 .112,. LJi liai,1 N' .

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3. i 1 1 1JL. -

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11* J2- .lb o LL. - -

11. 1 .11., 1 LL. _ - -

13. 9 tk o .J.JL. - - l

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rittu owtY

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Fit'th onY to.

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23. _ -

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26. _ _ _ _

!3. _ _ _

r TUIAL RT.,tAIR.S 9

$crH TU1ALS .11AL. .1],d S!1N 11.3 $CFM 16 i 511 A 100 BCrH ,,, ,,12.,,, 1 WTI.1 - -1 1 - Inboard. 0 natkard

1. If repair is required

3. Y e fes, N

• Noe y e linkmoim 4 & a BWN00 m>difications incorporated, N

• SWROG endtfications not thcorporated. NP

• new plant data

3. . . w r.i. not an ni %,.r Wiv. r, turn.d t. . tai. i.a ,.t.. .ta nor-i . int.nane.

practness. (Cata not included in initial 1.uage e,orage.)

t A-6

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11. A 1 i v41 ,,,,11 1

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26. J 1 0 M LL 1

25. e .ci, t Li. _

W TUIAL RD AIRS P 21UIAl3 .L11. lLA, Stim 11,S $CTH 13 9 brilw 100 SCTH 19 9 lor 13: 1. 1

• Inboard 0

• Atboard

2. If sepair is required

3. Y e fes, W e No, U e Unknown 4 B e BWROG sodifir.ations incerporated. p e BWR00 sodific.ations not thcorporated. NP e new plant.dat.a

6. *e taak rate not tael bewever, valve was returned to accettable leak rates with normal natntenance practices. (Lata not inclueed in initial leau4e average )

A7

i NEDO 31858 Rev. 1 t i

P r

Arvl810m t Paet 1 OF .11.

I anco Mt!Y f.RAEAGE (11'fi21 QMET;11 l

ptcEWT nrrY 12Atan scrf0H ANI' CDET RtWEFIT taf.A MsIV IWIUPA."JVRRR . ATWR1D-MDR111.L

• 1 ,

I NEX7 TUTAL INCRIA$DI int 0ARDI M MER ESEL W n'1ML OR- INTTIAL REFAIR OF ik1TIAL RDAIR RarAla ' *'Tu m . Det fA3XihBD .ASEL., 122L BEM&8 J 21L 5L,. IRS 121. I,la N831 M

1. .22.,,, .12,. f ., 1L1,. _ N  ;

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10. 9 .A4., B ,2J., _ W

11. 1

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13. - * .AA 1 e 1f W

13. .31, . .A1. . . e 1?

k

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1 1 -

- 16. .AL. ' .,4L , _ e 212L **

1' tJ. 1 JWRL 1190 62 1 [

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-- 17. - 1 , .3L 1 ,,L1, ,,,, t

18. 1 lk I 21 8 19.- a'- . JJ,,, f A ,,,,,,, _ t

30. 1 .34, f &.R. b ti. &

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2-1

13. .Ak, f ,LA. R

36. L,. .11. f ,,,,1.A. R

~ 2 5, 1 .12,. f ,.91 ,. R

• 1UIAL RD AIRS 6 SCPN TUIALS 1L1,, L1,,

stuW 11.5 SCFit - 10 6 Stuw 100 acFil .,,,11, 6 NOTES: 1. 1 + 1 aboard. O e oitboard

- 2. Il repair is required

. . . .. 3. T

• Tes, N e to. U e Unknown

4. 8 e DWROG and4ficattens incorporated, N e SWROG sodifications not incetporated, NP e new plant data -!

3. .ae taak rate not knowns hovveer,' valve was returned to acco; table leak rates with norest maintenance

practices. (tou not included in inattel tankage average.)

i A8

_ _ ._ .. _ _ _ _ __ _ . _ _. . _ _ u. u.

l NED0 31858 Rev. 1 Rfvtllts 1 PAGI 1 of ,,11.

D'P't M1v 1DJrt drffl intemu

[1Ltvf MiV 1.fAtr;t M117t1 Awm1,,11w11ft 11.'- A plalV PnM ACWJt AIET;ML,,1,

!KnuAM I ATTD If 5912 prxt 2 M AGE nu inat.Astti ems M Act

(* Ik!!IAL M1 AIR OF SW!?IAL fj) AIR PJ.7 AIR 11XTM tf3 ILQ 7.'IMe2 .8311.. .E2ji, f1VMI _171L_ .ft,, L'31 ,11*2,. 143 M glig,EQ" la 9 .11. t ,,,,,L1,, , , , , _ _ __ _ t l* 1

.h. 1 J*S. -  ?

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• A f ,,,.J.J. _ _ I

6. t ,,1L 0 ,,,,,1h1. ,,,,14, 1 1L ,,,J1. 91 ,,,,,,,,, 1 1 I. ,,12. .11. I 11.d , .JJ. 1

.1% Q.12 JE , A. 1"

e. 12. A 1 u2 LL i . 122. $Ji 12,. . 1 f *. M -

9. ,,,,12, 1 1 __ 6.2 ,,,,,,, , , , , , , , , , , , , , , ,

t .M

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• W

11. 12. 1 0 .2.12 LI. 1 122. 2d1 60 .L I*."

13. ,,,,1q. 1 0 st ,,,,,,, ,, n . n___.

11. 2 .11, 0 ,,,,,,,L9,,, ,,,,,,,,, ,,,,,, t.*t .

16 12., 1 0 00 1A 1 ,122. 241 it,., 1 u 2*M

15. 3

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16. 1

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10. ,,,,,1. .11,, 0 ..... L 1. ,,,,,,,,. ,,,,, t

11.  ? ,,12. 0 , L1 ,,,, ,,,,,, e t3. ,,, L .it 0 ,,,,,1J. ,,,,,. , _ _t

43. ,,,,.1. .11. 0 __ Lf ,,,,,, . I 16 3 .11,, 0 ,,,,i,J,, ,,,, _ 9 ,,

ll. 9 ,11,,, 0 6.B 9 MAL FIFAlu  ?

&Cnt MALS 121.4 .1L,1.

bl3AW 11.1 SCF'N IP  ?

Bi1IW 100 SCnt ?O i NOTES: 1.  ! = Intmord, O

• Outtmard

1. If repair is required

3. Ye f ea , N

• No, U e Uranaim

6. Be Swltc0 seeifications incorporated, a e BWX. modificattens not tacorporated, WP e sww plant data

$. ** 14aa rate not knowns howerot. valve was retust.ed to acceptable loat rates witfi hermal matattname practices. (Data not included in taattal laauge everage.)

A-9

NED0 31858 Rev. I l

trY181fm I iAGE .11, et Ji, reo m!v tlan c'>ni tutumt F1cDr! M1v tra A"t kitTotY AwD CDM ttwffit 141.A pti!V 8%MVTACTJFIR AhTDM*?f tl

• b pftXI I TUIAL 1>32A5til I 1kicARD I

Arfta postR OL'!Act HR3 WIAC,1 OR ik!!!AL ktfAIR Cf th!!1AL k!JAIR FIFAlk . LDt"H 10 0% IL68 11*IFA.D J21L. Jf22L FM'f.5 .E!)L. ,2tL, f2"J ,,113,,, 143 M d

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9. 1 9

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f it. 1,. f,1. 1 , LL ,,,,,

t it.  ? ti. 1 '1 *

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13. 6 1 , , , , , ,

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11. t ,,,ti, f _MIQL 0.7 1 M 164 .L,1 J,. f*

16. 6 ,,IL 1 ,21GtL 9.6 1 , JL ,,Lt .L f .'

17.  ?

JL 0 ,11.1. ,,JLt., 1 7.8 ,, 119,. J,J J,. I

15. T ,,,tL 0 ,?,J,, , , , , , ,,

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as. 1 ti, o s.? ,. ,,,

n 34  ? ,,gL o 1.6 _ _ _

E

16. 6 ,,,lL o 1J., ,,, ,

TUTAL RIIAIR$ 7 M TUIALS .11Lh .,1!J a.tuw 11,s scru ,,,,,11,,,,,, i SilAw 100 SCFH to '

WO!T.S t 1. 1. Inboard. O e Cutboarti

2. If repair is required 3 T e Tes. N e No. U e WJmerwn A. 8 = DVROG mo46fications incorporated, a e DVROG moelficationa.not tncorporated, kt e rev plant esta S. ** Leak rate not knossi tmover, valve was returmed to accettable les rates vttA normal mainter6ance practates. (Lata not included an initial leasage average.)

A-10 l 1_____ . . .. _

NED0 31858 Rey, 1 FIYlllL* 1 PAGE ,,1L OF J1, t#3) w!fy tMA"1 tientf net *Ig HTrf Mf f Y 11&JACY Mt!"PtY M (UM tEvif M t A1 A Msty Kor.7AC'!VFJJt ANffbMTJill

• 1,,

NIXI I ttf!AL 1hCMA%!D IkboARD I AFTD N'#9D UJ!ACI W3 Of1AG1 04 1N171AL PD Alt (,F IW111AL MJAlp PD A!R 1 tisi not m/t.2 .33., ,EDi. f M JR .EDL JtL. f.22 ,12L la,1,v tytidl stmit

1. ,, L,, ,,11. 0 ,,,JJ., ,,,,,, _ ,,,,,,

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• 11L LL 1 '

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a

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vn

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no.

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13. I 1 1 M ,,L1,, 1 154 Lb 1 __ L FP.*_

16. s .1L 0 ,,,,,LL ,,,,,,,, ,,,,,,

.P W

15.  ?

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W.W

16. t .1L 0 ,,,,LL _ ,,,, P.FF .

37. 9 1 0 m ,,,, _,,,,,,,,, _ ,,,,,

W .9

18. 9 ,,11,,, 0 ,,,,,,LL _ ,,,,

W.WP

19. S

'1 0 ,,,,,,LL _ ,,,,,, ,,,,, P."? ..

20. S ,,h., 0 ,,2iL,1, ,,,tJ,. _ L,,,, ,,2AL},,, ,,R!dL, LM 1 W E .*

11. ,,,,1,,,,, 1 0 J42.3_ ,,,,Li, 1 ,,11th L3 1 W M .*

11. ,,,,,, ,,,,,,, ,,,,,

II. - - -

16. - - -

15. ,,,, - -

TUIAL Rn atRs le

$CFM ttf!ALS J,1LA,,, .AL1,,

B!llw 11.S SCFM ,,,,12,,,,, 10 Stuv 100 scrH 1% 10 ICIts: 1. 1 - Inboard, C e cuttoard

1. If repetf is required

3. T

• Vos , N

• ko, U

• OrAnavn

4. $e BWWOG modific.ations incorporated M e $N modif' cations bot itworporated, KP

• new plant data ,

S. e* Laan tote not knowni however, volve was retmed to accept.Able leas tetes ytth hormal maintenance practites. (tasta not included in snittal leakags everage.)

A 11 1 i

i NE S 31858 Rev. 1  !

f arvtstos s a Past .AL or .11.  ;

anac un laat&Of CDEURE (IBMmt pener un taanat mistm m en ntnern rian t

M81V HANUFACR98R tsaur . 1 ,

t s

i 8  ;

Mexrl terAL Inca Aste INDOARD I AFTER lOteR CUTAGl Mit$ OUf608 - 6 on 1stf1AL RitAtt OF IN!?!AL REPAIR htPAlt isnt 2541 $L'IRDeRD .gZL. .3GDL $D$$$ .,8GDL. 3L. $$11 .113L QWGBXf1

1. '

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t. 1

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1.' 6 AL ,ALa., L1. 1 L2 ' $

2

8. 1

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39. 1 .AL ' f .,,.2 A ._ -

11. ' ,$, . ,,$ L - o .,,,,.Li. -

n. t

31. o Lt.

i

-13,..& & .

f L2.,, ,, i

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-[

26. . ,1 . .AL o L.f., _ . s

- 17. 6 .jk. f Li. , ..

18. '? AL i 1? _ _ l

' 19. . 6 .jL 'o 1J. .-

30. t ' ~,,AL c 1.? . , , , , .,,,,,

t

11. &

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tt.  ?

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- t 8. . & .Ah., o o _ --

to..  ?

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l 2S.- a ~ .Ak,, f e ., . , , , ,

191AL RtPA1A8 1

- M I0TALA.. .32LA A41. f Dt3AW 11.1 SCPH .

,22, M1AW 100 SCFH ,,1).,,.,,. 1 t

NOTEla '1,  ! . Inboard, o e htbeard ,

1.- If repair is required-

3. Y

• Tes W

• No. V

• Unknowh

,i s, S

• DWOG sodifications incorportledi N

• SA00 pdifit.ations not incorporated, MP e new plant data ,

5. - ** !aak rate not knowns howooer telee was returned to acceptable leak rates rith norest maintenance practicos. (Data not included in initial leakage aMrege.)

I  !

L 2

L r

'A.12 l u -

NED0 31858 Rev. 1 RfYll!0ff 3 FaGE.1L OF lL

.nu affy 11uxf etmf enemn Rf"IFY MifY 110Ef R11"f0PY AND ODM f fWIIM PA?A tt!!V 8WilfFACTVPIR CRAST

• 1 .. ,.

WlX7 I TUTAL INCRJ.ASIM INDOAADI AFT 13 NUMat3t w!Act hts OL'! AGE OR Ik!TIAL MPA!R OF IklTIAL MPAIR MPAlk 11CIM .

tCG1 H68 WIBDAF'D . EDL., ZDL BlE!11 SCni-- 3L. 12',1 .II.'5L kb 11.r21 (D+'I'!!'

1. t .1L i LL - -

4. 6 .,tL o e

3.  ?

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NED0 31858 Rev. 1 BWPOG HSIV LEAKAGE CLOSURE COMMITTEE RECENT MSTV LEAFACE HISTORY SlMtARY Atwood-MSIV Hanufacturer QC (.,, Morrill crane Totals NUMBER OF PLANTS .. j l 9 4 24 TOTAL INITIAL TESTS , _ M ,4,,_ _121 64 329 TOTAL INITIAL TESTS HIhTS TESTS WITH >X SCTH REPORTED _lf.,0,_, 96 64 300 TOTAL INITIAL SCTH Iff 58. 3 727.2 814.3 3399.8 AVERAGE INITIAL SCFH/ TEST 1) . 3 7.6 _12.7 11.3 TOTAL INITIAL TESTS >

ALLOWED LEAFAGE Ji 36 10 71 TOTAL AFTER REPAIR SCTH 31.7 95.3 44.4 171.4 AVERAGE AFTER REPAIR SCFH/ TEST _ 1.3 2.6 4.4 2.4 0F INITIAL TESTS: '

TOTAL <11.5 SCTH 117 83 54 254 (77%)

^

TOTAL <100 SCTH 138 95 295

_- _62 _ (190%)

A-15/A-16

NEDO 31858 Rev. 1 l

{

APPENDIX B BWROG MSly LEAKAGE CLOSURE COMMITTEE TYPE 1 EXAMPLE NRC SUBMITTAL FOR INCREASING MSIV LEAK RATE LIMITS AND EXEMPTION REQUEST TO 10CfR100, APPENDIX A 3

7 (f(; j , l_'i i

Type 1 B1-1

i NEDO 31858 Rev. 1 l 1

U.S. Nuclear Regulatory Commission l Attn.: Document Control Desk Washington, DC 20555 RE: [PlantStationName) l DocketNo.[xxxxx)  ;

-License No. [xxx)

Gentlemen:

[ Corporation) hereby transmits an application for amendment to the [' Plant ,

Name) Facility Operating License [xxx), and an applic.ation for exemption to  !

10CFR100 Appendix A. I

[ Corporation)_ requests an amendment to the Technical Specifications, set forth in Appendix A to the License, to permit an increase in the allowable  !

leak rate for the main steam isolation valves (MSIV). In addition.

[ Corporation): requests that the downstream main stea.n piping and condensor be i exempted from the seismic requirements specified in _ Appendix A of.10CFR100.

In support of the proposed change to the Technical Specifications, I enclosed for the Nuclear-Regulatory Commission's review are:.

(1). Application for Amendment to the Facility Operating Licenset i Techical. Specification pages affected by the enanges; and ,f supporting information and analysesLaddressing the changes. The analyses demonstrate that the-proposed changes do not involve a I

^

significant hazards consideration pursuant to 10CFR50.92.

f

. .. . j (2) 1 Application for' specific: exemption to 10CFR100 Appendix-Ai and.

supporting information and justification for this exemption.

[ Corporation) realizes that there is no provision in-100FR100 for exemption; however,:the Nuclear Regulatory Commission has- .f 4

+

Type 1- B1 2

.I

. .. . . = - . . _ - .. .. - .- - _ _ - - - - .

NEDO 31858 Rey. I the authority to grtnt this exemption. This exemption will not present undue risks to the public health and safety, and is consistent with the common defense and security. Furthermore, special circumstances are present which warrant issuance of this exemption request.

These proposed changes are supported by BWR Owners' Group (BWROG) work.

The BWROG formed a MSIV Leakage Committee in 1982 in response to Generic issue C 8 *HSly Leakage and LCS failure". Generic Issue C 8 addressed the safety concerns that reported MSIV Icakages are too high and that the Leakage Control System will not function at high MSly leakages. Based on the extensive, ongoing work performed by the BWROG to support resolution of the Generic issue, the BWROG has developed the technical justification for the proposed Technical Specifications change and associated exemption request.

The General Electric (GE) report, NED0 31858 Rev.1, describes the safety benefits and provides justification for the proposed change. With regard to increasing leakage ra e limit for the HSIVs this will avoid unnecessary maintenarice which has in the past adversely affected the operability of the HSIVs.

The proposed change involves implementing the reliable and effective main steam piping and condenser for NSIV leakage treatment. This treatment method is effective to treat MSlV leakage over an expanded operating rango without exceeding the off-site and control room dose limits. Except for the requirement to establish a pic,er flow path from the HSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for HSly leakage treatment.

Furthermore, the risk to the public health and safety will be reduced with the implementation of the proposed HSIV leakage treatment method. The Type 1 Bl 3

NE00 31858 Rev. 1 implementation will provide [ Plant Name) with a capability to process MSly leakage, and will also provide a uniform basis for establishing a plant specific HSIV leakage rate limit. From a safety perspective, the I proposed changes result in an increase in protection to the public.

l Pursuant to 10CfR50.91(b)(1), (Corporation) has provided a copy of this l license amendment request and the associated analysis regarding no significant l hazards consideration (s) to 'e appropriate state representative.

(Corporation) representatives will be available to discuss or meet witn the Nuclear Regulatory Commission staff at your convenience to address this matter.

Very truly yours, (CORPORATION)

INAME1 Vice President

?

Enclosures t

cc: Regioral Administrator, Region [x]

[Name';, Director (Name) Project Manager (Name),ResidentInspector

[Name), State Representative l

l l-l l

Type 1 B1 4 1

l NED0 31858 Rev. 1 l

l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of

[ Corporation) Docket No. [xxx)

(PlantName)

M!UDAY11 INamel , being duly sworn, states that he is Vice President of (Corporation]; that he is authorized on the part of said Corporation to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.

ISionaturel Subscribed and sworn to before me, a Notary Public in and for the State of

(. ] and County of [ ), this [ ), day of [ ),

1988.

ISianaturel Notary Public in and for (County, State)

My Commission expires:

Type 1 B1-5

._ _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ - - _ _- - - - _ - _ _ = _ _ _ _ _ _ _ . _ . _ _ _ - - - .

NED0 31858 Rev. 1 UN11ED STATES NUCLEAR REGULATORY COMMISSION in the Matter of ll

[ Corporation) )1 Docket No. (xxx)

[PlantName) )l APPLICATION FOR AMENDMrNT 10 OPERATING LICENSE Pursuant to Section 50.90 of the Regulations of the Nuc..,ar Regulatory Commission [ Corporation), holder of f acility Operating License No. [xxx),

hereby requests the Technical Specifications, set forth in Appendix A to the license, be amended to permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs), e Specifically, [ Corporation) requests that the allowable leak rate specified in the Technical Specifications Section 3.6.1.2* be changed from [x]

to [y] standard cubic foot per hour (sefh) per main steam line. This proposed change reflects a higher, but still conservative allowable leak rate for the MSIVs.

The proposed change to the Technical Specifications are set forth in Enclosure ~ 1. Also in concurrence with this application for license amendment.

[ Corporation) has applied for an exemption to 10CfR100, Appendix A.

The section identified is based on the BWR Standard lechnical Specifications for BWR/5 (NUREG 0123). Each utility should revise this section as applicable to their individual Technical Specifications.

Also, some plant sites may specify a maximum allowable MSly leakage rate per line (i.e.,100 scfh), but require that the total is less than four times the maximum allowable leakage rate for each line (i.e., 200 scfh total).

Type 1 B1-6 1

HED0 31858 Rev. I l

lhe proposed change is a result of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic issue C 8

'HSly leakage and LCS Failure". GE report NED0 31858 Rev. I provides technical justification on a generic basis to support this proposed change.

Based on the BWROG evaluation of MSIV Icakage performance, the current Technical Specification allowable HSly leak rate is extremely limiting and routinely requires the repair and re testing of MSlys. This unnecessary repair significantly impacts the maintenance work load, often contributes to outage extensions, and has in the past adversely affected the operability of the HSIVs. BWR outage planners routinely schedule several days of contingency to repair and retest the HSIVs. In addition, the needless dose exposure to maintenance personnel is inconsistent with As low As Reason 4bly Achievable (ALARA) requirements. There have also been many Licensee Event Reports g

written for HSIV leakages exceeding the Technical Specification limit.

The proposed change will reduce unnecessary MSIV repair costs, avoid unnecessary dose exposure to maintenance personnel, reduce outage durations. 3 and extend the effective service life of the HS!Vs. In addition, the proposed increase in the HSIV leakage limit has potential to significantly reduce recurring valve leakages, and minimize the possibility of needWss repair which can compromise plant safety.

The proposed change involves implementing the reliable and effective main steam piping and condenser for HSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off-site and control room dose limits. Except for the requirement to establish a proper flow path from the HSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for HSly leakage treatment.

Type 1 B1 7

N 00-31858 Rev. 1 Furthermore, the risk to the public health and safety will be reduced with the implementation of the proposed MSIV leakage treatment method. The implementation will provide [ Plant Name] with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a  !

plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to the public.

Plant-specifi; supporting information and results of the radiological analysis that justify the proposed chaage are included as Enclosure 2. As oncluded a the supporting information, the increased MSIV ellowable leak rate a' [y] scfh will not adversely affect the performance of the primary containment isolation function. A plant-specific radiological analysis has been performed utilizing the main condenser as an alternate treatment path for MSIV leakage. This radiological analysis demonstrates that the proposed changes results in an insignificant increase to the dose exposures previously calculated for a design basis Loss-of-Coolant Accident (LOCA). The revised LOCA dose exposures remain well within the guidelines of 10CFR100 for the off-site doses and 10CFR50, Appendix A (General Design Criteria 19) for the control room doses.

Upon approval of the proposed Technical Specification change,

[ Corporation] will institute into the MSIV maintenance and test program, the requirement that any MSIV exceeding the proposed [y] scfh limit, will be repaired and re-tested to meet a leakage rate of less than or equal to [x]

scfh. This will assure continuation of high quality repair and refurbishment efforts to improve the overall performance and reliability of the MSIVs.

The proposed change will require a minor plant design change to allow specified main steam drain valves to be opened even if off-site power is not available. Applicable normal plant operating proce6ures, and emergency operating procedures shail be reviewed and revi:cd accordingly

• Plant-specific Type 1 B1-8

_ , m . _ . - , -

, _ - , . _ . - .~ . . ~ . . . _.__-_m .- - _ - . ,

Q-- S MED0 31858 Rev. I_ .

Furthermore,-(Corporation) will incorporate the applicable leakage treatment methods;-_ consistent with GE ' document NEDO 30324 " Potential Operator LActions to. Control- MSIV. Leakage," into the Operational Procedures and

,  ! Emergency Operational Procedures at [ Plant Name).

[ Corporation]_willalsoperformaverificationof:seismicadequacyofthe ,

main' steam piping and condenser,-consistent with the guidelines discussed in -'

-Section-6.7 of NEDO 31858 Rev.1,_ to provide reasonable. assurance of the structural integrity of these components.

Pursuant to 10CFR50.92, an analysis which demonstrates.that the proposed change.does not-involve.a significant hazards corsideration,- is included as Enclosure 3.' The proposed change has been reviewed in accordance with Section 6.5 of the Technical Specifications The proposed change _will- not-authorize Lany change in the types ofL effluents or in the authorized power level-iof the facility. . Application.for exemption to Appendix A of'10CFR100 is-included-at Enclosure 4.

WHEREFORE, Applicant respectfully requests that Apnendix-A to the 1 Facility Operating License No;-[xxx] thereto be amended in-the-form attached *

h'ereto as Enclosure 1.-

-[ CORPORATION). O By: _

Vice President Subscribed and sworn to=before me-

_on;this  : day _of- .

o .

r-p L .

i, L Type l' B1-9 o q  ;

NE00-31858 Rev. 1 ENCLOSURE 1 ILORPORATIONJ LICENSE NO. Ixxxl DOCKET NO. fxxxl PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS Replace page [3/4 6-2] with the attached revised page.* This page is provided in its entirety with marginal marking to indicate the proposed change.

The page identifieu oere is based on the BWR Standard Technical Specifications for BWR/5 (NUREG-0123). Each utility should revise this page as applicable to tb31r individual Technical Specifications. If Technical Specification Bases are affected, the change to the Bases page should be included.

Type 1 B1-10

NED0 31858 Rev. 1 ENCLOSURE 2 ICORPORAT10N1 LICENSE NQ. Ixxxl DOCKET NO. Ixxxl SUPPORTING INFORMATION AND ANALYSES

1.0 INTRODUCTION

AND

SUMMARY

Of RESQL11 The proposed Technical Specification amendment involves an increase in the allowable leakage rate from [x] scfh to [y] scfh per main steam line.

In addition it-is requested that downstream main steam piping and condenser be exempted from the seismic requirements specified in Appendix A to 10CFR100.

Section 2.0 of this Enclosure provides a summary of background information; Section 3.0 discucsos the justifications for the proposed changes; Section 4.0 provides a summary of the plant-specific radiological dose assessment, and Section 5.0 summarizes the potential benefits for-a Technical Specification llSIV allowable leak rate of [y]

scfh.

The BWROG report, NED0-31858 Rev. 1, "BWROG Report for Increasing MSIV Leak Rate Limits and Elimination of Leakage Control Systems,"

-November 1991, provides the justification for increasing MSIV leakage limits. With concurrence from the valve manufacturers, this report concludes that MSIV leakage rates up to 200 scfh are not an indication of substantial mechanical defects in the valve which would challenge the isolation capability of the valve to fulfill its safety function.

Therefore, the proposed increase in the allowable leakage rate to [y]

scfh for the MSIVs will not inhibit the isolation capability of the valve.

Type 1 B1-11

NED0-31858 Rev. 1 The BWROG has evaluated several methods and has recommended the isolated condenser methods for MSIV leakage treatment. The isolated condenser method takes advantage of the large volume in the main steam lines and the condenser to hold-up the release of fission products leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser. [ Corporation) proposes to incorporate the isolated condenser for MSIV leakage treatment.

The BWROG has evaluated the availability of main steam system piping and condenser treatment pathways for processing MSIV leakage. The BWROG have reviewed the potential combinations of Loss-Of-Coolant Accidents GF.As) and seismic events of interest:

(1) LQCA WITHOUT NEAR COINCIDENT SEISMIC EVENT. For this occurrence, the pressure in the piping system downstream of the HSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the flow path through the main steam system piping to the condenser is assured.

(2) SEISMIC EVENT WITHOUT NEAR COINCIDENT LOCA. Without a LOCA and the potential associated core degradation, the radioactivity transported with MSIV leakage is of no radiologicai significance.

(3) LOCA WITH NEAR COINCIDENT SEISMIC EVENT. For this occurrence (also assuming significant core damage) the consequences are of interest because a seismic induced failure in the main steam or condenser system could allow MSIV leakage to bypass the treatment pathway. It has been previously well documented that the probability of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability is approximately 0.001 per reactor per year; core Type 1 B1-12

NEDO-31858 Rev. I melt probability is plant-specific and typically ranges from 0.00001 to 0.0001 per reactor per year). It is also noted that a LOCA does not induce a seismic event, and that a seismic event has a very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10-7 per reactor per year).

Considering that the probability of a near coincident LOCA and seismic event is much smaller than other plant-safety risks (less than 1 x 10-7 per reactor per year for coincident events, less than 5 x 10-7 per reactor per year for seismic induced LOCA), the concern for main steam piping or condenser damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design basis seismic events ANSI-831.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin. In addition, specific seismic design provisions have been incorporated into some newer BWR main steam and condenser systems.

In order to further justify the capability of the main steam system ~

piping and condenser treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes). The study summarizes data on-the performance of main steam piping and condensers in past strong-motion earthquakes and compares these piping and condensers with those in typical U.S. GE Mark I, II, and III nuclear plants. This

' limited earthquake experience data and similarity comparisons are Type 1 B1-13

NED0 31858 Rev. I then used to further strengthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in 6 design basis earthquake in conjunction with a LOCA occurring just prior to or after the seismic event.

The earthquake experience data are derived from an extensive database on the performance of industrial facilities and over 100 power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world from 1934 to the present.

The piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they are not designed to resist earthquakes. This is a common conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipment such as motor control centers, pumps, valves, structures, and so forth. That is, with limited exceptions, normal industrial construction and equipment typically have substantial inherent seismic ruggedness, even when they are not designed for earthquakes. No failures of main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tute leakage.

Comparisons of piping and condenser design in example GE Mark I, II, and III plants with those in the earthquake experience database reveal the GE plant designs are similar to or more rugged than those that exhibited good earthquake performance. in addition, [ Corporation] will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0-31858 Rev.1, to provide reasonable assurance of the structural integrity of these components. The BWROG concludes that (1) the l

Type 1 B1-14

. . _ - - ~ . -.- - - - - - - . . - - - .. - - .. - - -_.._ . . -

NEDO 31858 Rev. 1

- possibility of significant-failure in GE BWR main-steam piping.or

. condensers in the event.of-an eastern U.S. design basis earthquake is highly'unlikely and'(2) any such failure would also be contrary to a large. body of-historical earthquake experience data, and thus unprecedented. '

The-design basis LOCA has been re analyzed for radiological impacts utilizing the isolated condenser method for MSIV= leakage treatment. The  !

analysis deconstrates that a maximum MSIV leakage rate of (y) scfh per maia steam line.results in an acceptable increase to the dose exposures calculated for a design basis LOCA.. In addition,_the analysis demonstrates that MSiv leakage rates of approximately [z] scfh per main

. steam line will not: result in dose exposures in excess of the regulatoly i

' limits.

2.0 BACKGROUND

- The safety. function of the MSIVs is to isolate the reactor system in the 1

-event of a LOCA-or other events requiring containment isolation. The .

design of- the MSIVs and i_ts_-isolation requirements are described in ,

Sections-[5.4.5*]- and [6 2.4*) of the Final Safety Analysis' Report 1 (FSAR). '(The allowable leakage rate from MSIVs <is included:in the -LOCA '

- radiologicalLanalysis evaluated in Section [15.6.5*] of the-FSAR.)** I Operating experience indicates that MSIVs: frequently 1 exceed the Technical Specification: allowable leak rates. 'Some.of these valves repeatedly fail the loct1 leak rate . tests-despite frequent disassembly and refurbishment.

As a result >of increasing MSIV leakages and the potential consequences s

4. - - Plant-specific - . - .

Plant-specific. 'In some BWRs, only the containment design leak rate is

. included in the LOCA radiological analysis.

p:

Type 1- B1-15 u

-i.- -, ,~1_-, - - _ - .J-_,, - - -- - - - - ~,

NED0-31858 Rev. I following a LOCA, the Nuclear Regulatory Commission prioritized Generic issue C-8, "MSIV Leakage and LCS Failure" as a high priority item in January 1983. This issue was closed in 1990.

The BWROG formed a MSIV Leakage Committee in 1982 to address the increasing MSIV leakage rates, and a follow-on MSIV Leakage Closure Committee in 1986 to address alternate actions to resolve on going, but less severe MSIV problems. The MSIV Leakage Committee identified contributors which cause MSIVs to fail the leak rate tests by large margins, developed recommendations to minimize leakages, evaluated alternates for MSly leakage treatments, and compiled recent history of MSIV leakages and LCS operating experience.

3.0 JUSTIFICATIONS FOR THE PROPOSED CHANGE

[ Corporation) proposes to increase the Technical Specification allowable leakage rate for the MSIVs from [x] scfh to [y] scfh per main steam line.

The current Technical Specification MSIV leak rate limit is extremely limiting and routinely requires unnecessary repair and re-test of the MSIVs. This significantly impacts the maintenance work load during plant outages and often contributes to outage extensions. (The outage planners at [ plant station name) typically schedule reveral days of contingency to repair and re-test the MSIVs)* In addition, the needless dose exposures to maintenance personnel are inconsistent with As Low As Reasonably Achievable (ALARA). principles. (There have been many Licensee Event Report written for MSIVs failing to meet the current Technical Specification limit.)*

_(From a safety perspective, calculations using_ standard conservative assumptions for considering the off-site consequences of a postulated

+

Plant Unique i

Type-1 B1-16

NED0-31858 Rev. I design-basis LOCA confirm that off-site and control room doses will be within the regulatory _ guidelines for the allowable HSIV leakage rate.

This calculation is described in Section 15.6.5 for the FSAR. However, if MSiv leakages are only moderately higher than the allowable limit, the calculated doses will exceed the regulatory guidelines.)**

MSIV's failure to meet the current Technical Specification limit have been documentr? in response to surveys conducted by the Nuclear Regulatory C(o..ssion during the early 1980 and by the BWROG during the middle and late 1980s. As high as 50% of the total "as found" MSIV local leak rate test: were reported in the early NRC survey to cxceed the leakage rate limit.

The BWROG has studied the issues regarding MSIV leakage rates, their causes, and available alternatives. The results of the BWROG study are provided in NED0-31858 Rev. I and are also summarized in NUREG-1169. In response to Generic Issue C-8, the BWROG has recommended corrective actions and maintenance practices to reduce the MSIV leakage rates.

A recent survey conducted by the BWROG of MSIV leakage tests performed between 1984 and 1988 indicates that the implementation of industry and BW90G actions has been eifective in reducing the leakage rates, and, in particular, a reduction in the number of valves which experience substantial high leakage rates. The survey concludes that about 23% of the total "as found" HSIV leakages still exceed the limit of 11.5 scfh and about 10% exceed 100 scfh.

The HSIV-leakage performance at [ plant station name) is .....[to be provided by the Utility]. The leakage performance at [ plant station name] is consistent with the recent survey by the BWROG.

l l Plant-specific. In some BWRs, only the containment design leak rate is I included in the LOCA radiological analysis.

1 i

i Type 1 81-17

NE00-31858 Rev. 1 Despite the recent improvement in leakage performance, MSIV leakage rates l

still frequently exceed the current Technical Specification limit and the safety and maintenance problems related to high MSly leakage rates, although less severe, remain as a significant issue.

Furthermore, based on the extensive evaluation of valve leakage data, the BWROG has found that disassembling and refurbishing the MSIVs to meet very low leakage limits frequently contribute to repeating failure. By not ha ng to disassemble the valve 'nd refurbish them for minor leakage, the utility may avoid ' ~- one of the root causes of recurring valve leakage proble- :d to ter leak test failures and the possibility of 1 tt W ety.

(The current Technical Specification + rate is established by excessively conservative ' .. , cal analysis as described in Section 15.6.5 of the FSAR.)* The vai d s physical size and operability characteristics (large size and fast acting), and existing turbine building equipment were not considered at the time the leakage limit was established. Based on the in-depth evaluation of MSIV leakages, the BWROG has concluded the MSIV leakage rates up to 500 scfh are not an indication of substantial mechanical defects in the valve which would challenge the isolation capability of the valve to fulfill its safety function. Furthermore, valve manufacturers have stated that leakage rates up to 200 scfh can occur without having a major valve defect._ Therefore, the proposed increase from [x] to [y] scfh per main steam line will not inhibit the MSIV's performance of the isolation function and will not compromise the safety of [ plant station name]

Plant-specific. In some BWRs, only the containment design leak rdte is included in the LOCA radiological analysis.

Type 1 B1-18

NE00+31858 Rev, 1 This proposed increa ! provides a more realistic, but still conservative, limit for the MSIVs. Based on the BWROG study, the proposed increast in the allowable leak rate will increase the chance for a successful local leak rate test to greater than 90%, up from the 77% success rate at the current typical limit of 11.5 scfh, This increase in successful local leak rate testing will significantly reduce MSIV maintenance costs, reduce dose exposure to maintenance personnel, reduce outage durations, extend effective service life of the MSIVs, and minimize the potential for outage extensions at [ plant station name),

[ Corporation) proposes to implement the reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off-site and control room dose limits. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks, The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment. The implementation will provide [ Plant Name) with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to the public.

Furthermore, [ Corporation) will incorporate the' applicable leakage treatment methods, consistent with GE document NE00-30324 " Potential Operator Actions to Control MSIV Leakage", into the Operational Procedures and Emergency Operational Procedures.

l Type I B1-19 l

NE00-31858 Rev. 1 4.0 ANALYSIS OF MSIV LEAKAGE CONTRIBUTION TO RADIOLOGICAL DOSE CAL (VLATIONS 4.1 Selection of L.stakaae Treatment Methg_d The BWROG has evaluated several MSIV leakage treatment methods and has recommended the isolated condenser for MSIV leakage treatment. This leakage treatment method takes advantage of the large volume in the isolated main condenser to hold up the release of fission products leaking from the closed MSIVs, The main steam drain lines are employed to canvey leakage to the condenser.

As previously discussed in Section 1.0, the BWR0 has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage. The BWROG has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for other plant safety risks. The BWR0G has also determined that main steam piping and condenser designs are extremely rugged, and that the ANSI-831.1 design requirements typically used for nuclear plant system design contain h good deal of margin. In order to further justify the capability of the main steam piping and condenser treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes). This study concluded the possibility of a failure which could cause a loss of steam or condensate in BWR main steam piping

~

or condensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

This conclu:; ion is consistant with NUREG/CR 4407, " Pipe Break Frequency Estimation for Nuclear Power Plants", dated May 1987, which reported no observed failures in the main steam piping over 313 years of reactor operating years.

Type 1 B1-20

_~ . ..

NED0-31858 Rev. 1 Therefore, the isolated condenser MSIV leakage treatment path at (plant name] is considered appropriate for the reduction of radiological consequences of a design basis LOCA.

4.2 Radioloaical Analysis and Result The radiological dose methodology has been developed by General Electric for the BWROG.

Tlie radiological analysis calculates the effects of the proposed allowable MSIV leak rate in terms of control room and off-site doses.

The revised LOCA doses are the sum of the LOCA doses (as described in Section xxx of the FSAR) and the calculated MSIV leakage doses. (This method of calculating the revised dose exposures is very conservative since the'LOCA doses already include the dose contribution from MSIVs at the maximum leakage rate permitted in the current Technical Specifications.]*

Table 1** shows the calculated dose exposures from the BWROG radiological analysis for (plant name). Regulatory limits and calculated doses from

LOCA radalogical analysis are.also included in Table 1 for comparison purpose. This analysis demonstrates that a MSIV leakage rate of 200 scfh per main steam line results in an acceptable increase to the dose exposures previously calculated for the control room, EAB, and the LPZ.

The revised LOCA doses remain well within the guidelines of 10CFR100 for off-site doses and 10CFR50, Appendix A, (General Design Criteria 19) for the control room doses. Furthermore, the calculation shows that MSIV Plant-specific. In some BWRs, only the containment design leak rate is included in the original LOCA radiological analysis. For applicable BWRs, the utility may elect to replace the LOCA MSIV dose contribution with'the new doses using the leakage treatment method.

The following discussion on doses are based on as an example plant (Hope Creek)

Type 1 B1-21

NED0-31858 Rev. I t

leakage rates up to approximately [z] scfh per steam line would not exceed the regulatory limits. Therefore, the proposed method provides a substantial safety margin for mitigating the radiological consequences of MSIV leakage beyond the proposed Technical Specification leak rate limit of [y] scfh.

5.0 BENEFITS FOR THE PROPOSED CHANGE As discussed in NED0-31858 Rev. 1, recent MSIV leakage performance has significantly improved since the early BWROG survey in 1984 and the NRC survey in the early 1980s. L'espite the recent improvement, MSIV leakage rates exceeding the current T3chnical Specification limits still

_ frequently occur. The BWROG evaluation of the recent MSIV leakage performance concludes that the proposed change will improve the chance for a successful local leak rate test to greater than 90%, up from the 77% success rate at the current Technical Specification limit of 11.5 scfh.

Specifically, MSly leakage experience at (plant station name] are [to be provided by Utility on plant-specific leakage performance and any problems which may impact critical path, outage extensions, etc.).

Furthermore, the risk to the public health and safety will be reduced ~

with the implementation of the proposed MSIV leakage treatment method.

The implementation will provide [ Plant Name] with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to th' public.

Type 1 B1-22

NED0-31858 Rev. 1 Table 1 CONTF,IBUTION TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM HSIV LEAK RATE OF 200 scfh HOPE CREEK GENERATING STATION Whole Body Thyroid Beta Ltg.m.1 frem) (rem)

Exclusion Area A) 10CFR 100 Limit 25 300

• Boundary (2-Hour) B) Previous Calculated 0.6 76.7 Doses **

C) Contribution from 0.1 2.6 MSIVs at 200 scfh D) Ncw Calculated Doses 0.7 79.3 Low Population A) 10CFR 100 Limit 25 300

• Zone (30-Day)

B) Previous Calculated 0.08 7.7 Doses **

C) Contribution From 0.34 57.5 MSIV at 200 scfh D) New Calculated Doses 0.42 65.2 Control Room A) GDC-19 5 30 30/75***

(30-Day)

B) Previous Calculated 0.0* 0.26 .91 Doses **

C) Contribution From 0.10 2.71 1.64 MSIVs at 200 scfh D) New Calculated Doses 0.14 2.97 2.55

• No limit specified.

FSAR Section 15.6.5.5 and 6.4 (includes MSIV leak rate at a total of 45 scfh for the first 20 minutes; control room dose assumes 100% per day reactor building inleakage).

75 if prior commitment has been made to use protective clothing.

Type 1 B1-23

l NED0 31858 Rev. 1 BC.LQSVRE 3 L(QBPORAT10N1 LICENSE NO. Ixx21 DOCKET NO. Ixxx1 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS

[ Corporation) proposes an amendment to the Technical Specifications to revise Section 3.6.1.2* to permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs) from the current [x] standard cubic feet per hour (scfh) to [y] scfh per main steam line.

Pursuant to 100FR50.92, the proposed amendment involves no significant hazards considerations.

Ihg_poeration of IPlant Station Namel. in accordance with the cronosed amendment. will not involve a sianificant increase in the probability or conseauences of an accident oreviously evaluatg6 The proposed amendment to Section 3.6.1.2 does not involve a change to '

structures, components, or systems that would affect the probability of an accident previously evaluated in the Final Safety Analysis Report (FSAR).

Since MSIV leakage is included in the radiological analysis for the design-basis LOCA as described in Section 15.6.5 of the FSAR, the proposed amendments will not affect the precursors of other analyzed accidents. The proposed amendments result in acceptable radiological consequences of the design-basis LOCA previously evaluated in Section 15.6.5 of the FSAR.

The [ plant station name] has an inherent MSIV leakage treatment capability.

[ Corporation) proposes to use the main steam lines and condenser for MSIV 3 -

3 Type 1 B1-24

NE00-31858 Rev. I leakage treatment and will incorporate this method in the Operational Procedures and Emergency Operational Procedures.

The BWROG has evaluated-the availability of main steam system piping and condenser treatment pathways for processing MSly leakage, and has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for other plant safety risks. The BWROG has also determined that main steam piping and condenser designs are extremely rugged, and that the ANSI-831.1 design requirements typically used for nuclear plant system design contain a good deal of margin.

In order to further justify the capability of the main steam piping and condenser treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes). This study concluded that the possibility of a failure which could cause a loss of steam or condensate in BWR main steam piping or condensers in the event of a design basis earthquake is highly unlikely, and that such- a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

[ Corporation) will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NE00-31858 Rev.1, to provide reasonable assurance of the structural integrity of these components.

A plant-specific radiological analysis has been performed to assess the effects of the proposed increase to the allowable MSIV leak rate in terms of control room and off-site doses following a postulated design basis LOCA.

This analysis utilizes the hold-up volumes of the main steam piping and condenser for the MSIV leakages. As discussed earlier, there is reasonable assurance that the main steam piping and condenser will remain intact following a design basis earthquake. The radiological analysis uses standard conservative assumptions for the release of source terms consistent with Type 1 B1-25

NE00-31858 Rev. 1 l l

Regulatory Guide 1.3 Revision 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of a loss-Of-Coolant Accident for Boiling Water Reactors", dated April 1974.

