ML20195F813

Forwards Summary Re Proposed Generic Ltr to BWR Licensees Transmitting Technical Findings Related to MSIV Leakage & MSIV Leakage Control Sys.Review Re Resolution of Generic Issue C-8 Scheduled for CRGR Meeting 96 on 860917

ML20195F813
Person / Time
Issue date:	09/12/1986
From:	Taylor M NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Bernero R Office of Nuclear Reactor Regulation
Shared Package
ML20151C834	List: ML20151C832 ML20151C859 ML20151C868 ML20151C882 ML20151C904 ML20151J001 ML20151K338 ML20151S856 ML20154L830 ML20154L836 ML20195F797 ML20195F813
References
FOIA-87-714, RTR-NUREG-0800, RTR-NUREG-1169, RTR-NUREG-800, RTR-NUREG-CR-4330 NUDOCS 8609180240
Download: ML20195F813 (16)

Text

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SEP 121986 MEMORANDUM FOR: Robert M. Bernero, NRR Richard W. Starostecki, IE Richard E. Cunningham, NMSS Denwood F. Ross, RES Clemens J. Heltemes, Jr., AE00 Joseph Scinto, OGC THRU: John E. Zerbe. Director Regional Operations and Generic Requirements Staff F1t0M: Merrill A. Taylor, Senior Program Manager Regional Operations and Generic Requirements Staff

SUBJECT:

SUWARY AND ISSUE IDENTIFICATION FOR CRGR MEETING NO. 96 Enclosed for your informatiot; and use is the ROGR staff sumary associated with the proposed Generic letter to BWit licensees transmitting technical if ndings (i.e., NUREG.1169) related to MS!Y lakage and MSIV Leakage Control Systems.

This matter relates to the ultimate resolution of generic issue C.8 concerning MSIV Leakage Control Systems, and it is scheduled for CRGR review at Meeting No. 96 on Wednesday, September 17, 1986, in room 6507 MNBB.

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Merrill A. Taylor Regional Operations and Generic Requirements Staff

Enclosure:

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Sumary and Issue Identification on CRGR Review Item, Meeting No. 96 September 17, 1986 l

IDENTIFICATION I

1. Proposed Generic Letter to all Licensees of Boiling W6ter Reactors on l Availability of NUREG-1169 (Generic Issue C-8) l l

2. NUREG-1169, entitled "Technical Findings Report as Related to Generic Issue C-8, MSIV Leakage and Leakage Treatment Methods."

DESCRIPTION AND OBJECTIVE CRGR is being requested to recomend in favor of transmittal of a proposed generic letter to all BWR licensees. This letter would concern the availabil-ity and case-by-case use of technical findings to be published (NUREG-1169) toward resolution of generic issue C-8. Generic issue C-8 and the NUREG-1169 findings relate to the need for a safety-related Leakage Control System (LCS) to mitigate accident-caused leakage of radioactivity past the BWR MSIVs.

Transmittal of the proposed generic letter and the NUREG-1169 technical find-ings would not, according to the NRR staff, represent the NRC's ultimate reso-lution of generic safety issue C-8. It is expected that CRGR will be briefed at this meeting No. 96 on what should constitute the final resolution for this i generic issue. According to NRR, NUREG-1169 does not incorporate any new guidelines or requirements and the final resolution of generic issue is ex-pected to follow within about 12 months after publication of NUREG-1169. In the interim, the proposed generic letter is inviting a case-by-case relaxation (or relief) from existing regulatory requirements on the basis of NUREG-1169.

Additionally, publication cf NUREG-1169 would set forth a generally favorable staff evaluation of recomenuations provided in early 1984 by the BWR Owners Group (BWROG). The BWROG recomendations concern MSIV leak experiences and methods to enhance MSIV leak integrity. The proposed generic letter to BWR l licensee.s is, however, silent on the point of whether the letter constitutes a j specific NRC endorsement of the BWROG recomendations. It is also unclear if I any NRC followup action on requirements on licensees is intended as a result of the BWROG reconinendations.

The technical findings in NUREG-1169 are significant and can represent staff "

conclusions, interpretations and regulatory positions quite different from I those that have heretofore resulted in requirements on a number of the more  !

current BWRs to design, install, test, and maintain a safety-related LCS. The i findings in NUREG-1169 lead to the conclusion that the safety benefits of a safety-related LCS have been significantly overstated by past NRC positions taken pursuant to 10 CFR 50, Appendix A (GDC 54 and 55), Regulatory Guides 1.3, 1.5 and 1.96,10 CFR 100, Standard Review Plan (Sections 6.7, 15.6.4 and 15.6.5) and in current technical specification requirements. In essence, the i NUREG-1169 findings are founded on a considerably different modeling basis than l has traditionally been used (i.e., the DBA approach) in the licensing cecision i' process--the result being that staff finds that alternative MSIV leakage con-trol strategies can exist and that these alternatives can rely on use of exist-ing non safety-grade systems (as opposed to a safety-related LCS). It is unclear, however, whether the staff intends imposition of any regulatory i

requirements (e.g., fomal post-accident procedures) on the use of non safety-grade equipment as an alternative to use of a safety-related LCS. Through the use of probabilistic techniques, insights from nt.w source term work, more de-tailed fission product transport models, and risk-related consequence analyses, a key conclusion is reached in NUREG-1169 that "MSIV leakage would play no role in minimizing public exposure when considered with a core melt scenario and containment failure."

BACKGROUND See the proposed generic letter to BWR licensees, and NUREG-1169 (Section 2),

transmitted from H. Denton to J. Sniezek by memorandum dated August 21, 1986.

ISSUES

1. The proposed generic letter to licensees (and applicants?) would invite a case-by-case relaxation of existing LCS requirements as the staff has previously interpreted such a requirement to exist under GDC 54 and 55.

THis staff interpretation has been applied to some BWRs via the SRP and by specific regulatory positions in Regulatory Guide 1.96 (RG 1.96 is in Attachment I hereto for CRGR information. Note that rather prescriptive regulatory)

NUREG-1169positions

. CRGR mayarewish settoforth thatthe explore may now staff's be to intent negated use a NUREG by findings in document to implement new or differing regulatory interpretation positions or requirements contrary to existing EDO policy (Refer to the third from the bottom paragraph on page 2 of the proposed generic letter where a case-by-caese relaxation is invited as an "interim" regulatory action).

2. CRGR may wish to determine if the proposed case-by-case implementation of the NUREG-1169 findings by the staff would require (or is consistent with) use of the established process for exemptions to GDC 54 and 55 which is interpreted through use of Regulatory Guide 1.96.

3. CRGR may wish to determine staff's intent to use NUREG-1169 and the pro-posed generic letter as the vehicle to endorse the BWROG recommendations concerning MSly leak integrity and testing. For example, is staff in-tending any followup actions that would require licensees to implement the BWROG recomendations or is this expected to remain a completely voluntary action by applicants and licensees as indicated by the proposed generic letter? CRGR may also wish to explore staff's views on whether previous  !

MSIV testing strategies have resulted in biasing the test results on MSIV leak rates to the high-side--perhaps resulting in unnecessary leakage re-pairs for the velvas, plant downtime and occupational exposure increases.

If so, what are the staff's views on requiring improved MSIV test strate-gies by licensees to improve MSIV seating and leak test integrity?

4. SECY-86-76 dated February 28, 1986, addressed the staff's proposed imple-mentation plan for the Severe Accident Policy statement and for regulatory use of new source-tem information. Concerning the new source term infor- l mation, the staff has already proposed to make a number of changes to prior regulatory requirements. Attachment 2 hereto lists these from .

SECY-86-76 and CRGR has been previously briefed on this matter. CRGR l should note that Regulatory Guide 1.3 is one of those changes imediately i planned. The issue of MSIV leakage and the safety-related LCS requirement I

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also relates to RG 1.3 as well as RG 1.5 and 1.96 and the new source term I information. CRGR may wish to explore reasons why the ultimate resolution of generic issue C-8 is not planned for integration with those activities planned pursuant to SECY-86-76 or vice-versa.

5. RES has underway (pursuant to Commission policy guidance) an effort to examine existing rules and regulatory requirements that may have marginal importance to safety. The aim of this RES effort is to see if some of the regulatory requirements could be relaxed or eliminated to reduce regula- .

tory burdens without compromising public health and safety. The results from this RES pilot phase of work has been reported (NUREG/CR4330) and is the subject of a SECY paper (now at EDO and for transmittal to the Comission). One subject of the RES pilot work is the MSIV Leakage Con-trol System. PNL (who also provided the principal technical assistance for NUREG-1169 on this same subject) concluded in NUREG/CR-4330 that elim- )

ination of the MSIV-LCS requirement and by disabling the LCS in plants i currently having them could result in a significantly favorable balance i for cost-safety benefits (by order of magnitude or more relative to $1,000 per person-rem). CRGR may wish to understand how the RES and NRR staff plan to integrate these various initiatives in bringing about an ultimate resolution of generic issue C-8 and what this resolution may involve. For i example, will the staff accept the disabling or removal of a safety-  !

related LCS in any of the current BWRs or require this to be done? If so,  !

what burdens on licensees would be anticipated? If not, would the staff propose to continue to carry the LCS requirements in existing technical l specifications and enforce these? Does the staff plan to revise, delete or drop in entirety those RG 1.96 regulatory positions for an LCS? Does .

the NRR staff agree with those PNL resolutions presented by NUREG/CR-4330 l and if not why not? l 1

6. Resolution of Generic Issue C-8 was given a "high. priority" ranking based I on the GIMS and on the NRR prioritization schemes used in NUREG-0933 and NUREG/CR-2800 (PNL-4297) Supplement 1. CRGR may wish to explore the bases for NRR assigning a "high-priority" prioritization score to issue C-8 given that this ranking appears to have been considerably overstated (even with the degree of conservatism already introduced, e.g., factor of 100 by this prioritization scheme). CRGR may also wish to determine the totality l of NRC resources already expended to date on this generic issue, e.g., by  !

staff, by PNL for the prioritization work (NUREG/CR-2800), by PNL for the NUREG-1169 work, and by PNL (via RES) for the NUREG/CR-4330 work, and what -

further resources are anticipated to be spent through ultimate resolution. l l

7. CRGR should note that the first paragraph of the proposed generic letter '

states that LCS have been installed on most BWRs. This sounds inconsis-tent with NUREG-1169 (bottom of pg. 2-1) where it is pointed out that .

about ten BWR 4 and 5 plants and all BWR 6 plants have an LCS. Analyses l in NUREG/CR/4330 assumes that 25 BWR plants already have the sunk costs of an installed safety-related LCS. CRGR may wish to reconcile with staff the actual number of affected BWR plants.