The analysis demonstrates that dose contributions from the proposed MSiv leakage rate limit of [y] scfh result in an acceptable increase to the LOCA doses previously evaluated against the regulatory guidelines for the off-site doses and control room doses as contained in 10CFR100 and 10CFR50, Appendix A (General Design Criteria 19), respectively. The LOCA doses previously evaluated are discussed in Section [15.6.5]* of the FSAR. The revised LOCA doses are the sum of the LOCA doses previously evaluated in the FSAR and the additional MSIV doses calculated using the proposed treatment method. (This method of calculating the revised doses is very conservative since the LOCA doses previous evaluated already included dose contributions from MSIV at the maximum leakage rate permitted in the current Technical Specifications.)**

The attached table shows the previous calculated doses and the new calculated doses.

The whole body dose at the Low Population Zone (LPZ) and the control room is increased from 0.08 to 0.42 rem *** and from 0.04 to 0.14 rem, respectively.

These increases are acceptable because the revised doses are well within the Regulatory guidelines (0.42 versus 25 rem at the LPZ, and 0.14 versus 5 rem at the control room). The associated whole body dose at the exclusion area boundary (EAB) increased insignificantly from 0.6 to 0.7 rem.

Plant-specific Plant-specific, In some_ BW%, only the containment design leak rate is included in the previous LOCA radiological analysis. For applicable BWRs, the original LOCA dose from MSIV leakage may be replaced with revised dose.

All doses and comparisons shown here are Hope Creek specific.

l l

l Type 1 B1-26

NEDO-31858 Rev. 1 CONTRIBUTION TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV LEAK RATE OF 200 scfh HOPE CREEK GENERATING STATION Whole Body Thyroid Beta Irfjnl frem) (rem)

Exclusion Area A) 10CFR 100 Limit 25 300

• Boundary (2-Hour) B) Previous Calculated 0.6 76.7 Doses **

C) Contribution from 0.1 2.6 MSIVs at 200 scfh D) New Calculated Doses 0.7 79.3 Low Population A) 10CFR 100 Limit 25 300

• Zone (30-Day)

B) Previous Calculated 0.08 7.7 Doses **

C) Contribution from 0.34 57.5 MSIV at 200 scfh D) New Calculated Doses 0.42 65.2 Control Room A) GDC-19 5 30 30/75***

(30-Day)

B) Previous Calculated 0.04 0.26 .91 Doses **

C)- Contribution from 0.10 2.71 1,64 MSIVs at 200 scfh D) New Calculated Doses 0.14 2.97 2.55 No limit specified.

FSAR Section 15.6.5.5 and 6.4 (includes MSIV leak rate at a total of 45 scfh for the first 20 minutes; control room dose assumes 100% per day reactor building inleakage).

• • • ' 75 if prior ' commitment has been made to use protective clothing, i

Type 1 B1-27 i

NEDO 31858 Rev. I j The thyroid dost at the LPZ increased from 7.7 to 65.2 rem. This increase is acceptable because the revised dose of 65.2 rem is significantly less than the regulatory guideline (300 rem). The EAB thyroid dose increased slightly from 76.7 to 79.3 rem, whereas the control room thyroid dose increased from 0.26 tc 2.97 rem. The increase in control room thyroid dose is acceptable because the revised dose remains a small fraction (9.9%) of the limit. The control room beta dose is increased from 0.91 to 2.55 rem, which remains insignificant relativa to the regulatory guideline of 30 rem.

It is important to note that the resulting doses are dominated by the organic iodine fractions which occur because of the ultraconservative source term assumptions used in this analysis. For 200 scfh per steam line, more than 85%

of the off-site iodine and control room doses are due to organic iodine from the RG 1.3 source term and organic iodine converted from the elemental iodine deposited in main steam piping systems. If the actual iodine composition from the fuel release (cesium iodide) is used in the calculations, essentially all of this organic iodine dose would be eliminated, in summary, the proposed change results in an acceptable increase to the radiological consequences of a LOCA previously evaluated in the FSAR. The revised LOCA doses are well within the regulatory guidelines. Although the revised LOCA doses are slightly higher for low HSIV leakage rates, the effectiveness of the proposed method even for leakage rates greater than the proposed increased MSIV allowable leak rate, ensures that off-site and control room doses are not exceeded.

The ooeration of IPlant Station Namel. in accordance with the crocosed amendment. will not create the nossibility of a new or different kind of accident from any accident previou1]v evaluated.

The proposed amendment to Section 3.6.1.2 does not create the possibility for  !

a new or different kind of accident from any accident previously evaluated.

Type 1 B1-28

NE00-31858 Rev. 1 The BWROG evaluated MSIV leakage performance and concluded that MSIV leakage rates up to 200 scfh will not inhibit the capability and isolation performance of the valve to isolate the primary containment. There is no new modification which could impact the HSIV operability. The LOCA has been analyzed using the main steam piping and condenser as a treatment method to process MSIV leakage at the proposed maximum rate of [y] scfh. Therefore, the proposed change does not create any new or different kind of accident from any accident previously evaluated in the FSAR.

The ooeration of IPlant Station Namel, in accordance with the orocosed amendm3nt. will not involve a sianificant reduction in the marain of safety.

The proposed amendment to Section 3.6.1.2 does not involve a significant reduction in the margin of safety. [As discussed in the Bases of the Technical Specification 3/4.6.1.2, the allowable leak rate limit specified for the MSIVs is used to quantify a maximum amount of bypass leakage assumed in the LOCA radiological analysis.]* Results of the analysis are evaluated agair: t the dose guidelines contained in 10CFR100 for the off-site doses and 10CFR50, Appendix A (General Design Criteria 19) for the control room doses.

Therefore, the margin of safety is considered to be the difference between the calculated doses and the guidelines as contained in 10CFR100 and GDC 19.

Results of the radiological analysis demonstrate that the proposed change does not involve a significant reduction in the margin of safety. The whole body doses, in terms of margin of safety, are insignifi.:antly reduced by 1.4% at the LPZ, 2.0% in the control room, and 0.4% et the EAB. The thyroid dose margin of safety is reduced by 19.1% at the LPZ, 9.0% in the control room, and 0.9% at the EAB. The beta dose is insignificant 1y reduced by 5.5% in the control room in terms of margin of safety. The .aargins of safety are not significantly adversely affected because the absolute margins of safety remain Plant-specific Type 1 B1-29 l

)

e NED0;31858-Rev,.1 l

well below the guidelines (lowest whole body margin of safety is 97.2% at the EAB,-lowest thyroid margin-of safety is 73.6% at EAB).

In terms of MSIV leakage treatment capability, the overall margin of safety is increased.- (Corporation) proposes to implement the reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is-effective to treat MSIV leakage over an expanded operating range

- without exceeding the off-site and control room dose limits. Except for the

requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and

- interlocks.- The method is consistent with the philosophy of protection by

- multiple leak-tight barrders used in containment design for limiting fission product release to.the.. environment. Therefore, the proposed method is highly L reliable for MSIV leakage treatment. - The implementation will provide (Plant )

~

Name) with a capability to process MSIV leakage,~and will also provide a j uniform basis for establishing a plant-specific MSIV leakage rate limit. From

~

a. safety perspective,-the proposed changes result in an increase in protection to the public.

Furthermore,, the revised LOCA doses remain well within the regulatory limits for theisite and control room. -The calculation shows that MSIV leakage rates up to [z]. scfh- per steam line would not exceed the regulatory. limits.

--Therefore .the proposed method provides-a substantial safety margin for mitigating the _ radiological consequences of MSIV leakage beyond the proposed!

Technical Specification leak rate-limit of [y]_ scfh.- '

Therefore, the proposed amendment to the Technical Specifications-does not-involve a significant. hazards consideration.

O Type 1 81-30

NED0-31858 Rev. I DitLOSURE 4 ICORPORAT10N1 LICENSE NO. Ixxx1 DOCKET NO. Ixxx1 APPLICATION FOR EXEMPTION TO APPENDIX A OF 10CFR100 (CORPORATION], holder of Facility Operating License No. NPF-(xxx], hereby requests an exemption of the downstream main steam piping and condenser from the seismic requirements specified in Appendix A of 10CFR100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants.

Specifically, (Corporation] proposes to employ probability analysis, existing design capabilities, seismic exparience, and a plant specific seismic adequacy verification, as alternate methodology to the dynamic awlysis or qualification test specified ir. Paragraph VI(a)(1) of 10CFR100 Appendix A, to provide reasonable assurance that the existing main steam piping and condenser will remain functional following a design basis accident coincident with a significant seismic event.

The exemption would allow the ex' sting, non-seismically designed main steam piping and condenser to be used for mitigating the radiological consequences of MSIV leakage during the duration of a Design Basis Accident, such that the resulting doses are within the guidelines of 10CFR100.

(Corporation] recognizes that there is no provision in 10CFR100 for exemption; however, the Nuclear Regulatory Commission (NRC) has the authority to grant this exemption.

[ Corporation) proposes to implement the reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off-site and control room dose limits. Except for the Type 1 B1-31

NED0-31858 Rev. I requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and

' interlocks. -- The method is consistent with the philosophy of protection by ,

multiple leak-_ tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment. The implementation will provide [ Plant

-Name) with a capability to process MSIV 1(akage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes- result in an increase in protection to the?public.

In conjunction with this application for exemption request, [ CORPORATION] has transmitted to the NRC an application for a license amendment pursuant to 10CFR50.90. This license amendment involves a proposed change to Section 3.6.1.2 of the Technical Specifications to permit an increase in the allowable leak rate for the HSIVs from the current [x] standard cubic feet per hour

'(scfh); total to [y] scfh per main steam line. The safety analysis has been revised to assess'the radiological effects of MSIV leakage following a postulated design basis LOCA. [ CORPORATION] has demonstrated that the proposed change-does not involve a significant hazards consideration.

This proposed = exemption is a _ result of the extensive work performed.by the BWR Owners' Group:(BWROG)'in suppor+.-of the resolution of Generic Issue C-8 "MSIV Leakage and 'S Failure".

The fo' j discussion provides a detailed-justification and evaluation of the pt .d exemption. While recognizing this exemption criteria are=

specifi t 'v applicable to'10CFR50, [ Corporation) has evaluated the proposed exemption in accordance with the criteria specified inl10CFR50.12(a). -The proposed exemption will not'present an undue risk to the public health and safety and is consistent with the' common' defense and security. Furthermore, special' circumstances are present that warrant the granting of this- exemption..

Type 1 B1-32

l NED0 31858 Rev. I f-The proposed exemption will not cause additional operational activities that may significantly affect the environment, it does not result in a sigt;ificant increase in any adverse environmental impact previously evaluated in the Final Environmental Irpact Statement-0perating License Stage, result in a significant change in effluents or power levels, or affect any matter not previously reviewed by the NRC that may have a significant adverse environmental impact.

Upon he NRC approval of the license amendment and exemption request.

[Cc, ration] will perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0 31858 Rev. 1, to prnvide reasonable assurance of the structural integrity of these components.

Therefore, [C0APORAT10N] hereby requests an exemption to the seismic requirements af 10CFR100 Appendix A for (plant station name] to permit the use of existing, non-seismically designed main steam piping and condenser to mitigate the radiological consequences of MSIV leakage.

A. Justification Paragraphs Vl(a)(1) of 10CFR100 Appendix A requires that structures, systems and component:., which assure the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures of 10CFR100, be designed to remain functional following a safe shutdown earthquake (SSE) and concurrent loads. The engineering method used to assure that the required safety functions are maintained following the SSE shall involve the use of either dynamic analysis or a suitable qualification test to demonstrate that .ructure, systems, and components can withstand the seismic and other concurrent loads.

The BWROG has evaluated the capability of main steam piping and condensers to process MSIV leakage following a design basis accident coincident with a Type 1 B1-33

NED0-31858 Rev. 1 l

seismic event. Based on this comprehensive evaluation, the BWROG has concluded there is reasonable assurance that the main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the d?ign basis earthquake, to mitigate the radiological consequences of MSIV leakage. The following conclusions provide the bases for this assurance:

(1) Probability for which the resulting dose from MSly leakage is significant is extremely low. This requires a design basis LOCA, a degraded core where ECCS are not functional, and a significant seismic event.

(2) Main steam piping and condensers are designed to strict industrial standards and building codes; thus, significant design margin exists.

(3) Main steam piping and condensers exhibit substantial seismic ruggedness. Comparisons of piping and condenser design in GE plants L with those in the earthquake experience database reveal that the GE plant designs are similar to or more rugged than those that have exhibited good earthquake performance.

(4) Possibility of significant failure in GE BWR main steam piping or condensers in the event of design basis earthquakes is highly unlikely, and any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

(5) A plant-specific verification of seismic adequacy of the main steam piping and condenser will be performed to provide reasonable assurance of the structural integrity of these components.

In support of the above, the BWROG has reviewed the potential combinations of Loss-Of-Coolant Accidents (LOCAs) and seismic events of interest:

1 Type 1 B1-34

NED0-31858 Rev. 1 (1) LOCA WITHOUT NEAR COINCIDENT SEISMIC EVENT. -

for this occurrence the l pro sure in the piping system downstream of the MSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the flow path through main steam system piping to the condenser is assured.

(2) SEISMIC EVENT WITH0VLRLAR CQINCIDENT LOCA. Without a LOCA and the potential associated core degradation, the radioactivity transported with MSIV leakage is of no radiological significance.

(3) LOCA WITH NEAR COINCIDENT SEISMIC EVENT. For this occurrence (also assuming significant core damage) the consequences are of interest because a seismic induced failure in the main steam or condenser system could allow MSIV leakage to bypass the treatment pathway, it has been previously well documented that the probability of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability approximately 0.001 per reactor per year; core melt probability is plant-specific and typically ranges from 0.00001 to 0.0001 per reactor per year). It is also noted that a LOCA does not induce a seismic event, and that a seismic event has a very low probbility of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10-7 per reactor per year).

Considering that the probabil'y of a near coincident LOCA and seismic event is much smaller than other ant-safety risks (less than 1 x 10-7 per reactor per year for coincident events, less than 5 x 10-7 per reactor per year for seismic induced LOCA), the concern for main steam piping or condenser damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design asis seismic events.

Type 1 B1-35

NEDO 31858 Rev. 1 ANSI 831.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin. In cddition, specific seismic design provisions have been incorporated into sorW newer BWR main steam and condenser systems.

To furthor justify the capability of the main steam system piping and condenser treatment pathway, the BWROG has retsewed limited earthquike experience data on the performance of non sei mically designed piping and condensers (in past carthquakes). The study summarizes data on the performance of main steam piping and rondensers in past stror.g motico carthquakes and compares these piping and condensers with those in typical U.S. GE Miirk I, 11 and 111 nuclear plants. This limited eart q uake experience data and similarity comparisons are then used to further strengthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in a design basis earthquake in conjunction with a LOCA o: curring just prior to or ter the seismic eveht.

The earthquake expt ce data are derived from an extensive database on the performance of indus il facilities end over 100 power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world from 1934 to the present.

The piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they are not designed to resist earthquakes. This is a common conclusion in studies of this typt on other plant items such as n ided steel piping in general, anchored equipe. eat such as motor control centers, pumps, valves, structures, and so forth. Ttr t is, with limited exceptions, normal industrial corstruction and equipment typically have substantial inherent seismic ruggedness, even when they are not designed for earthquakes. No failures of main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.

Type 1 B1-36

NEE'.31850 Rev. 1 Comparisons of piping and condenser design in example GE Mark 1, 11, and ill plants with those in the earthquake experience database rtveal the GE plant designs are similar to or more rugged than those that exMbited gosd earthquake performance. The BWROL concludes that (1) the possibility of significant failure in GE BWR main steam piping or condensers in the event of an eastern U.S. design basis earthquake is highly unlikely and that (2) any such failure would also be contrary to a large body of historical earthquake

experience data, and thus unprecedented.

Earthquake experience methodology has been applied in seismic equipment qualification issues associated with Unresolved Safety issue A 46 (Seismic Qualification of Equipment in Operating Plants). Piping performance data are presented in NUREG 1061 (a report from the NRC Piping Review Committee), and this report proposes changes to criteria that era directed toward the recognition of the superior performance of riping in earthquakes and estab-lishing more realistic seismic criteria for piping qualification. The NRC has published NUREG 1030 and NVREG-1211 " Seismic Qualification of Equipment in Operating Nuclear Power Plants," which conclude that the seismic experience data approach provides the most reasonable and preferred alternative to other current equipment qualification methods.

The rapidly growing use of the seismic experience data approach is further illustrated by the fact that this method of analysis is now referenced in:

A. Draft RG 1.100, Revision 2

• Seismic Qualification if Electrical and Mechanical Equipment in Nuclear Power Plants"

8. Recent approved revision of IEEE Standard 344-1987, " Recommended Practice for Seismic Qualificstion of Class IE Equipment for Nuclear Power Generating Stations" C.- Draft report of ASME Standard ' Recommended Practice for Seismic Performance Qualification of Hechanical Equipment Used.in Nuclear Power Plants."

Type 1 B1-37

i HE00 31858 Rev. 1 The earthquake experience database includes a large number and variety of piping systems. In fact, piping is probably the strongest area in this regard (compared to areas like electrical or mechanical equipment, cable trays, etc.). It has been concluded that the earthquake experience data on piping, and in particular data on main steam piping, are applicable to main steam piping in BWRs.

In both nuclear and conventional power plants, the condenser is designed to reduce the low pressure turbine outlet pressure (thereby increasing turbine efficiency) and to condense the steam. The nuclear environment does not impose additional significant design considerations on the condenser. With the exception of hotwell size, a conventional plant and nuclear pl6,nt with similar per'ormance parameters have similar condensors.

None of the condensers within the seismic experience database has seismic design criteria. However, in view of the performance of the condensers within the database, it is concluded that the condensers have an inherent seismic ruggedness and that the earthquake experience data on condensers are applicable to condensers in BWRs.

Another recent study to develop, by data collection and statistical analysis, updated estimates of pipe breaks in commercial U.S. nuclear power plants was completed in 1987. This study evaluates both LOCA sensitive systems and non-LOCA sensitive systems. For BWR non-LOCA sensitive systems, ten pipe failures have occurred over 313 years -f operating experience. None of these failures occurred in the main steam piping. Based on the observed failure rates, this study estimated the failure rate for the main steam system piping to be 0.0007 failure / year /BWR with an upper bound of 0.0096 failures /

year /BWR. These results are consistent with the conclusion from the earthquake databases and NUREG ll69: BWR main steam piping is designed to withstand severe plant transients such as turbine trips and is expected to remain intact following accidents as severe as a design basis LOCA. Thus, the non-seismically designed main steam piping and the main condenser can be used to mitigate the consequences of MSly leakage.

Type 1 B1-38

NE00 31858 Rev. 1 l

(Corporation) will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines d'scussed in Section 6.7 of NED0 31858 Rev.1, to provide reasonable assurance of the structural integrity of these components.

In conclusion, there is reasonable assurance that the existing, non seismically designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage.

B. No Undue Risk to Public Health and Safely The BWROG has evaluated the capability of main steam piping and condensers to process MSly leakage following a design basis accident coincident with a seismic event. Based on this comprehensive evaluation, the BWROG has concluded there is reasonable assurance that the main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage. This assurance is based on methodology using probability analysis, margins in existing design codes, seismic experience, and a plant specific verification of seismic adequacy.

The treatment method for HSiv leakages is recommended by the BWROG in support of the resolution to Generic issue C-8 'HSIV Leakage and LCS failure".

[ Corporation) proposes to implement the reliable and effective main steam piping and condenser for HSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off-site and control room dose limits. Except for the requirement to establish a proper flow path from the HSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission Type 1 BI-39

NEDO 31858 Rev. 1 product release to the environment. Therefore, the proposed method is highly reliable for MS!V leakage treatment. The impicmentation will provide [ Plant Name) with a capability to process MSly leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit, from a safety perspective, the proposed changes result in an increase in protection to the public.

In conclusion, the proposed exemption presents no undue risk to public health and safety.

C. Consistent with Common Defense and Zacurity With regard to the " common defense and security" standard, the grant of the requested exemption is consistent with the common defense and security of the United States. The Commission's Statement of Considerations in support of the exemption rule note with approval the explanation of this standard as set forth in Lona Island Liahtina Comcany (Shoreham Nuclear Power Station, Unit 1), LBP 84 45, 20 NRC 1343, 1400 (October 29,1984). There, the term ' common defense and security" refers principally to the safeguarding of special nuclear material, the absence of foreign control over the applicant, the protection of Restricted Data, and the availability of special nuclear material for defense noods. The granting of the requested exemption will not affect any of these matters and, thus, such grants are a,,isistent with the common d3fense and security.

D. Special Circumstances Are preji.gni Special circumstances are present which warrant iscuance of this requested exemption. These special circumstances are discussed in accordance with the classificationcon'ainedin10CFR50.12(a)(2):

(ii) .Ar'lication of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule, 1

Type 1 81 40

i i

NE00 31858 Rev. 1 l

Compliance with Appendix A of 10CFR100 for the downstream main steam piping

{

and condenser is not necessary to achieve the underlying purpcse of the rule. }

The underlying purpose of the rule is to limit releases to within the off-site dose limits of 10CFR100. The regulation requires components that mitigate the  !

consequences of an accident to within the dose limits of 10CFR100 be designed

-f to the seismic requirements of 10CFR100 Appendix A. The regulation is i intended to nrovide a. reasonable assurance that tho' components will remain

-functional for the mitigating function. For the purpose of mitigating the I radiological consequences of MSIV. leakage, it is not necessary to apply the seismic requirements of 10CFR100 Appendix A to the main steam piping and condenser in order to achieve the underlying purpose of the rule because:

(1) There is: reasonable assurance that the existing, non seismically  ;

designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as -

great asLthe design basis earthquake, to mitigate the radiological ,

consequences of MSIV leakage. This assurance is based on methodology using probability. analysis, margins in the existing design codes,  ;

seismic experience, and a plant specific verification of seismic-adequacy.

(2)-The safety analysis has been revised to assess the' radiological consequences-of MSIV leakage following a design basis LOCA. The analysis has demonstrated that the revised doses are well within the i off-site dose guidelines of 10CFR100. .

1 i

Furthermore, the seismic approach is consistent with the current' resolution or

.the seismic and. equipment qualification issues. Earthquake experiences data

'have applied in seismic equipment qualification issues associated with '

-Unresolved Safety issues A 46 (Seismic Qualification of Equipment in Operating Plants). Piping. performance data have been presented in NUREG-1061, a report.

, from the NRC Piping Review Committee, which proposes changes to criteria that.

are' directed toward the recognition of the_ superior performance of piping in L

Type-1 .B1-41

. _ , - . . ~ . , [ ... ~ . . _ , . .._, ... ,J- ..,,a.._..-..~...._,.,,_.;. , . , - . _ . . . - ~ _ . . . . , . . _ , , , . . . . - _ . . . . -

1 NED0 31858 Rev. 1 earthquakes and establishing more raslistic seismic criteria for piping qualification. The NRC has published NUREGs 1030 and 1211 " Seismic ,

Qualification of Equipment in Operating Nuclear Power Plants," which conclude l that the seismic experience data approach provides the most reasonable and )

preferred alternative to other current equipment qualification methods.

(iii) Compliance would re ult in undue hardship or other costs that I are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.

The proposed MSIV leakage treatnent method utilizes the existing main steam piping and condenser for the mitigating function. Compliance with the seismic requirements of 10CFR100 Appendix A for the main steam piping and condenser would require significant upgrade of the existing equipment, lead to unnecessary long term plent shutdown for modification, and significantly increase maintenance requirements and the associated costs in order to meet seismic qualification requirements.

(iv) The exemption would tesult in benefit to the public health and safety that compensates for any decrease in safety that may result from the granting of the exemption.

[ CORPORATION) has transmitted to the NRC an application for a license amendment which involves proposed change to the Technical Specifications to increase the allowable MSIVs leak rate from [x] scfh total to [y] scfh per steam line. This application is partly based on the fact that the current limit is too restrictive, and results in excessive MSIV maintenance and roair, leading to additional MSIV failures which in turn result in higher L leakages. The proposed limit will benefit the public health and safety by l: reducing the potential for MSivs failures, and thus keeping the MSIV leakage within the radiological analysis values.

Type-1 B1-42

NEDO 31858 Rev. 1 (Corporation) proposes to implement the reliable and effective main steam l piping and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off site and control room dose limits. Except for the requirement to establish a proper flow path from the HSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consisteht with the philesophy of protection by multiple leak-tight barriers used in containment des!gn for limiting fission product relcase to the environtnent. Therefore, the proposed method is highly reliable for MSIV leakage treatment. The implementation will provide [ Plant Name) with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a plant specific MSIV leakage rate limit. From a safety perspective.-the proposed changes result in an increase in protection to the public.

The exemption from 10CFR100 Appendix A seismic requirements for the downstream piping and condenser is required so that (plant station name) can operate with the proposed Technical Specifications limit of [y] scfh and with a capability to process MSIV leakage. This benefit will compensate for any decrease in safety that may result from the granting of this exemption.

Thus, special circumstances exist warranting the granting of this exemption.

E. Environmental Impact The proposed exemption has been analyzed and determined not to cause additional construction or operational activities which may significantly affect the environment, it does not result in a significant increase in any

adverse environmental impact previously evaluated in the final Environmental L Impact St,c.

ment Operating License Stage, result in a significant change in effluents or power levels, or affect any matter not previously reviewed by the Nuclear Regulatory Commission which may have a significant adverse l

environmental impact.

I l

1 Type 1 B1-43

NEDO 31854 Rev. 1 The proposed exemption does not alter the land use for the plant, any water uses or impacts on water quality, air or ambient air quality. The proposed '

action does not affect the ecology of the site and vicinity and does not affect the noise emitted by station. Therefore, the proposed exemption does not affect the analysis of environmental impacts described in the environmental report. ,

I l

l

-Type 1 B1-44

NED0 31058 Rev. 1 APPENDIX 0 BWROG HSIV LEAKAGE CLOSURE COMMITTEE TYPE 2

, EXAMPLE NRC SUBMITTAL FOR INCRLASING MSIV LEAK RATE LIMITS, EXEMPTION REQUEST TO 10CFR50, APPENDIX J, AND EXEMPTION REQUEST 10 10CFR100, APPENDIX A

^

Type 2 02-1

NED0 31858 Rev. 1 U.S. Nuclear Regulatory Comission

• Attn.: Document Control Desk Washington, DC 20555 RE: [PlantStationName)

DocketNo.[xxxxx)

LicenseNo.[xxx)

Gentlemen

[ Corporation)herebytransmitsanapplicationforamendmenttothe[ Plant Name) Facility Operating License [xxx), and applications for specific exemption to 10CFR50 Appendix J and exemption to 10CFR100 Appendix A.

[ Corporation) requests an amendment to the Technical Specifications, set forth in Appendix A to the License, to permit an increase in the allowable leak rate for the main steam isolation valves (HSIV). In addition. [

(Corporation) requests that the HSIV leakage be exempted from the Type A and Type C test acceptance criteria specified in the Appendix J of 10CFR50. Also,

[ Corporation) requests that the downstream main steam piping and condenser be exempted from the seismic requirements specified in Appendix A of 10CFR100.

In support of the' proposed changes to the Technical Specifications, enclosed for the Nuclear Regulatory Commission's review are:

(1). Application for Amendment to the facility Operating License; Technical Specification pages affected by the changes; and supporting Information and analyses addressing the changes. The analyses demonstrate-that the proposed changes do not involve a significant hazards consideration pursuant to 10CFR50.92.

(2) Application for specific exemptier, to 10CFR50 Appendix J; and ,

supporting information and justification for this exemption.

Type 2 B2-2

NED0 31858 Rev. 1 Pursuant to 10CfR50.12, this exempt;)n request is authorized by law, will not present ur.due risks to the public health and safety, and is consifient with the common defense and security.

Furthermore, special circumstances are present which warrant issuance of this exmo61on request.

(3) Application for specific exemption to 10CFR100 Appendix At and supporting information and justification for this exemption.

[ Corporation) realizes that there is no provision in 10CFR100 for exemption; however, the Nuclear Regulatory Commission has the authority to grant this exemption. This exemption will not present undue risks to the public health and safety, and is consistent with the common defense and security. Furthermore, special circumstances are present wh: warrant issuance of this exemption request.

These proposed changes are supported by BWR Owners' Group (BWROG) work.

The BWROG formed a MSIV Leakage Committee in 1982 in response to Generic Issue C 8 'MSIV Leakage and LCS failure". Generic issue C 8 addressed the safety concerns that reported MSiv leakages are too high and that the Leakage Control System will not function at high MSIV leakages. Based on the extensive, ongoing work performed by the BWROG to support resolution of the Generic Issue, the BWROG has developed the technical justification for the proposed Technical Specifications changes and associated exemption requests The General Electric (GE) report, NED0 31858 Rev.1, describes the safety benefits and provides justification for the proposed changes. With regard to increasing leakage rate limit for the HSIVs, this will avoid unnecessary maintenance which has in the past adversely affected the operability of the MSIVs.

The proposed change-involves implementing the reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded opert, ting range without Type _-2 B2-3

NE00 31858 Rev. 1 l exceeding the off site and control room dose limits. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not reo' aire any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSly leakage treatment.

Furthermore, the risk to the public health and safety will be reduced with the implementation of the proposed MSly leakage treatment method. The implementationwillprovide[PlantName)withacapabilitytoprocessMSIV leakage, and will also provide a uniform basis for establishing a plant-specific MS!V leakage rate limit. From a safety perspective, the i proposed changes result in an increase in protection to the pubite.

Pursuant to 10CFR50.91(b)(1), (Corporation) has provided a copy of this license amendment request and the associated analysis regarding no significant hazards consideration (s) to the appropriate state representative.

[ Corporation) representatives will be available to discuss or meet with the Nuclear Regulatory Commission staff at your convenience to address this matter.

Very truly yours.

[ CORPORATION]

INAME1 Vice President Typc 2 82-4

NEDO 31858 Rev. 1 Enclosures i

1 cc: RegionalAdministrator, Region [xj

[Name), Director

[Name),ProjectManager (Name),Residentinspector (Name),StateRepresentative F

t l

Type 2- B2 5 l'.

NED0 31858 Rev. 1 UNITED STATES OF AMERICA NUCLEX REGULATORY COMMISSION In the Matter of

[ Corporation) Docket No. [xxx)

[ Plant.Name) aff_IDMll INamel , being duly sworn, states that he is Vice President of

[ Corporation); that he is authorized on the part of said Corporation to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.

ISianature)

Subscribed and sworn to before me, a Notary Public in and for the State of

) and County of ( ),this[ ), day of [_ ),

ISianaturel Notary Public in and for

[Coulity, State)

My Commission expires:

Type 2- B2 6

l NE00 31858 Rev. 1 UNITED STATES NUCLEAR REGULATORY COMMISSION in the Hatter of

[ Corporation) DocketNo.[xxx)

[PlantName)

APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the Regulations of the Nuclear Regulatory Commission,(Corporation),holderofFacilityOperatingLicenseNo.[xxx1, hereby requests the Technical Specifications, set forth in Appendix A to the license, be amended to permit an increase in the allomble leak rate for the mainsteamisolationvalves(MSIVs). Specifically, (Corporation) requests that:

1. Allowable leak rate specified in the Technical Specifications Section 3.6.1.2* be changed from [x] to [y] standard cubic foot per hour (scfh) per main steam line. This proposed change raflects a higher, but still conservative allowable leak rate for the MSIVs.

The sections identified are based on the BWR Standard .;ennical Specifications for BWR/5 (NUREG 0123). Each utility should revise these sections as applicable to their individual Technical Specifications.

Also, some plant. sites may specify a maximum allowable MSIV leakage rate

• per line (i.e.,100 scfh), but require that the total is less than four times the maximum allowable leakage rate for each line (i.e., 200 scfh total).

Type 2 B2-7

NE00 31058 Rey, 1

2. Section 3.6.1.2 to be administrative 1y amended to add a footnote exempting MSIV leakages from the overall integrated Icakage rate and from the combined leakage rate for all penetrations and all valves, in concurrence with this application for license amendment, (Corporation) has applied for specific exemption to Appendix J of 10CfR part 50.

The proposed changes to the Technical Specifications are set forth in . Also, in concurrence with this application for license amandment, (Corporation) has applied for an exemption to 10CFR100, Appendix A.

The proposed changes are a result of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic issue C 8 "HSIV Leakage and LCS failure". GE report NED0-31858 Rev. 1 provides technical justification on a generic basis to support these proposed changes.

Based on the BWROG evaluation of MSlV leakage performance, the current Technical Specification allowable HSIV leak rate is extremely limiting and routinely requires the repair and re-testing of MSIVs. This unnecessary repair significantly impacts the maintenance work load, often contributes to outage extensions, and has in the past adversely affected the operability of the HSIVs. DWR outage planners routinely schedule several days of centingency to repair and retest tne HSIVs. In addition, the needless dose exposure to maintenance personnel is inconsistent with As low As Reasonably Achievable (ALARA) requirements. There have also been many Licensee Event Reports written for HSIV leakages exceeding the Technical Specification limit.

The proposed changes will reduce unnecessary MSIV repair costs, avoid unnecessary dose exposure to maintenance personnel, reduce outage durations, and extend the effective service life of the HSIVs. In addition, the proposed increase in the HSIV leakage limit has potential to significantly reduce recurring valve leakages, and minimize the possibility of needless repair which can compromise plant safety.

Type 2 B2 8

NEDO 31858 Rev. 1 The proposed change involves implementing the reliable and effective main steam piping and condenser for HSIV leakage treatment. This treatment method is effective to treat HSlV leakage over an expanded operating range without exceeding the off site and control room dose limits. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by  ;

multiple leak tight barriers used in containment design for limiting fission product release to the en/ironment. Therefore, the proposed method is highly reliable for HSIV leakage treatment, furthermore, the risk to the public health and safety will be reduced with the implementation of-the proposed MSly leakage treatment method. The <

implementation will provide [ Plant Name) with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. from a safety perspective, the proposed changes result in an increase in protection to the public.

Plant specific supporting information and results of the radiological analysis that justify the proposed changes are included as Enclosure 2. As concluded in the supporting information, the increased MSIV allowable leak rate of [y] scfh will not adversely affect the performance of the primary containment isolation function. A plant-specific radiological analysis has been performed utilizing the main condenser as a treatment path for MSIV leakage. This radiological analysis demonstrates that the proposed changes result in an insignificant increase to the dose exposures previously calcu-lated for a design basis Loss-of-Coolant Accident (LOCA). The revised LOCA dose exposures remain well within the guidelines of 10CFR100 for the off-site doses and 10CfR50, Appendix A (General Design Criteria 19) for the control room doses.

l Upon approval of the proposed Technical Specification changes (Corporation) will institute into the HSIV maintenance and test program, the I

i l

Type 2 82-9 l

NED0-31858 Rev. I requirement that any MSly exceeding the propored (y] scfh limit, will be repaired and re-tested to meet a leakage rate of less than or equal to [x]

scfh. This will assure continuation of high quality repair and refurbishment efforts to improve the overall performance and reliability of the HSIVs. '

The proposed change will require a minor plant design change to allow specified main steam drain valves to be opened even if off-site power is not available. Applicable normal plant operating procedures, and emergency ,

operating procedures shall be reviewed and revised accordingly.*

Furthermore, (Corporation) will incorp . ate the applicable leakage treatment methods, consistent with GE document NED0 30324 " Potential Operator Actions to Control MSIV Leakage," into the Operational Procedures and Emergency Operational Procedures at (Plant Name).

(Corporation) will also perform a verificacion of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Sectica 6.7 of NEDO 31858 Rev.1, to provirie easonable assurance of the structural integrity of these components.

Pursuant to 10CFR50.92, an analysis which demonstrates that the proposed changes do not involve a significant hazards consideration, is included as Enclosure 3. The proposed changes have been reviewed in accordance with Section 6.5 of the Technical Specifications. The proposed changes will not authorize any change in the types of effluents or in the authorized power level of the facility. Enclosures 4 and 5 are applications for exemption to Appendix J of 10CFR50 and Appendix A of 10CFR100, respectively.

Plant-specific 1

l Type 2 B2-10

NEDO 31858 Rev. 1 ,

WHEREFORE, Applicant respectfully requests that Appendix A to the Facility Operating License No. [xxx) thereto be amended in the form attached hereto as Enclosure 1.

[ CORPORATION)

W By:

Vice President subscribed and sworn to before me on this day of .

Type 2 B2 Il

NED0 31858 Rev. 1 i

11CLD1VELl.

((0RPORATIONJ Li[ENSE NO. Ixxxl DOCKET NO. lxxx1  !

PROPOSED CHANGES TO TECHNICAL SPEClfl_ Call 0)H Replace the following pages with the attached revised page(s)*. These pages l

are provided in its entirety with marginal marking to indicate the changes.

l l

1. Page 3/4 6-2

2. Page 3/4 6-3 l i

?

The pages identified here are based on the BWR Standard Technical ,

Specifications for BWR/5 (NUREG-0123). Each utility should revise these pages as' applicable to their individual Technical Specifications.

l l

Type,2 82-12

.. . . _ ._ . .a. _ _ _ _ . _ : 2_ _ . . . _ . _ . _ . . .

! NEDD-31858 Rev. 1 l

l DiCLQSMRL2 100RPORAT_lQfil LILUISLliQm...Ixxx1 DOCKE1 NO. lxxxl SUPPORTING INFORMATION AND ANALYSES

1.0 INTRODUCTION

AND

SUMMARY

OF.RESULTS The proposed Technical Specification amenduent involves an increase in the allowable leakage rate from [x] scfh to (y) scfh per main steam line, and exemption of the MSIV leakages from the Type A and Type C acceptance criteria specified in Appeadix J of 10CFR50. In addition it is requested that downstream main stehai piping and condenner be exempted from the seismic requirements specified in Appendix A to 10CFR100.

Section 2.0 of this Enclosure provides a summary of background information; Section 3.0 discusses the justifications for the proposed changest Section 4.0 provides a summary of the plant-specific radiological dose assessment, and Section 5.0 summarizes the potential benefits for a Technical Specification MSIV allowable leak rate of [y]

scfh.

The BWROG report, NED0 31858 Rev. 1, "BWROG Report for increasing MSIV Leak Rate Limits and Elimination of Leakage Control Systems," November 1991, provides the justification for increasing MSIV leakage limits.

With concurrence from the valve manufacturers, this report concludes that HSIV leakage rates up to 200 scfh are not an indication of substantial mechanical defects in the valve which would challenge the isolation capability of the valve to fulfill its safety function. Therefore, the proposed increase in the allowable leakage rate to [y] scfh for the MSIVs will not inhibit the isolation capauility of the valve.

Type 2 B2-13

NE00 31858 Rev. 1 l

The BWROG has evaluated several methods and has recommended the isolated cone nser method for MSIV leakage treatment. The isolated condenser metL 1 takes advantage of the large volume in the main steam lines and )

the condenser to hold up the release of fission products leaking from the l closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser. [ Corporation)proposestoincorporatetheisolated condenser for MSly leakage treatment.

The BWROG has evaluated the availability of main steam system piping and condenser treatment pathways for processing MSIV leakage. The BWROG have reviewed the potential combinations of Loss Of-Coolant Accidents (LOCAs) and seismic events of interest:

(1) LOCA WITHOUT NEAR COINCIDENT SEISMIC EVENT. For this occurrence, the pressure in the piping system downstream of the HSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the flow path through the main steam system piping to the condenser is assured.

(2) SEISMIC EVENT WITHOUT NEAR COINCIDENT LOCA. Without a LOCA and the potential associated core degradation, the radioactivity transported with MSly leakage is of no radiological significance.

(3) LOCA WITH NEAR COINCIDENT SEISMIC EVERI. For this occurrence (also assuming significant core damage) the consequences are of interest because a seismic induced failure in the main steam or condenser system could allow MSIV leakage to bypass the

- treatment pathway. It has been previously well documented that j the probability of a near coincident LOCA and seismic event is l extremely small (design basis earthquake probabili'" is approximately 0.001 per reactor per year; core L

Type 2 B2 14 1

- - . ~ - - . - . .

j NED0 31858 Rev. 1 I

melt probability is plant specific and typically ranges from 0.00001 to 0.0001 per reattor per year), it is also noted that a LOCA does not induce a seismic event, and that a seismic event has a very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10 7 per reactor per year).

Considering that the probability of a near coincident LOCA and seismic event is much smaller than other plant-safety risks (less than 1 x 10-7 per reactor per year for coincident events, less than 5 x 10 7 per reactor per year for seismic induced LOCA), the concern for main steam piping or condenser damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design basis seismic events.

ANSI-831.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin. In addition, specific seismic design provisions have been incorporated into some newer BWR main steam and condenser systems.

In order to further justify the capability of the main steam system piping and condenser treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes). The study summarizes data on the performance of main steam piping and condensers in past strong-motion earthquakes and compares these piping and condensers with those in typical U.S. GE Mark I, II, and !!! nuclear plants. This limited earthquake experience data and similarity comparisons are Type 2 B2 15

NE00 31858 Rev. I then used to further strengthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in a design basis earthquake in conjunction with a LOCA occurring just prior to or after the seismic event.

The earthquake experience data are derived from an extensive database on the performance of industrial facilities and over 100 power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world from 1934 to the present.

The piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they are not designed to resist earthquakes. This is a common conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipment such as motor control centers, pumps, valves, structures, and so forth. That is, with limited exceptions, normal industrial construction and equipment typically have substantial inherent seismic ruggedness, even when they are not designed for earthquakes. No failures of main steem piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.

Comparisons of piping and condenser design ii example GE Mark I, 11, and til plants with those in the earthquake experience database reveal the GE plant designs are similar to or more rugged than those that exhibited good earthquake performance, in addition, (Corporation) will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NEDO-31858 Rev.1, to provide reasonable assurance of the structural integrity of these components. The BWROG concludes that (1) the Type 2 B2-16

NE00 31. > Rev. I possibility of significant failure in GE BWR main steam piping or condensers in the event of an erstern U.S. design basis earthquake is highly unlikely and (2) any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

The design basis LOCA has been re-analyzed for radiological impacts utilizing the isolated condenser method for MSlv leakage treatment. The analysis demonstrates that a maximum MSly leakage rate of [y] scfh per main steam line results in an acceptable increase to the dose exposures calculated for a design basis LOCA. In addition, the analysis demonstrates that MSIV leakage ratt:s of approximately [z] scfh per main steam line will not result in dose exposures in excess of the regulatory limits.

2.0 MCKGROUND The safety function of the MSIVs is to isolate the reactor system in the event of a LOCA or other events requiring containment isolation. The design of the MSJVs and its isolation requirements are described in Sections [5.4.S*]and[6.2.4*)oftheFinalSafetyAnalysisReport (FSAR). {The dooi>1e leakage rate from MSIVs is included in the LOCA radiologvM mf 0 6is evaluated in Section [15.6.5*) of the FSAR.)**

Operating experience indicates that MSIVs frequently exceed the Technical Specification allowable leak rates. Some of these valves repeatedly fail the local leak rate tests despite frequent disassembly and refurbishment.

As a result of increasing MSIV leakages and the potential consequences Plant-specific

• • Plant-specific. In some BWRs, only the containment design leak rate is included in the LOCA radiological analysis.

Type 2 B2-17

NEDO 31858 Rev. I following a LOCA, the Nuclear Regulatory Commission prioritized GLneric Issue C 8 "HS!V Leakage and LCS failure" as a high priority item in  ;

January 1983. This issue was closed in 1990.

The BWROG formed a MSly Leakage Committee in 1982 to adoress the increasing MSly leakage rates, and a follow on MSIV Leakage Closure Committee in 1986 to address alternate actions to resolve on going, but l 1ess severe H51V problems. The HSIV Leakage Committee identified contributors which cause llSIVs to fail the leak rate tests by large margins, developed recommendations to minimize ieakages, evaluated alternates for HSly leakage treatments, and compiled recent history of HSIV leakages and LCS operating experience.

3.0 JilSilf,l[ATIONS FOR THE PROPOSED CHANGL (Corporation) proposes to increase the Technical Specification allowable leakage rate'for the HSIVs from [x] scfh to [y] scfh per main steam line.

The current Technical Specification MSly leak rate limit is extremely limiting and routinely requires unnecessary repair and re-test of the HSIVs. This significantly impacts-the maintenance work load during plant outages and often contributes to outage extensions. (The outage planners at (plant station name) typically schedule several days of contingency to repair and re-test the HSIVs)* In addition, the needless dose exposures to maintenance personnel are inconsistent with As low As Reasonably Achievable (ALARA) principles. (There have been many Licensee Event Report written for HSIVs failing to meet the current it:chnical Specificationlimit.)*

Plant Unique Type 2 B2-18

NED0 31858 Rev. 1 (from a safety perspective, calculations using standard conservative assumptions for considering the off site consequences of a postulated design basis LOCA ccnfirm that off-site and control rocm doses will be within the regulatory guidelines for the allowable HSly leakage rate.

This calculation is described in Section 15.6.5 for the FSAR. However, if HSIV leakages are only moderately higher than the allowable limit, the calculated doses will exceed the regulatory guidelines.)*

MSIV's failure to meet the current Technical Specification limit have been documented in response to surveys conducted by the Nuclear Raulatory Commission during the early 1980 and by the BWROG during the middle and late 1980s. As high as 50% cf the total "as found" HSIV local leak rate tests were reported in the early NRC survey to exceed the leakage rate limit.

The BWROG has studied the issues regarding MSly leakage rates, their causes, and available alternatives. The results of the BWROG study are provided in NED0-31858 Rev. I and are also summarized in NUREG ll69. In response to Generic issue C 8, the BWROG has recommended corrective actions and maintenance practices to reduce the HSly leakage rates.

A recent survey conducted by the BWROG of MSIV leakage tests performed between 1904 and 1988 indicates that the implementation of industry and BWROG actions has been effective in reducing the leakage rates, and, in particular, a reduction in the number of valves which experience substantial high leakage rates. The Jurvey concludes that about 23% of the total "as found" HSly leakages still exceed the limit of 11.5 scfh and about 10% exceed 100 scfh.

Plant-specific, in some BWRs, only the containment design leak rate is included in the LOCA radiological analysis.

Type 2 B2-19

l NE00-31859 Rev. 1 The MSIV-leakage performance at (piant station name) is .....[to be provided by the Utility). The leakage performance at [ plant station name) is consistent with the recent survey by the BWROG.

Despite the recent improvement in leakage perfnrmance, MSIV leakage rates still frequently exceed the current Technical Specification limit and the cafety and maintenance probie related to high MSly leaktge rates, although less severe, remain as a significant issue.

Furthermore, based on the extensive evaluation of valve leakage data, the BWROG has found that disassembling and refurbishing the MSIVs to meet very low leakage limits frequently contribute to repeating failure. By not having to disassemble the valves and refurbish them for minor g leakage, the utility may avoid introducing one of the root causes of Q recurring valve leakage problems that may lead to later leak test failurcs and the possibility of compromising plant safety.

(The current Technical Specificat-;on allawable leakage rate is established by excessively conservative LOCA radiol)gical analysis as described in Section 15.6.5 of the FSAR.)* The valve's physical size and e operability characteristics (large size and fast atting), and existing turbine building equipment were not considered at the time the leakage limit was t tablished. Based on the in-depth evaluation of MSIV leakages, the BWROG has concluded the HSIV leakcge rates up to 500 scfh are not an inmation of substantial mechanical defects in the valve which would challenge the isolation capability of the valve to fulfill

}';

it: safety function. Furthermore, valve manufacturers have stated that leakage rates up to 200 scfh can occur without having a major valve defect. Therefore, the proposed increase from [x] to [y] scin per main steam line w"i not inhibit the MSIV's performance of the isolation Plant-specific. In some BWRs, only the containment design leak rate is included in the LOCA radiological analysis, lype 2 B2-20 j

NE00-31858 Rev. I function and will not c,mpromise the safety of (plant station name].

This proposed increase provides a more realistic, but still conservative, limit for the HSIVs. Based on the BWROG study, the proposed increase in the allowable leak rate will increase the chance for a successful local leak rate test to greater than 90%, up from the 77% success rate at the current typical limit of 11.5 scfh. This increase in successful local leak rate testing will significantly reduce MSiv maintenance costs, reduce dose exposure to maintenance personnel, reduce outage durations, extend effective service life of the MSIVs, and minimize the potential for outage extensions at [ plant station name).

[ Corporation) proposes to implement the reliable and effective main steam piping and condenser for MSIV leakage treatment. T!is treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off-site and control room dose limits. Except for the requirement to establish a proper flow path from the HSIVs to the condenser, the proposed method is passive and does nrJ require any logic control and interlocks. The method is consistent with the philosophy of

. protection by multiple leak-tight barriers used in centainment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment. The implementation will provide [ Plant Name] with a capability to process "SIV leakage, and will also provide a uniform basis for establishing a

-plant-specific MSIV leakage rate limit. From a safety perspective, the l proposed changes result in an increase in protection to the public.

l Furthermore, [ Corporation] will incorporate the applicable leakage treatment methods. consistent with GE document NED0-30324 " Potential l Operator Actions to Control MSIV Leakage",. into the Operational l Procedures and. Emergency Operational Procedures.

i Type 2 B2-21

ncv0-31858 Rev. 1 4,0 ANALYSIS OF MSIV LEAKAGE CONTRIBUTION TO RADIOLOGICAL DOSE CALCULATIONS 4.1 Selection of leakane Trtatment Method The BWROG hhs evaluated several MSIV leakage treatment methods and has recommended the isolated condenser for MSIV leakage treatment. This leakage <.reatment method takes advantage of the large volume in the isolated main condenser to hold up the release of fission products leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser.

As previously discussed in Section 1.0, the BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage. The BWROG has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for other plar., safety risks. The BWROG has also determined that main steam piping and condenser designs are extremely rugged, and that the ANSI-B31.1 design requirements typically used for nuclear plant system design contain a good deal of margin. In order to further justify the capability of the main steam piping and condenser treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes). This study concluded the possibility of a failure which could cause a loss of steam or condensate in BWR main steam piping or condensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

This conclusion is consistent with NUREG/CR 4407, " Pipe Break Frequency Estimation for Nuclear Power Plants", dated May 1987, which reported no observed failures in the main steem piping over 313 years of reactor operating years.

Therefore, the isolated condenser MSIV leakage treatment path at (plant name) is considered appropriate for the reduction of radiclogical Type 2 B2-22

.e . . - . - - . - - . - - . , _ . ~ . . - - - - _ . - ,

-NED0 31858 Rev.L1 .

consequences of a design basis LOCA.

4.2._ Radiolooical Analysis and Result

• The radiological dose' methodology has been developed by General Electr'c

.for the BWROG.

-_The-radiological- analysi_s calculates _ the effects of the proposed allowable-MSIV leak rate in terms of control room and off-site doses.

The revised LOCA doses are the sum of the LOCA' doses _(as described in Section xxx of the FSAR) and the calculated MSIV leakage doses. [This method of calculating the revised dose exposures ~ is very conservative since the LOCA-doses already include the dose contribution from MSIVs at

- the maximum leakage rate _ permitted in the current Tecnnical Specifications.]* ,

-Table 1** shows :thi calculated dose exposures from the. BWROG radiological' '

- analysis for-[ plant name). Regulatory limits and calculated doses' from

-LOCA radiolsgical analysis are also included'in Table 1 for comparison-- ,

purposen -This analysis' demonstrates that a-MSIV. leakage rate of 200_sefh-

- per main _ steam ;line results-in an acceptable . increase- to..the dose

' exposures previously calcuiated for the control room, EAD, and the'LPZ.

zThe revised LOCA doses remain well:within the guidelines of 10CFR100 for off-site doses-and 10CFR50, Appendix A,:(General Design Criteria 19) for I the control room' doses. - Furthermore, the calculation shows that- MSIV : -

leakage rates up to approximately [z] scfh per steam line would.not- '

. exceed;the regulatory limits. Therefore, the proposed method provides a substantial safety margin- for mitigating-the radiological consequences- of:

• 4 LMSIV leakage beyond the' proposed . Technical: Specification leak rate limit ofi[y] scfh..

o Plant-specific. :In some_ BWRs, only the containment- design leak rate is

' included:in the original <LOCA radiological analysis. For applicable

BWRs, the' utility lmay elect to-replace the LOCA MSIV dose contribution

with the-new doses using the leakage. treatment method.

The following discussion;on doses are based on an example plant (Hope-Creek)

L o

I  : Type 2 B2-23

. - - - _ _ - . .. _ _ _ _ _ _ __ _ - ~ _ . _ . , _ - _ _ _

lNEDO-31858Rev.-1

-l Table l CONTRIBUTION T0-THE LOCA DOSE EXPOSUdES FOR A MAXIMUM MSIV >

LEAK RATE OF 200 scfh HOPE CREEK GENERATING STATION r

Whole Body Thyroid Beta frem) (rem) frem)

': Exclusion Area. A) 10CFR'.100 Limit. 25 300

• Boundary.

-(2-Hour),

. B)-Previous-Calculated 0.6 76.7 -

Doses **~

C) Contribution from 0.1 2.6-

_ MSIVs at 200 scfh D) New Calculated Doses 0.7 79.3 Low Population. .- A): .10CFR -100 Limit 25 300 *

• Zone ~(30-Day)

. B) Previous Calculated 0.08 7.7 4

Doses **-

, _C) Contribution From 0.34 57.5 MSIV at 200 scfh 4

% .0) LNew Calculated Doses !

' O.42'- 65'. 2 -

Control Room iA)'GDC'-19: - 5;  :

30 30/75***

(30-Day)-t . -

B)-l 4. Previous Calculated. 0.04- 0.26- .911

' Doses **

C) ' Contribution from 0.10 2.71 1.64 3

~ MSIVs at'200 scfh-

, D)-New Calculated Doses. - 0.14 - 2.97 2.55 <

.J -No limit specified. '

~ **'

- . FSAR Section.15.6.5.5 andI6.4 (includes MSIV leak rate at"a' total of-45 scfh for the:first 20 minutes; control- room dose assumes 100% per day. '

reactor building inleakage). _. .

• • • ' 75 if prior commitment-has-been made to use protective clothing.-

IType 24 B2-24

• , - . , ,e

NE00-31858 Rev. 1 5.0 BENEFITS FOR THE PJ0 POSED CHANGES As discussed in NED0 31858 Rev. 1, recent MSIV leakage performance has significaatly improved since the early BWROG survey in 1984 and the PAC survey in the early 1980s. Despite the recent improvement, MSIV leakage rates exceeding the current Technical Specification limits still frequently. occur. The BWROG evaluation of the recent MSIV leakage performance concludes that the proposed change will improve the chance l for a successful local leak rate test to greater than 90%, up from the )

77% success rate at the current Technical Specification limit of 11.5 l scfh.

Specifically, MSIV leakage experience at [ plant station name] are (to be provided by Utility on plant-specific leakage performance and any problems which may impact critical path, outage extensions, etc.). )

i Furthermore, the risk to the public health and safety will be reduced with the implementation of the proposed MSIV leakage treatment method.

The implementation will provide [ Plant Name] with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to the public.

Type 2- B2-25

NED0-31858 Rev, 1 ,

i ENCLOSURE 3

[CORPORAT10N1 LICENSE NO. ixxxl QQCKET NO. Ixxxl NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS (Corporation] proposes an amendment to the Technical Specifications as follows:

1) Revise Section 3.6.1.2* to permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs) from the current [x]

standard cubic feet per hour (scfh) to [y] scfh per main steam line.

2) Revise Section 3.6.1.2* to add a footnote stating that the MSIV leakage is exemoted from the acceptance criteria of Appendix J of 10CFR50.

The proposed amendment to add a footnote to Section 3.6.1.2* clarifies that MSIV leakages are exempted from the overall integrated containment leakage rate limit and from the combined local leak rate limit as specifi2d in the test acceptance criteria in Appendix J of 10CFR50. As defined in the Bases Section' 3/4.6.1.2* of tne Technical Specifications, the purpose of these requirements is to ensure that the overall integrated containment leakage or the combined leakages from individual containment penetrations will not exceed the designed containment leak rate assumed in the design-basis Loss-Of-Coolant Accident (LOCA) radiological analysis. Since an allowable leak rate is specifically allocated for the MSIVs in the Technical Specifications, and that the radiological analysis has been revised to analyze MSIV leakage path separately from those of the containment leakage rates, the proposed exemption Plant-specific Type 2 82-26

NED0-31858 Rev. 1 is, therefore, appropriate and justified. This proposed change does not exempt the MSIVs from the test schedules as required in the Technical

' Specifications and 10CFR50 Appendix J.

This application also provides a detailed justification for exempting the MSIV leakages from the acceptance criteria of 10CFR50 Appendix J, and demonstrates that the proposed exemption will not present an undue risk to the public health and safety. Therefore, based on the above consideration, the proposed amendment to add the footnote to Section 3.6.1.2* is considered an administrative change.

Pursuant to 10CFR50.92, the proposed amendment involves no significant hazards considerations.

The operation of IPlant Station Namel. in accordance with the cropngd amendment will not involve a sianificant increase in the orobability or consecuences of an acciaent oreviously evaluated.

The proposed amendment to Section 3.6.1.2 does not involve a change to structures, components, or systems that would affect the probability of.an accident previously evaluated in the Final Safety Analysis Report (FSAR).

Since MSIV leakage is 1icluded in the radiological analysis for the design-basis LOCA as described in Section 15.6.5 of the FSAR, the proposed amendments will not affect the precursors of other analyzed accidents. The

proposed amendments result in acceptable radiological consequences of the design-basis LOCA previously evaluated in Section 15.6.5 of the FSAR.

The (plant station name] has an inherent MSIV leakage treatment capability.

[ Corporation) proposes to use the main steam lines and condenser for MSIV L leakage. treatment and will incorporate this method in the Operational Procedures and Emergency Operational Procedures.

Plant-specific Type 2 B2-27

NED0-31858 Rev. 1 The BWROG has evaluated the availability of main steam system piping and condenser treatment pathways for processing MSIV leakage, and has determined that the probsoility of a near coincident LOCA and a seismic event is much smaller than for other plant safety risks. The BWROG has also determined that main steam piping and condenser designs are extremely rugged, and that the ANSI-B31.1 design requirements typically used for nuclear plant system design contain a good deal of margin.

In order to further justify the capability of the main steam piping and condenser treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non seismically designed piping and condensers (in past earthquakes). This study concluded that the possibility of a failure which could cause a loss of steam or condensate in BWR main steam piping or condensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

[ Corporation) will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0-31858 Rev.1, to provide reasonable assurance of the structural integrity of these components.

A plant-specific radiological analysis has been parformed to assess the effects of the proposed increase to the allowable MSIV leak rate in terms of control room and off-site-doses following a postulated design basis LOCA.

This analysis utilizes the hold-up volumes of the main steam piping and condenser for the MSIV leakages. As discussed earlier, there is reasonable assurance that the main steam piping and condenser will remain intact

-following a design basis earthquake. The radiological analysis uses standard conservative. assumptions for the release of source terms consistent with Regulatory Guide 1.3 Revision 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of a loss-Of-Coolant Accident for Boiling

-Water Reactors",. dated April 1974.

i l

Type 2 B2-28

NE00-31858 Rev. 1 The analysis demonstrates that dose contributions from the proposed MSIV leakage rate limit of [y] scfh result in an acceptable increase to the LOCA doses previously evaluated against the regulatory guidelines for the off-site doses and control room doses as contained in 10CFR100 and 10CFR50, Appendix A (General Design Criteria 19), respectively. The LOCA doses previously evaluated are discussed in Section (15.6.5]* of the FSAR. The revised LOCA doses are the sum of the LOCA doses previously evaluated in th FSAR and the additional MSIV doses calculated using the proposed treatment method. (This method of calculating the i sised doses is very conservative since the LOCA doses previous evaluated already included dose contributions from MSIV at the maximum leakage rate permitted in the current Technical Specifications.)**

The attached table shows the previous calculated doses and the new calculated doses.

The whole body dose at the low Population Zone (LPZ) and the control room is increased from 0.08 to 0.42 rem *** and from 0.04 to 0.14 rem, respectively.

These increases are acceptable because the revised doses are well within the Regulatory guidelines (0.42 versus 25 rem at the LPZ, and 0.14 versus 5 rem at

.the control room). The associated whole body dose at the exclusion area boundary (EAB) increased insignificantly from 0.6 to 0.7 rem.

The thyroid dose at the LPZ increased from 7.7 to 65.2 rem. This increase is acceptable because the revised dose of 65.2 rem is significantly less than the regulatory guideline (300 rem). The EAB thyroid dose increased slightly from 76.7 to 79.3 rem, whereas the control room thyroid dose increased from 0.26 to 2.97 rem.- The increase in control room thyroid dose is acceptable because the rf rised dose remains a small fraction (9.9%) of the liniit. The control room bs2 dose is increased from 0.91 to 2.55 rem, which remains insignificant relative to the regulatory guideline of 30 rem.

• Plant-specific

• • Plant-specific. In some BWRs, only the containment design leak rate is included in the previous LOCA radiological analysis. For applicable BWRs, the original LOCA dose from MSly leakage may be replaced with revised dose.

• • • All doses and comparisons shown are Hope Creek specific.

Type 2 B2-29

-_m . _ _ _> . . , __

l NEDO-31858 Rev. 1 CONT!!IBUTION 10'THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV LEAK RATE OF 200 scfh HOPE CREEK GENERATING STATION Whole Body Thyroid Beta (rem) (rem) frem)

Exclusion Area' A) 10CFR 100 Limit 25 300 Boundary:

(2 Hour): B) Previous Calculated 0.6 76.7

' Doses **

~

C) Contribution From 0.1 2.6 MSIVs at 200'sefh LD)f ,New Calculated-Dases 0.7 79.3 ,

. Low Population- 300

• A)- 10CFR 100 Limit 25  ;

Zone:(30-Day)'

, ' B) Previous Calculated 0 0B

. 7.7 Doses **

C)- Contribution From 0.34 57.5

._MSIV at-200 scfh a

D) New Calculated Doses 0.42 65.2

+ Control-' Room : 1A):GDC-191 5- 30 -30/75***

l(30-Day)_

B)' Previous; Calculated L _ 0.04- 0.26 .91

Doses **:

C)

Contribution From. 0.10 - 2. 71.: 1.641 o 'MSIVs at 200 sefh-D) New Calculated Dosesi 0.14' 2.97 2.55 -

c 6- _

..:* :No limit =specified..

FSAR-Section.15.6.5.5 and '6.4'(includes MSIV: leak ~ rate atia: total of-

-45 scfh forithe first 20-minutes; control room dose assumes-100% per day -

P: -

l reactor building inleakage). ..

• • • l 75 if- prior- commitment- has been made _ to use protective clothing.

l.

lJ -

li

- Type -'2 - B2-30

NED0-31858 Rev. 1 It is important to note that the resulting doses are dominated by the organic iodine frar.tions which occur because of the utltraconservative source term asstnptions used in this analysis. For 200 scfh per steam line, more than 85%

of the off-site iodine and control room doses are due to organic iodine from the RG 1.3 source term and organic iodine converted from the elemental iodine deposited in main steam piping systems, If the actual iodine composition from the fuel release (cesium iodine) is used in the calculations, essentially all of this organic iodine dose would be eliminated.

In summary, the proposed changes result in an acceptable increase to the radiological consequences of a LOCA previously evaluated in the FSAR. The

revised LOCA doses are well within the regulatory guidelines. Although the revised LOCA doses are slightly higher for low MSIV leakage rates, the effectiveness of the proposed method even for leakage rates greater than the proposed increased MSIV allowable leak rate, ensures that off-site and control room doses are not exceeded.

The proposed change to add a footnote to Section 3.6.1.2 is administrative in nature and has_no effect on any accident.

The operation of IPlant Station Namel. in accordance with the L onosed amendment, will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed amendment to Section 3.6.1.2 does not create the possibility for a new or different kind of accident from any accident previously evaluated.

The BWROG evaluated MSIV leakage performance and concluded that MSIV leakage rates up to 200 scfh will not inhibit the capability and isolation performance of the valve to isolate the primary containment. There is no new modification E which could impact the MSIV operability. The LOCA has been analyzed using the main steam piping and condenser as a treatment method to process MSIV leakage  ;

l l

Type 2 B2-31 l

NEDO-31858 Rev. I

~at the proposed maximum rate of [y] scfh. Therefore, the proposed change does not create any new or different kind of accident from any accident previously evaluated in the FSAR.

The proposed change to add a footnote to Section 3.6.1.2* is administrative in nature, and does not create a possibility of a new or different kind of accident from any accident previously evaluated in Chapter 15* of the FSAR.

The operation of IPlant Station Namel. in accordance with the proposed amendment will not involve a sianificant reduction in the maroin of safety.

The proposed amendment to Section 3.6.1.2 does not involve a significant reduction in the margin of safety. [As discussed in the Bases of the Technical Specification 3/4.6.1.2, the allowable leak rate limit specified for the MSIVs-is used to quantify a maximum amount of bypass leakage assumed in the LOCA radiological analysis.]* Results of the analysis are evaluated against the dose guidelines contained in 10CFR100 for the off-site doses and 10CFR50, Appendix A (General Design Criteria 19)- for the control room doses.

Therefore, the margin of safety is considered to be the difference between the calculated doses and the guidelines as contained in 10CFR100 and GDC 19.

Results of the radiological analysis demonstrate that the proposed change does not involve a significant reduction in the margin of safety. The chole body doses, in terms of margin of safety, are insignificant 1y reduced by 1.4% at the LPZ, 2.0% in'the control l room, and 0.4% at the EAB. The thyroid dose margin of safety is reduced by 19.1% at the LPZ, 9.0% in the control room, and 0.9% at the EAB. The beta dose is insignificantly . reduced by 5.5% in the control room in terms of margin of safety. The margins of safety are not significantly adversely affected because the absolute margins of safety remain l- well below the guidelines (lowest whole body margin of safety is 97.2% at the EAB, lowest thyroid margin-of safety is 73.6% at EAB).

l=

l Plant-specific.

Type 2 B2-32 l

NE00-31858 Rev. 1 in terms of MSIV leakage treatment capability, the overall margin of safety is increased. [ Corporation] proposes to implement the reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off-site and control room dose limits. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed methed is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment. The implementation will provide [ Plant Name] with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to the public.

Furthermore, the revised LOCA doses remain well within the regulatory limits for the site and control room. The calculation shows that MSIV leakage rates up to [z] scfh per steam line would not exceed the regulatory limits.

Therefore, the proposed method provides a substantial safety margin for mitigating the radiological consequences of MSIV leakage beyond the proposed Technical Specification leak rate limit of [y] scfh.

The proposed change to add a footnote to Section 3.6.1.2 is administrative in nature.and does not affect the margin of safety.

Therefore, the proposed amendment to the Technical Specifications does not involve a significant hazards consideration.

Type 2 B2-33

NE00 31858 Rev. 1 ENCLOSURE 4

((DRPORATION1 L101NSE NO. IXXXI QQCKEl NO. fXXX1 APPLICATION FOR EXEP.PTION TO APPENDIX J OF 10CFR50 Pursuant to Section 50.12(a) of the Regulations of the Nuclear Regulatory Commission, (Corporation), holder of facility Operating License No. [xxx],

hereby requests specific exemptions to Appendix J of 10CFR Part 50 " Primary Reactor Containment Leakage Testing For Water-Cooled Power Reactors".

Specifically, [ Corporation) requests that leakages from the main steam isolation valves (MSIVs) be exempted from the acceptance criteria for:

(1) the overall integrated leak rate test (Type A), as defined in the regulations of 10CFR50, Appendix J, Pt.*agraphs III,A.5(b)(1) and e Ill.A.5(b)(2),and (2) the combined local leak rate test (Type B and C), as defined in the regulations of 10CFR50, Appendix J, Paragraph Ill.C.3.

The purpose of the test acceptance criteria is to ensure that the measured leak rate from the containment volume will not exceed the designed containment leak rato assumed in the safety analysis for a postulated design basis Loss-Of-Coolant Accident (LOCA).

In conjunction with this application for exemption request, (Corporation) han transmitted to the Nuclear Regulatory Commission an application for a license Technical Specificatio- sections are based on the BWR Standard Technical Specification for BWR/5 (NUREG-0123). Each utility shculd review the sections applicable to their Technical Specifications.

Type 2 S2-34 i

1

-m

NEDO 31858 Rev. I amendment pursuant to 10CFR50.90. This license amendment involves a proposed change to Section 3.6.1.2* of the Technical Specificaticns to permit an incre u4 in the allowable leak rate for the MSIVs from the current [x]

standard cubic feet per hour (scfh) to [y] scfh per main steam line. The safety analysis has been revised to assess the radiological effects of MSIV leakage following a postulated design basis LOCA. [ Corporation) has demonstrated that the proposed change does not involve a significant hazards consideration.

This proposed exemption is a result of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic issue C-8 "MSIV ,

leakage and LCS Failure".

The following discussion provides a detailed justification and evaluation of the proposed exemption. The proposed exemption is found to be authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security. Furthermore, special circumstances are present that warrant the granting of this exemption.

The proposed exemption will not cause additional operational activities that may significantly affect the environment. It does not result in a significant '

increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-0perating License Stage, result in a significant change in effluents or power levels, or affect-any matter not previously reviewed by the Nuclear Regulatory Commission that-may have a significant adverse environmental-impact.

Therefore, pursuant to 10CFR 50.12(a), [ Corporation] hereby requests an exemption for (plant station name] for MSIV leakages from the acceptance test criteria specified in Appendix J of 10CFR50.

1 Plant-specific il Type 2 B2-35

l NED0 31858 Rev. 1 A. Justificitt_ inn The regulation of 10CFR50, Appendix J, Paragraphs Ill.A.5(b)(1) and Ill.A.5(b)(2) requires the overall integrated leakage rate, as measured during containment pressure tests (Type A), to meet the acceptance criterion of less than or equal to 0.75 of the maximum allowable containment leak rate.

Paragraph Ill.C.3 of the regulation requires the combined leakage rate for all penetrations and isolation valves, as measured during local leak rate tests (Type B and Type C), to meet the acceptance criterion of less than or equal to 0.60 of the maximum allowable containment leak rate. Paragraphs Ill.C.3(a) and III.C.3(b) define the acceptance criteria for the exclusion of containment isolation valves from the acceptance criterion for Type B and C tests.

As described in the Bases Sections B 3/4.6.1.2* of the Technical Specifications, the limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure. As an added conservatism, the measured leak rate is further limited to less than or equal to 0.75 of the maximum allowable leak rate during the performance .,f the periodic tests to account for possible degradation of the containment leakage barrier between leakage tests.

The maximum containment leakage rate was included in the radiological analysi:;

of a postulated design basis LOCA as evaluated in Section 15.6.5* of the Final Safety Analysis Report (FSAR). The radiological analysis calculated the effect of the maximum leakage rate from the containment volume in terms of control room and off-site doses, which were evaluated against the dose guidelines of 10CFR50, Appendix A (General Design Criteria 19) and 10CFR100, respectivel). Leakages from the containment volume were contained in the reactor building (secondary containment), filtered by the Standby Gas Plant-specific information.

Type 2 B2-36

)

,~ .- --

NEDO 31858 Rev. 1 ,

Treatment' System, and released to the environment through the elevated release stack.* The maximum containment leakage-rate includes leakages through

. structures, all penetrations-identified as Type B, and all containment isolation valves identified as Type C. ,

i The safety analysis has been revised to account for the radiological effect from MSIV leakages and from those of other containment leakages following a postulated design basis LOCA, Unlike the treatment path for other containment-leakages, the treatment of MSIV leakages employs the main steam drain piping and the' condenser. Fission products are removed by plate-out and hold up in '

-..the relatively large_ volumes of the main steam piping and condenser.

i The- treatment- method for MSIV leakages-is recommend.d by the BWROG in support of the, resolution to Generic Issue C-8. The BWROG has evaluated the

-avai. lability' of main steam system piping and condenser alternate treatment

' pathways.for processing MSIV--leakage, and has determined that the_ prob' ability 1

-of a.near-coincident LOCA and a seismic event is much smaller than for other iplant safety risks. The BWROG has also determined that main steam piping and condenser designs are extremely-rugged,-and that the ANSI-831.1 design requirementsctypically used for nuclear plant' system design contain a good dea 1Lof margin.

Inorderto_-furtherjustifyithecapabilityofthemainsteam-pipingand condenser treatment pathway, the BWROG has' reviewed limited earthquakeL -

Lexperience data on the-performance of-non-seismically designed piping.and condensers-(in:past earthquakes), This study concluded the possibility of a failurelwhich couldicause;a loss.of steam or condensate-in- '

lBWR/6 plants may not have an elevated stack. Furthermore, . some plants

, .-may have " bypass: leakages"- defined in- their- radiological analysis. These

! are.the' leakages that-bypass.the reactor building and.' release to the- '

t environmental unfiltered. Each utility should correct the. text herein l' -consistent with their radiological analysis, p

I' i

le Type 2 B2-37

NE00-31858 Rev. 1 BWR main steart piping or condensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

Leakage from the MSIVs should not be included in the Type A acceptance criterion because the treatment path for MSly leakage is different from that of containment leakages. Potential leakage from the containment is contained in the reactor building (secondary containment), treated by the SGTS, and released via the main stack. MSIV leakage is contained, plated-out, and delayed in the main steam piping and the condenser, and released via the turbine building.

Furthermore, leakage from the MSIVs should not be included in the combined local leak rate test (Type B and Type C) acceptance criteria because a specific allowable leak rate has been allocated for the MSIVs in Section 3.6.1.2* of the Technical Specifications.

As discussed earlier, the basis for the containment leakage tests and the acceptance criteria is to ensure that the measured lea'.; rate will not exceed the maximum leak rate assumed in the safety analysis. The safety analysis for a design basis LOCA has been revised to include the maximum MSIV leak rate separately from the maximum containment leak rate. MSly leakages will be tested as part of the local leak rate test in accordance with the requirements in Section 3.4.6.1.2 of the Technical Specifications. This test ensures that the measured MSIV leak rate will not exceed the allowable leak rate assumed in the safety analysis.

There is sufficient conservatism in the allowable MSIV leak rate to account for possible degradation of the MSIV leakage barrier between leakage tests.

As discussed in the application for the license amendment, [ Corporation]

proposes a maximum allowable MSIV leak rate of [y] scfh per main steam line; Plant-specific information.

Type 2 B2-38

NE00-31858 Rev. 1 I

whereas, the analysis demonstrates that MSIV leakage rates up to approximately

[z] scfh per main steam line will not result in dose exposures in excess of the regulatory limits. Thus, a safety margin exists. Furthermore,

[ Corporation) will institute into the MSIV maintenance and test program, the requirement that any MSIV exceeding the proposed [y] scfh limit, will be repaired and re-tested to meet a leakage rate of less than or equal to (x) scfh. This will assure continuation of high quality repair and refurbishment efforts to improve the overall performance and reliability of the MSIVs.

Therefore, the proposed exemption from the acceptance criteria of 10CFR50, Appendix J will not defeat the underlying purpose of the regulation, and is consistent with the safety analysis.

(

B. Authorized By law The proposed exemption is consistent with Section 3.6.1.2 of the Standard Technical Specification (NUREG 0123). The reason for this exemption is provided in the Technical Specifications Baaes B 3/4.b.1.2. A review of the Technical Specifications for BWRs indicates that such an exemption has been granted to the following plants: Fermi 2, Hatch I & 2, Hope Creek, Limerick 1, Shoreham, LaSalle 1 and 2, Hanford, Clinton, Grand Gulf 1, Perry, Dresden 2 and 3, Monticello, Quad Cities 1 and 2, Brunswick 1 and 2 and Nine Mile Point 2.

Threfore, the proposed exemption is authorized by law.

C. No Undue Risk to Public Health and Safety The proposed exemption presents no undue risk to public health and safety.

The revised MSIV leakage rate has been incorporated in the radiological analysis for a postulated LOCA as an addition to the designed containment leak rate. The analysis demonstrates an acceptable increase to the dose exposures previously calculated for the control room and off-site. The revised LOCA Type 2 B2-39 1

l NE00-31858 Rev. I doses ' main well within the guidelines of 10CFR100 for off site doses and 10CFR50, Appendix A, (General Design Criteria 19) for the control room doses.

In addition, Section 3.6.1.2* of the Technical Specification has provided for allowable MSly leak rates, which assure that the MSIVs isolation function is not compromised. Finally, potential MSIV leakage is subjected to plate-out, and hold-up in the main steam piping and condenser, thus minimizing their effect on the total dose released. As discussed in Section F of this application, the proposed change will not adversely affect the conclusiens of the previously issued FES-OL. Therefore, the proposed exemption presents no undue risk to public health and safety.

Furthermore, the risk to the public health and safety will be reduced with the implementation of the proposed MSIV leakage treatment method. The implementation will provide [ Plant Name) with a capability to process MSlv leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to the public.

D. Consistent with Common Defense and Security With regard to the " common defense and security" standard, the grant of the requested exemption is consistent with the common defense and security of the United States. The Commission's Statement of Considerations in support of the exemption rule note with approval the explanation of this standard as set forth in Lona Island liahtina Comoany (Shoreham Nuclear Power Station, Unit 1), LBP-84-45, 20 NRC 1343, 1400 (October 29,1984). There, the term " common defense and security" refers principally to the safeguarding of special nuclear material, the absence of foreign control over the applicant, the protection of Restricted Data, and the availability of special nuclear material for defense needs. The granting of the requested exemption will not affect any of these matters and, thus, such grants are consistent with the common defense and security.

Type 2 B2-40

NED0 31838 Rev. 1 E. Eplclitl Circumstances Are Present Special circumstances are present which warrant issuance of this requested exemption. These special circumstances are discussed in accordance with the classification contained in 10CFR50.12(a)(2):

(ii) Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

The underlying purpose of the rule is to limit releases to within the off-site and control room dose guidelines of 10CFR100 and 10CFR50, Appendix A (GDC 19),

respectively. Compliance with Appendix J of 10CFR50 for Type A test acceptance criteria is not necessary to achieve the underlying purpose of the rule since HSIV leakage is not directed into the reactor primary containment.

Instead, the HSIV's leakage is directed through the main steam drain piping into the condenser. Since Type A tests are intended to measure the primary containment overall integrated leak rate (ILRT), the HSIV's leakage rate should not be included in the measurement of the ILRT. Compliance with Appendix J of 10CFR50 Type C test acceptance criteria is not necessary since a specific HSIV leak rate limit is already specified in Section 3.6.1.2* of the Technical Specifications.

The safety analysis tas been revised to assess the radiological consequences of MSIV leakage following a design basis LOCA. The analysis has demonstrated that the revised LOCA doses are well within the off-site and control room dose guidelines of 10CFR100 and GDC 19.

(iii) Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are signifd t .tly in excess of those incurred by others similarly situated.

Type 2 B2-41 l

NEDO-3185@ Rev. 1 Compliance with Appendix J of 10CFR50 Type A and Type C test acceptance criteria results in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted. The proposed increase in the MSIV allowable leak rate will not be possible if the MSIV leak rate results are included in the Type A and Type C test acceptance criteria.

Compliance requires unnecessary repair and re-testing of the MSIVs. This significantly imphets the maintenance work load during plant outagos and often contributes to outage extensions. The frequent MSIVs disassembly and refurbishing, which is required to meet the low leakage limits contributes to repeated failures.

Examples of these maintenance induced defects include machining-induced seat cracking, machining of guide ribs, excessive pilot valve seat machining, and mechanical defects induced by assembly and disassembly. By not having to disassemble the valves and refurbish them for minor leakage, [ plant name) avoids introducing one of the root causes of recurring leakage. Industrial experience suggests that, by attempting to correct non-existing or minimal defects in the valves, it is likely that some actual defects may be introduced that lead to later leak test failures.

In addition, the frequent maintenance work results in needless dose exposures to maintenance personnel. leading to additional eccnomical burdens, and are inconsistent with As low As Reasonably Achievable (ALARA) principles.

l l (iv) The exemption would result in henefit to the public health and l safety that compensates for any decrease in q fety that may result from the grant of the exemption.

[ Corporation) has transmitted to the NRC an application for a license amendment which involves a proposed change to the Technical Specifications to

increase the allowable MSIVs leak rate from [x] to [y] scfh. This application is partly based on the fact that the current limit is too restrictive, and Type 2' B2-42

NED0 31858 Rev. 1 l

l results in excessive MSly maintenance and repair, leading to additional MSIV failures, which in turn result in higher leakage. The proposed limit will benefit the public health and safety by reducing the potential for MSIVs failures, and thus keeping the MSIV leakage within the radiological analysis values.

[ Corporation] proposes to implement the reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off-site and control room dose limits. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Th9refore, the proposed method is highly reliable for MSiv leakage treatment. The implementation will provide [ Plant Name] with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to the public.

The exemption from Appendix J requirements for MSIV leakage rates is required so that [ plant name] can operate with the proposed Technical Specifications value of [y] scfh and with a capability to process MSIV leakage. This benefit will compensate for any decrease in safety that may result from the granting of the exemption.

Thus, special circumstances exist warranting the grant of the exemption.

F. Environmental Impact The proposed exemption has been analyzed and determined not to cause additional construction or operational activities which may significantly

(

Type 2 B2-43 l

NE00-31858 Rev. 1 affect the environment. It does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-0perating License Stage, result in a significant ct.ange in effluents or power levels, or affect any matter not previously reviewed by the Nuclear Regulatory Commission which may have a significant adverse '

environmental impact.

The proposed exemption does not alter the land use for the plant, any water uses or impacts on water quality, air or ambient ait' quality. The proposed action does not affect the ecology of the site and "icinity and does not affect the noise emitted by station. Terefore, the proposed exemption does not affect the analysis of environmental impacts described in the environmental report.

Type 2 B2-44

l NE00-31858 Rev. 1 L

ENCLOSVRE 5 iCORPORAt10N1 UEENSE NQ. Ixxxl DOCKET NO. ixxxl APPLICATION FOR EXEMPTION TO APPENDlX A 0F 10CFR100 (CORPORATION], holder of Facility Operating License No. NPF-[xxx], hereby requests an exemption of the downstream main steam piping and condenser from the seismic requirements specified in Appendix A of 10CFR100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants."

Specifically, (Corporation] proposes to employ probability analy.,is, existing design capabilities, seismic experience, and a plant specific seismic adequacy verification, as alternate methodology to the dynamic analysis or qualification test specified in Paragraph VI(a)(1) of 10CFR100 Appendix A, to provide reasonable assurance that the existing main steam piping and condenser will remain functional following a design basis accident coincident with a significant seismic event.

The exemption would allow the existing, non-seismically designed main steam piping and condenser to be used for mitigating the radio k gical consequences of MSIV leakage during the duration of a Design Basis Accident, such that the resulting doses are within the guidelines of 10CFR100.

[ Corporation] recognizes that there is no provision in 10CFR100 for exemption; however, the Nuclear Regulatory Commission (NRC) has the authority to grant this exemption.

(Corporation) proposes to implement the reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the off-site and control room dose limits. Except for the Type 2 B2-45 i

NED0-31858 Hev, I requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any looic control and interlocks. The method is consistent with the philosophy f protection by multiple leak tight barriers used in containme,t design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment. The implementation will provide [ Plant Name) with a capability to process MSIV leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the proposed changes result in an increase in protection to the public.

In conjunction with this application for exemption request, [ CORPORATION] has transmitted to the NRC an application for a license amendment pursuant to 10CFR50.90. This license amendment involves a proposed change to Section 3.6.1.2 of the Technical Specifications to permit an increase in the allowable leak rate for the MSIVs from the current [x] standard cubic feet per hour (scfh) total to.[y] scfh per main steam line. The safety analysis has been revised to assess the radiological effecte of MSIV leakage following a

- postulated design basis LOCA. [ CORPORATION] has demonstrated that the proposed change does not involve a significant hazards consideration.

This proposed exemption is a resu't of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic Issue C-8 "MSIV Leakage and LCS Failure".

The following discussion provides a detailed justification and evaluation of j the. proposed exemption. While recognizing-this exemption criteria are l specifically applicable to 10CFR50, [ Corporation) has evaluated the proposed exemption in ;. 'cordance with the criteria specified in 10CFR50.12(a). The L proposed exemption will not present an undue risk to the public health and safety and is consistent with the common defense and security. Furthermore, special circumstances are present that warrant the granting of this exemption.

1 I

Type 2 B2-46

NED0-31858 Rev. 1 i

The proposed exemption will not cause additional operational activities that may significantly affect the environment, it does not result in a significant increase in any adverse environmental impact previously evaluated in th? Final Environmental Impact Statement-0perating License Stage, result in a significant a mge in effluents or power levels, or affect any matter not previously reviewed by the NRC that may have a significant adverse environmental impact.

llpon the NRC approval of the license amendment and exemption requests,

[ Corporation] will perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0-31858 Rev.1, to provide reasonable assurance of the

• • ructural integrity of these components. ,

Therefore. [ CORPORATION] hereby requests an exemption to tqe seismic requirements of 10CFR100 Appendix A for (plant station name) to permit the use .

af existing, non seismically designed main steam piping and condenser to mitigate the raciological consequences of MSIV leakage.

A.- Justification Paragraphs VI(a)(1) 'CFR100 Appendix A requires that structures, systems and components, wnit ure the capabilit.y to prevent or mitigate the consequences of accidents which could result in potential off-site exposures of 10CFR100, be desip ed to reinain functional-following a safe shutdown earthquake (SSE) and concurrent loads. The engineering method used to assure that the required safety functions are maintained following the SSE shall involve the use of either dynamic analysis or 6 suitable qualification test to demonstrate that structure, systems, and components can withstand the seismic and other concurrent loads.

The BWROG has evaluated the capability of main steam Diping and condensers to

. process MSIV leakage following a design basis accident coincident with a Type 2 B2-47 l

l NED0 31858 Rev. 1 ,

i seismic even*, Based on this comprehensive evaluation, the BWROG has concluded there is reasonable assurance that the main steam piping and condenser will remdin function 01 following a design basis accident coincident with a seismic event, as great 7; the design basis carthquake, to mitigate the radiological consequences of MS!V leakage. The following conclusions provide the bases for this assurance:

(1) Probability for which the resulting dose from MSly leakage it.

significant is extremely low. This requires a design basis LOCA, a degraded core where ECCS are not functional, and a significant seismic event.

(2) Main steam piping and condensers are designed to strict industrial standards and buildin0 cod:s; thus, significant design margin exists.

(3) Main _ steam piping and condensers exhibit substantial seismic ruggedness. Comparisons of piping and condenser design in GE plants with those in the earthquake experience database reveal that the GE plant designs are similar to or more rugged than those that have exhibited good tarthquake performance.

(4) Possibility of significant failure in GE BWR main steam piping or condensers in the event of design br, sis earthquakes is highly unlikely, and any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

(5) A plant-sper:ific verification of seismic adequacy of tha main steam piping and condenser will be performed to provide reasonable assurance of the structural integrity of these components.

In support of the above, the BWROG has reviewed the potential combinatirns of loss Of Coolant Accidents (LOCAs) 'and seismic (vs.nts of interest:

L 1

L Type.2 82-48 l

l

NEDO 11858 Rev. 1 (1) LQC/LWEllEUT NEAR ColNCIDENT SEISMIC EVENI. For this occurrence the pressure in the piping system downstream of the MSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the flow path through main steam system piping to the condenser is assured.

1 (2) SEISMit EVIE ...WITH0tt! NEAR COINCIDENT LOCA. Without a LOCA and the )

potential associated core degradation, the radioactivity transported l

with MSIV leakage is of no radiological significance.

(3) LOCA WITH NEAR COINCIDENT SEISMIC EVENT. For this occurrence (also I assuming significant core damage) the consequences are of interest '

because a seismic induced failure in the main steam or condenser system could allow HSIV leakage to bypass the treatment pathway. It has been previously well documented that the probability of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability approximately 0.00) per reactor per year; core melt probability is plant specific and typically ranges from 0.00001 to 0.0001 per reactor per year), it is also noted that a LOCA does not induce a seismic event, and that a seismic event has a very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10-7 per reactor per year).

Considering that the probbnity of a near coincident inca and seismic event i

is much smaller than other plant-safety risks (less than 1 x 10-7 per reactor l per year for coincident events, less than 5 x 10 7 per reactor per year for seismic _ induced LOCA), the concern for main steam piping or condenser damage is of little significance. Nevertheless, because main steam piping 5 1 condenser systems designs are extremely rugged, this eqcipment is exped ed to l remain intact following design basis seismic events.

Type 2 82-49

NLD0 31858 Rev. 1 ANSI 831.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin. In addition, specific seismic design provisions have been incorporated into some newer BWR main steam and condenser systems.

To further justify the capability of the main steam system piping and l condenser treatment pathway, the BWROG has reviewed limited earthquake I experience data on the performance of non-seismically designed piping and condensers (in_pastearthquakes). The study summarizes data on the performance of main steam piping and condensers in past strong motion earthquakes and compares these piping and condensers with those in typical U.S. GE Mark I, 11, and til nuclear plants. This limited earthquake experience data and similarity comparisons are then used to further strangthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in a design ba;;is earthquake in conjunction with a LOCA occurring just prior to or after the seismic event.

The earthquake experie.1ce data are derived from an extensive database on the performance of industrial facilities and over 100 power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world from 1934 to the present.

The piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they are not designed to resist earthquakes. This is a comon conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipment such as R motor control centers, pumps, valves, structures, and so forth. That is, with limited exceptions, normal industrial construction and equipment typically l have substantial inherent seismic ruggedness, even when they are not designed l for earthquakes. No failures of main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to l minor internal tube leakage.