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CRGR Meeting No. 96 Attachlments to CRGR Issue Sumary

1. RegulatoryGuide1.96, Revision 1(June 1976)

2. Excerpt from SECY-86-76 (Table 2.3 of Enclosure)

3. Excerpts from NUREG/CR-4330 l

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REGULATORY GUIDE 1.96 DESIGN OF MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEMS FOR BOILING WATER REACTOR NUCLEAR POWER PLANTS A. INTRODUCTION isolate the reactor system in the event of a break in a steem line outside the primary containment, a design-General Design Criterion 54, Piping Systems Pene- basis LOCA, or other events requiring containment trating Containment," of Appendix A "General Design isolation. In the casc of a steam line break, the isolation Criteria," to 10 CFR Put 50,"Licensing of Production valves would terminate the blowdown of reactor coolant and Utilization Facilities," requires,in part, that piping in sufficient time to prevent an uncontrolled release of systems penetrating primary containment be prodded radion'etivity from the reactor ves:elto the environment.

with leak detection, isolation,and containment capabill- In the case of a LOCA, the valves would isolate the ties having redundancy, reliability, and performance reactor from the en.ironment and prevent the direct capabilities that reflect the importance to safety of iso- release of fission products from the containment.

lating these piping systems.This guide describes a basis The valves aae part of the reactor coolant pressure acceptable to the NRC staff for implementing General boundary. As such, they are Quality Group A compo-Design Criterion $4 with regud to the design of a leakage control system for the main steam isolation nents and their integrity must be maintained by strict

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valves of boiling water reactor (BWR) nuclear power inservice inspection and testing requirements. However, 4

• operating e :perience has indicated that degradation has plants to ensure that total site radiological effects do not exceed guidelines of 10 CFR Part 100, "Reactor Site occasionally occurred in the leak tightness of main steam Criteria," in the event of a postulated design basis isolation valves, and the specifed low leakag require.

loss-of coolant accident (LOCA). If an applicant pro- ments have not alwsys been maintained. .

-e:e: to use a method different from that described in this guide for implementing General Design Criterion 54 the staff has considered the need to provide addi-tional featuns to enarre the low-leakage characteristics

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with regud to the control or limitation ofleakage past the main steam isolation valves of a boiling water of the main suam isolation valves in the event of a Postulated design. basis loss-of coolant accident.1 The reactor, the acceptability of the attemative rnethod will use of a leakage co trol system would reduce direct be determined by the staff on a case by cue basis. The untreated leakage from the isolation valves when isola-Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the tion of the primary sptem and the containment is required.

regulatory position.

B. DISCUS $lON The results of staff analyses have indicated that calculated doses resulting from the maximum leakage Direct cycle boding water nuclear pour plants supply steam directly from the reactor vessel to the j l in its letters on the construction permit reviews of the Duanc turbine via main steam lines. The main steam lines i installed on current BWR plants ne provided with dual Arnold and shoreham plants (Demnber 18, 1949) and the l Jane A. nam M Uanuan M,1% the Ad% '

qulek closing isolation valves. These valves function to Committee on Reactor Safeguards acted that additknal features to control main neam isolatice v Jve leakage should be

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necessary, considering effects resulting from a LOCA, 10. The plant should be designed to permit testing of ancluding (a) missiles that may result from equipment the operability of controls and actuating devices of the failures,(b) dynamle effects associated with pipe whip LCS during power operation to the extent practical and and jet forces, and (c) normal operating and accident-testing of the complete functioning of the system during caused local environmental conditions consistent with plant shutdowns.

the design-basis event. Further, any portion of the LCS that is Ouality Group A and is located outdde the

11. The LCS should be designed so that (a) any primary containment structure should be protected from effects resulting from use of a Duld sealing medium c.g.,

missiles, pipe whip, and jet force effects originating thermal stresses, pressures associated with flashing, and outside containment so that containment integrity is thermal deformations under the loading conditions maintained. associated with the activated system, will not affect the structuralintegrity or operability of the main steam lines

3. The LCS should be capable of performing its or main steam isolation valves and (b) any deformation safety function following a LOCA and assumed sinde of isclation valve internals wul not induce leakage of the acthe fauure (including fauure of any one of the main main steam line isolation valve beyond the capacity or steam isolation valves to close). capability of the LC3.

4. The LCS should be designed so that effects 12. Equipment should be provided, as part of the resulting from faDure of a single active component of the LCS or other systems, to prevent or control valve stem leakage control system will not affect the integrity or packing leakage or other direct leakage from main steam operability of the main stean lines or main steam line isolation valves outside containment. If such equip.

teclation valves. ment is not part of the LCS, it should meet the same design standards as the LCS.

5. The LCS should be capable of performing its safety function foDowing a loss of all offshe power D. IMPLEMENTATION coincident with a postulated design. basis LOCA.

The purpose of this section is to provide information

6. The LCS should be designed with sufficient to applicants and licensees regarding the NRC staff's

, capacity and capability to controlleaksge from the main plans for using this regulatory guide.

, steam lines for as long as postulated accident conditions require containment lategrity to be maintained. Thh guide refleets cuntnt regulatory practice.These.

fore, except in those cases in which the applicant

. ... 7. The LCS may be manually or automatically Proposes an acceptable altemative metho! for comply.

actuated and should be designed to permit actuation ing with specified portions of the Commission's regula.

within about 20 minutes after a postulated design-basis tions, the rnethod described herein is being and will LOCA. Thu time period b ccaside ed to be consistent continee to be used in evaluating submittals for con.

with loading requirements of the emergency electrical struction permit and operating license applications, buses and with tersonable times for operator action. Although this guide may recomrner,d backfitting in certain cases that have already been docketed, as

8. Instrumentation and circuits necessary for the described below, it does not require it. Such require.

functioning of the LCS should be designed in accordance ments will be formulated on an Individual basis pursuant with standards applicable to an engineered safety to 10 CFR $50.109.

feature.

1. In the case of boiling water reactor plants for

9. The LCS controh should include interlocks to which construction permits were bsued prior to March prevent inadvertent operation of the LCS. In particular, I,1970 (see Table 1), applicants and licerisees should laterlocks should be provided to prevent damage to the continue the established inservice inspection programs to LCS or possibly to the main steam system due to ensure that isolation valves are maintained in such a inadvertent opening of any LCS isolation valves when. manner that leaksge is within Technical Specification evr the pressure in the connecting main steam piping' limits. If the valve inspections show recurring problems exceeds LCS design pressure and to prevent significant with excessive leakage, the staff recommends that releese of radioactbity to the environment. Where consideration be given toinstallation of a supplementary appropriate to the LCS design, interlocks should be leakage centrol system.

psovided to preclude direct communication with the post.LOCA conts'nment atmosphere in the event that 2. In the case of bouing water teactor plants of the inboard main steam isolation valve does not fully designs preceding the BWR 6/ Mark til design and for close. AU such controls and interlocks should be which construction perTrJts have been issued after March activated from appropriately designed safety systems or 1,1970 (see Table 1), the staff recomrnends that circuits. applicants und licensees instaU a supplernental leakage 1.96 3

I APPENDIX A SUPPLEMENTAL DESIGN FEATURES FOR OVALITY GROUP A PORTION OF LEAKAGE CONTROL SYSTEM PIPING This appendix provides supplerantal design features If the calculated maxirnum stnes range of Equation for any portion of pipbg for a leakage control system (10) exceed: 35 , the stress ranys calculated by both (LCS) that connects to steam system piping betnen Equation (12) and Equation (13) in Paragraph NB-3653 inner and outer containment isolation valves of the main should not exceed 2.4Smand the cumulative usage steam system for either single or dualbarrier contain- factor should be less than 0.1.

ment structures. Such piping, up to and including the l

fhst isolatfor valve in the LCS piping, should be constructed to meet the requirements ot' the ASME

2. Welded attachments to this portjon of piping for pipe supports or other purposes should be avoided.

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Code in Subarticle NE 1120 of Section til, supple.

mented by the following:

3. The number of circumferential and longitudinal

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1. The following design stress and fatigue limits should not be exceeded:

4. The portloa of piping extending to the first

a. ".e maximum stress range should not exceed uto e as sort as pmM.

5. The design of piping restraints should not require

b. The maximum stress range between any two welding directjy to the outer surface of the piping.

load sets (including the zero load set) should be calculated by Equation (10) in Paragraph NB 3653, 6. The design of this portion of the leakage control ASME Cooe, Section 1)), for upset plant conditions and system should permit the conduct of inservice examina-an operating basis earthquake transient. tions required by the rules of Section X] of the ASME Boiler and Pressure Vessel Code, and the extent of I If the calculated maxirnum stress range of Equation examinations during each inspection interval should (10) exceeds 2.45, but la not greater than 3Sm . the provide 100 percent volumetric examination of the cumulative usage factor should be less than 0.1. piping welds within this portion of piping.

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STA.NDARD OFFICE OF NUCLEAR REACTOR REGULATION REVIEW PLAN 6.7 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM (BWR)

REVIEW RESPONSIBILITIES  !

Primary - Auxiliary Systems Branch (ASB) i Secondary - None l I. AREAS OF REVIEW l 1

Direct cycle boiling water reactor (BWR) plants have redundant quick-acting isola-tion valves on each main steam line from the reactor to the turbine. In the event of a loss-of-coolant accident (LOCA), any leakage of contaminated steam through l these valves is controlled by a leakage control system. The leakage control system must satisfy the requirements of General Design Criteria 2, 4, and 54.

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The review of the main steam isolation valve leakage control system (MSIVLCS) l covers the entire leakage control system including the source of the sealing medium, '

g- if any, and pumps, valves, and piping to the points of connection or interface with t

, the main steam supply system. Emphasis is placed on the components of the leakage L' control system that are required to remain functional following a design basis LOCA. )

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. 1. ASB reviews the design of the MSIVLCS and essential subsystems to assure their ability to function following a postulated LOCA including the loss of offsite power. The system is reviewed to determine that:

a. A malfunction or failure of an active component of the system, or loss of the source of sealing fluid, if any, will not impair the functional performance of the system. l

b. The failure of nor. seismic Category I equipment or components will not '

have an adverse effect on the ability of the system or components to I function.

c. The capability of toe system to perform its intended safety function is maintained assuming a single active failure of a main steam line isola-tion valve.

Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN star. deed review plans are prepared foe the guidance of the of fice of Nuclear Reactor Regulation staff toeponsible for the review of eppl. cations to construct end operate nuclear power plants. These doewments are made available to the public as part of the Commession's policy to nnform the nuclear Industry and the general public of regulatory procedures and poGcies, standeed review plane ete not substitutes for regwistory guides or the Commission's regulations ont comellance with them is not reguited. The standard review pfen sectione see heyed to the standard Format and Content of safety AnalyOs Reports for Nwctear Power Plants.

Not es sections of the standard Format have e corresponding toview plan.

Pwb4.shed standard review piene will be revised periodically. ee appropriate, to accommodate comments and to ref.ect now Irifoema-tion and eaperience.

Comments and suggestiene for improvement will be considered and should be sent to the U.s. Nuclear Reguistory Ceramtssion.

Ottece of Nwetear Reactoe Regulation. Washington o C XM$

Related review evaluations will be performed by other branches and the results will be coordinated by ASB to complete the overall evaluation of the system.