L Type 2 B2-50 W- * --- ' ' - -

l NE00 31858 Rev. 1 l

Comparisons of piping and condenser design in example GE Mark 1, II, and !!!

plants with those in the earthquake experience database reveal the GE plant designs are similar to or more rugged than those that exhibited good earthquake performance. The BWROG concludes that (1) the possibility of significant failure in GE BWR main steam piping or condensors in the event of an eastern U.S. design basis earthquake is highly unlikely and that (2) any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

Earthquake experience methodology has been applied in seismic equipment qualification issues associated with Unresolved Safety Issue A 46 (Seismic Qualification of Equipment in Operating Plants). Piping performance data are presented in NUREG 1061 (a report from the NRC Piping Review Committee), and this report proposes changes to criteria that are directed toward the recognition of the superior performance of piping in earthquakes and estab-lishing more realistic scismic criteria for piping qualification. The NRC has published NUREG 1030 and NUREG 1211 " Seismic Qualification of Equipment in Operating Nuclear Power Plants," which conclude that the seismic experience data approach provides the most reasonable and preferred alternative to other current equipment qualification methods.

The rapidly growing use of the seismic experience data approach is further illustrated by the fact that this method of analysis is now referenced in:

A. Draft RG 1.100, Revision 2 " Seismic Qualification of Electrical and Mechanical Equipment in Nuclear Power Plants"

8. Recent approved revision of IEEE Standard 344-1987, " Recommended Practice for Seismic Qualification of Class lE Equipment For Nuclear Power Generating Stations" C. Draft report of ASME Standard " Recommended Practice for Seismic Performance Qualification of Mechanical Equipment Used in Nuclear Power Plants."

Type 2- B2-51

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NED0 31858 Rev. 1 The earthquake experience database includes a large number and variety of piping systems. In fact, piping is probably the strongest area in this regard (compared to areas like electrical or mechanical equipment, cable trays, etc.). It has been concluded that the earthquake experience data on piping, and in particular data on main steam piping, are applicable to main steam '

piping in BWRs.

In both nuclear and conventional power plants, the condenser is designed to reduce the low pressure turbine outlet pressure (thereby increasing turbine efficiency) and to condense the steam. The nuclear environment does not impose additional significant design considerations on the condenser. With the exception of hotwell size, a conventional plant and nuclear plant with similar performance parameters have similar condensers.

None of the condensers within the seismic experience database has seismic design criteria. However, in view of the performance of the condensers within the database, it is concluded that the condensers have an inherent seismic ruggedness and that the earthquake experience data on condensers are applicable to condensers in BWRs.

Another recent study to develop, by data collection and statistic 6 analysis, updated estimates of pipe breaks in commercial U.S. nuclear pcwer plants was completed in 1987. This study evaluates both-LOCA sensitive systems and non-LOCA sensitive systems. For BWR non LOCA sensitive systems, ten pipe failures have occurred over 313 years of operating experience. None of these failures occurred in the main steam piping. Based on the observed failure rates, this study estimated the failure rate for the main steam system piping to be 0.0007 failure / year /BWR with an upper bound of 0.0096 failurer/

year /BWR. These results are consistent with the conclusion from the earthquake databases and NUREG-ll69: BWR main steam piping is designed to withstand stivere plant transients such as turbine trips and is expected to I remain intact following accidents as severe a= a design basis LOCA. Thus, the non-seismically designed main steam piping and the main condenser can be used to mitigate the consequences of MSIV leakage.

l l  !

1 1 Type.2 B2-52

NED0 31858 Rev. 1

[ Corporation) will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0 31858 Rev.1, to provide reasonable assurance of the structural integrity of these components.

1 In conclusion, there is reasonable assurance that the existing, non seismically designed main steam piping and condenser will remain j functional following a design basis a cident coincident with a seismic event,  !

as. great as the design basis earthquake, to mitigate the radiological consequences of MS!V leakage. l B. No Undue Risk to'Public Health and Safety i

The BWROG has evaluated the casability oi main steam piping and condtnsers to  ;

. process MSIV leakage followino a design basis accident coincident with a seismic event.. Based on teis comprehensive evaluation, the BWROG has concluded there is reasonable assurance that the main steam piping and f

condenser will remain functional following a design basis accident coincident ,

=< with a seismic event, as great as the design basis earthquake, to mitigate the -

radiological consequences of.MSIV leakage. 'This assurance is based on methodology.using probability analysis, margins in the existing design codes,  !

~

seismic experience, and a plant specific verfication of seismic adequacy. .

.The treatment method for MSIV leakages is recommended by the BWROG in support  ;

of the resolution to Generic issue C-8 "MSIV Leakage and LCS Failure".

(Corporation) proposes to implement the reliable and effective main steam L

piping and condenser for MSIV leakage treatment. :This treatment method is effective-to' treat- MSIV leakage over an expanded operating range without  !

exceeding the, off-site and control room dose limits. Except for the  ;

requirement.to. establish a proper flow path from the MSIVs to the condenser, -i

-the proposed method is passive and does not require'any logic control and- q interlocks. The method-is consistent with the philosophy of protection by

[

multiple leak tight barriers used in containment design for limiting fission r:

i h

Type 2 82-53

-,-._,_a._.-_. _ . , _ . _ . _ . _ . . . - . - . . . . . . . _ . . _ . - . _ _ _ _ _ . . . . . _ . , . - . _ . _ . . _ _ - - . . ._ ,- . a . . , .

l NEDO 31858 Rev. 1 product release to the environment, Therefore, the proposed method is highly I reliable for MSIV leakage treatment. The implementation will provide [ Plant Name) with a capability to process MSly leakage, and will also provide a

- uniform basis for establishing a plant specific MSIV leakage rate limit. from a safety perspective, the proposed changes result in an increase in protection to the public.

In conclusion, the proposed exempion presents no undue risk to public health and safety.

C. Consistent with Common Defense and Security l

With regard to the " common defense and security" standard, the grant of the requested exemption is consistent with the common defense and security of the United States. The Commission's Statement of Considerations in support of the exemption rule note with approval the explanation of this standard as set  :

forth-in Lona Island Lichtino comp.Any (Shoreham Nuclear Power Station, Unit ,

1),LBP8445,20NRC'1343,1400(October 29,1984). There, the_ term " common "

defense _and security" refers principally to the safeguarding of special nuclear material, .the absence of foreign control over the applicant, the

- protection of Restricted Data, and the silability of special nuclear material- for defense needs. - The granting of the requested exemption will not affect any of these matters and, thus, such grants are consistent with the common defense.and security.

D. Special Circumstances Are present-Special circumstances are present_ which warrant issuance of this requested

- exemption. These special circumstances are discussed in accordance with the classificationcontainedin10CFR50.12(a)(2):

4

_ (ii) Application of the regulation in the'particular. circumstances would not serve the underlying purpose of the rule or-is not'-

necessary to. achieve.the underlying purpose of the rule.

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NE00-31858 Rev. 1 Compliance with Appendix A of 10CFR100 for the downstream main steam piping and condenser is not necessary to achicve the underlying purpose of the rule.

The underlying purpose of the rule is to limit releases to within the off site dose limits of 10CfR100. The regulation requires components that mitigate the consequences of an accident to within the dose limits of 10CFR100 be designed to the seismic requirements of 10CFR100 Appendix A. The regulation is intended to provide a reasonable assurance that the components will remain functional for the mitigating function. For the purpose of mitigating the radiological consequences of MSIV leakage. It is not necessary to apply the seismic requirements of 10CFR100 Appendix A to the main steam piping and condenser in order to achieve the underlying purpose of the rule because:

(1) There is reasonable assurance that the existing, non seismically designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSly leakage. This assurance is based on methodology using probability analysis, margins in the existing design codes, seismic experience, and a plant specific verification of seismic adequacy.

(2)Thssafetyanalysishasbeenrevisedtoassesstheradiological consequences of MSIV leakage following a design basis LOCA. The analysis has demonstrated that the revised ,ses are well within the off-site dose guidelines of 10CFR100.

Furthermore, the seismic approach is consistent with the current resolution of the seismic and equipment qualification issues. Earthquake experiences data have applied in seismic equipment qualification issues associated with Unresolved Safety Issues A 46 (Seismic Qualification of Equipment in Operating-Plants) . Piping performance data have been presented in NUREG 1061, a report from the NRC Piping Review Committee, which proposes changes to criteria that are directed toward the recognition of the superior performance of piping in ,

Type 2 B2-55

I NEDO 31858 Rev. I  ;

l earthquakes and establishing more realistic seismic criterir, for piping qualification. The NRC has published NUREGs 1030 and 1211 Wismic I Qualification of Equipment in Operating Nuclear Power Plants," wich conclude that the seismic experience data approach provides the mo:,t reasonable and preferred alternati)e to other current equipment qualification methods.

(iii) Compliance would result in undue hardship or other costs that i are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in axcess of those-incurred by others similarly situated.

The proposed MSIV leakage treatment method utilizes the existing main steam piping and condenser for the mitigating function. Compliance with the seismic requirements of 10CTR100 Appendix A for the main steam piping and condenser would require significant upgrade of the existing equipment, lead to unnecessary long term plant shutdown fur modification, and significantly inctease maintenance requirements and the associated costs in order to meet seismic qualification requirements.

(iv) The exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the granting of the exemption.

[ CORPORATION) has transmitted to the NRC an application for a license amendment which involves proposed changes to the Technical Specifications to increase the allowable HSIVs leak rate from [x] scfh total to [y] scfh per steam-line. This application is partly based on the fact that the current limit is too restrictive, and results in excessive MSIV maintenance and repair, leading to Utditional MSIV failures, which in turn result in higher leakages. The proposed limit will benefit the public health and safety by reducing the potential for MSIVs failures, and thus keeping the MSIV leakage within the radiological analysis values, l

l

l. -Type 2 B2-56 l

l

NED0 31858 Rev. 1 (Corporation) proposes to implernent the reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range without exceeding the of f site and control room dose limits. Except for the requirement to establish a proper flow path from the MStVs to the condenser, the proposed method is passive a,ed does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly

)

reliable for MSIV Icakage treatment. The implementation will provide [ Plant i Name) with a capability to process MSlV leakage, and will also provide a uniform basis for establishing a plant-specific MSIV leakage rate limit. From a safety perspective, the pronsed changes rewit in an increase in protection to the public.

The exemption from 10CTR100 Appendix A seismic requirements for the downstream piping and condenser is required so that (plant station name) can operate with the proposed Technical Specifications limit of [y] scfh and with a capability to process MSIV leakage. This benefit will compensate for any decrease in safety that may result from the granting of this exemption.

Thus, special circumstances exist warranting the granting of this exemption.

E. Environmental imoact

-The proposed exemption has been analyzed and determined not to cause additional construction or operational activities which may significantly affect the environment. It does not result in a significant increase in any adverse environmental impact previously. evaluated in the final Environmental Impact Statement Operating License Stage, result in a significant change in I-effluents or power levels, or affect any matter not previously reviewed by the

-Nuclear Regulatory Commission which may have a significant adverse environmental impact.

[

Type 2- B2-57

l NED0 31858 Rev. 1 The proposed exemption does not alter the land use for the plant, any water uses or impacts on water quality, air or ambient air qutilty. The proposed action does not affect the ecology of the site and vicinity and does not i affect the noise emitted by station. Therefore, the proposed exemption does not affer.t the analysis of environmental impacts described in the environmental report.

Type'2 58 p -

NEDO 31858 Rev. !

'PENDIX B BWROG MSIV LEAKAGE CLOSURE COMMITTEE TYPE 3 EXAMPLE NRC SUBMITTAL FOR INCREASING MSIV LEAK RATE LIMITS, EllMINATING REQUIREMENTS FOR LEAKAGE CONTROL SYSTEMS, AND EXEMPTION REQUEST TO-100FR100 APPENDIX A l

l

' Type 3 B3-1

- NED0-31858 Rev. 1 U.S. Nuclear Regulatory Iommission

-Attn.: Document Control Desk j Washington, DC 20555-RE: ' [PlantStationName) '

DocketNo.'[xxxxx)  !

LicenseNo.[xxx)

Gentlemen:

[ Corporation) hereby transmits an application for amendment to the [ Plant Name) Facility L;.erating License [xxx), and an application for exemption to C 10CFR100 Appendix A. i

_ [ Corporation) requests an amendment to the Technical Specifications, set .

forthlin Appendix' A to the License, to. permit an increase in the allowable leak rate for the_ main steam isolation valves-(MSIV) and to delete the MSIV Leakage Control System (LCS). In addition [ Corporation) requests that the  ;

downstream main steam piping and condenser be exempted from the seismic  ;

requirements specified in Appendix A of 10CFR100.

In support .of.the proposed changes'to the Technical Specifications, enclosed for the Nuclear Regulatory Commission's review are:

(1). Applicationfor'AmendmenttotheLFacilityOperatingLicense;

[

- Technical: Specification pages affected by the, changes; and supporting information and analyses addressing the changes. The

' analyses demonstrate that the proposed changes do not involve a-  ;

- significant hazards consideration pursuant _to-10CFR50.92.-

. (2) Application for specific exemption to 10CFR100 Appendix A; and '

supporting information and justification for this exemption.

[ Corporation) realizes that-there-is no provision in--10CFR100 s

i E Type.3 p

B3-2

+- .m, wi- n s-r- ---r-p. 2- e- - ** *w--,.wew-.- m--a,-- w---c= -* --- s-- - -

NE00-31858 Rev. I for exemption; however, the Nucleer Regulatory Commission has the authority to grant this exemption. This exemption will not present undue risks to the public health and safety, and is consistent with the common defense and security, furthermore, special circumstances are present which warrant issuance of this exemption request.

These propcsed changes are supported by BWR Owners' Group (DWROG) work.

The BWROG formed a MSly Leakage Committee in 1982 in response to Generic Issue C-8 "MSIV Leakage and LCS failure". Generic issue C-8 addressed the safety concerns that reported MSIV leakages are too high and that the Leakage Control System will not function at high MSIV leakages. Based on the extensive, ongoing work performed by the BWROG to support resolution of the Generic Issue, the BWROG has developed the technical justification for the proposed Technical Specifications changes and associated exemption requests.

The General Electric (GE) report, NED0 315358 Rev. 1, describes the safety benefits and provides justification for the proposed chunges. With regard to increasing leakage rate limits for the MSIVs, this will avoid unnecessary maintenance which has in the past adversely affected the operability of the MSIVs.

With regard to the deletion of LCS, the proposed changes involve a replacement of the existing LCS with the more reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to reduce dose consequences of MSly leakage over an expanded operating range and will, thereby, resolve the safety concern that the LCS will not function at MSly leakage rates higher than the LCS design capacity.

Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable and effective for MSly leakage treatment.

Type 3 B3-3 i 1

NEDD 31858 Rev. 1 Pursuant to 10CFR50.9i(b)(1), [Corporatio7) has provid>td a copy _ of this license amendment request ai,d e ie ossociated analyais regarding no significant -

hazards censideration(s) to the appropriate state representative.

[ Corporation: representatives will be available to discur. or meet with j the Nuclear Regulatory Connission staff at your convenience tJ address this J matter.

Very truly ycurs, i

i

[ CORPORAL!0NJ 4

__. INAMEl___

Vice President '

Enclosures cc:, Regional,%dmin'.strator,kegion[x]

[Name), Director

[Name), Project fianager GNtme), Resident!nsp9etor (Name), State.R1presentative 1

I h-l l_ Type 3 B3-4 L

NEDO.31858 Rev. 1 l

UNITED STATES Of AMERICA NUCLEAR REGULA10RY COMMISSION l

In the Matter of 4- (( Corporation) DocketNo.Jxxx) i t 1

[PlantName) l Mf1D3111

_ _INamel , being duly sworn, s'.ates that he is Vice Prosident of i [ Corporation); that he is authorized on the part 01' said Corporation ta siga

.and file with the Nuclear Regulatory Consnission the documents attached hereto a*id that all such documents, are true anu correct to the best of hn knowledge, i information and belief.- '

. iSion31)ggl Subscribud and sworn to before me, a Notary Public in and for the State of

/i J ) and County of [ a),this[ _), day of [ _ _ _) ,

, 1988.

t ISionaturel I Notary Public in and for

[ County, State)

My Commission expires:

1 l

+

i l

Type.3 83 S

0 L NE00 31858 Rev. I 1

UNITED STATES NUCLEAR REGULATORY COMMISSION in the Matter of )

[Corporati .) Docket No. [xxx)

[PlantName) )

APPLICATION FOR AMENDMENT TO C

OPERATING LICENSE Pursuant to Sectior S0.90 of the Regulations of the Nuclear Rriqu't'or,v Commission,[ Corporation),holderoffacilityOperatingLicenseNo.[xxx),

hereby requests the Tenhnical Specifications, set forth in Appendix A to the license, be. amended to permit an increase in the allowable leak rr e for the main steam isolation valves (MSIVs) and deletion of the HSIV Leakage Control System (LCS), Specifically [ Corporation) requests that: (

l. Allowable leak rate specified in the Technical Specificaliois Section 3.6.1.2* be changed from [x] to [y] standard cubir, foot per tour <

(scfh) per main steam line. This proposed change Nels ts a higher, but still conservative allowable leak rate for the MSIVs.

2. Sections [3/4.6.1.4)* be amended to permit the d(1 tion of the10S from the Technical Specifications. [ Corporation) proppes an alternate to Regulatory Guide 1.96, " Design of Hain Sterm isolation Valve Leakage Control Systems for Sniling WC er Reactor Nuclu.r Pot er Plants." Utilizing the main stean lines and condenser, as an alternate method for HSIV leakage '; eatment, has been demonstrated to be more effective than the-lCS in terms of reliability, in concurrence with this application for license amendment, (Corporation) has applied for an exemption to 10C"R100, Appendix A.

Type 3 B3-6 i

, I

NEDO 31858 Rev. 1

3. Table 3.6.3 l* be amended to permit the deletion of the LCS isolation valves from the Technical Specifications. The LCS lines which connect to the main steam lines will be disconnected and welded / cap closed to preserve containment integrity.

4. The index, Sections 3/4.6.1.5* through 3/4.6.1.8* and bases sections B3/4.6.1.4* through B3/4.6.1.9* be administrative 1y amended to rearrange section and page numbering.

The proposed changes to the Technical Specifications are set forth in Enclosure 1.

i The proposed changes are a result of the extensive work performed by the r BWR Owners' Group (BWROG) in support of the resolutioh of Generic issue C 8 "MSIV Leakage and LCS Failure". GE report NE00 31058 Rev. I provides technical justification on a generic basis to support these proposed changes.

Based on thL BWROG evaluation of MSly leakage performance, the current Technical Specification allowable MSIV leak rate is extremely limiting and routinely requires the repair and re testing of MSIVs. This unnecessary repair significantly impacts the maintenance work load, often contributes to outage extansions, and has in the past adversely affected the operability of the HSIVs. BWR outage planners routinely schedule several days of contingency to rcpair and retest the MSIVs. In addition, the needless dose exposure to maintenance personnel is inconsistent with As low As Reasonably Achievable (ALARA) requirements. There have also been many Licensee Event Reports written for MSIV leakages exceeding the Technical Specification limit.

The sections identified are based on the BWR Standard Technical Specifications for BWR/5 (NUREG-0123). Each utility should revise these sections as applicable to their individual Technical Specifications.

Also, some plant sites may specify a maximum allowable HSIV leakage rate per line (i.e.,100 scfh), but require that the total is less than four times the maximum allowable leakage rate for each line (i.e., 200 scfh total).

Type 3 B3-7 4

m_._ __-_m___.------ - - - - - - - - - - - - - - - - - - - - - - _ _ - - - - _ _ - - - _ _

NED0 31858 Rev. 1 The proposed changes will reduce unnecessary MSiv repair costs, avoid unnecessary dose exposure to maintenance personnel, reduce outage durations, and extend the effective service life of the MSIVs. In addition, the proposed increase in the MSIV leakage limit has potential to significantly reduce recurring valve leakages, and minimize the possibility of needless repair which can compromise plant safety.

Based on BWROG evaluttion, the LCS is extremely difficult to maintain.

The system contains extensive logic and instrumentation, requiring frequent calibration to meet the Technical Specification requirements. The system has been declared inoperable, and required entry into the Limiting Condition of Operation (LCO)atleast(Plantspecific). Also, replacement of safety-related parts, such as blowers, heaters and flow elements, has become increasingly difficult, and lead times have been long.

The proposed changes involve a replacement of the existing LCS with the more reliable and effective main ateam drain line and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage

.over an expanded operating range. Except for the requirement to establish a proper flow path from the main steam isolation valves to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak tight barriers used in containment design for limiting fission product releasc to th) environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment.

The existing leakage control systems have limitations for mitigating MSIV leakage. Operation of the system induces higher MSly leakage by increasing the differential pressure across the valves, and is inconsistent with the philosophy of multiple barriers for limiting fission product releases (does not take advantage of the outboard MSIVs). (For positive pressure LCS, operation of the system increases containment pressure and thereby increases the containment leakage.]* The LCS requires multiple logic controls,

[ Type 3. B3-8

NEDO 31858 Rev. 1 interlocks, timers, containment isolation valves, and other equipment to ensure containment integrity and protection from the hiji pressure main steam lines. Based on plant operating experience, the LCS does not provide a high degree of reliability. Also the LCS has limited capacity and does not function at moderate HSIV leakage rates above 100 scfh.

Even though the resulting off-site doses may be slightly higher for low MSIV leakage rates, the effectiveness of the proposed method even for leakage rates greater than the proposed increased allowable limits, ensures offsite does limits to the public are not exceeded. Overall, the proposed treatment method can handle MSly leakage over an expanded operating range, and will thereby resolve the safety concern that the LCS will not function at MSIV leakage rates higher than the LCS capacity. Thus, a margin of safety exists.

Furthermore, it is c1carly a safety improvement to replace a r.ystem with know limitations with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability.

Plant-specific supporting information and results of the radiological analysis that justify '..- proposed changes are included as Enclosure 2. As concluded in the s'Jpporting information, the increased MSIV allowable leak rate of [y] scfh and the deletion of the LCS, will not adversely affect the performance of the primary containment isolation function. A plant-specific radiological analysis has been performed utilizing the main condenser as an alternate treatment path for MSIV leakage. This radiological analysis demonstrates that the roposed changes result in an insignificant increase to the dose exposures previously calculated for a design 1 basis Loss-of-Coolant Accident (LOCA). the revised LOCA dose exposures remain well within the guidelines of 10CfR100 for the off-site doses and 10CFR50, Appendix A (General Design Criteria 19) for the control room doses.

_[ Corporation)willinstituteintotheMSIVmaintenanceandtestprogram, the requirement that any MSIV exceeding the proposed [y] scfh limit, will be repaired and re-tested to meet a leakage rate of less than or equal to [x]

scfh. This will assure continuation of high quality repair and refurbishment efforts to improve the overall performance and reliability of the MSIVs.

L l

Type 3 B3-9

I NEDO 31858 Rev. 1 l The proposed change will require a minor plant desica change to allow specified main steam drain valves to be opened even if off-site power is not availabic. Applicable normal plant operating procedures, and emergency operating procedures shall be reviewed and revised accordingly.*

Upon approval of the proposed LCS deletion from the Technical Specifications [ Corporation) will disconnect [ remove) the LCS, Lines connecting to the main steam lines will be disconnected and welded / cap closed such that containment integrity is maintained [ plant-specific). In lieu of the safety-related LCS, MSIV leakage will be transported by the main steam drain lines to the condenser and processed as discussed in Section 4.3 of NE00-31858 Rev. 1.

Furthermore [ Corporation) will incorporate the applicable alternate leakage treatment methods, consistent with GE document NED0 30324 " Potential Operator Actions to Control MS!V Leakage," into the Operational Procedures and Emergency Operational Procedures at (Plant Name).

[ Corporation) will also perform a verification of seismic adequacy of the main it.eam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0-31858 Rev.1, to provide reasonable assurance of the structural integrity of these components.

Pursuant to 10CFR50.92, an analysis which demonstrates that the proposed changes do not involve a significant hazards consideration, is included as Enclosure 3. The proposed changes have been reviewed in accordance with Section 6.5 of the Technical Specifications. The proposed changes will not authorize any change in the types of effluents or in the authorized power level of the facility. An application for exemption to Appendix A of 100FR100 is included as Enclosure 4.

y' l

Plant-specific 1

Type 3 B3-10 l

NEDO 31858 Rev. !

WHEREFORE, Applicant respectfully requests that Appendix A to the facilityL,:eratingLicenseNo.[xxx)theretobeamendedintheformahached -

hereto as Enclosure 1.

(CORPORATION)

By:

Vice President i Subscribed and sworn to before me on this day of _ . .

)

l l

l l

-Type 3. 83-11 l.

NEDO 31858 Rev. 1 EECLQ11RL1 K0E0fA1L0ill LLCENSE NO. lxxx1 QQCKEI NO. Ixxx1 PROPOSED CHANGES TO TECHNICAL SPEClflCAILO!!S Replace the following pages with the attached revised page(s)*. These pages are provided in its entirety with marginal marking to indicate the changes.

1. Page vii

2. Page xiii

3. Page 3/4 62

4. Page 3/4 63

, 5. Page 3/4 6-7

b. PNe3/4 6 8a

7. Page 3/4 6-9

8. Page 3/4 6-10

9. Page 3/4 6-11

10. Page 3/4 6-12

11. Page 3/4 6-22

12. Page B 3/4 6-1

13. Page B 3/4 6-2

14. Page B 3/4 6-3 The pages identified here are based on the BWR Standard Technical Specifications for BWR/5 (NUREG 0123). Eac' atility should revise these pages as applicable to their individual T' .hnical Specifications.

Type 3 B3-12

NEDO 31858 Rev. 1 i ENCLOSURL2  :

LC_0BPDEhl10M LICENSE NO. fxxx1 DQf1LT NO. Ixxx1  ;

SUPPORTING INFORMATION AND ANALYSES ,

i 1.0 ~1NTRODUCTION AND

SUMMARY

OF RESULTS The proposed Technical Specification amendrant involves an increase in the allowable leakage rate from [x] scfh to u , scfh per main steam line. .

deletion of the LCS. In addition, it is requested that the downstream ,

-main. steam piping and condenser be exempted from the seismic requirements ,

specitid in Appendix A to 10CFR100. ,

Section 2.0 of this Enclosure provides a summary of background

information; Section 3.0 discusses the justifications for ti . proposed changes; Section 4.0 provides a summary of the plant-specific radiological; dose assessment, and Section 5.0 summari7es the potential benefits for a Technical Specification MSIV allowable ',ak rate of [y]

scfh and the deletion of the LCS.

The BWROG_ report, NEDO-31858 Rev. 1 "BWROG Report _for Increasing MSIV Leak Rate Limit s'and Elimination of_ Leakage Control Systems," November -;

1991, provides the justification- for increasing MSIV. leakage limits', and for eliminating the requirements for the LCS. With concurrence from the valve manufacturers, this report concludes that MSIV leakage-rates'up to-200 scfh are'not an~ indication of substantial mechanical defects in the

-valve which-would challenge the isolation capability of the valve to.

fulfill its safety. function. Therefore, the proposed increase-in the allevable leakage ' rate ~ to -[y] scfh for the MSivs will not inhibit the -

isolation capability of the valve. .

b (Type'3 B3 ,

m.., ~.,p.,. ,,,-,.s._ _ _ . , . , . . -.,,_.,..,m.-.,.4..m.- r,,,,%. _,-.......,.r 4_.., r

NED0-31858 Rev. 1 The BWROG has evaluated several methods and has recommended the isolated condenser as an alternate method to the LCS for MSIV leakage treatment. i The isolated condenser method takes advantage of the large volume in the j main steam lines and the condenser to hold up the release of fission products leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser. [ Corporation) proposes to j delete the LCS from the Technical Specification and to incorporate the isolated condenser as an alternate method for MSIV leakage treatment.

The BWROG has evaluated the availability of main steam system piping and  !

condenser alternate treatment pathways for processing MSIV leakage. The  !

BWROG have reviewed the potential combinations of Loss-Of Coolant i Accidents (LOCAs) and seismic events of interest:

(1) LOCA WITHOUT NEAR COINCIDENT SEISMIC EVENI. For this occurrence, the pressure in the piping system downstream of the MSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the alternate flow path through the main steam system piping to the condenser is assured.

(2) SEISMIC EVENT WITHOUT NEAR COINCIDENT [0CA. Without a LOCA and the potential associated core degradation, the radioactivity transported with MSIV leakage is of no radiological significance.

(3) LOCA WITH NEAR COINCIDENT SEISMIC EVENT. for this occurrence (also assuming significant core damage) the consequences are of interest because a seismic induced failure in the main steam or condenser system could allow MSIV leakage to bypass the alternate treatment pathway. It has been previously well documented that the probability of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability is approximately 0.001 per reactor per year; core Type 3 B3-14

NEDO 31858 Rev. I melt probability is plant-specific and typically ranges from 0.00001 to 0.0001 per reactor per year). It is also noted that a LOCA does not induce a seismic event, and that a seismic event has a very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NVREC/CR 4792 Volun.e 4  :

reported probability of seismic induced LOCA to be less than 5 ,

10*7 perreactorperyear).

Considering that the probability of a near coincident LOCA and seismic event is much smaller than other plant-safety risks (less than 1 x 10 7 per reactor per year for coincident events, less than 5 x 10-7 per reactor per year for seismic induced LOCA), the concern for main steam piping or condenser damage is of little significance. N3vertheless, because main steam' piping and condenser systems designs are extremely rugged, this equipeent is expected to remain intact following design basis seismic eunts.

' ANSI-031.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin, in addition, specific seismic design provisions have been incorporated into some newer BWR main steam and condenser systems.

In order to further justify the capability of the main steam system piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non seismically designed piping and condensers (in past earthquakes). The study summarizes data on the performance of main steam piping and condensers in past strong-motion earthquakes and compares these piping and condensers with those in. typical U.S. GE Mark I, II, and III nuclear plants'. This limited earthquake experience data and similarity comparisons are Type _3 B3-15

NEDO-31858 Rev. I then used to further strengthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in a design basis earthquake in conjunction with a LOCA occurring just prior to or after the seismic event.

The earthquake experience data are derived from an extensive database on the performance of industrial facilities and over 100 power plant units (tur'inea, a associated condensers, and main steam piping) in 19 j earthquakes around the world from 1934 to the present.

The piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they are not desior.ed to resist earthquakes. This is a common conclusion in r.tudies of inis type on other plant items such as welded steel piping in general, anchored equipment such as motor control centers, pumps, valves, structures, and ,

so forth. That is, with limited exceptions, normal industrial construction and equipment typically have substantial inherent seismic ruggedness, even when they are not designed for earthquakes. No failures of main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.

Comparisons of piping and condenser design in example GE Mark I; II, and III plants with those in the earthquake experience database reveal the GE plant designs are similar to or more rugged than those that exhibited l good earthquake performance. In addition, [ Corporation) will also

perform 6 verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NEDO 31858 Rev.1, to provide reasonable assurance of the structural integrity of these components. The BWROG concludes that (1) the L

Type 3 B3-16

NED0 31858 Rev. l  !

pussibility of significant failure in GE BWR main steam piping or j condensers in the event of an eastern U.S. design basis earthquake is l highly unlikely and (2) any such failure would also be contrary to a I large body of historical earthquake experience data, and thus unprecedented.

The design basis LOCA has been re analyzed for radiological impacts j' utilizing the isolated condenser as an alternate method for MSiv leakage treatment. The analysis demonstrates that a maximum MSIV leakage rate of

[y] scfh per main steam line results in an acceptable increase to the  ;

dose exposures calculated for a design basis LOCA. In addition, the ,

analysis demonstrates that MSIV leakage rates of approximately (z) scfh ,

per main steam line will not result in dose exposures in excess of the regulatory limits.

2.0 -McKGROUND i

The safety function of the MSIVs is to isolate the reactor-system in the  ;

i event of a LOCA or other events requiring containment isolation. The design of the MSIVs and'its isolation requirements are described in Sections (5.4.5*] and [6.2.4*] of the. Final- Safety Analysis Report (FSAR). The: allowable leakage rate from MSIVs is . included in the LOCA -

[

radiological analysis' evaluated in Section (15.6.5*)- of the FSAR.

The safety related LCS is designed to direct MSIV: leakages to the reactor building Standby Gas Treatment System (SGTS) for treatment of fission products. The LCS'is described in Section 6.7** of the FSAR, and-consists of inboard-and outboard subsystems. The inboard subsystem contains isolation' valves which provide for containment integrity in the i

eve'nt that the corresponding MSIV fails to close.

j. Plant-specific- .

Plant Unique. -This' draft describes .the negative pressure LCS.

Type 3- :B3 17 ,

, 4 \ W... ,- L_- . - ~ - . - _ . . . . -

. n. ._ -, - ._._,._u.-.

- - . ~ - - . - - . - - - . . - . . - - . - - - . . _ - . . - - - .

I NED0-31858 Rev.-~1 Operating experience indicates that MSIVs frequently exceed the Technical Specificatioa allowable leak rates. Some of these valvet, repeatedly fail the local leak rate tests despite frequent ' disassembly a'id refurbishment.

As _a-result of_ increasing MSIV leakages and the inability of the LCS function at high MSIV leakages, the Nuclear Regulatory Commission prioritized Generic Issue C-8, "MSIV_ Leakage and LCS Failure" as a high priority item in January 1983. This' issue was closed in 1990.

The BWROG formed a MSIV Leakage Committee in 1982 to address the increasing MSIV leakage rates, and a-follow-on MSIV Leakage Closure Committee in 1986 to ' address alternate-actions to resolve on-going, but less severe MSIV problems. - The MSIV Leakage Committee' identified contributors which cause MSIVs to fail- the leak rt4te tests by large margins, developed recommendations to minimize leakages, evaluated alternates for MSIV leakage treatments, and compiled recent history of MSIV leakages and LCS operating experience.

3.0 JUSTIFICATIONS FOR THE PROPOSED CHANGE

[ Corporation) proposes to increase the Technical Specification allowable leakage rate for the MSIVs-from (4. .ah to [y] scfh per main steam line and to delete the LCS requirements from the Technical Specifications.

1he current' Technical Specification MSIV le#k- rate limit is extremely L- limiting and routinely requires unnecessary repair and re-testLof the '

L _MSIVs. ThisWgnificantly-impacts the maintenance work load-during plant.

' outages and uf u:n contributes 1to outage extensions. (The outage planners-at [ plant station name] typically schedule several days of contingency to repair and re-tes't'the-MSIVs)*- In addition, the needless dose exposures

- to-maintenance personnel'-are inconsistent with As Low As-Reasonably l-i P1 ant Unique.

Type'3 B3-18

• , r--- = = - .-_m ___ _ __._.___m_ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - m

. NED0-31858 Rev. 1

- Achievable (ALARA) principles. (There have been many Licensee Event Report written for MSIVs failing to meet the current Technical

_ Specification limit'- )*

From a-safety perspective, calculations using standard conservative assumptions;for considering-the off-site consequences of a postuisted design-basis LOCA confirm that off-site and control room doses will be within the regulatory guidelines for the allowabb d!1V leakage rate.

This :alculation is described in 3ection 15.6.r rm he FSAR. However, if MSIV leakages are. only moderateL c..per than me allowable limit, the calcu' lated doses.will exceed the regulatory guidelines. Furthermore, as documented in Generic Issue C-8, the LCS will not function if MSIV exceeds the leakage limit by a moderate amount.

MSIVfs failure to meet the current Technical Specification limit have been documented .in response to surveys _ conducted by the Nuclear LRegulatory Commission'during the early 1980 and by the BWROG during the imiddle and late _1980s. As high as 50% of the total'"as found" MS V local leak rate tests were reporttd in the early-NRC survey to exceed the

-leakage l rate limit.

The BWROG has-studied the issuss regarding MSIV leakage rates, their causes, and available alternecives. The results of the BWROG study are provided'in NED0-31858 Rev. I and are also summarized in NUREG-1169.. In -

response to Generic Issue-C-8, the BWROG has recommended-corrective actions and maintenance practices to reduce the MSIV leakage rates.

A recent survey conduc'.ed-by the BWROG of MSIV-leakage tests performed between 1984. and 1988' indicates that the implementation of industry and BWROG actions has been effective in reducing the leakage rates, and,-in particular, a: reduction-in' the number of valves which experience

substantial high leakage rates. The survey concludes that about 23% of Plant Unique i

Type 3: 83-19

NE00-31858 Rev. I the total "as found" MSIV-leakages still exceed the limit of 11.5 scfh and about 10% exceed 100 scfh.

The MSIV leakage performance at (plant station name) is .....[to be provided by the Utility).- The leakage performance at (plan + station nime) is consistent with the recent survey by the BWROG.  !

1 Despite the recent improvement in leakage performance, MSIV leakage rates still frequently exceed the current Technical Specification limit and the -

safety and maintenance problems related to high MSIV leakage rates, although less severe, remain as a significant issue.

Furthermore, based on the extensive evaluation of valve leakage data, the BWROG has found that disassembling and refurbishing the MSIVs to meet-

.very low leakageLlimits frequently contribute to repeating failure. By not having to disassemble' the valves and refurbish them for minor leakage, the= utility may avoid introducing one of the root causes of-recurring valve leakage-problems that may lead to later leak test.

failures and the possibility of compromising: plant safety.

The current Technica1L Specification allowable -leakage rate is established by; excessively conservative LOCA radiological analysis as described in Section 15.6.5 of the FSAR. - The valve's physical size'and operability characteristics (large size and fast acting), and existing turbine building. equipment- were -not considered: at-the time the. leakage limit was established. Based on the in-depth' evaluation of MSIV leakages, the

- BWROG has concluded:the-MSIV: leakage rates up to 500 scfh are not an lindication of substantial mechanical defects in"the valve which would challenge the-isolation ~ capability' of the valve to-fulfill its safety function. Furthermore, ' valve manufacturers have stated that leakage ,

rates up to' 200 scfh can occur without having a major valve defect.

Therefore, the proposed increase from [x] to [y] scfh per main steam-line will'not inhibit the MSl_V's performance of the isolation function and

will not compromise the safety of [ plant station name),

l Type 3 - B3-20

-. . A

NED0-31858 Rev. 1 This oroposed increase provides a more realistic, but still conservative, limit for the MSIVs. Based on the BWROG study, the proposed increase in the allowable ' leak rate will increase the chance for a successful local leak rate test to greater than 90%, up from the 77% success rate at the current typical limit of 11.5 scfh. This increase in successful local leak rate testing will significantly reduce MSIV maintenance costs, reduce dose exposure to maintenance personnel, reduce outage durations, extend effective service life of the MSIVs, and minimize the potential for outage extensions at [ plant statico name).

A safety-related LCS was required by Regulatory Guide 1.96 in order to reduce the radiological consequences of MSIV leakage. As discussed earlier, Generic Issue C-8 identified the safety concern that MSIV leakage rctes, as determined by conservative local leak rate tests, were too high and that the LCS would not function at high MSIV leakage rates.

The 1981 NRC survey indicated that 33 percent of the total tests exceeded leakage rates of 100 scfh. Since the process capability of the LCS at (plant station name) is designed for MSIV leakage rates of no more than 100 scfh, the potential existed for the LCS not to function as analyzed for a oesign-basis LOCA as described in [Section 15.6.5]* of the FSAR.

Consequently, the conservatively calculated dose contribution from MSIV leakage would exceed the regulatory limits for off-site and control room doses.

[ Corporation] proposes to delete the LCS requirements from the Technical Specifications. The proposed changes involve a replacement of the existing LCS with the more reliable and effective main steam drain line and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV_ leakage over an expanded operating range. Except for the requirement to establish a proper flow path from the main steam drain line to the condenser, the proposed method is passive and does not Plant-specific Type 3 B3-21

i NEDO-31858 Rev. I require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the 1

environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment.-

The existing LCS has limitations for mitigating MSIV leakage. Operating the system induces higher MSIV leakage by increasing the differential pressure across the valves, and is inconsistent with the philosophy of multiple barriers for limiting fission product releases (does not take advantage of the outboard MSIVs). [For positive-pressure LCS, operating the system increases the containment pressure and thereby increases the containment leakage.]* The LCS requires multiple logic controls, interlocks, timers, containment isolation valves, and other equipment to ensure containment integrity and protection from the high pressure main steam lines. Based on plant operating experience, the LCS does not provide a high degree of reliability. Also the LCS has limited capacity and dcas not function at moderate MSIV leakage rates above 100 scfh.

Even though the resulting off-site doses may be slightly higher for low MSIV leakage rates, the effectiveness of the proposed method even for leakage rates greater than the proposed increased allowable limits,

. ensures off-site dose limits to the public are not exceeded. Overall, the proposed treatment method can handle MSIV leakage over an expanded operating range, and will thereby resolve the safety concern that the LCS will not function at MSIV leakage rates higher than the LCS capacity.

-Thus, a margin of safety exists. Furthermore, it is clearly a safety improvement to replace a system with known limitations with the alternate main steam piping and condenser treatment pathway, which has been shown

)

to have excellent reliability.

Furthermore, [ Corporation) will incorporate the applicable alternate l leakage treatment methods, consistent with GE document NED0-30324 l

" Potential Operator Actions to Control MSIV Leakage", into' the '

Type 3 B3-22

a1 4 _ .a ~ , a +s+ x4 NEDO-31858 Rev. 1 Operational Procedures and Emergency Operational Procedures.

In addition to improving plant safety, the proposed deletion of the LCS requirements in the Technical Specifications will result in significant operational and maintenance benefitss The LCS has extensive logic and instrumentation and calibration of this instrumentation is frequently requirad to meet the Technical Specification requirements. The BWROG has evaluated recent LCS performance data. Results of this evalu. tion is summarized in NE00-31858 Rev. 1. The evaluation indicates that the LCS is extremely difficult to maintain and that plant shutdowns and start-up delays frequently occur. The LCS performance at (plant station name) is

.....[to be provided by the Utility]. The LCS performance at [ plant station name] is consistent with the recent survey by the BWROG.

In conclusion, [ Corporation) proposes to increase the Technical Specification allowable HSIV ' leakage rate from [x] to [y] scfh per main steam line, and to eliminate the requirements for LCS in the Technical Specifications. The proposed increase in the HSIV leakage limit should significantly reduce recurring valve leakage problems, and will minimize needless valve repair which can compromise plant safety. The proposed deletion of the safety-related LCS and the proposed alternate method (main steam lines and condenser) resolve the safety concern regarding LCS-effectiveness at higher MSIV leakage rate.

4.0 ANALYSIS OF MSIV LEAKAGE CONTRIBUTION TO RADIOLOGICAL 00jiE CALCULATIONS 4 .- l Selection of Alternate leakaae Treatment Method The BWROG has evaluated several alternate MSIV leakage treatment methods and has recommended the isolated condenser for MSIV leakage treatment.

This leakage treatment method takes advantage of the large volume in the 1solated main condenser to hold up the release of fission products leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser..

Type 3 B3-23

.NEDO-31858 Rev. 1-1 As previously discussed-in Section 1.0,.the BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage. The BWROG has determined that the probability of a near coincident LOCA and a seismic event is much-smaller _ than for other plant safety risks. The BWROG has also determined that main steam piping and condenser designs are extremely rugged, and that the ANSI-831.1 design requirements typically used for nuclear plant system design contain a good deal of margin. In order to further justify the capability of the main steam piping and condenser

-alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers-(in past earthquakes). This study concluded the possibility of a failure which could cause a loss of steam or conda; sate in BWR main steam piping or condensers.in the evant of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

'This conclusion'is consistent with NUREG/CR 4407, " Pipe Break Frequency

= Estimation for Nuclear Power Plants", dated May 1987, which reported no observed failures in the main steam piping over 313 years of reactor

-operating years.

JTherefore, the'isol'ated condenser alternate MSIV leakage treatment path at (plant name) is considered -appropriate for. the reduction 'of radiological consequences of a design basis LOCA.

4.2- Radiolooical-Ansivsis-and Result

' The radiological' dose methodology has been developed by General Electric

'for the BWROG.-

The radiological analysis calculates the effects of the proposed allowable MSIV leakirate in terms of control room and off-site doses.

Typn 3 B3-24

NED0-31858 Rev.'l The revised LOCA doses are the sum of the LOCA doses (as described in Section xxx of the FSAR) and the calculated MSIV leakage doses. This method of calculating the revised dose exposures is very conservative since the LOCA doses already include the dose contribution from MSIVs at the maximum leakage rate permitted in the current Technical Specifications.*

Table 1** shows the calculated dose exposures from the BWROG radiological analysis for [ plant name]. Regulatory limits and calculated doses from LOCA radiological analysis are also included in Table 1 for comparison purpose. This analysis demonstrates that a MSIV leakage rate of 200 scfh per_ main steam line results in an acceptable increase to the dose exposures previously calculated for the control room, EAB, and the LPZ.

The revised LOCA doses remain well within the guidelines of 10CFR100 for off-site doses and 10CFR50, Appendix A, (General Design Criteria 19) for the control room doses. Furthermore, the calculation shows that MSIV leakage rates up to approximately [z] scfh per steam line would not exceed the regulatory limits. Therefore, the proposed method provides a substantial safety margin for mitigating the radiological consequences of MSIV leakage beyond the proposed Technical Specification leak rate limit of-[y] scfh.

Plant-Specific. Other BWRs may elect to replace the LOCA MSIV dose

contribution with the new doses using the alternate leakage treatment

1. , method.

The following discussion on doses are based on as-an example plant (Hope Creek)

Type 3 B3-25

NED0 31858 Rev. 11

.l'">

Table 1 CONTRIBUTION.TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV l.EAK RATE OF 200 scfh HOPE CREEK GENERATING STATION Whole Body- Thyroid Beta

.(ram), (rem) .(r.RE).