The evaluations provided by other branches are as follows. The Structural Engineering Branch (SEB) determines the acceptability of the design analyses, procedures, and criteria used to establish the ability of seismic Category I structures housing the system and supporting systems to withstand the effects of natural phenomena such as the safe shutdown earthquakes (SSE), the probable maximum flood (PHF), and tornado missiles as part of its prin&ry review responsi-bility for SRP Sections 3.3.1, 3.3.2, 3.5.3, 3.7.1 through 3.7.4, 3.8.4, and 3.8.5. The Mechanical Engineering Branch (MEB) determines that the components piping and structures are designed in accordance with applicable codes and standards as part of.its primary review responsibility for SRP Sections 3.9.1 through 3. 9.3. The MEB also determines the acceptability of the seismic and quality group classifications for system components as part of its primary review responsibility for SRP Sections 3.2.1 and 3.2.2. The MEB also reviews the adequacy of the inservice testing program of pumps and valves as part of its primary review responsibility for SRP Section 3.9.6. The Materials Engineer-ing Branch (MTEB) verifies that inservice inspection requirements are met for system components as part of its primary review responsibility for SRP Section

, 6.6, and, upon request, verifies the compatibility of the materials of construc-tion with service conditions. The Equipment Qualification Branch (EQB) reviews the seismic qualification of Category I instrumentation and electrical equipment and the environmental qualification of mechanical and electrical equipment as part of its primary review responsibility for SRP Sections 3.10 and 3.11, respec-tively. The Instrumentation and Control Systems Branch (ICSB) and the Power Systems Branch (PSB) determine the adequacy of the design, installation, inspec-tion, and testing of all electrical components (sensing, control, and power) s required for proper operation as part of their primary review responsibility j for SRP Sections 7.1 and 8.0, respectively. The Containment Systems Branch (CSB) reviews the MSIVLCS to assure thst no malfunction can adversely affect containment integrity as part of its primary review responsibility for SRP Sections 6.2.1 and 6.2.4. The review for fire protection, technical specifica-tions, and quality assurance are coordinated and performed by the Chemical Engineering Branch, Licensing Guidance Branch, and Quality Assurance Branch as part of their primary review responsibility for SRP Sections 9.5.1, 16.0, and

' 17.0, respectively.

For those areas of review identified above as being the responsibility of other branches, the acceptance criteria and their methods of application are contained in the SRP sections corresponding to those branches.

II. ACCEPTANCE CRITERIA Acceptability of the MSIVLCS, as described in the applicant's safety analysis report (SAR), is based on specific general design criteria and regulatory guides.

An additional basis for determining the acceptability of the MSIVLCS is the degree of similarity of the design with that of previously reviewed plants.

The design of the MSIVLCS is acceptable if the integrated system design is in accordance with the following criteria:

1. General Design Criterion 2, as related to structures housing the system and the systen itself being capable of withstanding the effects of earth-quakes. Acceptance is based on meeting position C.1 of Regulatory Guide 1.29 and position C.1 of Regulatory Guide 1.96. l 6.7-3 Rev. 2 - July 1981

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d. Design provisions have been made that permit appropriate inservice inspection and functional testing of sptes components. It is accept-able if the SAR information delineates a testing and inspection program and if the system drawings show the necessary design provisions ,

to accomplish the testing program. I

2. The reviewer determines that the safety function of the MSIVLCS will be maintained, as required, in the event of adverse environmental p'1enomena such as earthquakes. The reviewer uses engineering judgment, the results of failure modes ar.d effects analyses, and the results of reviews performed under other SRP sections indicated in subsee. tion I of this SRP section to l determine that the failure of nonessential portions of the system or of other systems not designed to seismic Category I and located close to essen-tial portions of the system, or of nonseismic Categoty I structures located close to essential portions of the system, will not preclude operation of the essential portions of the MSIVLCS. Reference to SAR sections describir,g site features, the general arrangement and layout drawings, and the tabula- '

tion of seismic design classifications for systems and structures will be nece s sary. Statements in the SAR that the above conditions are met are acceptable.

3. If the leakage control system is one using a fltid sealing medium:

a. The system design is revi2wed to determine that the quantity of sealing fluid needed for an effective seal of the valves has been provided.

Independent analyses, using the pump performance curves in the SAR, are made to assure that the design and tne location of the pump and

. components are such as to maintain the appropriate net positive suction i head (NPSH) requirements and provide a continuous supply of sealing

-' fluid during the full course of an accident.

b. The system design is reviewed to determine that effects resulting

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from the sealing fluid, such as thermal stresses, pressures associated with flashing, thermal defomations, and other effects will not effect the structual integrity of the steam lines or the main steam isolation valves, or lead to excessive leaktge of the valves. This portion of the review is done on a case-by-case basis. The ASB also accepts ,

the system design if a statement in the SAR commits to performing j calculations or functional testing to demonstrata that the above l conditions are met. I l

4. The MSIVLCS is reviewed to verify that instrumentation, controls, and inter-locks designed to standards appropriate for an engineered safety feature l are provided to actuate the system in the event of a design basis LOCA, I and to prevent inadvertent actuation. Interlocks to prevent inadvertent  ;

operation of the leakage control system that are actuated by signals from the reactor protection, engineered safety feature, or containment isolation i systems are acceptable. A statement in the SAR that such instrumentation, I controls, and interlocks will be provided is acceptable for construction permit (CP) review.

5. The systen performance requirements, P& ids, MSIVLCS drawings, and the results of failure modes and effects analyses are reviewed to assure that the systes can function following a design basis LOCA assuming a concurrent single active failure, including the failure of a single main steam isola-tion valve to close. The reviewer evaluates the analyses presented in 6.7-5 Rev. 2 - July 1981

Y. IMPLEMEKTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plan for using this SRP section.

Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, .

the method described herein will be used by the staff in its evaluation of complianc2 with Commission regulations.

Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulatory guides.

VI. REFERENCES

1. 10 CFR Part 50, Appendix A, General Design Criterion 2, "Design Bases for Protection Against Natural Phenomena."

2. 10 CFR Part 50, Appendix A, General Design Criterion 4, "Environmental

, and Missile Design Bases."  :

3. 10 CFR Part 50, Appendix A, General Design Criterion 54, "Piping Systems Penetrating Containment."

4. Regulatory Guide 1.29, "Seismic Design Classification."

5. Regulatory Guide 1.96, "Design of Main Steam Isolation Valve Leakage Control

, Systems for Boiling Water Reactor Nuclear Peder Plants."

i l

- 1 6.7-7 Rev. 2 - July 1981

l

'A yy

• 4,ac o r 2 f*

. ,ts.rwif Je w " y S'A CA'M mac?sm 4. PG l

)

l Table 2.3 Sumary of Expected Accomolishments source Term Related Changes Issue For Coment Revised SRP Section 6.5.2 9/86 Specifying The Need For Spray Additives In PWRs Issue For Coment Regulatory Guide 1.3 And The 9/86. l AppropMate Section Of The SRP On Fission Product Scrubbing In Suppression Pools (BWRs)

Issue For Coment Proposed Changes To 10 CFR 50.47 2/87 .

and 10 CFR 50, Appendix E On Emergency Planning tevise MRR office Letter 16 With Respect To The 2/87 .

Use Of Source Terms In Safety Issue Evaluation .

. Issue For Cement Changes In Containment Leak -

3/87 Rate Requirements Including Potential Changes In 10 CFR 50 Appendix J Revise 10 CFR 50.49 And Regulatory Guide 1.89 6/87 With Respect To The Radiation Environment For Ecuipment Qualification, For Coment By Issue For Coment Revisions Of Siting CHteMa 10/87 (10 CFR 100) Based On New Source Ters Infor1 nation Issue For Coment Revised Regulatory Guide 1.97 12/87 l On Accident Monitoring And Management l

-- -,-,-.-e.,--, , . , _ - - _ - - g . ---

l .

l I

y),4(e) 20.102(b)

PART 20 e STANDARDS FOR PROTECTION AGAINST RADIATION

~

l (, (3) A dose of 0.1 rsd due to neutrons l cause any individual in a restricted I 20 6 Interpretations.

or high energy protons; 1 w area to neem in any pulod W one f (4) A dose of 0 05 rad due to parueles Except as specificauy authorised by 3 calendar quarter from radloscuve ma.

heavier than protons and with sufft j Comrnission in wriung, no inter- terial and other sources of rad 14Uon a

- putauon of W meaning of the ngu- [a total occuptuonal dose in excess of cient!!energy eye: It is more toconvenient reach thetolens of the ! lations in this part by any officer or 2 the standards specified in the foDow-meas, etnployee of the Commtmalon other W table:

uretothe than neutron determine the neutron flux,doseor equlvajent,8:

in an a writtenhereral interpntation Counsel by the wiU be recognised t rada, as provided in paragraph (cX3) o Reus na Cmutam ouwsa of this accuon, one rem of neuttoa rs _be binding upon the Commtwon.

diauon snay, for purposes of the regw e t******'"8 * * ' " "

MQ7 lauons in this part, be assume <* t4 be "*

,7~

equivalent to 14 million neutrons per

,a s., er .a.d"."T, i$

seer square centimeter ines$ent upn the I 20J Communtentions. (

body; or, if there exists sufficient in. Except where otheraise specified in - .

formauon to estimate sith reasonable this part, au corranunications and re- (b) A licensee may permit an ladivid-accuracy the approstmate distribution ports ecncerning the regulatans in < ual in a restricted area to reestu a in energy of the neutrons, the incident 2 this part should be addressed to the tcta) occupauonal cose f.o the whole number of neutrons per square centJ- E Executive Director for Opersuona, ,*: bNy greater than that permitted sneter estirnated equivalent to one tem from the following may bc a U.S. Nuclear Regulatory Cotemia*;n.

• un.4er paragraph (a) cf this secuon, table: Wuhington, D.C. 20555. Comeunica. I NevTnow FLur Dose Ecurymans gM ns, nports, ad appUeshona may be (1) *During provided. any etM.dar quarter W douvered in person at the Commis- tota] occuptuonal dose to the whole alon's offlees at l'f17 E Street NW , body thaD not exceed 3 rems; and I , ,.,,,,.r,, l. a m.~ Wash!aston, D.D.; or at 1920 Norfolk g *y,'g ~ Avenue, Bethesda. Maryland. - (2) The dose to the uhole body,

[ *""*'"**"

• • ='

• a when added to the accumulated occu.

se *

• i *'7 s

'78 -

patlonaJ done to the whole body, than l asA Interniseen eesessen not exceed 5 (N-18) rems where "N" r remdroments Otte appresuL equalt the indhidual's age in years at

,T 7,7. ' E (a)The Nuclear Regulatory his last birthday; and e

o or aos . _ ~~ are. e ano . e sm =='*ah has sabenitted the , (3) The licensee hu determined the suo informatioa ceDection '-- - -ts j individual's accurnulated occupauonal U% l,',g.

f.e s

contaloed in this part loY' OfSee of t and W (NB

,, dose to the whole body on Form NRC-4 or on a clear and legible record con.

as  ;

m.w{ a) approval u W by the PsW) for l taining

• an the information required in I,h U a that form; and has otherwise comptled N,, i 3,n. u Reduction Act of 1980 (44 UAC. 3801 et sith the requirements of I 20.102, As k OMB has approved he ,

l ornatioe coDection contained la this part d.control L rte s le bod dee ed to i (d) For determining exposures to X aumber 313(N1014, clude any dose to the whole body.

or gunma rays up to 3 Mev, the dose (b)1he approved laformetlos gonads, active blood forming orgsns, head and trunk, or tens of eye.

limits specified in il 20.101 to 20.104, .: couection requirements costaland la thb inclusjve. may be assumed to be equiv. 3 pan appaar is ll Riot. M108. M108, __ "

alent to the "air do:e". For the pur. '* 30.108, Nuos. E2)S E302. 2811. I H.le2 Determlastion of prior does.

pose of this part "att dose' rneans that l M401,30.402, M401, M408. E40r. (a) Each licensee shall require any the dose is appropriate calibrated measured byinstrument a properlF in*I 30.40s. and 30.408. individual, prior to first entry of the (c) This pari contalas laformation indhidual into the licensee's restricted air at or near the body surface in the couectice requirements in addition to area during each employment or work

, region of highest dosage rate. those approved under the control anaisnment under such circumstarses number specHied la paragraph (e) of thle that the individual wiu receive or is .

sectice. These Information conection likely to receive in any period of one  !