Exclusion 'Area A) 10CFR 100 Limit 25 300

• Boundary.

-(2-Hour) B) . Previous- Calculated 0.6 76.7

. Doses **

C).iContributionFrom' O.1 2.6 MS!Vs at 200 scfh

- D) - New Calculated = Doses -0.7 -79.3

-Low Population- A)l10CFR100 Limit- 25 300

• Zone.'(30-Day)

B). Previous, Calculated 0.08- 7.7

< Doses **

.C): Contribution From 0.34 ': 57.5' MSIV-at 200-scfh Dr New Calculated Doses 0.42 '65.2

' Control - Room .- . A) GDC 5 30 30/75***

- .'(30- Day) _

B) l Previous-Calculated-- 0.04 0.26 . 91 Dos.s**

L C): _ Contribution From 0.10 ' 2. 71 -. 1.64L

MSIVs at 200 scfh D);-New Calculated Doses'

-0.14 2.97- 2.55

• - - No limit specified. '

FSAR Section '15.6.' 5.5 and-6.4 (includes MSIV leak rate at a total of -

45 scfh- for the. first 20 ~ minutes; control room dose assumes 100% per day -

_ _ reactor building-inleakage).

" **" .75' if_ prior commitment'has been made to use protective clothing.

n.

-Type 3- B3-26

__.__._._.___.___m_..m.__.mm.-._____m- . _ _ . _ _ . _ _ - - ___.__._- - _--

NED0 31858 Rev. 1 5.0 BENEFITS FOR THE PROPOSED CHANGES As discussed in.NED0-31858 Rev. 1, recent MSIV leakage performance has significantly improved since the early BWROG survey in 1984 and the NRC survey in the early 1980s. Despite the recent iraprovement, MSIV leakage rates exceeding the current Technical Specification limits still frequently occur. The BWROG evaluation of the recent MSIV leakage performance concludes that the proposed change will improve the chance for a successful local leak rate test to greater than 90%, up from the 77% success rate at the current Technical Specification limit of 11.5 scfh.

Specifically, MSIV leakage experience at [ plant station name) are [***to be provided by utility on plant-specific leakage performance and any problems which may impact critical path, outage extensions, etc.***].

As discussed in detail in Section 3.0, deleting tN u S will reduce the overall dose rates, and eliminate the system's impact on refueling and maintenance outage activities at [ Plant Name]. [*** Provide specific plant data for support ***]. The proposed alternate method (main steam lines and condenser) for MSIV leakage treatment will also eliminate the safety concern regarding LCS effectiveness at higher MSIV leakage rates.

Although the revised LOCA doses are slightly higher for low MSIV leakage rate, the effectiveness of the proposed method, even for-leakage rates greater than the proposed increased MSIV allowable leak rate, ensures that off-site and control room doses are not exceeded.

l l

Type 3 83-27

k NEDO-31858 Rev. 1 INCLOSURE 3 fCORPORATION1 LICENSE NO. Ixxxl l DOCKET-NO.{yzx1 N0 SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS (Corporation] proposes an amendment to the Technical Specifications as follows:

1) Revise Section-3.6.1.2* to permit an increase in the allowable leak rate-for-the main steam isolation valves.(MSIVs) from the current [x]

e standard cubic _ feet per hour (scfh) to (y] scfh per main steam line.

2)- Delete Sections 3/4.6.1.4* and B3/4.6.1.4* to reflect the deletion of the Leakage Control System (LCS)_from the Technical Specifications.

3)! Revise Table 3.6.3-1* to delete the inboard LCS isolation valves

-[ plant unique identification).

L4) Revise-Index pages vii and xiii,_ renumber Sections'3.6.1.5*Ethrough 3.6.1.8*, 4.6.1.5* through 4.6.1.8*, pages number 3/4 6-8a* through 3/4 6-12*,;and-SectionsLB3/4.6.1.5* through B3/4.6.1.9* as a result' of deleting -Sections 3.6.1.4* and B3/4.6.1.4*.

'PursuaAt'to 10CFR50.92, .the proposed amendment: involves no significant hazards considerations..

r

- Pl ant-speci fic l-L L

Type.3 83-28

NE00 31858 Rev. 1 l

The operation of IPlant St ition Namel. in accordance with the arocosed amendment. will not involve a sianificant increase in the probability or conseouences of an accident previously evaluated.

The proposed amendment to Section 3.6.1.2 does not involve a change to structures, components, or systems that would affect the probability of an accident previously evaluated in the Final Safety Analysis Report (FSAR).

The proposed amendment to delete Sections 3/4.6.1.4 and Bases Sections 83/4.6.1.4 involves eliminating the LCS requirements from the Technical Specifications. As described in Section 6.7 of the FSAR, the LCS is manually initiated in about 20 minutes following a design-basis LOCA. Since the LCS is operated only after an accident has occurred, this proposed amendment has no effect on the probability of an accident.

Since HSIV leakage and operation of the LCS are included in the radiological analysis for the design-basis LOCA as described in Section 15.6.5 of the FSAR, the proposed amendments will not affect the precursors of other analyzed accidents. The proposed amendments result in acceptable radiological consequences of the design-basis LOCA previously evaluated in Section 15.6.5 of the FSAR.

The (plant station name] has an inherent MSIV leakage treatment capebility.

[ Corporation] proposes to use the main steam lines and condenser as an alternate to Regulatory Guide 1.96 " Design of Main Steam Isolation Valve Leakage Control System For Boiling Water Nuclear Power Plants" for MSIV leakage treatment. [ Corporation) will incorporate this alternate method in the Operational Procedures and Emergency Operational Procedures.

The BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways fur processing MSIV leakage, and has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for othar plant safety risks. The BWROG has also Type 3 B3-29

NED0-31858 Rev. I determined.that main steam piping and condenser designs are extremely rugged, and that:the ANSI-B31.1 design requirements typically used for nuclear plant system design contain a good deal of margin.

In order to further justify the capability of the main steam piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and. condensers (in past earthquakes). This study concluded that the possibility of a failure which could cause a loss of st:11m or condensate in BWR main steam piping or condensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also te contrary to a large body of historical earthquake experient e data, and t.hus unprecedented.

[ Corporation] will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0-31858 Rev.1, to provide reasonable assurance of the structural 'integr_ity._of these components.

A plant-specific radiological analysis has been performed to assess the effects of the proposed increase to the allowable MSIV leak rate in terias of controlcroom and off-site doses following a postulated design basis LOCA.

This analysis utilizes the hold-up volumes.of the main steam piping and condenser as an alternate method for.the MSIV leakages. As discussed earlier,

'there is= reasonable assurance that the main steam piping and condenser will

-remain intact-following a design basis earthquake. The radiological analysis

.uses: standard conservative assumptions for the release of source' terms-conv.' stent with Regulatory Guide 1.3 Revistor 2, " Assumptions Used for

' Evaluating. the Potential Radiological. Consequences of a loss-Of-Coolant ,

Accident for Boiling Water Reactors", dated' Apri' 1974.

The Analysis demonstrates that dose contributions from the proposed MSIV leakage rate limit'of [y] scfh and from the proposed deletion of the LCS

/> result.in an acceptable increase to the LOCA doses previously evaluated 1

1

-Type 3 B3-30 i V _ .. . . .

NE00.31858 Rev. I against the regulatory guidelines for the off-site doses and control room doses as contained in 10CFR100 and 10CFR50, Appendix A (General Design Criteria 19), respectively. The LOCA doses previously evaluated are discussed in Section [15.6.5]* of the FSAR. The revised LOCA doses are the sum of the LOCA doses previously evaluated in the FSAR and the additional MSIV doses calculated using the alternate treatment method. This method of calculating the revised doses is very conservative since the LOCA doses previous evaluated already included dose contributions from MSIV at the maximum leakage rate permitted in the current Technical Specifications.** The attached table ; hows the previous calculated doses and the new calculated doses.

The whole body dose at the Low Population Zone (LPZ) and the control room is increased from 0.08 to 0.42 rem *** and from 0.04 to 0.14 rem, respectively.

These increases are acceptable because the revised doses are well within the Regulatory guidelines (0.42 versus 25 rem at the LPZ, and 0.14 versus 5 rem at the control room). The associated whole body dose at the exclusion area boundary (EAB) increased insignificantly from 0.6 to 0.7 rem.

The thyroid dose a the LPZ increased from 7.7 to ,.,.2 rem. This increase is acceptable because the revised dose of 65.2 rem is significantly less than the regulatory guideline (300 rem). The EAB thyroid dose increased slightly from 16.7 to 79.3 rem, whereas the control room thyroid dose increased from 0.26 to 2.97 rem. The increase in control room thyroid dose is acceptable because the revised dose remains a small fraction (9.9%) of the limit. The control room beta dose is increased from 0.91 to 2.55 rem, which remains insignificant relative to the regulatory guideline of 30 rem.

Plant-specific Plant-specific, original LOCA dose from MSIV leakage may be replaced with revised dose.

• • ~

All doses and comparisons indicated here-on are Hope Creek specific.

Type 3 B3-31

NEDO-31858 Rev. 1 l

~ ~

CONTRIBU.* ION TO THE LOCA' DOSE EXPOSURES FOR A MAXIMUM MSIV LEAK RATE OF 200 scfh- ,

HOPE CREEK GENERATING STATION Whole

, Body lnyroid Beta 4

frem) (rem) (rem) ia' ' Exclusion l Area A) 10CFR 100 Limit' 25- 300 *

Boundary-c _

(2-Hour) 18);: Previous Calculated 0.6 76.7 Doses **

-C)- Contribution From 0.1 2.6 iMSIVs at'200 scfh ,

D) _- New Calculated Doses 0.7 79.3' Low Population:. L))10CFR'100 A _ Limit - 25 300 *

Zonet(30-Day)-

' ~

Previous Calculated ' 0.08 7.7 B)' Doses **

C)l Contribution.From '0.34 57.5 MSIV at 200-scfh-lD) - New Calculated > Doses - 0.42 65.2 $

J a,. LControl1 Room A)JGDC-19; 5 30 30/75***

'(30-Day)l

B). Previous Calculated - 0.04 0.26 .91 Doses **

C)E. contribution 1From 0.10 2.71: - 1.64.

=MSIVs at-200 scfh D) New Calculated' Doses - 0.14 2.97 2.55:

• - :No limit specified.

' **~

FSAR-Section 15.6.5~5 and 6.4 (includes MSIV-leak ratelat a total of L45:sefh:for the first E20 minutes;' control room dose" assumes 100% per. day reactor; building-inleakage).

• • • f '75 :if prior commitment has been made to use protective clothing.-

+n

! Type:3 B3-32 L

NED0-31858 Rev. 1 It is important to note that the resulting doses are dominated by the organic iodine fractions which e because of the u:traconservative source term assumptions used in this analysis. For 200 scfh per steam line, more than 85%

of the off-site iodine and control room doses are due to organic iodine from the RG 1.3 source term and organic iodine converted from the elemental iodine deposited in main steam piping systems. If the actual iodine compositic, from the fuel release (cesiuu iodide) is Jsed in the calculations, esstntially all of thia organic iodir.e dose would be eliminated.

In summary, the proposed changes resc't in an acceptable increase to the radiological consequences of a LOCA previously evaluated in the FSAR. The revised LOCA doses are well within the regulatory guidelines. Although the revised LOCA doses are slightly higher for low MSIV leakege rates, the effectiveness of the p*oposed method even for leakage rates greeter than the proposed ina.reased MSI' allowable leak rate, ensures that off-site and control roca doses are not exceeded.

The piaposed amendment to Table 3.6.3-1 af the Technical Specifications

. involves the deletica of LCS valves [plar.t aique identification) from the list of containment isolation valves. This proposed change is consistent with the prot 1 sed deletion o/ the LCS. The LCS lines which are connected to the main steam piping will be welded / cap close* to assure containment integrity is maintained. Tht welding and examination procedurts will be in accordance with ASME Section III requirements.

This oroposed chcnge does not involve an increase in the probobility of an accident previously evalucted in the FSAR. In fact, this propcsed change reduces the probability of an accident since, with this proposed change, the plant will je operating with le:s isolation valves subjected to postehted failure. This proposed change has no offert in the consequences of an

• Plant unique input Type 3 B3-33 L-

3EDO-31853 Rev. 1 accident since the LCS lines will be wolded/ cap close, thus assuring that the containment int grity, isolation, and ieak tast capabilicy are not compromised for the postulated accident.

The proposed changes to the indrx pag?s, renun. bering of Sections 3.6.1.5 through 3.6.1.8, 4.6.1.5 through 4.6.1.8, page:, .1u1ber 3/4 6 Ba through 3/4 6-12, and Section B3/4.6.1.5 through B3/4.6.1.5 is adminictrative in nature

/ and has no effect on any accident. These changes provide new section and page number designations due to the deletion of Sections 3.6.1.4 and Bases Sections 3 S/4.6.1.4.

The operation of [P,lant S1ation Namel. in .ggpr_d3Ld.Gf witFdhe oroogjd jLnigndment. will not create _,the Dessihility,oLa.new or_di[Legart e

kind of Agttdf.nt from any accidfnt,previously gyjt]uated.

Tht proposed amendment to Section 3.6.1.2 does not create the possibility for

_; a new or different kind of accident from any accident previously evaluated.

The BWROG evaluated' MSIV leakage perforti.ance and concluded that MSIV leakage rates up to 200 scfh will not inhibit the capability and isolation ptrformance of the valve to isolate the primary containme'.it. There is no new modification which could impact the MSIV operability. The LOCA has been analyzed using the main steam piping and condenser as a treatment method to process MSIV leakage at th6 proposed maximum rate of [y] scfh, Th w rore, the proposed change does not create any new or different kind of ar ident from any accident previously evaluated in the FSAR.

The proposed amendment to delete Sections 3/4.6.1.4 and Bases Sections B3/4.6.1.1 does not create the possibility of a new or different kind of accident- from any eccident previously evaluated because the removal of the LCS does not affect a: of the remaining systems at [ plant name) and the LOCA has been analyzed using the alternate method to process MSIV leakages.

Type 3 B3-34

NED0-31858 Rev. 1

)

The proposed amendment to delete the LCS isolation valves from Table 3.6.3 1 does not create the possibility of a new or different kind of accident from to the main steam piping will be welded / cap closed to assure that the primary containment integrity, isniation, and leak testing capability are not compromised, therefore eliminaticg the possibility for any new or different kind of accident.

The proposed changes to the index pages, and the revision of section numbers are administrative in nature, and do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The operation of IPlant Station Namel. in accordance e:ith the oroposed amendment. will not involve a sionificant reduction in tha Jnaroin of safety.

The proposed amendment to Section 3.6.1.2 does not involve a significant reduction in the margin of safety. As discussed in the Bases of the Technical Sptcification 3/4.6.1.2, the allowable leak rate limit specified for the MSIVs is used to quantify a maximum amount of bypass leakage assumed in the LOCA radiological analysis. Results of the analysis are evaluated against the dose guidelines contained in 10CFR100 for the off-site doses and 10CFR50, Appendix A (General Design Criteria 19) for the control room doses. Thtrefore, the margin of safety is considered to be the difference between the calculated doses and the guidelines as contained in 10CFR100 and GDC 19.

Results of the radiological analysis demonstrate that the proposed change does not involve a significant reduction in the margin of safety. The whole body doses, in terms of margin of safety, are insignificantly reduced by 1.4% at the LPZ, 2.0% in the control room, and 0.4% at the EAB. The thyroid dose margin of safety is reduced by 19.1% at the LPI, 9.0% in the control room, and 0.9% at the EAB. The beta dose is insignificantly reduced by 5.5% in the centrol room in terms of margin of safety. The margins of safety are not significantly adversely affected because the absolute margins of safety remain well below the guidelines (lowest whole body margin of safety is 97.2% at the EAB, lowest thyroid margin of safety is 73.6% at EAB).

Type 3 B3-35

NEDO.31858 Rev. 1-

.Therefore, the proposed amendment does'not involve a significant reduction in the overall margin of safety at (plant station name).

The ~ proposed amendment 'co delete Sections 3/4.6.1.4, and B3/4.6.1.4 does not reduce margin of safety. In fact. the overall margir. of safety is increased.

The' function of the LCS for MSIV leakage treatment will be replaced by alternate main steam drain lines and condenser equipment. This treatment method is effective to reduce-dose consequences of Msiv leakage over an expanded' operating range and will, thereby, resolve the safety concern that >

the-LCS will not function at MSlV leakage rates higher than the LCS design capacity. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require ,

-any . logic control and interlocks. The method is consistent with the philosophy of ~ protection-by-multiply-leak-tight barriers used in containment design for. limiting fission pmluct release to the environment. Therefore, the proposed method _ is_ highly-reliable and effective for MSIV leakage treatment.

The revised LOCA doses remain well within the regulatory limits for the site

.and control room, furthermore, the calculation shows that MSIV leakage rates

up to-[z] scfh per steam line would not exceed the regulatory limits.

Therefore, the proposed method provides-a substantial safety margin for

-_ mitigating the radiological consequences-of MSIV leakage bevond the proposed

-Technical? SpecificationLleaf rate limit;of [y] scfh.

- The' proposed amendmentito' delete LCS isolation valves [ plant unique

' identification] from' Table 3.6.3'-1 doos not reduce- the margin lof safety.

Weldeo/ cap closure of the LCS lines. assures that-the primary containment-integrity,and: leak testing capability are not compromised; therefore, it does:

not resultiin'a reduction in-the margin of safety, f

L Type 3' B3-36

NEDO-31858 Rev. 1 The proposed amendment to the index pages, and the revision of section and page numbers is administrative in nature, and does not have any impact on the m.trgin of safety.

Therefore, the proposed amendment to the Technical Specifications does not involve a significant hazards consideration.

Type 3' B3-37

NED0 31858 Rev. 1 ENCLOSURE 4 iCORPORAfION1 LICENSE NO. Ixxxl DOCKET NO. Ixxxl APPLICATION FOR EXEMPTION TO APPENDIX A 0F 10CFR100

[ CORPORATION), holder of Facility Operating License No. NPF-[xxx), hereby requests an exemption of the downstream main steam piping and condenser from the seismic requiremerts specified in Appendix A of 10CFR100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants."

Specifically, [ Corporation) proposes to employ probability analysis, existing design capabilities, seismic experience, and a plant specific seismic adequacy

. verification, as alternate methodology to the dynamic analysis or qualification test specified in Paragraph VI(a)(1) of 10CFR100 Appendix A, to provide reasonable assurance that the existing esin steam piping and condenser will remain functional following a design basis accident coincident with a significant seismic event.

The exemption would allow the existing, non-seismically designed main steam piping and condenser to be used for mitigating the radiological consequences tof MSIV leakage during the duration of a Design Basis Accident, such that the resulting doses are within the guidelines of 10CFR100.

[ Corporation]' recognizes that there is no provision in 10CFR100 for exemption; however, the Nuclear Regulatory Commission (NRC) has the authority to grant this exemption.

[ Corporation) proposes to replace the existing LCS with the more reliable and effective main steam drain line and condenser for MSIV leakage treatment.

This treatment method is effective to treat MSIV leakage over an expanded Type 3 B3-38

NED0 31858 Rev. I operating range. Except for the requirement to establish a proper flow path from the main steam isolation valves to the condenser, the proposed method is pessive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Therefora, the proposed method is highly reliable for MSIV leakage treatment.

The existing leakage control systems have limitations for mitigating MSIV leakage. Operation of the system induces higher MSIV leakage by increasing the differential pressure across the valves, and is inconsistent with the philosophy of multiple barriers for limiting fission product releases (dces not take advantage of the outboard MSIVs). (For positive pressure LCS, operation of the system increases containment pressure and thereby increases the containment leakage.]* The LCS requires multiple logic controls, interlocks, timers, containment isolation valves, and other equipment to ensure containment integrity and protection from the high pressure main steam lines. Based on plant operating experience, the LCS does not provide a high degree of reliability. Also the LCS has limited capacity and does not function at moderate MSIV leakage rates above 100 scfh, Even though the resulting offsite doses may be slightly higher for low MSIV leakage rates, the effectiveness of the proposed method even for leakage rates greater than the proposed increased allowable limits, ensures off-site does limits to the public are not exceeded. Overall, the proposed treatment method can handle MSIV leakage over an expanded operating range, and will thereby resolve the safety concern that the LCS will not function at MSIV leakage-rates higher than the LCS capacity. _Thus, a margin of safety exists.

Furthermore, it is clearly _ a safety improvement to replace a system with know limitations'with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability.

In conjunction with this application for exemption request, [ CORPORATION] has transmitted to the NRC an application for a license- amendment pursuant to Type 3 B3-39

NE00-31858 Rev. I 10CFR50.90. This license amendment involves a proposed change to Section l 3.6.1.2 of the Technical Specifications to permit an increase in the allowable )

leak rate for the MSIVs from the current [x] standard cubic feet per hour (scfh) total to [y] scfh per main steam line, and a proposed change to Section 3.6.1.4 for eliminating the requirements for the LCS. The safety analysis has been revised to assess the radiological effects of MSIV leakage following a postulated design basis LOCA. [ CORPORATION) has demonstrated that the i proposed change does not involve a significant hazards consideration.

This proposed exemption is a result of the extensi"e work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic Issue C-8 "MSIV Leakage and LCS Failure".

Tha following discussion provides a detailed justification and evaluation of the proposed exemption. While recognizing this exemption criteria are specifically applicable to 10CFR50, (Corporation] has evaluated the proposed exemption in accordance with the criteria specified in 10CFR50.12(a). The proposed exemption will not present an undue risk to the public health and safety and is consistent with the common defense and security. Furthermore, special circumstances are present that warrant the granting of this exemption.

The proposed exemption will not cause additional operational activities that may significantly affect the environment. It does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-0perating License Stage, result in a significant change in effluents or power levels, or affect any matter not previously reviewed by the NRC that may have a significant adverse environmental impact.

Upon the NRC approval ci- *he license amendment and exemption requests,

[ Corporation] will perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0-31858 Rev.1, to provide reasonable assurance of the structural integrity of these components.

Type 3 B3-40

.- . ,- - . - . - ~ ~ - . - - . -

NE00-31858 Rev. 1.

Therefore,[C'RPORATION)hereby:requestsanexemptiontotheseismic O

requirements of 100FR100 Appendix. A for (plant stat _1on name) to permit the use ,

-ofl existing,4non-seismically designed main steam piping and condenser to mitigate the' radiological consequences of MSIV leakage.

A. - Justification-Paragraphs Vl(a)(1) of 10CFR100 Appendix A requires that structures, systems-and components, which assure the capability to. prevent or mitigate the consequences of accidents which could result in potential off-site exposures- L ofl10CFR100, be designed to remain-functional-following a safe shutdown a

earthquake (SSE) and concurrent loads. The engineering method used to assure that;the required safety functions are maintained following the SSE shall

' involve the use of either; dynamic analysis or a suitable: qualification test to

-demonstrate-that structure;-systems, and components can withstand'the seismic and other concurrent loads.

The-BWROG has evaluated the. capability of main steam. piping and. condensers to

. process:MSIV leakage following a design basis. accident coincident with a .

. seismic event. Based on this comprehensive. evaluation, the BWROG has concluded' there is: reasonable assurance that .the main steam piping and -

~

condenser will remain functional following a design basis accident coincident with;a seismic. event,-as great as.the design: basis earthquake, to mitigate the

' radiological; consequencesL of. MSIV leakage. .The following conclusions' provide

~

1the bases.for. this assurance:: '

(1) Probability ~ for which the resulting dose from MSIV leakage is significant is extremely low. This requires a design basir.LOCA, a

degraded core where ECCS are not Lfunctional, and a' significant 1 seismic event.- ,

(2). Main steam piping andLcondensers'are designed to strict industrial s standards and building codes _;. thus, significant design margin exists.

Type 3 83-41

. ye , - . v---- - --- ,wyw

" 1 NED0-31858 Rev -1 (3)' Main' steam piping and condensers exhibit substantial seismic ruggedness. Comparisons of piping and condenser design in GE plants with'those in the earthquake experience database reveal that the GE plant designs are similar to or more rugged than those that have exhibited good earthquake performance.

_(4) Possibility of significant failure in GE BWR main steam piping or condensers-in the event of design basis earthquakes is highly unlikely, and any such failure would also be contrary to a large body

of historical earthquake experience data, and thus unprecedented.

(5). A plant-specific verification of seismic adequacy of the main steam piping _ and condenser will be performed to prov_ide reasonable assurance of the structural integrity of these components.

In support of,the above, the BWROG has reviewed the potential-combinations of Loss'-Of-Coolant Accidents (LOCAs) and seismic events of interest:

(1):LOCA WITHOUT NEAR COINCIDENT SEISMIC EVENT. For this occurrence the pressure in the' piping system downstream of the MSIVs is rapidly reduced to atmospheric pressure; and since there'is no seismic event,

- the alternate flow' path through main steam syst' piping to the

. condenser!is um.rd.

-(2)' SEISMIC EVENT WITHOUT NEAR COINCIDENT LOCA. .Without.a LOCA and the.

potential. associated core-degradation,-the radioactivity transported with MSIV leakage :is of no radiological' significance.

(3) LOCA WITH NEAR COINCIDENT SEISMIC EVENT. For this occurrence (also Lassuming sigriificant core damage) the consequences are of interest because a seismic induced failure in the main steam or condenser system could allow MSIV leakage to bypass the-alternate treatment pathway. -It has been previously wel_1 documented that the probability Ty'pe 3 B3-42

k NED0 31858 Rev. 1-4 of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability approximately 0.001 per reactor per year; core melt probability is plant-specific and typically ranges from 0.00001 to 0.0001 per reactor per year). It is also j noted that a LOCA does not induce a seismic event, and that a seismic event has a very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10-7 per reactor per year).

Considering that the probability of a near coincident LOCA and seismic event I is much smaller than other plant-safety risks (less than 1 x 10-7 per reactor per year for coincident events, less. than 5 x 10-7 per reactor per year for seismic induced LOCA), the concern for main steam piping or condenser damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design basis seismic events.

ANSI-831.2 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin. In addition, specific seismic design provisions have been incorporated into some newer BWR main steam and condenser systems.

To further justify the capability of the main steam system piping and condenser alternate treaiment pathway, the BWROG has reviewed limited earthquake experience ccta on the performance of non-seismically designed piping and condensers O n past earthquakes). The study summarizes data on the

- performance of main steam piping and condensers in past strong-motion earthqua w and compares these piping and condensers with those in-typical U.S. GE Mark I, II, and III nuclear plants. This limited earthquake experience data and similarity comparisons are then used to further strengthen the con,:1usions on how the GE Type 3 B3-43

NED0-31858 Rev.'l piping _and condensers would maintain their pressure retention function in a design basis earthquake in conjunction with a LOCA occurring just prior to or ,

after the seismic event.

The earthquake experience data are derived from an extensive database on the performance of industrian facilities and over 100-power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world from-1934 to the present.

-The piping and condensers-in the earthquake experience database exhibited substantial seismic-ruggedness, even when they are not designed to resist-

-earthquakes. This is~ a common conclusion in studies of this type on othcr plant. items such as-welded steel piping in general,_ anchored equipment such as moNr control centers, pumps, valves, structures,: and so forth. That is, with

-limited exceptions, normal industrial construction and equipment typically have substantial inherent seismic ruggedness, even when they are not designed for earthquakes._ No failures.of main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal _ tube leakage.

Comparisons of piping and condenser design'in example GE Mark _l, II, and !!!

plants with those in the earthquake experience database reveal the GE plant

-designs are similar.to-or more rugged than those'that exhibited good earthquake-performance. -The-BWROG concludes that-(1) the possibility of.

'significant; failure in GE BWR main steam piping or condensers in the event of an eastern U.S. design basis earthquake is_ highly unlikely and that (2).any such failure would also-be contrary to.a large body of historica1 ' earthquake experience' data, and thus unprecedented.

' Earthquake experience methodology _has been applied in seismic equipuint qualification issues associated with Unresolved. Safety Issue A-46 (Seismic

' Type-3 B3-44

NEDO-31858 Rev. -Qualification of- Equipment' in Operating Plants). Piping performance data are

_ presented in NUREG 1061 (a-report from the NRC Piping Review Committee), and this report proposes changcs to criteria that are directed toward the recognition of the superior performance of piping in earthquakes'and estab-lishing more realistic seismic criteria for piping qualification. The NRC has l published NUREG-1030 and NUREG-1211 " Seismic Qualification of Equipment in Operating ~ Nuclear Power-Plants," which conclude that the seismic experience data approach provides the most reasonable and preferred alternative to other current equipment qualification methods.

The rapidly growing use_of the seismic experience data approach is further illustrated by the fact that this method.of analysis is now referenced in:

A. Draft RG 1.100, Revision 2 " Seismic Qualification of Electrical and

- Mechanical Equipment in Nuclear Power Plants" '

B. Recent approved revision of IEEE Standard 344-1987, " Recommended Practice for Seismic Qualification of Class IE Equipment For Nuclear ,

Power Generating Stations"

.C.-JDraftl report of ASME Standard " Recommended Practice for Seismic Performance Qualification of MecPical _ Equipment Used in- Nuclear Power Plants."

Th'e earthquake experience database includes a large number and-variety of-piping. systems. - In fact,- piping is probably the strongest area in this regard-(compared-to areas like electrical, or mechanical equipment, cable trays, etc.). It has'.been concluded,'that the earthquake experience data on piping,: .

and:in particular data on main steam piping, are applicable to main' steam piping _in:BWRs.

In both nuclear and. conventional power plants, 'the. condenser is designed to-reduce the low pressure turbine _ outlet pressure-(thereby increasing turbine L

l' V.

-Type 3 -B3-45 r

- r - ,

NED0 31858 Rev. I efficiency) and to condense the steam. The nuclear environment does not impose additional significant design considerations on thi ;ondenser. With the exception of 50twell size, a conventional plant and nuclear plant with similar performance parameters have similar condensers.

flone of the condensers within the seismic experience database has seismic design criteria. However, in view of the performance of the condensers within the database, it is concluded that the condensers have an inherent seismic ruggedness and that the earthquake experien.e data on condensers are applicable to condensers in BWRs.

Another recent study to develop, by data collection an' natit: .a1 analysis, updcted estimat4s 9f pipe breaks in commercial U.S. nuc1Mr p ar plants was completed in 1987. This study evaluates both LOCA sensitive systems and non-LOCA sensitive systems, for BWR non LOCA sensitive systems, ten pipe failures have occurred over 313 years of operating experience. None of these failures occurred in the main steam piping. Based on the observed failure raret, this study estimated the failure rato for the main steam sysi.em piping to be 0.0007 failure / year /BWR with an upper bound of 0.0096 failures /

year / 6WR. These results are censistent with the conclusion from the carthquake databases and NUREG 1169: BWR main steam piping is designed to withstand severe plant transients such as turbine tript and is expected to remain intact following accidents as severe as a design basis LOCA. Thus, the non-seismically designed main steam piping and the main condenser can be used to mitigate the consequences of MSly leakage.

[ Corporation) will also pwform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelir,es discussed in

'Section 6.7 of NED0 31858 Rev.1, to provide reasonable assurance of the structural integrity of these components.

in conclusion, there is reasonable assurance that the existing, non-seisnically designed main steam piping ar4 u ndenser will remain l

Type 3 B3-46 1

NE00 31858 Rev. 1 i i

functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitip + the radiological consequences of MSIV .. ge.  !

D. Eg_h due Ris o imblic Health and Safety the BWROG has evaluated the capability of main steam piping and condensers to '

process MSIV leakage following a design basis accident coincident with a scismic event Based on this comprehensive evaluation, t.'e BWROG has concluded ther,, is reasonable assurance that the main steam piping and condenser will remain functional following a design basis accident coincident

- with a seismic event, as great as th? design basis earthquake, to mitigate the i radiological conwquences of MSIV leakage. This assurance is based on methedology.using probability analysis, margins in the ex! sting design codes, ,

saismic experience, and a plant pecific verification of seismic idequacy.

i The treatment method for MSIV leakages is recommended by the BWROG in support ,

of the resolution to-Generic Issue C 8 "MSIV Leakage and LCS failure". The proposed changes involve a-replacement of the existing LCS with the more reliable and effective main steam piping nd condenser for MSIV leakage treatment.: This treatment method is effective to-reduce dose consequences of MSIV leakage over an expanded operating range and will, thereby, r m ive the i

safety concern that the LCS will-not function at MSIV leakage rates h!6her than.the LCS design capacity. Except for the requirement to establish a

- proper flow path from the HSIVs to the candenser, the proposed athod-is [

passive andL does not require any logic control and interlocks. - The method 'is-consistent with the philosophy of protection.by multiple leek tight barriers used-in-containment design for_ limiting fission product release to-the '

environment.- Therefore, the proposed method. is highly eff ,tive and reliable

- for MSIV leakage treatmint.

-In conclusion, the proposed exemption presents no undue risk.to public health and safety.

Type.3 83-47 4

, C..~.i .,n. ._.em,-m. ---w-~.,,-ww,. ..

r-.- - - ,,,---w,,-- ,--,,p -

y -w, v+3 .-4..m , --.e., ..-- ,,

.g g... -wn.- e ,3-... -- + , y

NE00 31858 Rev. 1 C. Consistent with Conmon Defense and Security With regard to the

• common defense and security" standard, the grant of the requested extmotion is consistent with the comon defense and security of the United States. The Commission's Statement of Considerations in support of the exemption rule note with approval the explanation of inis standard as set ,

forth in Lon1 sland 1 Liahtina comoany (Shoreham Nuclear Power Station, Unit 1),LBP-84-45,20NRC1343,1400(October 29,1984). There, the term " common

-defense and security" refers principally to the safeguarding of special nuclear material, the absence of foreign controi over the applicant, the protection of kestricted Data, and the availability of spec 41 nuclear

'terial for defense needs. The granting of the requetted exemption will not affect any of these matters and, thus, such grants are consistent with the common defense and security.

D. Special Cirq11r i mtes Art ' resent Special circumstances are present which warrant issuance of this requested exemption. These special circumstances are discussed in accordance with the classification contained in 10CFR50.12(a)(2):

(ii) Appilcation of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

Compliance with Appendix A of 10CFR100 for the downstream win steam piping and condenser is not necessary to achieve the underlying purpose of the rule.

The underlying purpose of the it'le is to limit releases to within the off-site dose limits of 10CFR100,- The regulation requires components that mitigate the

' consequences of an accident-to within the dose limits of 10CFR100 be designed to the. seismic requirements of 10CFR100 Appendix A. The regulation is intended to provide a reasonable assurance that the components will remain functional for the' mitigating function. For the purpose of mitigating the L

L -Type 3 B3-48

l NED0-31858 Rev. I radiological consequences of MSly leakage, it is not necessary to apply the seismic requirements of 10CfR100 Appendix A to the main steam piping and condenser in order to achieve the underlying purpose of the rule because:

(1) There is reasonable assurance that the existing, non seismically designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthqucke, to mitigate the radiological consequences of MSIV leakage. This assurance is based on methodology using probability analysis, margins in the existing design codes, seismic experience, and a plant specific verification of seismic adequacy.

(2) The safety analysis has been revised to assess the radiological consequences of MSIV leakage following a design basis LOCA. The analysis has demonstrated that the revised doses are well within the off site dose guidelines of 10CfR100.

Furthermore, the seismic approach is consistent with the current resolution of the seismic and equipment qualification issues. Earthquake experiences data have applied in seismic equipment qualification issues associated with Unresolved Safety issues A 46 (Seismic Qualification of Equipment in 0,)erating Pl..nts ) . F1 ping performance data have been presented in NVREG 1061, a report from the NRC Piping Review Committee, which propcces changes to criteria that are directed toward the recognition of the superior performance of piping in earthquakes and establishing more realistic seismic criteria for piping qualification. The NRC has published NUREGs 1030 and 1211 " Seismic Qualification of Equipment in Operating Nuclear Power Plants," which conclude that the seismic experience data approach provides the most reasonable and preferred alternative to other current equipment qualification methods.

1 Type 3 B3-49

NED0 31858 Rev. 1 l (iii) Compliance would result in undue hardship or other costs that j are significantly in excess of those contemplated when the l regulation was adopted, or that are significantly in excess of i those incurred by others similarly situated.

The proposed MSIV leakage treatment method utilizes the existing main steam piping and condenser for the mitigating function. Compliance with the seismic ,

requirements of 10CFR100 Appendix A for the main steam piping and condenser would require significant upgrade of the existing equipment, lead to unnecessary long term plant shutdown for modification, and significantly

-increase maintenance requirements and the associated costs in order to meet  ;

seismic qualification requirements. I (iv) The exemption would result in ber< fit to the public health and safety that compensates for any decrease in safety that may result from the granting of the exemption.

[ CORPORATION) has transmitted to the NRC an application for a license amendment which involves proposed changes to the Technical Specifications to -

increase the allowable MS!Vs leak rate from [x] scfh total to (y)'scfh per .

steam line and to delete the requirements of LCS. For the MSIV leak rate l limit,-t.his application is partly based on the fact that the current limit is ,

too restrictive, and results-in excessive MSly maintenance and repair, leading-to af! *nal MSIV failures which in turn result in higher leakages. The propos 11mit will benefit-the public health and safety by reducing the  !

potential for MSIVs failures, and thus keeping the MSly leakage within the

-radiological analysis values. i for the LCS,'the proposed changes involve a replacement of the existing LCS-with the more reliable and effective main steam piping and condenser method-for MSly leakage treatment. The effectiveness of the proposed method even for

leakage _ rates greater than the proposed increased allowable limits, ensures.

off-site dose limits to the public are not exceeded. _0verall, the p oposed '

L L

!- . Type'3- _B3-50 t

< -- m-v w .,,,,-.~,,,v,,,n-,,,,, ,,,,,w v----,-,,--,n, , , , , - , . , , , ,-n, n, , - , --, ,--,.,n.- ,r,,n,-av--, , . . . -

NCDn.31358 Rev. I treatment method can handle HSIV leakage over an expanded operating range, and will thereby resolve the safety concern that the LCS will not function at MSIV leakage rates higher than the LCS capacity. Thus, a margin of safety exists.

Furthermore, it is clearly a safety improvement to replace a system with known limitations with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability.

The exemption from 10CFR100 Appendix A seismic requirements for the downstream piping and condenser is required so that (plant station name) can operate with the proposed Technical Specifications limit of [y] scfh and with the alternate MSIV leakage treatment method. This benefit will compensate for any decrease in safety that may result from the granting of this exemption.

Thus, special circumstances exist warranting the granting of this exemption.

E. Environmental Imnici The proposed exemption has been analyzed and determined not to cause additional construction'or operational activities which may significantly affect the. environment. It does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement 0perating License Stage, result in a significant change in effluents or power levels, or affect any matter not previously reviewed by the Nuclear Regulatory Commission which may have a significant adverse environmental impact. 4 The proposed exemption does not alter the land use for the plant, any water uses or impacts on water quality, air or ambient air quality. The proposed action does not affect the ecology of the site and vicinity and does not affect the noise emitted by station. Therefore, the proposed exemption does-not affect the analysis of environmental impacts described in the environmental report.

1 p.

Type 3 83-51/B3-52 l

i NEDO 31858 Rev. 1 l

APPENDIX B BWROG HSIV LEAKAGE CLOSURE COMMITTEE TYPE 4 EXAMPLE NRC SUBMITTAL FOR INCREASING MSIV LETiK RATE LIMITS, EllHINATING REQUIREMENTS FOR LEAKAGE CONTROL SYSTEMS, EXEMPTION REQUEST TO 10CFR50, APPENDIX J EXEMPTION REQUEST 10 100FR100. APPENDIX A Type 4- B4-1

. NED0 31858 Rey, 1 U.S. Nuclear Regulatory Commission  !

Attn.: Document Control Desk Washington, DC. 20555 ,

RE: [PlantStationName)

DocketNo.'[xxyxx)

License No. [xxx)

L

= Gentlemen:  ;

+

[ Corporation) hereby transmits an application for amendment to the (plant Name) Facility Operating License [xxx), and applications for specific -

-exemption to 10CFR50 Appendix J and exemption to 100FR100 Appendix A.

[ Corporation) requests an amendment to the lechnical Specifications, set l

r

-forth in Appendix A to the License, to permit an increase in the allowable

. leak rate for the main steam isolation valves (MSIV) and to delete the MSIV Leakage Control System (LCS). Inaddition,(Corporation)-requeststhatMSIV ,

leakage be exempted from the Type A and Type C test acceptance criteria specified in the-' Appendix J of 10CFR50. Also, [ Corporation) requests that the downstream main ~ steam-piping and condenser be exempted.from the' seismic requirements specified in Appendix A of 10CFR100.  :

In support of the proposed changes to the Technical Specifications,

onclosed for the Nuclear Regulatory Commission's review are:  :

(1 ) Appilcation for Amendment to the Facility Operating License; Technical Specification pages affected by the-changes; and supporting ~information and' analyses addressing the changes. The-

-analyses demonstrate that the proposed changes do not involve a significant hazards consideration pursuant-to 10CFR50.92 - '

(2) -Application'for specific exemption to 10CFR50 Appendix J; and >

supporting information and justification for this exemption.

i Type 4' B4-2

.=. . - . . . . _ - - . , - -

Nf00 31858 Rev. 1 Pursuant to 10CTR50.12, this exemption request is authorized by law, will not present undue risks to the public health and safety, and is consistent with the common defen,e and security.

Furthermore, special circumstances are present which warrant issuance of this exemption request.

(3) Application for specific exemption to 10CfR100 Appendix A; and supporting information and justification for this exemption.

(Corporation) ealizes that there is no provision in 10CFR100 for exemption; however, the Nuclear Regulatory Commission has the authority to grant this exemption. This exemption will not present undue risks to the public health and safety, and is consistent with the common defense and security. Furthermore, special circumstances are present which warrant issuance of this exemption request.

These proposed changes are supported by DWR Owners' Group (BWROG) work.

The BWROG formed a MSIV Leakage Committee in 1982 in response to Generic Issue C 8 "HSIV Leakage end LCS failure". a1eric issue C 8 addressed the safety concerns that reported M51V leakages are too high and that the Leakage Control System will not function at high MSIV leakages. Based on the extensive, ongoing work performed by the BWROG to support resolution of the Generic Issue, the BWROG has developed the technical justification for the proposed Technical Specifications changes and associated exemption requests.

The General Electric (GE) report, NEDO-31858 Rev.1, describes the safety benefits and provides justification for the praposed changes. With regard to increasing leakage rate limits for the MSIVs, this will avoid unnecessary maintenance which has in the past adversely affected the or rability of the HSIVs.

With regard to the deletion of LCS, the proposed changes invoin a

- replacement of the existing LCS with the more reliable and effective main steam piping and condenser for MSIV leakage treatment. This treatment method is effective to reduce dose consequences of MSIV leakage over an expanded Type 4- B4-3

NED0 31858 Rev. 1 operating range and will, thereby, ren!se the safety concern that the LCS will not function at HSIV leakage rates higher than the LCS design capacity.

Except for the requirement to establish a proper flow path from the HSIVs to the condenser, the proposed method is passive and does not require any logic l control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for ,

limiting fission product release to the environment. Therefore, the proposed method is highly reliable and effective for HSIV leakage treatment.

Pursuant to 100fR50.9)(b)(1), (Corporation) has provided a copy of this license amendment request and the associated analysis regarding no significant hazards consideration (s) to the appropriate sta',e representative.

(Corporation) representativas will be available to discuss or meet with '

the Nuclear Regulatory Commission staff it your convenience to address this matter.

Very truly yours.

(CORPORATION)

INAME1 Vice President l

Enclosures j cc: Regional Administrator, Region (x)

(Name), Director (Name),ProjectManager (Name),Residentinspector l [Name), State Representative.

Type 4 B4-4 1

NE00 31858 Rev. 1 1

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of

[ Corporation) Docket No. [xxx]

[ Plant Name) bUlDMIT INamel .,__, being dly sworn, states that he is Vice President of

[ Corporation); that he is aut:. ized on the part of said Corporation to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that_ all such documents are true and correct to the best of his knowledge, information and belief.

1Signaturel Subscribed and sworn to before me, a Notary Public in and for the State of

[ ] and County of [ ), this [. ), day of ( ),

1988.

ISianaturel Notary Public in and for

[ County, State)

My Commission expires:

Type 4 B4-5 e

NE00+31858 Reu. 1 UNITED STATES NUCLLAR RECULA10RY COMMISSION In the Matter of

[ Corporation) Docket No. [xxx)

[PlantName) .

APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the Regulations of the Nuclear Regulat:Jry Commission, [ Corporation), holder of facility Operating License No. [xxx),

hereby requests the Technical Specifications, set forth in Appendix A to the license, be amended to permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs) and deletion of the MSIV Leakagt Control System (LCS). Specifically,-[ Corporation) requests that:

1. Allowable leak rate specified in the Technical Specifications section 3.6.1.2* be changed from [x] to [y] standard cubic foot per hour (scfh)permainsteamline. This proposed change reflects a higher, but still conservative allowable leak rate for the MSIVs.