I 20.5 t' nits of radioactidy. requirements and the control sambere calmdar quarter an occupadonal dose (a) Radioactivity is commonly, and for purposes of the regulations in this under which they an approved are as fob

(,",y Qjt , ,

part shall be, meuured in terms of dis. (1)la { { 30.101 and 30.102. Form and i 20.104(a), to disejose in a writ.

A integrations per unit time or in curies. NRC-4 la approved under control ten, signeo statement, elther: (1) That E "  : the individual had no prier occupa.

U nal dose during the current calen.

' One curie = 3.1 x 10" disinte:Tations ' (2) fm orm NRC 4 W approud under control number N '8 dar quarter, or (2) the nature and

[' per tions second (dps)= (domi per minute 2.2x10" disintegra.

Commonly arne unt of any occupational dose used submultiples of the curie are the 0008.

which the individual may have re.

millicurie and the microcurie: #

C ceived during that specifically identl.

(1) One millicurie (mCH '= 0.001 Puucsstat. Doses,1xvna, AJrD , fled Current calendar quarter from curie (C1) '= 3.1 x 10 ' dpa. sources of radlaUon possened or con.

(2) One tnicrocurie (pCD '=0.000001 Cdnenons trolled by other persons. Each licensee curie = 3.7 x 10 ' dps. shall maintain records of such state.

~ g 6 20.101 Radiation dose standards for ta.

I s didduals in restricted arena, mmts untD W Commiasion suhr.

ites thelt disposition.

iI g ,,  ; (a)In accordsace sith the provtstons (b) Before permitting, pursuant to

• of I 20.102(a), and except a.s provided i 20.10lf b), any individuaj in a restrict.

(4 (Deeted 39 Fa 21990-1 in paragraph (b) of this section, no 11 cen.see shnJ1 possess, use, or trsnsfer !.l. ed area to rteelve an occupations) ra.

censed material in such a manner as to dlation dwe in escess of the standards specifW in i M101(a), neh licmee shall:

2M May 31, igg 4

. 20.102(b) PART 20

• STANDARDS FOR PROTECTION AGAINST RADIATION l

/ (1) Obtain a certificate on Forrn

( NRC-4. or oc a clear and legible record containing au the informacon required in that form, signed by the .

todtvidual showing each period of time after the indjvidual attained the age of 18 to whkb the individual received

. an -pauonal dose of radiacon: and .

3) Ca3culate on Form NRC-4 in ac-l co(rdanne with the instructJons appear.

2 Ing therein, or on a clear and legible ('f

)

l g record containing all the informauper required in that form, the previodaly accursulated oocupaUenal dose re- .

ceivrd b.T the indJviduaj r.nd the addj.

ticeal dose aDowed for that individual -

under i 20.101(b).

(cX1) In the preparadon of Form NRC-4. or a clear and legible record conta'nin g all the informauon re-quired in'that form, the licensee shall ,

i make a reasonable effort to obtain re-ports of the tod!vidual's previously ac.

cumulated occupadonal dose, Por each ,

period for which the licensee obtains

• 1 1

i i

I g

l l

i

l l l 1 \

l 20.l o2(c) -

20.103(c) l PART 20 e STANDARDS FOR PROTECTION AGAINST RADIATION ,

1 such reports the licensee shall use the manner as to permit any indJvicuaJ in would result from inhalation of euch done shown in the report in prepa.rtng a restricted area to inhale a quanuty materla] fer 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at the uniform the form. !n any case shere a licensee , of such mater 1&J in excess of the concentrations specified in Appendix is unable to obtain reports of the indj. . Intake 11rnits specified in Appendix B, B. Table 1. Column 1 as la reasonably v) dual's occupational dose for a previ.  ! Table 1. Coluinn 1 of this part. If such .schjevable. Whenever the intake of rs-ous complete cajendar quarter,it shall l soluble uranium is of a form such that a dioacUve material by any indjyldual be assumed that the indJvidual has re. i absorption through the skin is likely, a exceeds this 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> control measure, utved the occupationd dose specified individual exposures to such material a the licensee shall make such evalua-in whichever of the follosing columns $ hall be controlled so that the uptake i tjons and take such actions as are Dec-apply. of such rnaterial by any organ from

• esaary to nasure askinst recurrence.

either inhalation or absorption or The licensee shall maintain records of )

cue s. C"* r-both routes of intake

• does not exceed such occurrences, evaluations, and ac-

'" $!.7, that s hich would result from inhs!!ng tions taken in a clear and readily iden-I ," "  %* such matenal at the limits speelfled in tifiable form suitable for summary review and evajuauon.

"* O"

• Appendix B. Table 1. Column 1 and w e

! *OY yJ footnote 4 thereto.