2. Sectftn 3.6.1.2 to be administratively amended to add a footnote exempting MSIV leakages from the overall integrated leakage rate and from the combined leakage rate for all penetrations and all valves.

In concurrence with this appilcation for-license amendment,

[ Corporation) has applied for specific exemption to Appendix J of 10CFR Part 50.

i

, Type 4 B4-6 I

~ r - , - ,r---. -- g--

NED0 31856 Rev. 1

3. Sections [3/4.6.1.4]* be amended to permit the deletion of the LCS from the Technical Specifications. [ Corporation)proposesan alternate to Regulatory Guide 1.96, " Design of Hain Steam isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants." Utilizing the main steam lines and condenser, as an alternate method for MSIV leakage treatment, has been demonstrated to be more effective than the LCS in terms of reliability. In concurrence with this appilcation for license amendment.

[ Corporation) has applied for an exemption to 10CFR100, Appendix A.

4. Table 3.6.3 l* be amended to permit the deletion of the (CS isolation valves from-the Technical- Specifications. The LCS lines which j connect to the main steam lines will be disconnected and welded / cap closed to preserve containment integrity.

6. The index, Sections 3/4.6.1.5* thru A 3/4.6.1.8* and bases sections B3/4.6.1.4* through B3/4.6.1.9* be administrative 1y amended to rearrange section and page numbering.

The proposed changes to the Technical. Specifications are set forth in l Enclosure 1.

l The sections identified are based on the BWR Standard Technical Specifications for BWR/S (NVREG-0123). Each utility should revise these sections as applicable to their individual Technical Specificrtions.

Also, some plant sites may specify a maximun allowable HSIV leakage rate per line (i.e.,100 scfh), but require that the total is less than four times the maximum allowable leakage rate for each line (i.e., 200 scfh L

total).

l l

Type 4 B4-7 E

_ . . . _ _ _ __ _ _ _ _ - _ . _ - _ . _ _ _ _ _ _ _ _ _ . ~ . _

NED0 31858 Rey. I l The proposed changes are a result of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic Issue s L,

'MSly Leakage and LCS Failure". GE report NE00 31858 Rev. I provides 7 technical justification on a generic basis to support these proposed changes.

Based on the BWROG evaluation of MSIV leakage perform '.2, the current Technical Specification allowable MSIV leak rate is extremely limiting and routinely requires the repair and re-testing of MSIVs. This unnecessary repair significantly impacts the maintenance work load, often contributes to outage extensions, and has in the past adversely affected the operability of the MSIVs. BWR outage planners routinely schedule several days of contingency i to repair anc retest the MSIVs. In addition, the needless dose exposure to maintenance personnel is inconsistent with As low As Reasonably Achievable (ALARA) requirements. There have also been many Licensee Event Reports written for MSIV leakages exceeding the Techhical Specification limit.

The proposed changes will reduce unnecessary MSIV repair costs, avoid unnecessary dose exposure to maintenance personnel, reduce outage durations, and extend the effective service life of the MSlvs. 'n addition, the proposed increase in the MSIV leakage 1;mit ha;. P tential to significantly reduce recurring valve leakages, and minimize the possibility of needless repair which can compromise plant safety.

Based on BWROG evaluation, the LCS is extremely difficult to maintain-The system contains extensive logic and instrumentation, requiring frequent

-calibration to meet the Technical Specification requirements. The system has been declared inoperable, and required entry into the Limiting Condition of Operation (LCO' at least (Plant specific). Also, replacement of safety-related parts, such as blowers, heaters and flow elements, has become increasingly difficult, and lead times have been long.

The_ proposed changes involve a replacement of the existing LCS with the more reliable and effective main steam drain line and condenser for MSIV leakage treatr'it. This treatment method is effective to treat MSIV leakage Type 4 B4-8 j

NED0 31858 Rev. 1 over an expanded operating range. Except for the requirement to establish a proper flow path from the main steam isolation valves to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosopny of protection by multiple leak-tight barriers used in containment esign for limiting fission product reelease to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment.

The existirg leakage control systems have limitations for mitigating MSIV leakage. Operation of the system induces higher MSIV leakage by increasing the differential pressure across the valves, and is inconsistent with the philosophy of multiple barriers for limiting fission product relea as e joes L not take advantage of the outboard MSIVs). [For positive pressure LCS, operation of the system increases containment pressure and thereby increases the containment leakage.)* The ICS requires nultiple logic controls, interlocks, timers, containment isolation valves, and other equipment to ensure containment integrity and " otection from the high prenure inain steam lines. Bar9d on plant operating experience, the LCS does not provide a high degree of reliability. Also the LCS has limited capacity and does not function at moderate MSIV leakage rates above 100 scfh.

Even though the resulting off-site doses may te slightly higher for low MSIV leakage rates, the effectiveness of the preposed method even for leakage rates greater than the proposed increased allowable limits, ensures offsite does limits to the public are not exceeded. Overall, the propcted trealment method can haadle MSIV leakage over an expanded operating range, and will thereby resolve the safety concerr, that the LCS will not function at MSIV leakage rates higher than the LCS capacity. Thus, a margin of safety exists.

Furthermore, it is clearly a safety improvement to replace a system with know limitations with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability.

Plant-specific supporting information and results of the radiological an:1ysis that justify the proposed changes arn included as Enclosure 2. As Type 4 B4 9

N(00 31858 Rev. 1 )

l ll concluded in the supporting information, the hereased MSIV allowable leak rate of [y] scfh and the deletion of the LCS, will not adversely affect the performance of the primary containment inlation function. A plant specific radiological analysis has been performed utilizing the main condenser as an alternate treatment path for MSly leakage. This radiological analysis demonstrates that the proposed changes result in an insignificant increase to the dose exposures previously calculated for a design basis Loss-of Coolant Accident (LOCA). The revised LOCA dose exposures remain well within the guidelines of 10CFR100 for the off-site doses and 10CFR50, Appendix A (General Design Criteria 19) for the control roc.n ocses.

[ Corporation) will institute into the MSly maintenance and test program, i the requirement that any MSIV exceeding the proposed [y] scfh limit, will be repaired and re tested to meet a leakage rate of less than or equal to [x]

scfh. This will assure cot.tinuation of high quality repair and refurbishment efforts to improve the overall performance and reliability of the MSIVs.

The proposed change will require a minor plant design change to allow specified main steam drain valves to be opened even if off-site power is not available. .Solicabic normal plant operating procedures, and emergency operatine rir. -dures shall be reviewed and revised accordingly.*

Upt " w roval of the proposed LCS deletion from the Technical Specifications, [ Corporation) will disconnect [ remove) the LCS. Lines ,

connecting to the main steam lines will be disconnected and welded / rap closed such that containment integrity is maintained (plant specific). In lieu of the safety-related LCS, MSIV leakage will be transported by the main steam drain lines to the condcnser-and processed as discussed in Section 4.3 of NED0 31858 Rev. 1.

Plant-specific

-Type 4 B4-10

NED0 31858 Rev. I furthermore,[ Corporation)willincorporatetheapplicabicalternate leakage treatment methods, consistent with GE do.ument NE00 30324 " Potential Operator Actions to Control MSlV Leakage,* into the Operational Procedures and Emergency Operational Procedures at (Plant Name).

[ Corporation) will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NED0-31858 Rev. 1, to provide reasonable assurance of the structural integrity of these components.

Pursuant to 10CFR50.92, an analysis which demonstrates that the proposed changes do not involve a significant hazards consideration, is included as Enclosure 3. The proposed changes have been reviewed in accordance with Section 6.5 of the Technical Specifications. The proposed changes will not authorize any change in the types of effluents or in the authorized power level of the facility. Enclosures 4 and 5 are applications for exemption to 1 Appendix J of 10CFR50 and Appendix A of 10CfR100, respectively.

WHEREFORE, Applicant respectfully requests that Appendix A to the facility Operating License No. [xxx) thereto be amended in the form attached hereto as Enclosure 1.

[ CORPORATION) ,

4 Dy:

Vice President Subscribed and sworn to before me on t..is day of _ .

lype 4 B4-ll l

n J

NE00 31858 Rev. 1 i IRC1DSMRL1 RORE0MI. LOR 1 LLCENSE NO. Ixxx]

DOCKET NO. Ixxx1

\

PROPOSED CHANGES TO TECHNICAL SPECIF' O.10ft$

Replace the following nages with the attached revised page(s)*. These pages are provided in its entirety with marginal marking to indicate the changes.

1. Page vii

2. Page xiii

3. . Page 3/4 6-2

4. Page3/4 63

5. Page 3/4 6-7 >

6. Page3/4 6 8a

7. Page3/4 69

8. Page 3/4 6 10

9. Page 3/4 6 11

10. Page 3/4 6 12

11. Page3/4 6-22

12. Page B 3/4 6-1

13. Page B 3/4 6 2

14. Page B 3/4 6 3 The pages identified here are based on the BWR Standard Technical Specifications for BWR/5 (NUREG 0123). Each utility should revise these pages as applicable to their individual Technical Specifications.

l l

Type.4 B4-12

NED0 33858 Rey, 1 IE.LQMRL2 1 CORPORA 110N1 LICENSE NO. lxx11 DOCKET NO. lxxxl l

, SMPPORTING INFORMATION AND ANALYSES

1.0 INTRODUCTION

AND SM MARY OF RESUlls i

The proposed Technical Spectfication amendment involves an increase in the allowable leakage rate from [x] scfh to [y] scfh per main steam line.

-deletion of the LCS, and exemption of.the MSIV leakages from the Type A -

and Type C acceptance criteria specified in Appendix J of 10CFR50. In '

addition it is requested that downstream main steam piping and condenser i be exempted from the seismic requirements specified in Appendix A to 100fR100. ,

i Section 2.0 of this Enclosure provides a summary of background information: Section 3.0 discusses the justifications for the proposed changes;. Section 4.0 provides a summary of the plant specific 1 radiological dose assessment, and Section-5.0 summarizes the potential benefits for a Technical Specification.MSIV allowable leak rate of [y]

scfh, and the deletion of the LCS.

The BWROG report,.NED0 31858'Rev. 1, "BWROG Report for increasing-MSIV

-Leak. Rate Limits and Elimination of Leakage Control Systems," November 1991, provides the justification for increasing MSIV leakage limits, and for' eliminating the requirements for the LCS, With concurreneti from the

valve manufacturers, this report concludes that MSIV leakage rates up to 3 1200 scfh 'are not an indication of substantial mechanical' defects in the l valve which would challenge the isolation-capability of the valve to -

fulfill its safety function. Therefore, the proposed increase in the fallowableleakagerateto-[y]-scfhfortheMSIVswill-not'inhibitthe ,

isolation capability of the valve.

l-  ?

Type 4 -B4-13

{ g- y- p g '

p-i+ g m-e.-,-4., n m 7_y~ --y- p yg-g.-t.. y y

,,9p,- p,y.3-min _t . yrePg-.w e. .iyw m,e, arr-g-ma y- ee.,5e,an, -w-vewm.-----_,, w,i.

NEDO 31858 Rev. 1 The BWROG has evaluated several methods and has recommended the isolated condenree as an alternate method to the LCS for MSIV leakage treatment.

The l ulated condenser method takes advantage of the large volume in the '

main steam lines and the condenser to hold up the release of fission products leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser. [ Corporation) proposes to delete the LCS from the Technical Specification and to incorporate the isolated condenser as an alternate method for MSIV leakage treatment.

The BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage. The BWROG have reviewed the potential combinations of loss Of-Coolant Accidents (LOCAs) and seismic events of interest:

(1) LOCA WITHOUT NEAR COINCIDENT SEISMIC EVENT. For this occurrence, the pressure in the piping system downstream of the HSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the alternate flow path through the -

main steam system piping to the condenser is assured.

(2) SEISMit EVENT WITHOUT NEAR COINCIDENT LOCa. Without a LOCA and the potential associated core degradation, the radioactivity transported with MSIV leakege is of no radiological significance.

(3) LOCA WITH NEAR COLNCIDENT SEISMIC EVENT. for this occurrence (also assuming significant core damage) the consequences are of interest because a seismic induced failure in the main steam or

' condenser system could allow MSIV leakage to bypass the alternate treatment pathway. It has been previously well

documented that the probability of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability is approximately 0.001 per reactor per year; core l

Type 4 B4-14 l

l _ __ _, _

. ~ _ _ _ - . _ _ _ . _ -_._ _ _. _ _ . . _ . -

NEDO 31858 Rev. I melt probability is plant specific and typically ranges from 0.00001 to 0.0001 por rm. tor per year). It is also noted that  !

a LOCA does not induce a seismic event, and that a seismic event has a very low probability of causing a LOCA because the primary pressure boundary and (mo lency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 i

reported probability of seismic induced LOCA to be less than 5 x {'

10 7 per reactor per year).

1 Considering that the probability of a near coincident LOCA and seismic j event is much smaller than other plant safety risks (less than 1 x 10 7 l per reactor per year for coincident events, less than 5 x 10*7 per )

reactor per year for seismic induced LOCA), the concern for main steam piping or condenser. damage is of little significance. Nevertheless, because ma'n steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design basis seismic events.

ANSI 831.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin. In addition, specific seismic design provisions have been

• incorporated into some newer BWR main steam and condenser systems.

In order to further justify _the capability of the main steam system piping and condenser alternate treatrent pathway, the BWROG has reviewed limited earthquake experience data on the performance of non seismically designed piping and condensers (in past earthquakes). The study summarizes data on the performance of main steam piping and condensers in past strong motion earthquakes and compares these piping and condensers with those in typical U.S. GE Mark I, II, and 111 nuclear plants. This limited earthquake experience data and similarity comparisons are Type 4_ B4-15

NEDO-31858 Rev. I then used to further strengthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in a design basis earthquake in conjunction with a LOCA occurring just prior to or after the seismic event.

The earthquake experience data are derived from an extensive database on the performance of industria' facilities and over 100 power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world from 1934 to the present.

The piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, egen when they are not designed to resist earthquakes. This is a common conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipment such as motor control centers, pumps, valves, structures, and so forth. That is, with limited exceptions, normal industrial construction and equipment typically have substantial inherent seismic ,

ruggedness, even when they are not designed for earthquakes. No failures of main steam piping were found. Anchored condensers have also performed well in past earthquaku 6.ith damage limited to minor internal tube leakage.

Comparisons of piping and condenser dt. sign in example GE Mark I, II, and 111 -lants with those in the earthquake experience database reveal the GE plant designs are similar to or more rugged than those that exhibited good earthquake performance. In addition, [ Corporation) will also perform a verification of seismic adequacy of the main steam piping und condenser, consistent with the guidelines discussed in Section 6.7 of NEDO-31858 Rev. 1, to provide reasonable assurance of the structural integrity of these-components. The BWROG concludes that (1) the possibility of significant failure in GE BWR main steam piping or l Type 4 B4-16

NED0 31858 Rev. I condensers in the event of an eastern U.S. design basis earthquake is highly unlikely and (2) any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

The design basis LOCA has been re-aulyzed for radiological impacts utilizing the isolated condenser as an alternate method for MSIV leakage treatment. The analysis demonstrates that a maximum MSIV leakage rate of

[y] scfh per main steam line results in an acceptable increase to the dose exposures calculated for a design basis LOCA. In addition, the analysis demonstrates that MSIV leakage rates of approximately [z] scfh per main steam line will not result in dose exposures in excess of the regulatory limits.

2.0 BACKGROUND

The safety function of the HSIVs is to isolate the reactor system in the event of a LOCA or other events requiring containment isolation. The design of the MSIVs and its isolation requirements are described in Sections [5.4.5*] and [6.2.4*] of the Final Safety Analysis Report (FSAR). The allowable leakage rate from MSIVs is included in the LOCA radiological analysis evaluated in Section [15.6.5*] of the FSAR.

The safety-related LCS is designed to direct MSly leakages to the reactor building Standby Gas Treatment System (SGTS) for treatment of fission products. The LCS is described in Section 6.7** of the FSAR, and consists of inboard and outboard subsystems. The inboard subsystem contains isolation valves which provide for containment integrity in the event that the corresponding MSIV fails to close.

1 3

.. Plant specific elant Unique. This draft describe, the negative pressure LCS.

Type 4 B4-17

NED0 31858 Rev. 1 Operating experience indicates that MSIVs frequently exceed the Technical Specification allowable leak rates. Some of these valves repeatedly feil the local leak rate tests despite frequent disassembly and refurbishment.

As a result of increasing MSIV leakages and the inability of the LCS function at high MSIV leakages, the Nuclear Regulatory Commission prioritized Generic issue C 8, *MSly Leakage and LCS failure" as a high ,

priority itern in January 1983. This issue was closed in 1990.

The BWROG formed a MSIV Laakage Committee in 1982 to address the increasing MSIV leakage rates, and a follow on MSIV Leakage Closure Committee in 1986 to address alternate actions to resolve on going, but less severe MSly problems. The MS1V Leakage Committee identified contributors which cause MSIVs to fail the leak rate tests by large margins, developed recommendations to minimize leakages, evaluated alternates for MSlV leakage treatments, and compiled recent history of MSIV leakages and LCS operating experience.

3.0 JUSTIFICATIONS FOR THE PROPOSED CHANGE

[ Corporation) proposes to increase the Technical Specification allowable leakage rate for the MSIVs frra [x] scfh to [y] scfh per main steam line and to delete the LCS requirements from the Technical Specifications.

The current Technical. Specification MSIV Icak rete limit is extremely limiting and routinely requires unnecessary repair and re-test of the MSIVs. This significantly impacts the maintenance work load during plant Type'4 B4-18

NED0 31858 Rev. 1 i

outages and often contributes to outage extensions. (The outage planners at[plantstationname)typicallyschedulesoveraldaysofcontingencyto repair and re test the HSIVs)* In addition, the needless dose exposures to maintenance personnel are inconsistent with As low As Reasona'ly Achievable (ALARA) principles. (There have been many Licensee Event I Report written for MSIVs failing to meet the current Technical Specification limit.)*

From a safety perspective, calculations using standard conservative assumptions for considering the off site consequences of a postulated design basis LOCA r.onfirm that off-site and control room doses will be j within the regulatory guidelines for the allowable MSIV leakage rate.

This ca,culation is described in Section 15.6.5 for the FSAR. However, if MSIV leakages are only moderately higher than the allowable limit, the calculated doses will exceed the regulatory guidelines. Furthermore, as documented in Generic issue C 8, the LCS will not function if MSIV exceeds the leakage limit by a moderate amount.

hn V's failure to meet the current Technical Specification limit have been documented in respunse to surveys conducted by the Nuclear Regulatory Commission during the early 1980 and by the BWROG during the middle and late 1980s. As high as 50% of the total "as found" HSIV local leak rate tests were reported in the early NRC survey to exceed the leakage rate limit.

The DWROG has studied the issues regarding MSIV leakage rates, their causes, and available alternatives. The results of the BWROG study are provided in NED0-31858 Rev. I and are also summarized in NUREG-ll69. In response to Generic Issue C 8, the BWROG has recommended corrective actions and maintenance practices to reduce the HSIV leakage rates.

Plant Unique Type 4 B4-19

.-_e_ . . . _ . a y 9,t. ,,_,i

NED0 31858 Rev. 1 A recent survey conducted by the BWROG of HSIV leakage tests performed between 1984 and 1988 ir.dicates that the impicmentation of industry and BWROG actions has been effective in reducing the leakage rates, and, in particular, a reduction in the number of valves which experience substantial high leakage rates. The survey concludes that about 23% of the total *as found" HSIV leakages still exceed the limit of 11.5 scfh and about 10% exceed 100 scfh.

The HSIV leakage performance at (plant station name) it .....[to be provided by the Utility). The leakage performance at (plant station name) is consistent with the recent survey by the BWROG.

Despite the recent improvement in leakage performance, MSIV leakage rates still frequently exceed the current Technical Specification limit and the safety and maintenance problems related to high HSIV leakage rates, although less severe, remain as a significant issue.

Furthermore, based on the extensive evaluation of valve leakage data, the BWROG has found that disassembling and refurbishing the HSIVs to meet very low leakage limits frequently contribute to repeating failure. By not having to disassemble the valves and refurbish them for minor leakage, the utility may avoid introducing one of the root causes of recurring valve leakage problems that may lead to later leak test failures and the possibility of compromising plant safety.

The current Technical Specification allowable leakage rate is established by excessively conservative LOCA radiological analysis as described in Section 15.6.5 of the FSAR. The valve % physical size and operability characteristics (large size and fast acting), and existing turbine building equipment were not considered at the time the leakage limit was established. Based on the in-depth evaluation of MSIV leakages, the BWROG has concluded the HSly leakage rates up to 500 scfh are not an indication of substantial mechanical defects in the valve which would Type 4 B4-20

NEDO 31858 Rev. I challenge the isolation capability of the valve to fulfill its safety function. Furthermore, valve manufacturers hr.ve stated that leakage rates up to 200 scfh can occur withnut having a major valve defect.

Therefore, the proposed increase from [x] to [y] scfh per main steam line will not inhibit the MSIV's performance of the isolation function and will not compromise the safety of [ plant station name).

This proposed increase provides a more realistic, but still conservative, limit for the MSIVs. Based on the BWROG study, the proposed increase in the allowable leak rate will increase the chance for a successful local leak rate test to greater than 90%, up from the 77% success rate at the current typical liHt of 11.5 scfh, This increase in successful local leak rats testing will significantly reduce MSly maintenance costs, reduce dose exposure to maintenance personnel, reduce outage durations, extend effective service life of the MSIVs, and minimize the potential for outage extensions at [ plant station name).

A safety releted LCS was required by Regulatory Guide 1.96 in order to reduce the radiological consequences of MSIV leakage. As discussed earlier, Generic Issue C-8 identified the safety concern that MSIV leakage rates, as determined by conservative local leak rate tests, were too high and that the LCS would not function at high MSly leakage rates.

The 1981 NRC survey indicated that 33 percent of the total tests exceeded leakage rates of 100 scfh. Since the process capability of the LCS at

[ plant station name) is designed for MSIV leakage rates of no more than 100 scfh, the potential existed for the LCS not to function as analyzed for a design basis LOCA as described in [S9ction 15.6.5)* of the FSAR.

Consequently, the conservatively calculated dose contribution from MSIV leakage would exceed the regulatory limits for off-site and control room doses.

Plant-specific Type 4 B4-21

NEDO 31858 Rev. 1

[ Corporation) proposes to delete the LCS requirements from the Technical Specifications. The proposed changes involve a replacement of the existing LCS with the more reliabic and effective main steam drain line and condenser for HSIV leakage treatment. This treatment method is effective to treat HSly leakage over an expanded operating range. Except for the requirement to establish a proper flow path from the main steam drain line to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for HS!V leakage treatment.

The existing LCS has limitations for mitigating MSly leakage. Operating the system induces higher MSlV leakage by increasing the differential pressure across the valves, and is inconsistent with the philosophy of multiple barriers for limiting fission roduct releases (does not take tdvantage of the outboard MSIVs). [For ositive-pressure LCS, operating the system increases the containment pressure and thereby increases the containment leakage.1* The LCS requires multiple logic controls, interlocks, timers, ct..tainment isolation valves, and other equipment to ensure containment integrity and protection from the high pressure main steam lines. Based on plant operating experience, the LCS does not provide a high degree of reliability. Also the LCS has limited capacity and does not function at moderate HSIV leakage rates above 100 scfh.

Even though the resulting off-site doses may be slightly higher for low NSIV leakage rates, the effectiveness of the proposed method even for leakage rates greater than the proposed increased allowable limits, ensures off-site dose limits to the public are not exceeded. Overall, the proposed treatment method can handle HSIV leakage over an expanded operating range, and will thereby resolve the safety concern that the LCS will not function at HSIV leakage rates higher than the LCS capacity. l ll Type 4 B4-22 1

NED0-31858 Rev. I am sin u ety exists. Furthermore, it is clearly a safety I'  :. en c t ' + .. ce I system with known limitations with the alternate s- 3'p ag and condenser treatment pathway, which has been shown

' ca' lent reliability.

[

3

~

. v- nion) t it Pus porate the applicable alte

aate

}' ,

en+ i r. hods, c.W f.ent with GF docutaent NED0-30324

. ar ier Actions a control MSIV Leakage", into the Opt -

Proredures and Emergency 0parational Procedures.

t h

In W6ttion <o improving plaia safety, the proposed deletion of the LCS reemrements in the Technical Specifications will result in significant operational and maint enance benefits. The LCC has extensive logic and instrumentation ar.G libration of this instrumentatx. is frequently required to meet tiie Technical Specification requirements. The BWROG has evaluated recent LCS performance dats. n its of this evaluation is summarize. n NE00-31858 Rev, 1. The evaluation indicates that the LCS is extremeij difficult ta raintain and that plant shutdowns and start-up delays frequently occur. The LCS performance at [ plant station name) is

.....[to oe provided by the Utility). The LCS performar.c? at [ plant station riame) is consistent with the recent survey by the BWROG. -

\

In conclusion, [ Corporation] proposes to increase the Technical Specification allowable HSIV leakage rate fro:a tx] to [y]  % 1ain -

ster line, and to eliminate the requireaents for LCS in tt. .echnical

!pecifiestions. The preposed increase in the MSIV leakage limit should significantly reduce recurring valve leakage problems, and will minimize neeoless valve repair which can compromise plant safety. The proposed deletion of the safety-related LCS and the pro;,osed alternate method (main steam. lines and condenser) resol;e the safety concern regarding LCS effectiveness at higher MSIV leakage cate.

Type 4 B4-23

NED0-31858 Rev. I l l

4.0 ANALYSIS OF MSIV LEAKAGL(ONTRIBUTION TO RADIOLOGICAL DOSE CALCVLATIONS 4.1 Selection'of Alternate leakaae Treatment Method The BWR0G has evaluated several alternate MSIV leakage treatment methods and has recommended the isolated condenser for MSIV leakage treatment.

This leakage treatment method takes advantage of the large volume in the isolated main condenser to hold up the release of fission products leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser.

As previously discussed in Section 1.0, the BWROG has evaluated the availability of main steam system piping and condenscr alternate treatment pathways for processing MSIV leakage. The BWROG has determined that the prc' ability of a near coincident LOCA and a seismic event is much smaller than for other plant safety r'sks. The BWROG has also determined that main ste&m piping and condenser designs are extremely rugged, and that 'the ANSI-B31.1 design requirement- typically used for nuclear plant system design contain a good deal of margin. In order to further justify the capability of the main steam piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes). This study concluded the possibility of a failure which could cause a loss of steam or condensate in BWR main steam pipine ondens'ers in the event of a design basis earthquake is highly unl; e N ,d that such a failure would also be contrary to a large body of :.istorical earthquake experier.ce data, and thus unprecedented.

This conclusion is consistent with NUREG/CR 4407, " Pipe Break Frequency Estimation for Nuclear Power Plants", dated May 1987, which reported no observed failures in the main steam pipi g over 313 years of reactor operating years.

l l Therefore, the isolated condenser alternate MSIV leakage treatment path l-l Type 4 B4-24

NED0 31358 Rev. I l at ,[ plant name) is considered appropriate for the reduction of

radiological consequences of a design basis LOCA.

1 l

(4.21 RadioloaiceLAnalysis and Result The radiological dose methodology has been developed by General Electric

.for the BWROG.

The radiological analysis calculates the effects of the proposed allowable' MSIV leak rate in terms 'of control room and off-site doses.  ;

_ The revised LOCA doses are the sum of the LOCA doses (as described in  !

iSection xxx of the FSAR) and the calculated MSIV leakage doses. This method of calculating the revised dose exposures is very conservative

. since the' LOCA'dcses already include the dose contribution from MSIVs at the maximum leaktpe rate permitted in the cur rent Technical

~Specifir-tions.*

' Table 1** shows the calculated dose exposures from the BWROG radiological- '

analysis for [ plant name). Regulatory limits and' calculated doses from LOCA radiological; analysis are'also included in Table 1 for comparison

~

purpose.- -This analysis demonstrates that-a MSIV leakage-rate of_ 200 sefh per mainisteam line-results in an acceptable increase to the dose exposures- previously calculated'for the control-room,- EAB, and the-LPZ.

The revised LOCA doses remain well within the guidelines _ of-10CFRlde for off-site doses and 10CFR50, Appendix A, (General Design Criteria 19) for

~

the control--room doses. Furthermore, the calculation shows that MSIV ,

1.eakageL rates up to approximately [z] scfh per steam line would not

-axceed4he regulatory limits.- Therefore, the proposed method provides a (substantial- safety margin for mitigating the radiological consequences of-MSIV leakage beyond the proposed Technical ~ Specification leak rate limit of=[y] scfh.

t

' Plant-Speci fic. : Othet BWRs may elect to replace- the LOCA MSIV dose contribution with the new doses ~using the alternate leakage treatment

.u method, '

ihe' following. discussion' on doses are based on as an example plant -(Hope -

Creek)-

LType 4- B4-25 4

NED0 31858 Rev. 1 l Table 1 CCNTRIBUTION TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV LEAK RATE OF 200 scfh HOPE CREEK GENERATlhG STATION Whole Body Thyroid Beta '

frem) _Jrem) Ltgm).

Exclusion Area A) 10CFR 100 Limit 25 300 Boundary (2-Hour) B) Previous Calculated 0.6 76.7 Doses **

C) Contribution from 0.1 2.6 MSIVs at 200 scfh D) New Calculated Doses 0.7 79.3 Low Population A) 10CFR 100 Limit 25 300 Zone (30-Day)

B) Previous Calculated 0.08 7.7 Doses **

C) Contribution From 0.34 57.5 MSIV at 200 scfh D) New Calculated Doses 0.42 65.2 Control Room A) GDC-19 5 30 30/75***

(30-Day)

B) Previous Chlculated 0.04 0.26 .91 Doses **

C) Contribution From 0.10 2.71 1.64 MSIVs at 200 scfh D) New Calculated Doses 0.14 2.97 2.55

• No limit specified.

FSAR Sectinn 15.6.5.5 and 6.4 (includes MSIV leak rate at a total of 45 scfh for the first 20 minutes; control room dose assuues 100% per day reactor building inleakhge).

• • • 75 if prior commitment has been made to use protective clcthing.

1 Type 4 B4-26 I

C NED0-31858 Rev. 1

-5.0; BENEFITS FOR THE PROPOSED CHANGES As discussed in NEDO-31858 Rev. 1, recent MSIV leakage performance has

=-

significantly improved since the early BWROG survey in 1984 and the NRC survey in the early 1980s. Despite the recent improvement, MSIV leakage

.n rates exceeding the current Technical Specification limits still ,

frequently occur. The BWROG evaluation of f he recent MSIV leakage performance concludes that the proposed change will improve the chance for a successful local leak rate test to greater than 90%, up from the ,

77% success rats at the current Technical Specification limit of 11.5 scfh..

Specifically, MSIV leakage experience at [ plant- station name) are [***to be provided by utility on plant-specific leakage performance and any problems which may impact critical path, outage extensions, etc.***].

As discussed in-detaillin Section 3.0, deleting the LCS will reduce the overall dose rates, and. eliminate the system's impact on refueling and maintenance outage activities at (Plant Name]. [*** Provide specific plant data for supr rt ***]. The proposed alternate method (main steam lines- and condenser) for MSIV leakage treatment will also eliminate the safety concern regarding LCS effectiveness at higher MSIV leakaga rates.

Although:the revised LOCA doses are slightly higher for low MSIV leakage-rate, theEeffectiveness of the proposed method,:even for leakage rates ,

l greater than the proposed increased MSIV allowable leak rate,; ensures that off-site and control room doses are not exceeded.

V L

=

L

- Type 41 84-27 L

NED0-31858 Rev. I j ENCLOSURE 3-ICORPORAT10N1 LICENSE NO. Ixxxl DOCKET NO. Ixxx1

-NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS (Corporation) proposes an amendment to the Tec.hnical Specifications as Lfollows:

L1) - Revise Section 3.6.1.2* to permit an increase in the allowable leak rats for the main steam isolation valves (MSIVs) from the current [x]

standard cubic feet per hour (scfh) to-[y] scfh per main steam 'ine.

2) Delete-Sections 3/4.6.1.4* and B3/4.6.1.4* to reflect the deletion of the _ Leakage' Control System (LCS) from the Technical Specifications.

3) Revise Table 3.6.3-l* to delete the inboard LCS isolation valves

[ plant unique identification).

4) ' Revise Section 3.6.1.2* to add'a footnote stating that the MSIV leakage'is exempted from the. acceptance criteria of Appendix J of 10CFR50.

-5)' Revise Index;pages"vii and xiii, renumber Section; 3.6.1.5* through 3.6.1.8*, 4.6.1.5* - through 4.6.1.8*, pages' number 3/4 6-Ba* througij-

-3/4 6-12*, and Sections 83/4'.6.1.5* through B3/4.6.1.9*'as a result' of deleting Sections-3.6.1.4* and.83/4.6.1.4*,

The proposed amendment to add a footnote to Section 3.6.'l.2* clarifies that.

1

-MSIV leakages are exempted from the overall integrated _ containment-leakage rate limit and from the combined local leak rate limit as specified in the itest acceptance criteria'

in Appendix J of 10CFR50. As defined in'the Bases.

"Section 3/4.6.1.2*, of the Technical Specifications, the purpose of these Type 4 B4-28

NE00 31858 Rev. I requirement!, is to ensure that the overall integrated containment leekage or the combined leakages from individual containment penetrations will not exceed the designed containment leak rate assumed in the design-basis Loss-Of-Coolant Accident (LOCA) radiological analysis. Since an allowable leak rate is specifically allocated for the HSIVs in the Technical Specifications, and that the radiological analysis has been revised to analyze MSly leakage path separately from those of the containment leakage rates, the proposed exemption is, therefore, appropriate and justified. This proposed change does not exempt the MSIVs from the test schedules as required in the Technical Specifications and 10CFR50 Appendix J.

This application also provides a detailed justification for exempting the MSIV leakages from the acceptance criteria of 10CFR50 Appendix J, and demonstrates that the proposed exemption will not present an undue risk to the public health and safety. Therefore, based on the above consideration, :he panposed amendment to add the footnote tc '.'nttion 3.6.1.2* is considered an administrative change.

Pursuar.t to ' 150.92, the proposed amendment involves no significant hazards considerations.

The operation of IPlant Station Namel. in ar.ardance with the pronosed .

amendment. will not involve a sianificant increase in the probablity or ccnsecuences of__an accident previously evaluated.

The proposed amendment to Section 3.6.1.2 does not involve a change to structures, components, or systems that would affect the probability of an accident previously evaluated in the Final Safety Analysis Report (FSAR).

Plant-specific Type 4 B4-29

.._._y____.__.._.

-NED0-31858'Rev. 1.

.i The; proposed. amendment to delete Sections 3/4.6.1.4 and Bases Sections  :

~

B3/4.6.1.4 involves eliminating the LCS requirements from the Technical

_ Specifications. As described in Section 6.7 of the FSAR, the LCS is manually initiated in about.20 minutes following a design-basis LOCA. Since the LCS is operated.only after an accident has occurred, this proposed amendment has no effect on the probability of:an accident.

i Since MSIV leakage and operation of the LCS are included in the radiological -

analysis for the design-basis LOCA as described in Section 15.6.5 of the FSAR,-

the proposed amendments will not affect the precursors of other analyzed accidents.- The- proposed amendments result in acceptable radiological consequences of the design-basis LOCA previously evaluated in Section 15.6.5

-of the FSAR.

The (plant station name] has an inherent MSIV leakage treatment capability.

-[ Corporation) proposes to= use the main steam lines and condenser as an alternate to Regulatory Guide 1.96 " Design of Main Steam Isolation Valve Leakage Control System For Boiling Water Nuclear Power Plants

• for MSIV leakage treatment. .(Corporation] will incorporate this alternate method in the-Operational 4 Procedures and Emergency Operational Procedures- .

The BWROG has evaluatedLthe availability _of main steam system piping and-condenser alternate iireatment pathways for processing MSIV leakage, and has determinedithat the probability of a near coincident LOCA and a: seismic event

is much smaller'than for other plant safety _ risks. The BWROG has also

~

a . determined that main steam piping and conderser designs are extremely rugged, and that the ANSI-831.'1 design; requirements: typically used for nuclear: plant -

system design contain~ a good deal'of margin.

In order to further-justify the capability of the main steam piping-and~

condenser alternate treatment pathway,- the BWROG has reviewed limited earthquake. experience data on the_ performance of. non-seismically designed:

piping and condensers. (in past earthquakes). This study concluded that;the.

Type 4- B4-30 T - - v.- - - - ----------___---__L__'-___-___.'_--..

NED0-31858 Rev. 1 o

possibility of a failure which could cause a loss of steam or condensate in BWR main steam piping or condensors in the event of a design basis earthquake

-is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

[ Corporation) will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NEDO-31858 Rev. 1, to provide reasonable assurance of the structural integrity of these components.

A plant-specific radiological analysis has been performed to assess the j effects of the proposed increase to che allowable MSIV leak rate in terms of  !

control room and off-site doses following a postulated design basis LOCA.

This analysis utilizes the hold-up volumes of the main steam piping and condenser as an alternate method for the MSIV leakages. At discussed earlier, there is reasonable assurance that the main steam pioing and condenser will remain intact following a design basis earthquake. The radiological analysis uses standard conservative assumptions for the release of source terms consistent with Regulatory Guide 1.3 Revision 2, " Assumptions Used for Evalueting the Potential Radiological Consequences of a Loss-Of-Coolant Accident for Boiling Water Reactors", dated April 1974.

The analysis-demonstrates that dose contributions from the prcposed MSIV leakage rate limit of [y] scfh and from the proposed deletion of the LCS Tesult in an acceptable increase to tLi LOCA doses previously evaluated against the regulatory guidelines for the off-site doses and control room

-doses as contained ir. 10CFR100 and 10CFR50, Appendix A (General Design Criteria 19), respectively. The LOCA doses previously evaluated are discussed in Section [15.6.5)* of the FSAR. The revised LOCA doses are the sum of the LOCA doses previously evaluated in the FSA3 and the additional MSIV doses calculated using the ai:.ernate treatment method. This method of calcul.ating Plant-specific Type 4 B4-31

NED0-31858 Rev. I the revised doses is very conservative since the LOCA doses previous evaluated already included dose contributions from MSIV at the maximum leakcge rate permitted in the current Technical Specification!..

• The attached table shows the previous calculated doses and the new calculated doses, The whole body dose at the Low Population Zone (LPZ) and the control room is itcreased from 0.08 to 0,42 rem ** and from 0.04 to 0.14 rem, respectively.

These increases are acceptable because the revised doses are well within the Regulator,/ guidelines (0.42 versus 25 rem at the LPZ, and 0.14 versus 5 rem at the control room), The associated whole body dose at the exclusion area boundary (EAB) increased insignificantly from 0,6 to 0.7 rem.

Tho thyroid dose at the LPZ increased from 7.7 to 65,2 rem. This increase is acceptable because the revised dose of 65,2 rem is significantly less than the regulatory guideline (300 rem). The EAB thyroid dose increased slightly from 76.7 to 79,3 rem, whereas the control room thyroid dose increased from 0.26 to 2.97 rem. The increase in control room thyroid dose is acceptable because the revised dose remains a small fraction (9.9%) of the limit. The control room beta dose is. increased from 0.91 to 2.55 rem, which remains insignificant relative to the regulatory guideline of 30 rem, it-is important to note that the resulting doses are dominated by the organic iodine fractions which occur because of the ultraconservative source term assumptions used in this analysis. For 200 scfh per steam line, more than 85%

of the off-sito iodine and cont al room doses are due to organic iodine from the RG 1.3 source term and organic iodine converted from the elemental iodine deposited in main steam piping systems, if the actual iodine composition from the fuel release (cesium iodide) is used in the calcu; 'ons, essentially all of this organic iodine dose would be eliminater.,

Plant-specific, original LOCA dose from MSIV leakaae may be replaced with revised dose.

• • ~ All doses and comparisons indicated here-on are Hope Creek specific.

Ty;a 4 B4-32 l

NEDO-31858 Rev. 1 CONTRIBUTION TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV LEAK RATE Oi 200 scfh HOPE CREEK GENERATING STATION Whole Body Thyroid Beta IIff!!1 (rem) IIcm).

Exclusion Area A) 10CFR 100 Limit 25 300

• Boundary (2-Hour) B) Previous Calculated 0.6 76.7 Deses**

C) Contribution from 0.1 2.6 MSIVs at 200 scfh D) New Calculated Doses 0.7 79.3 Low Population A) 10CFR 100 Limit 25 300

• Zona (30-Day)

B) Previous Calculated 0.08 7.7 Doses **

C) Contribution From 0.34 57.5 MSIV at 200 scfh D) New Calculated Doses 0.42 65.2 Control Room A) GDC-19 5 30- 30/75***

(30-Day)

B) Previous Calculated 0.04 0.26 .91 Doses **

C) Contribution From 0.10 2.71 1.64 MSIVs at 200 scfh D) New Calculated Doses 0.14 2.9) 2.55

• No limit specified.

FSAR Section 15.6.5.5 and 6.4 (includes MSIV leak rate at a total of 45 scfh for the first 20 minutes; control room dose assumes 100% per day >

reactor building ir. leakage).

• • • 75 if prior commitment has been made to use protective clothing.

Type 4 B4-33

FED 0-31858 Rev. 1 In summary, the proposed changes -result in an acceptable increase to the

' radiological; consequences of a LOCA previously evaluated in the FSAR. The revised LOCA doses are well within the regulttory guidelines. Although 'Fe

revised LOCA doses. are slightly higher for low MSIV leakage rates, the effectiveness of_ the proposed method even for leakage rates greater than the proposed increased MSIV allowable leak rate, ensures that off-site and control 1 room doses 1are not exceeded.

The-proposed amendmer.t to Table 3.6.31 of the Schnical Specifications  ;

involves the deletion of LCS valves [ plant unique identification] from the-list of containment isolation _ valves. This proposed change is consistent with

' the proposed deletion of_ the LCS. The LCS lines which are connected to the

~

main steain piping will be welded / cap close* to assure containmat integrity is maintained.- The welding and_ examination procedures will be in accordance with "A3ME Section III_ requirements.

This proposed change does not-involve an increase-in the probability of an accident'previously. evaluated iti the FSAR. In fact, this. proposed change Lr'edeces thel probability,of sn accident since, with this proposed change, the plantLwill'bc operating with less isolation' valves _ subjected to postulated o

ifa'ilure.1 This proposed change has no effect in:the consequences of an -

k _

1 accident sincesthe!LCS lines will be_ welded / cap close, thus assuring that the .

1 containment integrity, isolation, and leak test capab'.lity are not compromised for the; postulated -accident.

The proposed change to_ add a footnote to Section 3.6.1.2'is administrative in=

y . nature-and~has no effect on-any. accident.

P1 ant unique input-j:

i:

~

k LType 4; B4-34 1_._

NE00-31858 Rey, 1 The proposed changes to the index pages, renumbering of Sections 3.6.1.5 through 3.6.1.8, 4.6.1.5 through 4.6.1.8, pages number 3/4 6-8a through 3/4 12, and Section B3/4.6.1.5 through B3/4.6.1.9 is administrative in nature and has no effect on any accident. These changes provide new section and page number designations due to the deletion of Sections 3.6.1.4 and Bases Sections B3/4.6.1.4.

The operation of IPlant Station Namel. in accordance with the proposed amendment. will not create the possibility of a new or different_ kind of accident from any accident previously evaluated.

The proposed amendment to Section 3.6.1.2 does not create the possibility for a new or different '.ind of accident from any accident previously evaluated.

The BWROG evaluated MSIV leakage performance and concluded that MSIV leakage rates up to 200 scfh will not inhibit the capability and isolation performance of the valve to isolate the primary containment. There is no new mcdific:stion which could impact tne MSIV operability. The LOCA has been analyzed using the main _steei piping and condenscr as a treatment method to process MSIV leakage at the proposed maximum rate of [y] scfh. Thereforet the proposed change does no+. create'any new or different kind of accident from any accident previously evaluated in the FSAR.

The proposed amendment to dctlete Sections 3/4.6.1.4 and Bases Sections B3/4'.6.1.4 does not create the possibility of a new or different-kind of

' accident from any accident previously evaluated because the removal of the LCS does not affect any of the remaining systems at (plant name] and the LOCA has been analyzed using the alternate method to process MSIV leakages.

Ti.e proposed amendment to delete the LCS isolation valves from Table 3.6.3-1 does not create the possibility of a new or different kind of accident from to the main steam piping will be welded / cap closed to assure that the primary Type 4 B4-35

NED0 31858 Rev. I containment integrity, isolation, and leak testing capability are not compromised, therefore eliminating the possibility for any new or different kind of accident.

The proposed change to add a footnote to Section 3,6.1.2* is administrative in nature, and does not create a possibility of a new or different kind of accident from any accident previously evaluated in Chepter 15* of tlie FSAR.

The proposed changes to the index pages, and the revision of section numbers are administrative in nature, and do not create the possibility of a new or different kind of accident fra gny accident previously evaluated.

The operation of IPlant Station Namel. in accor_d_ance with the ororosed Amendment. will not involyf a sianificant reduction in the marain of safety.

The proposed amendment to Section 3.6.1.2 does not involve a significant reduction in the margin of safety. As discussed in the Bases of the Technical Specification 3/4.6.1.2, the allowable leak rate limit specified for the MSIVs is used to quantify a maximum amount of bypass leakage assumed in the LOCA radiological analysis. Results of the analysis are evaluated against the dose guidelines contained in 10CFR100-for the off-site doses and 10CFR50, Appendix A-(General Design Criteria 19) for the control room doses. Therefore, the margin of' safety is considered to be the difference between the calculated dsses and the guidelines as contained in 10CFR100 and GDC 19.

Results of the radiological analysis demonstrate that the proposed cha.nge does net involve a significant reduction in the margin of safety. The whole body doses, in terms of margin of safety, are insignificantly reduced by 1.4% at the LPZ, 2.0% in the control room, and 0.4% at the EAB. The thyroid dosc margin of safety is reduced by 19.1% at the LPZ, 9.0% in the control room, and 0.9% at the EAB. The beta dose is insignificantly reduced by 5.5% in the control room in. terms of margin of safety. The margins of safety are not significantly adversely affected because the absolute margins of safety remain l

4 l

Type 4 B4-36 i

~

NED0 31858 Rev. 1 l

well below the guidelines (lowest whole body margin of safety is 97.2% at the EAB, lowest thyroiti margin of safety is 73.6% at EAB).

Therefore, the proposed amendment does not involve a significtnt reduction in the overall margin of safety at (plant station name).

} The proposed amendment to delete Sections 3/4.6.1.4, and 83/4.6.1.4 does not reduce margin of safety, in fact, the overall margin of safety is increased.

The function of the LCS for MSIV leakage treatment #111 be replaced by alternate main stesa drain lines and condenser equipment. This treatment method is effective to reduce dose consequences of MSIV leakage over an expanded operating range and will, thereby, resolve the safety concern that the LCS will not function at MSIV leakage rates higher than the LCS design capacity. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is pan ive and does not require any logic control and interlocks. The nethod is consistent with the philosophy of protection by multiply leak-tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable and effective for MSIV leakage treatment.

Tne revised LOCA doses remain well within the regulatory limits for the site and control roca. Furthermore, the calculation s.;ows that MSIV leakage rates up to [z] scfh per steam lina would not exceed the regulatory limits.

Therefore, the proposed metimd provides a substantial safety margin for mitigating the radiological unsequences of MSIV leakage beyond the proposed Technical Specification leak rate limit of [y] scfh.

The proposed amendment to delete LCS isolation valves (plant unique identification) from Table 3.6.3-1 does not reduce the margin of safety.

Welded / cap closure of the I.CS lines assures that the primary containment integrity, and leak testing capability are not compromised; therefore, it does not result in a reduction ir: the margin of safety.

Type 4 B4-37  !

m .- - . - . . . .

1 NED0-31858 Rev. 1 The proposed change to add a footnote to Section 3.6.1.2 is administrative in nature and does not affect the margin of safety.

The proposed amendment to the index pages, and the revision of section and ,

page numbenu is administrative in nature,- ano does not have any impact on the margin of safety.

Therefore, the proposed amendment to the Technical Specifications does not involve a significant hazards consideration.

j ffype 4 B4-38 I

NED0 31858 Rev, 1 ENCLOSURE 4 100RPORATI0fil ROINSE NO. IXXXI DOCKET NO. IXXX1 APPLICATION FOR EXEMPTION TO APPENDIX J OF 10CFR50

- Pursuant to Section 50.12(a) of the Reo"latior.s of the Nuclear Regulatory Commission, (Corporation], holder of facility Operating License No, [xxx],

he k V requests specific exemptions to Appendix J of 10CFR Part 50 " Primary Reactor Containment Leakage Testing For Water-Cooled Power Reactors".

Specifically, [ Corporation] requests that leakages from the main steam isolation valves (MSIVs) be exempted from the acceptance criteria for:

(1) the overall integrat_ leak rate test (Type A), as defined in the regulations of 10CFR50, Appendix J, Paragraphs Ill.A.5(b)(1) and III.A.5(b)(2),and (2) the combined local leak rate test (Type B and C), as defined in the regulations of 10CFR50, Appendix J Paragraph III,C.3.

The purpose of the test acceptance criteria is to ensure that the measured leak rate from the containment volume will not exceed the designed containnnt leak rate. assumed in the safety analysis for a postulated design basis 1.oss-Of-Coolant Accident (LOCA).

In conjunction with this application for exemption request, [ Corporation] has transmitted to the Nuclear Regulatory Commission an applict. tion for a license Technical Specification sections are based on the BWR Standard Technical i- Specification for BWR/S (NUREG-0123). Each utility should review the sections applicable to their Technical Specifications.

Type 4 84-39

- _ _ ~ _ --

NED0 31858 Rev. 1

-amendment pursuant to 10CFR50.90. This license amendment involves a proposed change to Section 3.6.1.2* of the Technical Specifications to permit an increase in the allowable leak rate for the MSIVs from the current [x]

standard cubic feet per hour (scfn) to [y] scfh per main steam line, and a proposed change to Section 3.6.1.4 for eliminating the requirements for the Leakage Control System. The safety analysis has been revised to assess the radiological effects of MSIV leakage following a postulated design basis LOCA.

[ Corporation) has demonstrated that the proposed change does not involve a significant hazards consideration. '

This proposed exemption is a result of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic Issue C-8 "MSIV Leakage ano LCS Failu e".

The following discussion provides a detailed justification and evaluation of the proposed exemption. The proposed exemption is found to be authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security. Furthermore, special circumstances are present that warrant the prnting of this exemptio'.

The proposed exemption will not cause additional operational activities that may significantly affect the environment. It does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-Operating License. Stage, result in a significant cFange in effluents or power levels, or affect any matter not previously reviewed by the Nuclear Regulatory Commission that may have a significant adverse environniental impact.

Therefore, pursuant to 10CFR 50.12(a), (Corporation] hereby requests an exemption for [ plant station name] for MSIV leakages from the acceptance test-criteria specified in Appendix J of 10CFR50.

Plant-specific t

L l Type 4 B4-40

NEDO-31858 Rev. 1 A. Justification The regulation of 10CFR50, Appendix J Paragraphs III.A,5(b)(1) and III.A.5(b)(2) requires the overall integrated leakage rate, as measured during containment pressure tests (Type A), to meet the acceptance criterion of less than or equal to 0.75 of the maximum allowable containment leak rate.

Paragraph III.C.3 of the regulation requires the combined leakage rate for all penetrations and_is'lation valves, as measured ouring local leak rate tests (Type B and Type C), to meet the acceptance criterion of less than or equal to 0.60 of the maximum allowable containment leak rate. Paragraphs III.C.3(a) and III.C.3(b) define the acceptance criteria for the exclusion of containment isolation valves from the acceptance criterion for Type B and C tests.

As described in the Bases Sections B 3/4.6.1.2* of the Technical Specifications, the limit ations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure. As-an added conservatism, the measured leak r4te is furthei limited to less than or equal to 0.75 of the maximum allowable leak rate during the performance of the periodic tests to account for possible degradation of the containment leakage barrier between leakage tests.

The maximum containment-leakage rate was included in the radiological analysis of a postulated design basis LOCA as evaluated in Section 15.6.5* of the Final Safety Analysis Report (FSAR). The radiological analysis calculated ti effect of the maximum leakage rate from the containment volume in terms of

- control room and off-site doses, which were evaluated against the dose l guidelines of 10CFR50, Appendix A (General Design Criter ia 19) and 10CFH00, respectively. Leakages from the containment volume were contained in the reactor building (secoadary containment), filtered by the Standby Gas Plant-specific information.

I 1

I l Type 4 B4-41

UED0-31858 Rev. 1 Treatment System, and released to the environment through the elevated release stack **' The maximum containment leakage rate includes leakages through structures, all penetrations identified as Type 0, and all co tainment isolation valves identified as Type C.

The safety anIlysis has been revised to account for the radiological effect from MSIV leakages and from those of other containment leakages following a postulated design basis LOCA. Unlike the treatment path for other containment leakages, the treatment of MSIV leakages employs the main steam drain piping and tha. condenser. Fission products are removed by plate-out and hold-up in the relatively large volumes of the main steam piping end condenser.

The treatment method for MSIV leakages is recommended by the BWROC in support of-the resolution to Generic Issue C-8. The BWROG has evaluated the availability of main steam system pipin; and condenser alternate treatment pathways for processing MSIV leakage, and has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for other plant safety risks. The BWROG nas also determined that main steam piping and condenser designs are extremely rugged, and that the ANSI-B31.1 design requirements tvaically used for nuclear plant system design contain a good deal of marc .

In order to further ,1ustify the capability of the main steam piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes). This study concluded the possibility of a failure which could cause a loss of steam or condensate in b Plant-specific information.

BWR/6 plants may not have an elevated stack. Furthermore, some plants may have " bypass leakages" defined in their radiological analysis. These are the leakages that bypass the reactor building and release to the environmental unfiltered. Each utility should correct the text herein consistent with their radiological analysis.

Type 4 B4-42

NE00 31858 Rev. 1 BWR main steam piping or cc,ndensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

Leakage from the MSIVs should not be included in the T,pe A acceptance criterion because the treatment path for MSIV leakage is different from that of containment leakages. Potential leakage from the containment is contained in the reactor building (secondary containment), treated by the SGTS, and released via the main stack. MSIV leakage is contained, plated-out, and delayed in the main steam piping and the condenser, and released via the turbine building.

Furthermore, leakage from the MSIVs should not be included in the combined local leak rate test (Type B and Type C) acceptance criteria because a specific allowable leak rate !.as been allocated for the MSIVs in Section 3.6.1.2* of the Technical Specifications.

The deletion of the LCS is proposed partly-ii response to the safety concern identified by Generic Issue C-8 that the LCS would not function at high MSIV leakage rates since the process capability of_the LCS at (plant name)'is designed for MSIV leakage rate of no mere than (100 scfh]*. MSIV leakage is treated separately from other containment leakages, therefore any exemption

.which was previously granted in accordance with Paragraph III.C.3 of Appendix J-of 10CFR50 should remain applicable.

As discussed earlier, the basis fo- the containment leakage tests and the acceptance criteria is to ensure that the measured leak rate will not exceed the maximum leak rate assumed in the safety analysis. The safety analysis for -

a design basis LOCA has been revised to include the maximum MSIV leak rate separately from the maximum containment leak rate. MSIV leakages will be tested as part of the local leak rate test in accordance with the requirements Plant-specific inforraation.

Type 4 B4-43

l NE00 31858 Rev. 1 l in-Section 3.4.6.1.2 of the Technical Specifications. This test ensures that the measured MSl> 1eak rate will not exceed the allowable leak rate assumed in the safety analysis.

1There is sufficient conservatism in the allowable MSly leak rate to account for possible degradation of the MSly leakage barrier between leakage tests.

As. discussed-in the appli:ation for the license amendment, (Corporation]

proposes a maximum allowable MSIV leak rate of [y] scfh per main steam line; whereas, the analysis demonstrates that MSIV leakage rates up to approximately '

(z] scfn per main--steam line will not result in dose exposures in excess of the regulatory limits. Thus, a safety margin exists. Furthermora, (Corporation) will institute into the MSIV maintenance and test program, the

.requirecent that any MSIV exceeding the proposed (y] scfh limit, will be repaired and re-tested to meet a leakage rate of less than or equal to [x]

-scfh. This will assure continuation of high quality repair and refurbishment efforts to improve the overall performance and v aliability of the MSIVs. -

Therefore, the proposed exemption from the acceptance criteria of 10CFR50, ,

Appendix J will not defeat the underlying purpose of the regulation, and is consistent with the safety analysis.

B. _ : Authorized By Law

~

The proposed exemptio'n is consistent with Section 3.6.1.2 of the Standard Technical Specification (NVREG-0123). The reuon for this exemption is provided in the Technical Soecifications Bases BL3/4.6.1.2. A review of the Technical Specifications for BWRs indicates that such an exemption has been granted to the following plants: Fermi'2, Hatch I'& 2, Hope Creek, Limerick.1, Shoreham, LaSalle 1 and 2, Hanford, Clinton, Grand Gulf 1, Perry, Dresden 2 and 3, Monticello, Quad Cities 1 and 2, Brunswick 1 and 2 and Nine- .

Mile Point 2.

'Therefore, the proposed exemption-is authorized by law.

' Type-4=

B4-44

, NEDO 31858 Rev. 1 I

C. No Undue Risk to Public Health and Safety The proposed exempt hn presents no undue risk to public health and safety.

The revised MSIV leakage rate has beer incorporated in the radiological analysis for a postulated LOCA as en addition to the designed containment leak rate. The analysis demonstrates an acceptable in,:rease to the dose exposures previously calculated for the control room and off-site. The revised LOCA doses remain well within the guidelines of 10CFR100 for off-site doses and 10CFR50, Appendix A, {'ieneral Design Criteria 19) for the control room doses.

In addition, Section 3.6.1.2* of the Technical Specification has provided for allowable MSIV leak rates, which assure that the MSIVs isolation function is not compromised. Finally, potential MSIV leakage is subjected to plate-out, and hold-up in the main steam piping and condenser, thus minimizing their effect on the total dose released, As discussed in Section F of this application, the proposed change will not adversely affect the conclusions of the previously issued FES-OL. Therefore, the proposed exemption presents no undue risk to public health and safety.

D. tonsistent with Common Defense and Security With regard to the " common defense and security" standard, the grant of the requested exemption is consistent with the common defense and security of the United States. The Commission's Statement of Coasiderations in support of the exemption rule note with approval the explanation of this standard as set forth in Lona Island Liahtina comoany (Shoreham Nrlear Power Station, Unit 1), LBP-84-45, 20 NRC 1343, 1400 (October 29,1984). There, the term " common defense and security" refers principally to the safeguarding of special nuclaar material, the absence of foreign control over the applicant, the protection of Restricted Data, and the availability of.special nuclear material for defense needs. The granting of the requested exemption will not affect any of these matters and, thus, such grants are consistent with the coanon defense and security.

Type 4 84-45

.- - n -- - - - - - . _ - - = . - . - - - -. - - .

N -NED0-31858 Rev. 1-ll E..  : Soecial Circumstances ' Are Present  ;

=

Special. circumstances are present which warrant issuance of this requested exemption. These special circumstances are discussed in accordance with the classification contained'in 10CFR50.12(a)(2): -i (ii) Application of the regulation in'the particular circumstances would not serve the underlying purpose of the rule or is not necessary to- achieve the underlying purpose of the rule. ,

The underlying purpo_se of-the rule is to limit releases to within the off-site

and control room dose guidelines of 10CFR100 and 10CFR50, Appendix A (GDC 19),

respectively. Compliance with Appendix J of 10CFR50 for Type A test

^

acceptance criteria is not necessary to achieve the underlying purpose of the

~

ru',e-since MSIV:leaktge is not_ directed into the reactor primary containment.

Instead,- the MSIV's leakage is directed through the main steam drain piping into the condenser. Since: Type A tests are intended to measure the primary containment 1overall integrated leak rate (ILRT), the MSIV's leakage rate

~s hould not.be included in-the measurement of the ILRT. Compliance with Appendix J of 10CFR50 Type C test acceptance criteria is not necessary since a specific MSIV leak rate limit is already specified-in Section 3.6.1.2* of the Technical Specifications.

l

• :.e safety analysis has been revised to assess the radiological consequences -

of._MSIV. leakage following a. design basis LOCA. The analysis has demonstrated

-that the revised LOCA doses-are well within the off-site and control room dose guidelines ~of 10CFR100 and GDC 19.

(iii) . Compliance would result in undue hardship.or.other costs that are-significantly in excess of those contemplated when'the-y, regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.

1 l

l

' Type 4 46

_M__. _ _ . . . _ _ _ __ _

_ J

NED0 31858 Rev. 1 Compliance with Appendix J _of 10CFR50 Type A and Type C test acceptance criteria results.in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted. The proposed increase in the MSIV allowable leak rate will not be possible if the MSIV leak rate results are included in the Type A and Type C test acceptance criteria.

Compliance requires unnerassary repair and re testing of the MSIVs. This significantly imnacts the maintenance work load during plant outages and often contributes to outage extensions. The frequent MSIVs disassembly and refurbishing, which is required to meet the low leakage limits contributes to rapeated failures.

Examples of these maintenance induced defects include machining-induced seat crtcking, machining of.gtide ribs, excessive pilot valve seat machining, and mechanical defects induced by assembly and disassembly. By not having to disassemble the valves and refurbish them for minor leakage, [ plant name) avoids introducing one of the root causes of recurring leakage. Industrial experience suggests thal, by attempting to correct non-existing or minimal defects in tne va?ves, it is likely that some actual defacts may be introduced that lead to later leai test failures.

In addition, the frequent maintenance work results in needless dose exposures to maintenance personnel leading to additional economical burdens, and are inconsistent with As Low As Reasonably Achievable (ALARA) principles.

-(iv) The-exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may

. result from the grant of the exemption.

(( Corporation)hastransmittedtotheNRCanapplicationforalicense amandment which involves. proposed changes to the Technical Specifications to increase the allowable MSIVs leak rate from [x] to [y] scfh and to delete the

- requirements for the LCS. For the MSIV leak rate limit, this application is

= Type 4 B4-47 C

c NEDO 31853 Rev. 1 partly based on the fact that thn current limit is too restrictive, and Jesults in excessive MSly maintenance and repair, leading to additional MSiv failures which in turn result in higher leakage. The proposed limit will benefit the public health and safety by reducing the potential foc MSivs failures, and thus' keeping the MSIV

• eakage within the radiological analysis  ;

values.

For the LCS, the proposed changes in C '.e ' replacement of the existing LCS with the more reliable and effective main steam piping and condenser method for MSIV leakage treatment.. The effectiveness of the proposed method even for leakage rates greater than the proposed. increased allowable limits, ensures off-site dose limits to the public are not exceeded. Overall, the proposed

-treatment: method can handle MSly leakage over an expanded operating range, and will thereby resolve the safety concern that the LCS will not function at MSIV leakage rates. higher than the LCS capacity Thus, a margin of safety exists.

Furthermore,-.it is clearly a' safety improvement to replace a system with known limitations with the alternate mai_n steam piping and condenser treatment pathway, which has been shown to have excellent reliability.

The exemption from Appendix J requirements for MSIV leakage rates is required so that (plant name) can operate with the propoced Technical Specifications #

valueif-[y] scfh. This' benefit will compensate for any decrease in safety that may result-from the granting of the exemption.

-Thus, s,scial circumstances exist warranting the grant of the exemption.

F. - Eny'ironmental Impact The proposed exemption has been analyzed and determined not to cause

! additional construction or operational activities which may significantly -

affect the environment. It does not result in a significant incr se in any

. adverse environmental impact previously evaluated in the Final Environmental

.lmpact Statement-0perating License Stage, . result in a significant change in u

Type 4 89-48

NEDO-31858 Rev. I effluents or power levels, or affect any matter not previously reviewed by the Nuclear Regulatory Commission which may have a significant adverse environmental impact.

The proposed exemption does not alter the land use far the plant, any water uses or impacts on water quality, air or ambient air quality. The proposed action does not affect the ecology of the site and vicinity and does not affect the noise emitted by station. Therefore, the proposed exemption does not 1*'ect the analysis of environmenta'l impacts described in the environmental report.

Type 4 B4-49

l NEDO 31858 Rev. 1 I (

ENCLOSURE 5 iCORPORAT10N1 LICENSE NO. I m l DOCKET NO. Ixxx1 APPLICATION FOR EXEMPTION TO APPENDIX A 0F 10CFR100

[ CORPORATION], holder of Facility Operating License No. NPF-[xxx], hereby requests an exemption of the downstream main steam piping and condenser from the seismic requirements specified in Appendix A of 10CFR100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants,*

Specifically, [ Corporation] proposes to employ probability analysis, existing design capabilities, seismic experience, and a plant specific seismic adequacy verification, as alternate methodology to the dynamic analysis or qualification test specified in Paragraph Vl(a)(1) of 10CFR100 Appendix A, to provide reasonable assurance that the existing main steam piping arid condenser will remain functional following a design basis accident coincident with a significant seismic e ant.

The exemption would allow the existing, non-seismically designed main steam piping and condenser to be used for mitigating the radiological consequences of MSIV leakage during the duration of a Design Basis Accident, such that the resulting doses are within the guidelines of 10CFR100.

[ Corporation] recognizes that there is no provision in 10CFR100 for exemption; however, the Nuclear Regulatory Commission (NRC) has the authority to grant this exemption.

Type 4 B4-50

NE00 31858 Rev. 1 (Corporation) proposes to replace the existing LCS with the more reliable and effective main steam drain line and condenser for MSly leakage treatment.

This treatment method is effective to treat MSIV leakage over an expanded operating range. Except for the requirement to establish a proper flow path from the main steam isolation valves to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment.

The existing leakage control systems have limitations for mitigating MSIV leakage. Operation of the system induces higher MSIV leakage by increasing the-differential pressure across the valves, and is inconsistent with the philosophy of multiple barriers for limiting fission product releases (does not take advantage of the outboard MSIVs). [For positive-pressure LCS, t,peration of the system increases containment pressure and thereby increases the containment leakage.)* The LCS requires multiple logic controls, interlocks, timers, containment isolation valves, and other equipment to ensure containment integrity and protection from the high pressure main steam

'W. Based on plant operating experience, the LCS does not provide a high ogree of reliability. Also the LCS has limited capacity and does not function at moderate MSIV leakage rates above 100 scfh.

Even though.the resulting offsite doses may be slightly higher for low MSIV leakage rates, the effectiveness of the proposed method even for leakage ratas greater than the proposed increased allowable limits, ensures off-site does limits-to the public are not exceeded. Overall, the proposed treatment method ran handle MSIV leakage over an e(panded operating range, and will thereby resolve the safety concern that the LCS will not function at MSIV leakage rates higher than.the LCS capacity. Thus, a margin of safety exists.

Furthermore,-it-is clearly a safety improvement to replace a system with know limitations with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability.

Type 4 B4-51

_ . ~.

NEDO 31858 Rev. 1

-In conjunction with this application for exemption request, [ CORPORATION) has transmitted to the NRC an application for a license amendment pursuant to iOCFR50.90. This license amendment involves a proposed change to Section 3.6.1.2 of the Technical Specifications to permit an increase in the allowable leak rate fcr the MSIVs from the current [x] standard cubic feet per hour (scfh) total to [y] tcfh per main steam line, and a proposed change to Section 3.6.1.4 for eliminating the requirements for the LCS. The safety analysis has been revised to assess the radiological effects of MSIV leakage following a postulated design basis LOCA. [ CORPORATION) has demonstrated that the proposed change does not involve a significant hazards consideration.

This proposed exemption is a result of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic Issue C-8 "HSIV Leakage and LCS Failure".

The following discussion provides a detailed justification and evaluation of the proposed exemption. While recognizing this exemption criteria are specifically applicable to 10CFR50, [ Corporation] has evaluated the proposed exemption in accordance with the criteria specified in 10CFR50.12(a). The proposed exemption will not present an undue risk to the public health and safety and is consistent with the common defense and security. Furthermore, special circumstances are present that warrant the granting of this exemption.

The proposed exemption will not_cause additional operational activities that may significantly affect the environment. It does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-Operating License Stage, result in a significant change in effluents or power levels, or affect any matter not-previously reviewed by the NRC that may have a significant adverse environmental impact.

Type 4 B4-52

NED0 ,71858 Rev, 1 Upon the NRC approval of the license amendment and exemption requests, l (Corporation]_will perform a verification of seismic adequacy of the main l steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NE00 31858 Rev.1, to provide reasonable assurance of the structural integrity of these componants.

Therefore, [ CORPORATION) hereby requests an exemption to the seismic requirements of 10CFR100 Appendix A for (plant station name) to permit the use of existing, non-seismically designed main steam piping and condenser to mitigate the radiological consequences of MSIV leakage.

_A. Jgstification Paragraphs VI(a)(1) of 10CFR100 Appendix A requires that structures, systems and components, which assure the capability to prevent or mitigate the  ;

. consequences of accidents which could result in potential off-site exposures of_10CFR100, be designed to remain functional following a safe shutdown earthquake-(SSE) and concurrent loads. The engineering method used to assure  ;

that the required safety functions are maintained following the SSE shall ,

involve the use of'either dynamic analysis or a suitable qualification test to demonstrate that structure, systems, and components can withstand the seismic .

and other concurrent loads.

The-BWROG has evaluated the' capability of main steam piping and condensers to process MSIV leakage following a design basis accident coincident with a seismic event. Based on this comprehensive evaluation, the BWROG has concluded there is reasonable assurance that.the main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage. The following conclusions provide

.the bases for this assurance:

Type 4 B4-53

NED0 31858 Rev. 1 (1) Probability for which the resulting dose from MSlv leakage is significant is extremely low. This requires a design basis LOCA, a degraded core where ECCS are not functional, and a significant seismic event.

(2) Main steam piping and condensers are designed to strict industrial standards and building codes; thus, significant design margin exists.

(3) Main steam piping and condensers exhibit substantial seismic ruggedness. Comparisons of piping and condenser design in GE plants with those in the earthquake experience database reveal that the GE plant designs are similar to or more rugged than those that have exhibited good earthquake performance.

(4) Possibility of significant failure in CE BWR main steam piping or condensers in the event of design basis earthquakes is highly unlikely, and any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

(5) A plant-specific verification of seismic adequacy of the main steam piping and condenser will be performed to provide reasonable assurance of the structural integrity of these components.

In support of the above, the BWROG has reviewed the potential combinations of Loss-Of-Coolant Accidents (LOCAs) and seismic events of interest:

(1) LOCA VITH0VT NEAR COINCIDENT SEISMIC EVENT. For this occurrence the pressure in the piping system downstream of the MSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the alternate flow path through main steam system piping to the condenser is assured.

(2) SEISMIC EVENT WITHOUT NEAR COINCIDENT LOCA. Without a LOCA and the potential associated core degradation, the radioactivity transported with MSIV leakage-is of no radiological significance.

Type 4 B4-54 l

NE00 31858 Rev. 1 (3) LOCA WITH NEAR COINCIDENT SEISMIC EVENT. For this occurrence (also assuming significant core damage) the consequences are of interest because a seismic induced failure in the main steam or condenser system could allow MSIV leakage to bypass the alternate treatment pathway, it has been previcusly well documented that tha probability of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability approximately 0.001 per reactor per year; core melt probability is plant-specific and typically ranges from 0.00001 to 0.0001 per reactor per year). It is also noted that a LOCA does not induce a seismic event, and that a seismic event has a very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10-7 per reactor per year).

Considering that the probability of a near coincident LOCA and seismic event is much smaller than other plant-safety risks (less than 1 x 10-7 per reactor per year for coincidcnt events, less than 5 x 10-7 per reactor per year for seismic induced LOCA), the concern for main steam piping or condenser damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design basis seismic events.

ANSI-B31.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin. In addition, specific seismic design provisions have been incorporated into some newer BWR main steam and condenser systems.

To further justify the capability of the t.ain steam system piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed Type 4 B4-55 )

NED0-31858 Rev. 1 i

-piping and condensers (in past earthquakes). The study summarizes data on the performance of main steam piping and condensers in past strong motion earthquakes and compares these piping and condensers with those in typical U.S. GE Mark I, II, and III nuclear plants. This limited earthquake experience data and similarity comparisons are then used to further strengthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in a design basis earthquake in conjunction with a l LOCA occurring just prior to or after the seismic event.

The earthquake experienco data are derived from an extensive database on the performance of industrial facilities and over 100 power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the '

world from 1934 to the present.

The piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they are not designed to resist earthquakes. This is a common conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipment such as motor' control centers, cumps, valves, structures, and so forth. That is, with limited exceptions, normal industrial construction and equipment typically l have substantial inherent seismic ruggedness, even when they are not designed

! for earthquakes. No failures of main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.

Comparisons of piping and condenser desigr. in example GE Mark I, II, and III 1 plants with those in the earthquake experience database reveal the GE plant designs are similar to or more rugged than those that exhibited good earthquake performance. The BWROG concludes that (1) the possibility of significant failure in GE'BWR main steam piping or condensers in the evsnt of an eastern U.S. design basis earthquake is highly unlikely and that (2) any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

Type 4 B4-56

NEDO 31858 Rev. 1 Earthquake experience methodology has been applied in seismic equipment qualification issues associated with Unresolved Safety Issue A-46 (Seismic Qualification of Equipment ir, Operating Plants). Piping performance data are l presented in NUREG-1061 (a report from the NRC Piping Review Committee), and this. report proposes changes to criteria that are directed toward the recognition of the superior performance of piping in earthquakes and estab-lishing more realistic seismic criteria for piping qualification. The NRC has published NUREG-1030 and NUREG-1211 " Seismic Qualification of Equipment in Operating Nuclear Power Plants," which conclude that the seismic experience i data approach provides the most reasonable and preferred alternative to other I current equipment qualification methods. -

The rapidly growing use of the seismic experience data approach is further illustrated by the fact that this method of analysis is now referenced in:

A. Draft RG 1.100, Revision 2 " Seismic Qualification of Electrical and

' Mechanical Equipment in Nuclear Power Plants"  ;

j B. Recent approved revision of IEEE Standard 344-1987, " Recommended Practice for Seismic Quelification of Class IE Equipment For Nuclear Power Generating Stations" C. Draft report of ASME Standard " Recommended Practice for Seismic Performance Qualification of Mechanical Equipment Used in Nuclear Power Plants."

The earthquake experience database includes a large~ number and variety of piping systems. In fact, piping is probably . he strongest area in this regard

-(compared to areas like electrical or mechani ;1 equipment, cable trays,

-etc.). Lit has been concluded that the earthquake experience data on piping, and in particular data on main steam piping, are applicable to main steam

-piping in BWRs.

' Type 4 B4-57

NED0-31858 Rev. 1 ,

I in both nuclear and conventional power plants, the condenser is dt : inned to reduce the low pressure turbine outlet pressure (thereby increasing turbine efficiency) and to condense the steam. The nuclear environment does not impose additional significant design considerations on the condenser. With the exception of hotwell size, a conventional plant and nuclear plant with similar performance parameters have similar condensers.

None of the condensers within the seismic experience database has seismic design criteria. However, in view of the performance of the condensers within the database, it is concluded that the condensers have an inherent seismic ruggedness and that the earthquake experience data on condensers are applicable to condensers in BWRs.

Another recent study to develop, by data collection and statistical analysis, updated estimates of pipe breaks in commercial U.S. nuclear power plants was completed in 1987. This study evaluates both LOCA sensitive systems and non-LOCA sensitive systems. For BWR non-LOCA sensitive systems, ten pipe failures have occurred over 313 years of operating experience. None of these failures occurred in the main steam piping. Based on the observed failure rates, this' study estimated the failure rate for the main steam system piping to be 0.0007 failure / year /BWR with an upper bound of 0.0096 failures /

year /BWR.- These results are consistent with the conclusion from the

-earthquake databeses and NUREG-1169: BWR main steam piping is designed to withstand severe plant transients such as turbine trips and is expected to remain intact following accidents as severe as a design basis LOCA. Thus, the non-seismically designed main steam piping-and the main condenser can be used to mitigate the consequences of MSIV leakage.

[ Corporation) will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NEDO-31858.Rev.1, to provide reasonable assurance of the structural integrity of these components.

Type 4 B4-58

NE00-31858 Rev. I In conclusion, there is reasonable assurance that the existing, non-seismically designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event,

'as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage.

B. No Undue 31sk to Public Health and Safety The BWROG has evaluated the capability of main steam piping and condensers to process MSIV leakage following a design basis accident coincident with a seismic event. Based on this comprehensive evaluation, the BWROG has concluded there is reasonable assurance that the main steam piping and condenser will -emain functional following a design basis accident coincident with;a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage. This assurance is based on methodology using probability analysis, margins in the existing design codes, seismic experience, and a plant specific verification of seismic adequacy.

The treatment method for MSIV leakages is recommended by the BWROG in support of the' resolution to Generic Issue C-8 "MSIV Leakage and LCS Failure". The proposed changes involve a replacement of the existing LCS with the more reliable and effective maire steam piping and condenser for MSIV leakage treatment. This treatment method is effective to reduce dose consequences of MSIV leakage over an expanded operating range and will, thereby, resolve the safety concern that-the LCS will not function at MSIV leakage rates higher than the LCS dnign capacity. Except for the requirement to establish a -

proper flow pat from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak-tight barriers

-used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly effective and reliable for.MSIV leakage _ treatment.

In conclusion, the proposed exemption presents no undue risk to public health and safety.

Type 4- B4-59

NEDO-318b8 Rev. 1 l

C. Consistent with Common Defense and Security With regard to the " common defense and security" standard, the grant of the requested exemption is consistent with the common defense and security of the United States. The Commission's Statement of Considerations in support of the exemption rule note with approval the explanation of this standard as set

' orth in lona Island liahtina Compa.ny (Shoreham Nuclear Power Station, Unit 1), LBP-84-45, 20 NRC 1343, 1400 (October 29,1984). There, the term " common defenu and security" refers principally to the safeguarding of special nuclear material, the absence of foreign control over the applicant, the protection of Restricted Data, and the availability of special nuclear material for defense needs. The granting of the requested exemption will not affect any of these matters and, thus, such grancs are consistent with the common defense and security.

v D. jip_efial e Circumstances Are Present Special circumstances are present which warrant issuance of this requested exemption. These special circumstances are discussed in accordance with the classification contained in 10CFR50.12(a)(2):

(ii) Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

Compliance with Appendix A of 10CFR100 for the downstream main steam piping and condenser is not necessary to achieve the underlying purpose of the rule.

The underlying purpose of the rule is to limit releases to within the off-site dose limits of 10CFR100. The regulation requires components that retigate the consequences of an accident to within the dose limits of 10CFR100 be designed to the seismic requirements of 10CFR100 Appendix A. The regulation is intended to provide a reasonable assurance that the components will remain functional for the mitigating function. For the purpose of mitigating the Type 4 B4-60

NEDO 31858 Rev. I radiological consequences of MSIV leakage, it is not necessary to apply the seismic requirements of 10CFR100 Appendix A to the main steam piping and condenser in order to achieve the underiying purpose of the rule because:

(1) There is reasonable assurance that the existing, non-seismically designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSiv leakage. This assurance is based on methodology using probability analysis, margins in the existing design codes, seismic experience, and a plant specific verification of seismic _

adequacy.

(2) The safety analysis hes been revised to assess the radiological consequences of MSIV leakage following a design basis LOCA. The analysis has demonstrated that the revised doses are well within the off-site dose guidelines of 10CFR100.

Furthermore, the seismic approach is consistent with the current resolution of the seismic and equipment qualification issues. Earthquake experiences data have applied in seismic equipment qualification issues associated with Unresolved Safety Issues A-46 (Seismic Qualification of Equipment in Operating Plants) . Piping performance data have been preser.ted in NUREG 1061, a report from the NRC Piping Review Committee, which proposes changes to criteria that -

are directed toward the recognition of the superior performance of piping in earthquakes and establishing more realistic seismic criteria for piping qualification. The NRC hcs published NUREGs 1030 and 1211 " Seismic Qualification of Equipment in Operating Nuclear Power Plants," which conclude (

that the seismic experience data approach provides the most reasonable and preferred alternative to other current equipment qualification methods.

(iii) Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.

Type 4 B4-61

w-L J-w,-i --Am LA&.

NEDO 31858 Rev. 1-The proposed MSIV leakage treatment method utilizes the existing main steam piping and condenser for the mitigating function. Compliance with the seismic requirements of 10CFR100 Appendix A for the main steam piping and condenser would require significant upgrade of the existing equipment, lea 3 to unnecessary long-term plant shutdown for modification, and significantly incresso_maintenanc6 requireriienti, au the associated costs in oroer to meet seismic qual'fication requirements, (iv) The exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the granting of the exemption.

[ CORPORATION) has transmitted to the NRC an application for a license amendment which involves proposed changes to the Technical Specifications to increase the' allowable MSIVs leak rate from [x] scfh total to [y] scfh per steam line _and to delete the requirements of LCS. For the MSIV leak rate -

limit, this application is partly based on the fact that the current limit is too restr k tive, and results in excessive MSIV maintenance and repair, leading to additional MSIV' failures, which in turn result in higher leakages. The

- proposed limit will benefit the public health and safety by reducing the potential for MSIVs failures, and thus keeping the MSIV leakage within the radiological _ analysis values.-

For the LCS,-the proposed changes invcive a replacement of the existing LCS with the more reliable and effective main steam piping and condenser method for-MSIV leakage treatment. 'The effectiveness of the proposed method even_for leakage rates greater than the proposed increased allowable '. imits, ensures off-site dose limits to the public are not exceeded. Overall, the proposed treatment method can handle MSIV leakage over an expanded operating range, and will thereby resolve the safety concern that the 1.CS will not function at MSIV leakage rates higher than.the LCS capacity. Thus, a margin of_ safety exists.

Furthermore, it is' clearly a ' safety improvement to replace a system with knowa-limitations with the alternate main steam piping and condenser treatment pathway, which has been shown-to have excellent reliability.

Type 4 B4-62

NEDO-31858 Rev. 1 The exemption from 10CFR100 Appendix A seismic requirements for the downstream piping and condenser is required so that [ plant station name) can operate with the proposed Technical Specifications limit of (y) scfh and with the alternate MSIV leakage treatment mett'od. This benefit will compensate for any decrease in safety that may result from the granting of this exemption.

Thus, special circumstances exist warranting the granting of this exemption.

E. Environmental Impaq1 The proposed exemption has been analyzed and determined not to cause additional construction or operational activities which may significantly affect the environment, it does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-0perating-License Stage, result in a significant change in effluents or power levels, or affect any matter not previously reviewed by the Nuclear Regulatory Commission which may have a significant adverse environmental impact.

The proposed exemption does not alter the land use for the plant, any water uses or impacts on water quality, air or ambient air quality. The proposed action does not affect the ecology of the site and vicinity and does not affect the noise emitted by station. Therefore, the proposed exemption does nct affect the analysis of environmental impacts described in the environmental. report.

Type 4 B4-63/B4-64

.s NED0-31858 Rev. 1 NEDO-3030t.

95!ED026 Class I September 1995-APPENDIX C-POTENTIAL OPERATOR ACTIONS TO ColfTRO.', MSIV II.AKAGE Revision 0 BWR 2 throuah 6 PREPARED TOR BVR00 ftSIV LEAKAGE C0lrTROL CottfITTEE-BY THE GEkF.RAL ELICTRIC CottPANY -

. Approved by: d. w "1-l3 fj

~

-T, .

W. Craig U Generic Programa NUCLSAR ENERGY SU$1NE58 OPERATIONS

• CENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 9$125 GENERAL $ ELECTRIC C-1

_. _ i-t - ~ - -

J4s

NE00-31858-Rev l-NEDO- 3032 I.

1 DISCLAI!ER OF RESPONSIBILITY This document was prepared by or for the General Electric Cenpany.

Neither the General Electric Company nor i ly of the contributors 1

to this documents i I

A. Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the I information contained in this document, or that the use of any inf ormation disclosed in this document may not inf ringe privately owned rights, or B. Assumes any responsibility for liabil:ty or damagt of any kind which may result from the use of any infomation disclosed in this document.

11 C-2

N NE00 31858 Rev. 1 NEDO-30324 i

CORTENTS i

' P,_ a gg ABBREVIATIONS vil

' ABSTRACT- is

1.0 INTRODUCTION

l' ,

2.0

SUMMARY

2 -- i 3.0 DISCUSSION 5 ,

3.1 Assessment.of Operator Actions for Control of MSIV Leakage 5 3.1.1; Operation of the.0ffges System : 6 3.1.2 -Operation of the Mechanical Vacuus Pump 8

-3.1.3 Containment within the Main Stese System 10-3.1.4 RPV Depressurization to the Suppression Pool ~12

' 3.1. 5 ' Flooding of.the Main Steam Lines 14 3.1.6 Pressurization between the MSIVs 16 3.1.7 Treatment / Containment within the Plant Buildings 18 3.1.8 Utilization of the MSIV-LCS 21 3.1.9 Flooding of the' Primary-Ccatatament ~23 3.2 Recommended Operator Actions - for NSIV Leakage Control 24 3.3  : Conditions Requiring Control of MSIV Leakage 25 3.3.1? . Evaluation'of Conditione Requiring Control of MSIV Leakage 26-3.3.2 Classification of Conditions Requiring Control of MSIV Leakage 26 --

40 -LREFERENCES 38 APPENDICES A. RECOMMENDED EPG CONDITIONS ~AND STEPS TO ADDRESS

-MSIV LEAKAGE. CONTROL A-1

8. 'PARTICIPATINC VIILITIES g.g til/iv C-3

NEDO-31858 Rev. I s

NEDO-30324 LIST OF TABLES Table Title ~ Para 2-1 Summary of Potentia! Symptomatic Conditions for which I Conte:L of MSIV. Leakage May Be Appropriate 3 2-2 Summary of Operation Actions Considered for MSIV Leakage I Control 4 l

LIST OF ILLUSTRATIONS 1

l

-Flaure Title P a p,e 3-1 Potential Release Flow Path for Operation of the Offgas System 31 3-2 Potentiel Release Flow Path for Operation of Mechanical Vacuum Pump 32 3-3 Potential Release Flow Path for Containment within the Main Steam System 33 3-4 Potential Release Flow Path.for RPV Depressuriaation to the Suppression Poel 34 3-5 Potential Release Flow for Flooding the Main Steam Lines Pressurization Between the MSIVs 35 3-6 Potential Release Flow Path for Treatment / Containment within Plant Buildings 36 1-7 Potential Release Flow Path for Operation of MSIV-LCS- 37 v/vi C-4

NE00 31858 Rev. 1 NEDO-30324 ABBREVIATIONS BWR -- Boiling Water Reactor BWROG --

Boiling Water Peactor Owners Group EPG --

Emergency Procedure Guidelines

, HEPA --

High Efficiency Particulate Absolute (Pilters) 4 HVAC --

Heating, Ventilating and Air Conditioning LCO --

Limiting Condition for Operation LOCA --

Loss-of-Coolant Accident MSIV --

Main Steamline Isolation Valve M51V-LCS --

Main Steamline Isolation Valve Leak ge Control System RPV- --

Reactor Pressure Vessel RWCU --

Reactor Vater Cleanup SBGT --

Standby Gas Treatment SJAE --

Steam Jet Air Ejector villvlii C-5

NEDO 31858 Ree, 1 NEDO-30324 ABSTRACT This report assesses potential operator actions to limit radioactive gas release through the Main Steamline Isol. tion Valves (MSIVs), It is concluded that opsrator actions can minimize the MSIV leakage dose contribution from the plant by:

• Control and treatment of any MSIV leakage,

• Contaicaent of any MSIV leakage within the main steam system, and

• Controlled release of any MSIV leakage when containment is not possible.

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NEDO-31858 Rev. 1 NEDO-30324 1.0 IhTRODUCTION In accordance with General Design Criterion $$ (Reference 1) the MSIVs in the main steam lines of a BVR plant are designed to isolate the Reactor Pressure Vessel (RPV) in the event of a break in the steam line avtside the primary centainment , a design basis Loss of-Coolant Accident (LOCA), or other events requiring containment isolation. The redundant fast-closing MSIVs would terminate the RPV coolant blowdown outside the primary containment in suf ficient time to prevent an uncontrolled release of radioactivity from the RPV and ensure that totw1 offsite radiological consequ'ences do not exceed the guidelines of 10 CTR Part 100, " Reactor Site Criteria." Although the MSIVs are designed to provide a leak-tight barrier, it is recognized that some leakage through the valves will occur. The current Technical Specification liett (typically 11.5 scfh) for MSIV leakage is believed to be very conservative. An MSIV leakage dose model based on more realistic analytical assumptions has been developed as part of a separate Boiling Water Reactor Owners Group (BVROG) study. To supplement that model, the BVROG .tSIV Leakage Com-mittee authorized an assessment of potential operator actions to limit the dose consequences of radioactivity releases through the MSIVs.

The purpose o' this report is to evaluate selected potential operator Tetions for preventing MSIV leakage or mitigating its effects. These actions are evaluated relative to their feasibility of implementation, effect on reduction of MSIV leakage, effect on dose consequences and consistency with instructions in the Emergency Procedure Guidelines (EPGs). Other benefits and drawbacks for each potential action are also discussed. Reccannended operator actions for possible inclusion in the EPGs are presee.t-d.

In addition, this report discusses the symptomatic conditions and operator guidelines for which control of MSIV leakage is appropriate.

Based on evaltation of these considerations, a set of reconsnended EPG conditions and steps to addr ss MSIV leakage control was generated (Appendix A) for possible consideration by the Emergency Procedures Committee of .se BVROG.

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-NED0 31858 Reu. 1

l

=' NED0-30324 2.0'

SUMMARY

i Effective operator actions are possible to minimize the dose contribu-tion from MSIV leakage following an event that results in significant i core damage. Two symptomatic conditions for which control of MSIV l leakage is appropriate are recommended. Several additional conditions-for which control of MSIV leakage may be appropriate are also discussed.

These are summarized in Table 2-1. 1

-Nine pot::tial operator actions to limit radioactivity releases through l the MSIVs and consequent dose contributions from the plant were evalu-ated. These are summarized in Table 2-2. Based'on the. discussion in Section 3.0 of this report, operator guidelines are recommended in Appendix A tor-

• - Control and treat any MSIV leakage,

'* Contain any MSIV leakage in the main steam system, and !

• Control the release of any MS!V leakage when containmeat is not possible.

L It'is recommended that-the MSIV Leakage Committee evaluate the dose contribution from the options - recommended- and determine the ' potential . .

for incorporation into normal plant procedures and the EPGs. Should >

this be_decernined appropriate,-the recommended conditions'andl steps to >

' address MSIV leakage control,should be evaluated by the' BWROG Emergency Procedures Committee to ensure the MSIV leakage control actions are

--fully-integrated with_the EPGs, 7

2

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NED0 31858 Rev. I NED0-30324 Table 2-1

SUMMARY

OF POTENTIAL SYMOTOMATIC CONDITIONS FOR iiHICH CONTROL OF MSIV LEAXACE MAY BE APPROPRIATE l Group 1 l Group 2 l Group 3 l l May Be i Not Symptomatic Condition Considered 1 Appropriate l Appropriate { Appropriate

!  ! 1 l l 1

• High main steam line radiation j X l l l l l 1

• High area radiation level near l l l the main steam lines l X 'l l

! l l 1

• High control room air intake l l l radiation l l X l l I i I

• High off-site radioactivity l l l release rate l l X l l l 1 1

• High secondary containment HVAC l { l exhaust radiation l l X l l l l 1 i  !

• High-turbine building HVAC l l l exhaust radiation l l X l l l I i l l

• High offgas pretreatment l l l radiaticn l l X l l

l

• High drywell pressure- l l l X l i I I

• High drywell radiation i l l X l I I I

• Lev RPV water level I l l X l l l 3

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NEDO 31858 Rev. 1 NE00-30324 Table 2-2

SUMMARY

OF OPERATOR ACTIONS CONSIDERED FOR MSIV LEAKACE CONTROL i l Not Operator Action Consfiered l Recommended j Recommended l

• Operation of the offgas system i Plant-Specific l l

• Operation of tba mechanical vacuum ptap l l X l l i I ,

• Containment within the main steam system l l 1

- Closure of main turbine stop, control and l l bypass valve. and isolate main steam lines i X i

- Isolate main cendenser and establish l l turbine seals i Plant-Specific l l

l

• RPV depressurization to the suppression pool I X l l l I i ,

• Flooding of the main steam lines l l

- Upstream of the MSIVs l l o By flooding the RPV  ! l X o With clean water source l X i o Back-filling from main condenser i Plant-Specific l

- Downstceam of the MSIVs i l o With clean water source l Plant-Specific l o'Back-filling from main condenser l Plant-Specific l 1 I I i

• Pressurization between the MSIVs ) i

- With air I l X

- With nitrogen l X l

- With clean water source l X l 1 l l I

• Treatment / Containment within-the plant l l buildings l l

- Operate secondary containment and turbine l i building HVAC. On high exhaust radiation, l i confirm HVAC isolation and start SBGT. l X l On high air intake radiation, start l l control room HVAC pressurization mode. I X i l i i i

• Utilization of the MSIV-LCS l Plant-Specific l I I i 1

• Flooding the primary containment l l X l I 4

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NE00 31852 Rey. 1 NEDO-30324 I

I 3.0 DISCUSSION The approach taken was to identify condations potentially requiring MSIV leakage control. These conditions were evaluated to determine which  ;

would be symptomatic of radioactivity releanes through the MSIVs. )

Similarly, a list of potential operator actions for mitigating MSIV i leakage and its consequences was identified. These ntions were evalu-ated relative to their effect on reduction of radioactivity releases through the MSIVs and f rom the plant, their feasibility of implementa-tion, other benefits, drawbacks and ce .sistency with the EPGs.

3.1 Assessreent of Operator Actions for Control of MSIV Leakage The following operator actions were considered for control of MSIV Leakage:

• Operation of the offgas system

• Operation et the mechanical vacuum pump

• Containment within the main steam system RPV depressurization to the suppression pool

• Flooding of the main steam lines.

Upstream of t.he MSIVs Downstream of the MSIVs Pressurization between the MSIVs Treatment / Containment within the pla su11 dings Utilization of the MSIV-Leakage Control System (LCS)

• Flooding of the Primary Containment C-ll L

I NEDD 31858 Rev. 1 t

NE D0-30324 4

!bese operator acticas were evaluated with respect to the following factors to determine the appropriateness of the action for control of MSIV leakage.

• The flow path for any prtential release

• Benefits of the action a Drawbacks of the action

• Magnitutv of modification required to enable implementation of the action

• Feasibility and merit of integration with the curr?nt EPCs.

The following Sectiocs discuss each of the above operator actions taking into account each of the above considerations. It 4s assumed that the MSIVs have already been closed.

3.1.) Operation of the Offgas System

1. Potential Operator Action This attien maintains main condenser vacuum using the Steam Jet Air Ejectors (SJAEs). Any MSIV leakage would be processed through the offgas system by opening the sain turbine bypass valves to the main cocdenser. SJAEs and turbine seals must be provided with an auxiliary steam source. If available, the circulating vster system is operated to candense any steam passing the leaking MSIVs and reaching the mai< .ondenser.

2. Flow Path for any Potential Release (risure 3-1)

MSIV leakage would travel through the main steam linen and to the main condenser via ths main turbine bypass valves. The non-l condensible radioactive gases would be evacuated from the main 6

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NEDO 31858 Rev. I NED0-30324 condenser by the SJAE and processed through the offgas system recombiner, condensers, charcoal beds or delay tanks, High Effi-ciency Particulate Absolute (REPA) filters and out the main L ack.

This option is most effective for those cases in which the integrity of the piping downstream of the M51Vs is preserved.

Should there be a break in the mein steam line, the r dtoactive gas leaking past the MSIVs could escape into the secondary contaarnent ,

or turbine building depending on the location and severity of the break. This could also cause gas flow rates to exceed the capacity of the SJAEs or offgas system.

3. Benefits of the Action Unless there is a latte steam line break, this action would provide the optimum treatment of the radioactive gases prior to release from thtt plant. With the main condenser vW er vacuum and the turbine b> Tass valves open, any MSIV leakage would be evacuated to Lbe main condenser thus precluding or minimizing any leakage from elsewhere in the main steam system. This will be true even for steam line breaks up to some small size (typical $JAE air removal capabity is 40 cfm at 1.0 in. Hg. ba vpressure).

This flow path would also permit the cold trapping of iodine and ,

volatiles plus servbbing ans platcout that would occur in the main steam lines and main condenser and condensation of any steam in the Irakage. This would continue in the offgas system recombiner, condensers, holdup piping, charcoal adsorbers, and EPA filters.

This action would maximite the holdup time, other than containment within the plant, of the radioactive gases prior to release from the plant at an elev.ted point. For example, a typical charcoal adsorber system would provide an estimated 6.8 day Xenen and 9.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> Kr>Ttoo holdup in the absence of a steam line break. ,

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HLD0 31858 Rev. 1 WEDO 30J24 '

4. Erawbacks of the Action A major steam line break would make this action ineffective since  ;

it relies on a relatively intact flow path to the main condenser, This action requires auxiliary steam, since SJAE drive steam and turbine seal steam are normally supplied from the main steam lines

$. Ability ef Plant to implement Many plants do not have an auxiliary steam supply designed to operate the SJAEs without modification. In addition, some plants do not have an auxiliary steam supply designed to operate the  ;

turbine seals.

6. Tessibility and Merit of EPG lategration Since the main condenser 's the preferred beat sink in the EPCs, this action should not be in conflict with the EPGs.

3.1.2 Operation of the Mechanical Vacuus Pump

1. Potential operator Action This action maintains main condenser vacuum using the mechanical vacuum pump. Any MS!V leakage is drawn into the main condenser through the turbine bypass valves and is pumped directly to the main stack. Turbine seals are maintained with an auxiliary steam source. If available, the circulating water system is operated to condense any steam passing the leaking MSIVs and reaching the main condenser.

2. Flow Path- for any Potential Release (rigure- 3-2)

In the absence of a steam 14,ne break M31V leakage would travel through the main steam lines and to the main condenser via the main turbine bypass valves. The radioactive g.ses would be evacuated 8

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NED0 31858 Rev. 1 l

ACD0-30324 i i

from the main condenser by the mechanical vacuum pump and pumped  ;

through a holdup volume (typically 1.7$ minute) and out the main stack for the plant.

Should there be a break in the main steam line, the radioactive gas leaking past the MS!Vs could escape into the secondary containment or turbine building depending on the location and severity of the break.

3. Benefits of :he Action Unless there is a steam 11 e ireak outside the capacity of the mechanical vacuum pump (typical mechanical vacu2m pump air removal capacity is 2I00 cfm at 15 in. Hg. absolute), this action would permit the cold trapping of iodine and volatiles plus scrubbing and plateout that would occur in the main steam lines and main con-denser and condensation of any steam in the leakage. This flow path would also provide an elevated release from the plant.

4. Drawbacks of the Action This actica provides essentially no boldup (typically 1.7$ minutes) prior to release from the plant. There are usually no condensers, filters or g charcoal adsorbers in the flow path to reduce the radioactivity release from the plant. The mechanical vacuum pump also isolates on high steam line radiation making it unavail.

able when MSIV leakage control is needed. Conceivably, following 1SIV isolation, the natn steam line radiation trip sight clear

>;ending on the amount of MSIV leakage and the radioactivity level of the leakage. A major steam line break would make this action ineffective since it relies on a relatively intact flow path to the -

main condenser.

This action requires auxiliary steam to maintain turbine seals since seal steam is normally supplied from the rain steam lines.

This is necessary to maintain sain condenser vacuum and protect the main turbine shaft when the main condenser is under vacuum.

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_ - _ - . _ _ - - _ - _ . . _- ~ - . - _ - _ . _ _ . - _ . - . __ _ _ _ _ _ - -

NEDO 31858 Ree, 1 NErc-30324

$. Ability of Plaut to implement Taking into consideration the main steam line high radiation trip of the mechanical vacuws pump, most plants $bould be able to take this action witbout any plant modifications. The only limitation is having the capability t.o maintain turbine seals from an aux-111ary steam supply. $ose plants do not have an auxiliary bv11er supply to the turbine seals.

6. Feasibility and Merit of EPG Integration it should be feasible to integrate this action into the EPGs.

However, because of the many drawbacks the merit of this action is questionable.

3.1.3 Containment within the Main Steam System

1. Potential Operator Action This action isolates the main steam system by closing the main turbine stop, control, and bypass valves as well as all branch Itne valves connected to the main steam lines. The Boiling Vater Reactor (BVR) 6 shutoff valves between the MSIVs and the main turbine stop valves would also be closed. To contain any leakage past the main turbine valves, the main condenser is isolated by closing the main condenser vacuum breakers, estabitshing turbine seals with an auxiliary steam source and operation of the circulating water system to condense any steam in the leakage.

Verification o' no steam line break is desirable for the success of i these actions.

I

2. Flow Path for any Potential Release (Figure 3-3)

MSIV leakage would be contained within the main steam lines, tur-bine and main condenser. Any leakage from these boundaries would travel into the turbine building and out the turbine building Heat-ing, Ventilating and Air Conditioning (HVAC) exhaust system to an i elevated release point. If the turbine building INAC exhaust f i

10 l

C-16 )

i l

-- . - ~ _ .

1[DO 31858 Rev. 1 NED0-30324 system is isolated, the radioactive gases will be contained within the turbine building to a degree dependent upon the leakage con-straints of the building.

Should there be a break in the main steam line, the radioactive gas would be discharged into the secondary containment or turbine building depending on the locatiou and severity of the break.

3. Benefits of the Action Depending on the'1eak tightness of the valves in all lines con-nected to the main steam lines and the leak tightness of the main condenser, this action would contain any MSIV leakage within the sin steam lines and main condenser (providing turbine seals are established) and no MSIV leakage will be released to the environ- ,

ment. Unless there is a steam line break, this action would also peruit the cold trapping of iodine and volatiles plus scrubbing and platecut that would occur in the main steam lines and main con- a denser and condensation of any steam in the leakage. This action would delay release of non-condensible radioactive gases from the plant. ,

4. Drawbacks of the Action A steam line break would make this action ineffective since it relies on intact main steam lines and main condenser.

5. Ability of Plant to Implement Depending on the plant design several options are available to most plants and the majority of these actions can be taken from the main control room. BWR 6 plants have double disc gate shutoff valves downstream of the MSIVs which could be operated from the main control room. Although there is no leak tightness criterion for these valves, they vould delay the transport of any radioactive

gases. All lines connected to the main steam lines including the main turbine stop and control valves could also be isolated, most of these from the main control som.

11 C-17

NEDO 31858 Rev. 1 NEIO*30324 If the plant ban the capability to maintain turbine seals from an auxiltary steam supply, the main turbane seal. thould be maintained to improve the leak tightness of she main condenser. for this cena dttien the main condenser vacuum breakers shot 0d be closed. With the main condenser intact, seals being maintained and vacuute breakers *losed the main turbine bypass valves could be left open j to provide a larger volume for any MSIV leakage and pertait additional cold trapping of iodine and volatiles and steam con-densation. Should the integrity of the main condenser be lost, the main turbine bypass valves could be closed to try and contain any non condensible radioactive gases within the main steam lines alone. All of these actions depend on the integrity of the main steam lines and main condenser being intact.

l

6. Feasibility and Merit of EPG Integration It should be feasible to integrate this action into the EPGs and may merit incorporation into the Radioactivity Release Control Guideline.

3.1.4 RPV Depressurization to the Suppression Pool

1. Potential Operator Action This action opens one or more SRVs to depressurize the RPV to the suppression pool.

2 Flow Path for Any Potential Release (rigure 3-4)

Radioactive gases will be discharged into the suppression pool through the open SRVs as the RPV is depressurized. MSIV leakage will be contained within the constraints ry what other mechanism is being used to control MSIV leakage. For e'zample, if " Containment within the Main Steam System" was being used to control MSIV leakage the potential radioactive gas release path would be as l' described in Section 3.1.3 Item 2 (rigure 3-3).

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NtD0 31858 Rev. I h NE!O-30324 i

3. Benefits of the Action Reducing RPV pressure wet.d reduce the leakage flow through the MSIVs. Depressuttring the RTV to the suppression pool would contain within tne primary containment the radioactive gases released f rom the RPV through tho $RVs.

4. Drawbacks of the Action Depressurizing the RPV to the suppression pool adds heat to the pool which is undesirable under emergency conditions. However. RPV depressurization is the correct action for high suppression pool temperature conditions.

Some plants have SRVs which reclose at approximately 50 psig.

Depressurization would be limited to this RPV pressure for plants with this type of SRV.

$. Ability of Plant to implement All BWR/2 through 6 plants have the capability to depressurize the RPV to the suppresston pool from the main control room ustng SRVs.

6. Feasibility and Merit of EPG Integration The EPGs currently provide guidelines for RPV depressuritation with the SRVs and guidelines for handling high suppression pool tempera-tures. RPV depressurization to minimize MSIV leakage may merit incorporation into the Radioactivity Release Control Guideline.

This action would require integration with the current guidelines to depressurire the RPV in the RPV Control Guideline.

The Radioactivity Release Control Guideline currently requires Emergency RPV Depressurization if offsite radioactivity release rate approaches or exceeds the release rate which requires a General Emergency. The source of the offsite radioactivity release rate could be from MS!V leakage and therefore this guidance is already provided in the EPGs.

13 C-19

NE00 31858 R;v, 1 NED0-30324 3.1.5 Flooding of the Main Steam Lines i

1. Potential Operator Action This action floods the main steam lines either upstream or down.

stream of the MSIVs. The main steam lines can be flooded upstream or downstream of the MSIVs by:

• Tilling the main steam lines from a clean water external s~irce through test connections, or

• T1ooding the main condenser and backfilling the main steam lines through the main steam line drains (capability is plant-specific depending on elevation relationship between the main condenser and the MSIVs).

Additionally, the main steam lines can be flooded upstream of the MSIVs by flooding the RPV above the main steam nor:1es.

2. Flow Path for any Potential Release (Figure 3-5)

MSIV .eakage would be directed to the main condenser via the main steam line drains or contained within the main steam lines.

3. Benefits of the Action flooding would delay the transport of any radioactive gases prior to any release from the plant. Depending on the location of the flooding and the RPV pressure, flooding could retain the non.

condensible radioactive gases within the RPV or primary rontain.

ment. Flooding upstream of the MS1Vs such that flow of water would be into the RPV would optimize retention of the non-condensible radioactive gases within the RPV or primary containment.

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NL00 31858 Rev. I NEDO-30324 l

4. Drawbacks of the Action Since the main steam llarb are insulated and there is a great amount of steel support structure near the M51Vs, the MSIVs and  ?

steam lines in this ^4ciotty will remann at elevated temperatures for some time. Flos g the main steam lines with cooler water could lead to warpage of the seats of the MSIVs and greater MSIV leakage.

Tlovding the RPV to fill the main steam lines upstream of the hSIVs would result in leakage of water with much higher specific activity than the gas which was leaking before. Therefore, it would be desirable to flood with a clean source of water. If there were a large LOCA, it might not be possible to flood the main steam lines from the RPV.

Tlooding the main steam lines downstream of the MSIVs would have to i take into consideration the capability of the hanget system of the main steam lines to support flooded main steam lines without failure. Those plants which can support 3 flooded main steam line may require the hangers be planed to accomplish this. Accessia bility of the hangers may be a problem due to their locrtion or area radioactivity levels. A significant amount of water would also be required to flood the maan steam lines dovnetream of the MSIVs and this might not he possible if there is a large steam line break. Pressurizing the flooded steam line downstream of the MSIVs would tend to open the pilot valve on the hSIVs and could lead to increased MSIV leakage. .

5. Ability of Plant to implement The main steam lines can be flooded upstream of the MSIVs by ,

flooding the RPV from the main control room providing there is not a large LOCA which precludes ficoding of the RPV. Anu water leakage by the MSIVs could be drained to the main , er by openiog the main steam line drain valves downstres. <: *tboard 15 C-21

NE00 31858 Rev. 1 NEDO-30324 MSIVs. This would preclude filling the main steam lines downstrese of the MSIVs and direct the MSIV leakage to the main condenser rather than into the turbine building should there be a main stese line break.

Some plants could. fill the main steam lines either upstream or downstream of the MSIVs by flooding the main condenser and back.

filling the main steam lines through the main steam line draitis,

, This depends an the elevation of the main condenser telative to the >

elevation of the MS!Vs, the elevation to which the main condthser structurally can be flooded and the piping arrangement of the main steam lines and steam line drains.

Most plants could flood the main steam lines either upstream or downstream of the MSIVs from an external source through test connections on the main steam lines. These typically might not be accessible due to radiation levels, but by having these connections made up in advance to handle this emergency situation the MSIVs ,

could be flor.Jed in this manner using a clean water source,

6. Tessibility and Merit of EPG Integration

-It should be- feasible to integrate this action into the EPGs and may merit incorporation into the Radioactivity Release Control Guideline as one of the actions to take under dograded conditions.

Utilisation of EPC Contingency 6. RPV F1,00 DING, may be possible by including some actions to maintain RPV vater level above the top of the main steam not:1es if this contingency were to be used for this reason.

3.1.6 Pressurization Between the MS!Vs

1. Potential Operator Action This action pressurites the steam line between the inboard and outboard MSIVs and maintains this pressure at some small pressure L- 16 C-22

NED0 31858 Rev. 1 NE D0- 303 24 above RPV pressure. Air, nitrogen and water were considered as the pressurtzing media introduced through test connections on the main steam line between the MSIVs.

2. Tlev Path for any Potential Release (rigure 3-5)

MS1V leakage would be directed to the main condenser via the main steam line drains or contained within the main steam lines.

3. Benefits of the Action Air, nitrogen, and water were considered as potential pressurtaing media since all three are typically available in the plant. The relatively small volume between the MSIVs enhances the ease of pressurization. Pressurization between the MSIVs cabances the possibility of containment of the radioactive gases within the RPV and primary containment. The lower heat transfer of a gas (air or nitrogen) for e pressurtaing fluid by comparison to using water, reduces the possibility of MS1V seal warpage and increased leakage.

4 Drawbacks of the Action Unless the pressure is controlled at some small value above RPV pressurc, this action will tend to open the MSIV pilot valve on the inboard MSIV thereby causing a backflow of pressurizing fluid into the RPV. This is undesirable if the supply of pressurizing fluid is itaited or if there is a pipe break inside the drywell (leakage into the RPV vould pressurize the RPV and primary contain-ment if there was a primary system break inside the primary con-tainment). The use of air and leakage into the RPV as described above could lead to a detonable mixture of hydrogen and oxygen in the RPV or primary containment if there were a severe LOCA.

Pressurization with water would increase the possibility of MSIV seal warpage and increased MSIV leakage as described previously.

17 i

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NE00-31858 Rev. 1 NE D0- 30324 Any leakage past the outboard MSIV could be drained back to the main condenser by opensag the main steam 1sne drain valves down-stream of the outboard MSIVs. This would preclude filling the main steam lines downstream of the MSIVs and preclude leakage to the turbine building through a steam line break.

5. Ability of Plant to implement Currantly, there are no provisions to pressurize between the M51Vs.

Either temporary or permanent provisions would have to be installed to enable ese of this action. A simple system could be connected into the test connections on the main steam lines between the MSIVs.

6. Peasibility and Merit of EPG Integration It should be feasible to integrate this action into the EPGs and may merit incorporation into the Radioactivity Release Control Guideline if plant modifications were made to accomplish this action. Adding non-condensibles to the primary containment should minimize the constraints on use of dryvell sprays in the EPGs.

3.1.7 Treatment / Containment within the Plant Buildings

1. Potential Operator Action This action confirms that the secondary containment and turbine building KVAC exhaust systess are in operation or places them in operation. Upon receipt of high KVAC exhaust radiation, confirm that these systems isolate and the Standby Gas Treatment (SBCT) system is started. In addition, the control room KVAC system is started in its self-contained recirculation or pressurization mode when habitability of the control room is threatened due to MSIV leakage.

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2. Flow Path for Any Potential Release (rigure 3-6)

MSIV leakage would travel through the main steam lines and could leak into the secondary containment or turbtne building tarough any one of the following mechanisms:

• Steam line break

• Valve packing leakage

• Main condenser leakage

• Main turbine leakage if the secondary containment or turbine building HVAC exhaust systems are operating, the radioactive gases would be processed through the charcoal beds and REPA filters of the secondary con-tainment or turbine building HVAC exhaust

• systems and discharged out the elevated KVAC release point for the plant.

If the secondary containment HVAC system is not operating, any radioactive gases leaking from the main steam lines into the secondary containment would be directed to the SBGT system and processed through the bBGT charcoal beds, HEPA filters and out the main stack for the plant.

I f tt.t turbine building HVAC system is not operating, any radio-active gases leaking from the main steam lines , main condenser or main turbine would be contained within the turbine building subject to the leakage constraints of this building. Building leakage would be a direct, untreated release from the plant.

• Some plant designs may not have charcoal beds or HEPA filters.

19 C-25

1 NE00 31858 Rev. 1 1

NEDO-30324

3. Benefits of the Action This flow path would allow the cold trapping of iodine and vola-tiles plus scrubbing and plateout that would occur in the main steam lines and main condenser (provided the turbine bypass valves are open). Tae holdup time vould be increased by processing the radioactive gases through the charcoal beds and HEPA filters of the HVAC exhaust systems prior to release at an elevated point from the plant. If the HVAC exhaust system is inoperable or isolates on high exhaust radiation, the radioactive gases would be processed thrcugh the charcoal beds and HEPA filters cf the SBGT system when they are released to the secondary containment or tLty would be contained within the turbine building subject to the leakage constraints of this building. Containment within the turbine building would provide some holdup within this building prior to release from the plant. This might be the only release control of -

radioactive gases possible for a steam line break outside the primary containment.

4. Drawbacks of the Action Any releases from the main steam lines or main condenser would create a personnel access dose problem in the secondary contatnment or turbine building. This could also lead to a habitability problem in the main control room. Loss of the HVAC exhaust systems could result in an untreated, uncontrolled release from the plant if MSIV leakage is entering the turbine building.

5. Ability of Plant to Implement Most plants have some treatment system in their secondary contain-ment and turbine building HVAC exhaust systems. Even those which may not have an KVAC exhaust treatment system do have an HVAC isolativo on high exhaust radiation. All BWR/2 through 6 plants br < : 4 SBb7 system.

20 C-26

NED0 31858 Rev. 1 l

! WEN 30324 i

Containment and access to the secondary contatn.sent or turbane building could be controlled for excessive high radiation levels to these buildings. The control room HVAC systee vould require switching to the self contained recirculatico or pressurtration mode (depending on plant design) to allow habitability of the eatn control room under these conditions.

6. Teasibility and Merit d EPO Integration Secondary containment HVAC and SBGT operation plus isolation of secondary containment HVAC when secondary containment HVAC exhaust radiation levels exceed the isolation set point are already covered in the Secondary Containtient Control Guideline of EPG, Rev. 3 and therefore should not provide a conflict. Similar operator actions for the turbine building HVAC should be feastble to integrate into the LPGs and may merit incorpt, ration into the kadioactivuy kelease Control Guideline. Operation of the main centrel room HVAC in the pressurization mode should be fetaible to integtste into the EPGs and may merit incorporation into the Radioactivity Release Control Guideline.

3.1.8 Utiltestion of the M51Y LCS

1. Potential Operator Action This action initiates the inboard and outboard MSIV LCS and the SBOT system when the MSIV-LCS interlocks clear and high radio-activity release due to MSIV leakrge is indicated.

2. Fitw Path for Any Potential Releases (rigure 3-7)

Any inboard MSIV leakage would be evacuated through the main steam line drains between the MSIVs by the Anboard MSIV-LCS to the SBGT system. The radioactive gases would be processed through the SBCT charcoal beds and HEPA filters, and out the main stack of the plant. Likewise, any MSIV leakage through both MSIVs would be 21 C 27

NED0 31858 Rev. 1 NE D0-303 24 t

evacuated through the downstreat main steam line dratus by the outboard MSIV LCS, through the SBGT system and out the main stack of the plant.

Should there be a break in the main steam line, the radioactive gas could be dischstged into the secondary containment or turbine building depending on the location and severity of the break.

1

3. Benefits of the Action this action would provide some treatment and holdup of the radio-active gases by the SBGT system charcoal beds and KEPA filters, and provide an elevated release from the plant. Except for steam line bre' s (typical MSIV LCS blower capacity is 100 scfm at 60 in.

wi .r suction) downstream of the MSIVs, this action would enhance avcess to the secondary containment and turbine building due to reduction in airborne radioactivity levels from MSIV leanage. Use of this system under these conditions would also reduce the risk of the main control room becoming uninhabitable. The inboard system isolates at high flow (100 scfh maximum) to minimize the offsite release ccatribution from this system,

4. Drawbacks of the Action The inboard system is a low flow system which isolates at flows exceeding 100 scfh per main steam line. The inboard system would therefore not be available for treating large MSIV leakages. The inboard and outboard systems cannot be operated unless the RPV pressure and main steam line pressure are below 20 psig. For plants with the negative pressure type MSIV-LCS (present plants have this type system), there is also some question for the non-steam line break case whether treatment through the HSIV LCS and SBGT system is more beneficial in reducing offsite doses than containment within the main steam and main condenser systems.

22 C-28

f1ED0 31358 Rev. 1 l NEIC-30324

5. Ability of Plant to Implement bWR/2 and 3 plants do not have this system. Approxte4ately 10 BWR/4 and 5 plants have this systete. All domestic and some foreign BWR/6 plants have this system.

6. Feasibility and Merit of F.PG Integration It should be feasible to integrate this action into the FJ'Gs and may merit incorporation into the Radioactivity Release Control Guideline. The offsi.te dose contribution due to operation of this

.em should be evaluated to determine if it is desirable to use this system as a backup when other methc.ds to control MS!V leakage are unsuccessful.

3.1.9 Flooding of the Primary Csntainesent

1. Potential Operator Action This action floods the primary contair.msnt to a level above the main steam lines to fill the main stean lines upstream of the MSIVs through a large prisary system break inside the primary contain-ment. This is accomplished by discharging some high volume water source (e.g., fire water) into the primary containment.

2. Flow Path for Any Potential Release (Figure 3-5)

MSIV leakage would be directed to the main condenser via the main steam line drains or contained within the main steam lines.

3. Benefits of the Action This was considered as a possible last resort alternate means to flood the main steam lines (discussed in Section 3.1.5) under LOCA conditions. For example, should there be a break of sufficient size to prevent flooding of the RPV, then flooding of the primary containment might be the only way to flood the RPV. In so doicg the main steam lines could also be flooded. The benefits discussed in Section 3.1.5 would also be applicable for this actaon.

23 C-29

tied 0 31858 Rev, 1 NEDO-30324

4. Drawhacks of the Action l

If there is not sufficient water to flood the RPV, there may not be suf ficient water to flood the primary containment either. However, l

plant service water or fire water might be available in the primary '

cont 6inment but not to the RPV making this alternate feasible. In eddition to the drawbacks to flooding the main steam lines dis-cussed in Section 3.1.5, this method would take considerable time to reach its intended objective.

5. Ability of Plant to Implement Most plants have source of fire water and/or service water to the primary containtent which could be used for flooding the primary containment under extreme emergency conditions. Many plants are also capabic of witlttanding the hydrostatic loads that result from flooding tL3 t riuary containeen as high as the RPV flange as evidenced by a w rvey taken during the development of the EPGs, 6 Feasibility and Merit of EPG Integration The EPGs currently address bigt ster loads in the primary con-t sinment in the Primary Contatt f t Control Guideline. This action i would reqeire integration with these actions. However, due to the relatively long time required to benefit from this action, this degraded case may not merit consideration,

^2 i, Recoramended Operator Actions for Mt!V Leakage Control based on the previous assessment, .he following are recommended operator actions for MSIV leakage control '4Lcb could be incorporated intc, the EPGs if MSIV leakage coutrol guir ines were to be addressed:

t

• Operation of the Offgas Systra

• Containment within the Main llteam System l

24 C 30 l

NED0 31858 Rey, 1 3 NED0-303:4

• RPV depressurization to the suppression pool

• Pressurization between the MSIVs with nitrogen or water

• Treatment / containment within the plant buildings

• Operation of the MSIV-LCS

• Tlooding of the main steam lines with a clean water source 3.3 Conditions Requiring Control of MSIV '.eakage if a condition which requires MSIV isolation does not exist, then actions to limit radisactive gas release through the MSIVs are not required. Consequently, the prlmary prerequisite for control of MSIV leakage is a condition which requires !!SIV isolation.

In addition to this prersquisite, the following conditions for which control of M51V leakage may be appropriate were evaluated:

• High main steam line radiation

• High area radiation level near the main steam lines

• High control room air intake radiation --

• High off-site radi *rtivity release rate

• High secondary containment KVAC exhaust radiation

• High turbine building HVAC exhaust radiation

• High offgas pretreatment radiation

• High drywell pressure 25 C-31

__ _ ___ _ _ _ ___ _ _9

I NED0 31858 Rev. 1 NE DO-303 24

• High drywell radiation

• Low RPV water level 3.3.1 Evaluation of Conditions Requiring Control of MSIV Leakage The conditions in the above list were categorized into three groups:

Group 1 Those directly symptomatic of conditions for which control of MSIV leakage is appropriate.

Group 2 Those symptomatic of ccaditions for whieb control of MSIV leakage may be appropriate.

Group 3 Those not appropriate for MSIV leakage control.

The Jecond category includes those conditions which might indicate a problem with MSIV leakage but could also result from a problem from some other source.

3.3.2 Classification of Conditions Requiring Control of MSIV Leakage This section discusses the classification of each of the conditions and the basis for this classification.

The " Recommended EPG Conditions and Steps to Address MS!V Leakate Control" (Appendia A) Contains the Group 1 and 2 conditions w'.th a suggested action level in a guideline format.

1. Group 1 Conditions The following conditions fall into Group I where control of MSIV leakage is appropriate.

• Main steam line high radiation -- Group 1 High radiation in the main steam lines downstream of the MS1Vs with the MSIVs closed is directly indicative of radioactive ses leakage through the MSIVs. If the flow (leakage) and 26 C 32

NED0 31858 Rev. 1 NED0-30324 acttvity level are high enough, control of MSIV leakage would be warranted. The action level in Appendix A of {KXX mr/br (MSlY leakage Dose Contribution)]* is enytstoned as a level which could lead to offsite doses above 10 CFR Part 100 limits. The main steam line radiation monitors are typically set to trip at 3 timas normal background resding at 100%

reactor power. With the MSIVs closed, leakage through the MSIVs having significant activity level to warrant operator action should be discernible on this instrumentation in the main control room.

• High area radiation level near the main steam lines Group 1 High area radiation near the main steam lines with the MSIVs closed is also a direct indication of radioactive ses release through the MSIVs. The maximum normal operating radiation level in Appendix A should be considered as a level which could lead to offsite doses above 10 CTP,Part 100 Itmits if released from the plant. Area radiatior monitor locations near the main steam lines are plant spe-ific and some plants may not have area monitors close enough to the main steam lines to warrant action on this condition. Additilaal mont-tors could be added should these opera tor actions be judged warranted.

2. Group 2 Conditions A number of conditions fall into Group 2 where control of MSIV leakage may be appropriate. These conditions might indicate a problem that results from leakage of radioactive gases through closed MSIVs but could also result from a problem from some other source as discussed below and therefore might not always warrant control of MS!V leakage.

• The bracketed action level indicates it is a plant-specific value. This action level should be determined by radiological calculation.

27 C-33

NED0 31858 Rev. 1 NED0-30J 24

• High control room air tutake radiation -- Group 2 A break in the main steam lines or lear'se from the main condenser combined with radioactive gas leakage through closed MS1Vs could contribute to this condition. However, the high control room air intake radiation could also be from some other radioactive source in the turbine building near the control room ventilation intake. The action IcVel in Appendia A of [1 mr/hr (control room air intake radiation Limiting Condition for Operation (LCO)] was chosen so that the re-circulation mode of the control room ventilation system could be initiated to maintain habitability of the control room.

• Offsite radioactivity release rate above the Alert level --

Group 2 As in the case of high control room air intake radiation, the source may be through the closed MS1Vs, but could also result from a release from a source in some other building such as a release from the radwaste building. The action level in Appendix A of [3 Ci/sec (release rate which requires an Alert)) was chosen at a level where offsite release is a concern. This condition is consistent with the current entry condition to the Radioactivity Release Control Guideline of EPG, Revision 3.

• High secondary containment HVAC exhaust radiation -- Group 2 Here again the source may be through the closed MS1Vs with a break in the main steam lines in the secondary containment downstream of the MSIVs, but could also result from a break or leak in some other radioactive system like the Reactor Vater Cleanup (RWCU) system. The maximum normal operating radiation level in Appendix A she'ald be considered at a level where the secondary contairment HVAC would isolate to limit radioactive release from the plant. This condition is con

• sistent with the current entry condition to the secondary Containment Control Guideline of EPG, Revision 3.

28 C-34

NEDO 3!B58 Rey, 1 NE DO-303:4

• High turbine buildtog NVAC exhaust radiation Group 2 Once again, the source may be through the closed MSIVs but could also result from some other radioactive source in the turbine building. This source could result from some release in the KF.PA or charcoal treatment portion of the turbine building NVAC exhaust system. The maximum normal operating radiation level in Approdix A should be considered at a level wbere the turbine building KVAC system would isolate to limit radioactive release from the plant.

• High offgas pretreatment radiation -- Group 2 This would be a good indication of leakage of radioactive gases through closed MSIVs if the offgas system were in operation. However, for most plants the SJAls (which provide the driving force for this system) would be out of service with the MSIVs closed. Without system flow the pretreatment radiation instrumentation would not provide a meaningful reading. The maximum normal operati g radiation level in Appendix A should be considered as a level which could lead to offsite doses above 10 CTR Part 100 limits if released from the plant.

3. Group 3 Conditions These are conditions considered but deemed not appropriate for control of MSIV leakage.

• High drywell pressure -- Group 3 High drywell pressure could indicate a primary system break in the drywell which could lead to radioactive gas release through closed MSIVs if significant RPV coolant inventory is lost. However, this condition could also result from loss of dryvell cooling where no radioactive gas release would be 29 C-35

NED0 31858 Rev. I NE 0-30324 involved. In addition, closure of MS1Vs is not required on high dryvell pressure. Low RPV water level would be more likely to lead to conditions requiring control of MSIV leakage. High dryvell pressure is an entry condition to the RPV Control Guideline of EPG, Revision 3 in anticipation of an RPV coolant inventory threatening event and acticos appro*

priate to maintain RPV water level are adequately covered in this guideline.

• High drywell radiation - Group 3 High radiation level in the drywell is indi:stive of a primary system break in the drywell and potential problem with MSIV leakage. However, if there is no MSIV leakage and the radio-activity is contained in the drywell, there would not be a need to control MSIV leakage. An entry condition could be con-sidered at a level which could lead to offsite doses above 10 CFR Part 100 limits if released from the plant.

• Low RPV Water Level -- Group 3 Low water level of itself would not necessarily cause a MSlv leakage concern even if _below the isolation set point for the MSIVs as long as the RPV water level is maintained above the top,of the active fuel or adequate core cooling is otherwise assured. Low RPV water level is an entry condition to the RFY Control Guideline of EPG, Revision 3 to provide appropriate actions to aaintain adequate core cooling.

30 C-36

NEDO 31858 Rev, 1 NEDO-30324 d

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NE00-31858 Rev. 1 i NEDO-30324 4 RETERENCES

1. General Design Criteria $5, " Reactor Coolant Pressure Boundary Penetrating Containment, of Appendix A, " General Design Criteria,"

to 10 CFR Part 50, " Licensing of Production and Utilization Fecilities."

2. BWR 1 through 6, " Emergency Procedure Guidelines," Revision 3, December 8,1982.

'l 38 C-44

NEDO 31858 Rey, 1 NE Do-30324 APPENDIX A RECOMMENDED EPG CONDITIONS & STEPS TO ADDRESS MSIV LEAXAGE CONTROL PL'RPOSE

'The purpose of this document is to recommend enhancements to the EPGs to limit non-condensible' radioactive gas release through the MSIVs.

CONDITIONS The conditions for which control of MSIV leakage is appropriate are conditions which require MEIV isolation with either of the following:

Maint steam line radiation level above [XXX mr/hr (MSIV Leakage Dose Corcribution. An area radiation level near the main steem lines above the maximum normal operating radiation level. Additional contitions for which control of MSIV leakage may be appropriate are any of the following: Control room air intake radiation above [1 mr/hr (control room air intake radiation LCO)].

1. -s Offsite radioactivity release rate above [3 Ci/sec (release rate which requires an Alert)).*

A secondary containment HVAC exhaust radiation level above the maximum normal operating radiation level.* A turbine building HVAC exhaust radiation level above the maximum normal operating radiation level.

• Currently-part of EPG, Revision 3 A-1 C-45

4A NEDO-31858 Ree 1 NEDo-30324

• An offgas pretreatment radi ation level above the maxtmum normal operating radiation level.

OPERATOR ACTIONS The operator actions identified to date for control of MSIV leakage are as follows: If while executing the following steps secondary containment HVAC exhaust radiation level exceeds (20 mr/hr (secondary containment HVAC isolation setpoint)):

• Confirm or mant y initiate isolation of secondary containment HVAC, and Confirm initiation of or manually initiate SBGT [only when the space being t eacuated is below 212'F).

If while executing the following steps:

• Secondary containment HVAC isolates, and
• Secondary containment HVAC exhaust radiation level is below (20 mr/hr (secondary containment HVAC isolation setnoint)},

restart secondary containment HVAC. #24 If while executing the following steps turbine building HVAC 24haust radiation level exceeds [15 mr/hr (turbine building HVAC isolation setpoint)], confirm or manually initiate isolation of turbine build-ing HVAC.

• Currently part of EPG, Revision 3 A-2 C-46

NEDO-31858 Rev. 1 NED0-30324 If while executing the following steps: Turbine building HVAC isolates, and Turbine building HVAC exhaust radiation level is below (15 mr/hr (turbine building HVAC isolation setpoint)], restart turbine building HVAC. If while executing the following steps control room air antake radiation approaches or exceeds (1 mr/hr (Control room air intake radiation LCO)], confirm or manually initiate the control room HVAC pressurization mode.

1. Confirm or initiate MSIV isolation.

1.1 Operate available SJAE through the offgas system. 1.2 Depressurize the RPV and maintain cooldown rate below #14 (100'F/hr (RPV cooldown rate LCO)) vith SRVs only when 915 suppression pool water level is above (4 ft. 9 in. (eleva- #17 tion of top of SRV discharge device)]. If one or more SRVs are being used to depressurize the RPV and the continuous SRV preumatic supply is or becomes unavailable, depressurize with sustained SRV opening.* 1.3 When the SJAE and offgas system become unavailable, break condenser vacuum. 1.4 If the main steam line radiation level approaches or exceeds [XXX mr/hr (MSIV Leakage Dose Contribution)), close the: Main turbine stop valves. Main turbine control valves, Main turbine bypass valves Mawn condenaer vacuum breakers, and isolate all lines connected to the main steam lines.

• Currently part of EPG, Revision 3 A-3 1

C-47 '

   . . ~ . .           .                . _ . _ . . -         -        .            .__    ._ _ . _ _ . _

NEDO 31858 Rev. 1 , NEDO-30324 i Establish main turbine seals and initiate the main condenser-circulating water system, i 2.- When:

                      *:       The airborne radiation level near the main steam lines approaches or exceeds-[XXX ar/hr (MSIV Leakage Dote Contribution)), and                         -
                                                                                                                ^
                      * . -MSIV-LCS-interlocks clear, initiate the MSIV-LCS, IfIthe MSIV-LCS cannet be initiated or the airborne radiation level near the main steen lines continue to exceed [XXX tsr/hr (MSIV Leakage Dose                     ,

Contribution)].:[when RPV pressure is below YY psig (Maximum MSIV Leakage-

                    -Control'_RPV Pressure:)-

(2.1PressurizethespacebetweentheMSIVs~withnitrogen. 2.2 Maintain press ure between the M"IVs XX psis (Mintrum MSIV Leakage Control..Bachpressure) above RPV pressure.]

                       '2.3-Open (B21-T020 (main steam line. drain valve downstream of outboard                -;

_MSIVs)). [If pressure between the MSIVs cannot:be maintainediabove XX psig (Minimum MSIV Leakage Control Backpressure), fill' the r:ain steam lines

                    -betweentheMSIVs.with' water.)
              - 3.'  If offsite radioactivity release rate approaches'or exceeds-[91 Ci/see
                    .(release' rate which' requires a. General Emergency)) and-a primary system"
                                                    ~
                    -is~ discharging into an area =outside^the primary and secondary contain-K <

ments, EMERGENCY RPV DEPRESSURIZATION IS' REQUIRED enter (procedute developed from the RPV Control Guideline).at -[ Step RC-1) and execute it ' concurrently-with'this procedure.*

             '-* Currently part of EPG, Revision 3 A-4 C-48

NE00 31858 Rev. I l NEDO-30324 { SPECITIC CAUTIONS The following specific cautions are identified with the symbol # at the point noted in the guidelines. . CAUTION #14

• Do not depressurize the RPV below [100 psig (HPCI or RCIC low pres-sure isolation setpoint, whichever is higher)] unless motor driven pumps sufficient to maintain RPV vater level are running and available for injection.

CAUTION #15

• Open SRVs in the following sequence if possible: [SRV opening sequencej.

CAUTION 17

• Cooldown rates above [100*F/hr (RPV cooldown rate LCO)] may be required to conserve RPV water inventory, protect primary contain-ment integrity, or_ limit radioactive release to the environment.

CAUTION 24

• Bypassing high drywell pressure and low RPV water level secondary containment HVAC isolation interlocks may be required to accomplish this step.
• Currently part of EPG, Revision 3 A-5/A-6 C-49

NED0 31858 Rev, 1 NEDO-30324 APPENDIX B PARTICIPATING UTILITIES This report applies to the following plants, whose Owners participated in the report's development Utility Plants (s) Boston Edison Company Pilgrim 1 Detroit Edison Company Fermi 2 Hatch 1, 2-Georgia Power Comprav GPU Nuclear Corporation Oyster Creek 1 Iowa Electric Light & Power Company Duane Arnold Long Island Lighting Company Shoreham Mississippi Power & Light Company Grand Gulf 1 Niagara Mohawk Power Corporation Nine Mila Point 1,2 Northeast Utilities M111stona 1 Northern States Power Company Monticello Pennsylvania Power & Light Company Susquehanna 1,2 Philadelphia Electric Company Peach Bottom 2,3 & Limerick 1,2 Public Service E1cetric & Gas Hope Creek 1 TenLessee Valley Authority Browns Ferry 1,2,3 Washington Public Power Supply WP-2 l I l B-1/B-2 C-50

h. GE Nuclear Energy fl$"EUES I { I 4}}