0) For purposes of deterrnining

, l i. %

• (c) Wa rsspirstory ptotective compliance with the requirements of E *== ee* saan this section the licernee shall use suit- equipment is used to limit the inhalation

• " "No able measurements of concentrations of a% cme redjoectin material Z7"*~ n, a of radioactive materials in air for de. ~ Pumant to paragraph M2)of this

• section. the licensee shall use equipment

- tecting and evaluating airborne radjo

• that is certfed or had cemfication Lettvity in restricted tress and in addj-tion. as appropriate, shal] use meu E entended by the Nauonalinstitute for f, (2) The licensee shall retain and pre

• urements of radiotetJvity in the body, O Occupetional Safety and Health /Mine measurements of radjonctivity excret. Safety and Health Administrauon

- serve records used in preptrms Form ed from the body, or any combination (NIOSH/MSHA). The licensee may

!. NRC-4 until the Commission author

• of such measurements as may be nec- make allowanca for this use of

. tres their disposition.

essary for timely detection and usess- respiratory protective equipment in ,

tment of indjvidual intakes of radioac- nMmating exposms of tridividcals to j

5tivity by exposed individuals. It is as, this material provided that

Humed that an individual inhales ra-Ti calculation of the individuars accu- Lllotetive materiti at the airborne con- ,8' nee the concentrauon specified for tri-mulated occupaticnal dose for all pert- geentration in which he is present Umo r

[ ods prior to January 1,1961 yields a w result higher than the applicable accu.

unless he uses respiratory protecuve equipment pursuant to paragraph (c) u s al intake l

intale permitted La ts1ce that shkh sould I

{ mulated dose value for the indMdual of this sectiort pardeultr When assessment individual's intake of radlo- of a

• result from Inhalauon alone at the concen.

g as of that date, as specified in part. tranon specified for H 3 S in Appendix B.

gTaph (b) of I 20.101, the excess may active materit! ls necessary, intakes ; Table 1. Column I for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per seen for be dtsregarded. less than those t hich would result = 13 *eet2.

~ from inhdation for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in any ont ;; 'For radon.222. the lunluna quanuty is L - -

day or for to neurs in any one week at that u*aied in a period of one ca;endar 8 ft 163 Exposure of ladindaals to con. uniform concentrations specified in Appendix B. Table 1. Column 1 need

!j'"#[*' '*d oscun_ materials dea.gnated

'f etatrauons of red.oactne siatenats in the co centra nN nQh[ t#

air in restricted areas. not be included in such assessment, (ax1) No licensee shall possesa. use, provided thr.t for any assessment in upon exposure to t.he material u an exter.

excess of these amounts the entire naJ rad:auen sourn. Indnuuti exposures to or trtnsfer licensed materialin such a amount is included. Lhese materials may be accounted for as manner u to permit any indn1dualin part of the limitauon on individual ocee in a restricted area to inhale a quantity (bXla The licensee shall, u a precau- l to 101. These nuclides shall be subject to tionary procedure, use process or the precauuonary procedures required by of radioactive matenal in any pened other engineering controls, to the 12010Fbxil of one calendar quarter greater than extent practicable, to llrnit concentra.

the quantity s hich would result from 'Muluply the concentration values specs-Lnhalation ior 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per sett for tions of radioactive enateritis in air to fled in Appendix B. Table 1. Column 1, by levels belom those which delimit an 4.3 x to' m2 to obtain the quarterly quanuty 13 seeks at uniform concentrations of airborne radioacthity trea u defined !tmit. Multiply *.he concentrauen tajue spec. I radioactive material in air specified in ifwd in ADMndix B. Table I, Column to by l Appendix B. Table 1. Column 1. If in(2)i 20.203tdXIXil)'

When it is trn practicable to apply 3 Ifr]t f$$.f22 btain the annuaJ ovanuty tt the radioactive snaterial is of such process or other engineering controls 'Slanifkant intale by ingeauon or injec-form that intake by absorption to limit concentrations of radjotetive uon is presumed to occur only u a result of through the skin is likely, indMduaJ matenal in att belos those defined in circustances such u accident. inadver.

exposures to radletethe material shall i 20.203(d X I Xil), other precautionary tence. Poor prbeedure, or simJlar smetaJ be controlled so that the uptake of ra, procedures, such as increued survell. cond;tjera. Such intales must be essluated lance, llrnitation of working times, or and accounted for by techniques and proce.

g diotetjve rotterial by any organ from g either inhalation or absorptjon or both routes of lntake "in any calen.

provision of respiratory protective equipment, shall be used to maintain k,'e',(('#[ d " 'kPN $ M g esah.ated shall be included in determining m dar quarter does not exceed that intake of radioactive material by any s hether the limitation on indjvfdual expo.

5 whJch sould result from Inhaling such individual sithin any period of seven sures in i 20.103<ax1) hu been eacceded.

I radioactive material for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per consecutive days u f ar below that 'Rerulatory guidance on sueument of in.

seek for 13 weeks at uniform conetn. intake of radioactive materiti s hich dividuaJ intanes of radioactive material ts tratforu specified in Appendix then in Rervlatory Oulde 8 9. "Acceptable B- Concerta. Modets. Equations and Assump-Table I, Colurnn 1.

uons for a sicusay Prorism " single ceples (2) No licensee shall possess, use, or of thxh are asallable from the Offlee of (

transfer mixtures of U-234,0-235. and standtids Development. U.S. Nuclear Reru.

U-238 in soluble forts is such a latory Commtasiott, Wuhtngton. D.C. 20545, upon entten request.

September 1,1982 W

_ ~ - -

i 20.106<b) 20.202(b) 1 PART 20 e STANDARDS FOR PROTECTION AGAINST RADIATION e nt y d propose centrations of radonuclides. Parcactiomy P*oetncus hmits higher than those spedfled in (7) A descripuon of the waste treat-ment iteultles and procedures used to I N 291 Serv eys. .

paragraph (a) of this section The reduce the concentration of radionu. (a) As used in the regulations in this )

Commission will approve the peoposed clides in efPuents prior to their re* aart. sunev" means an evaluation of limjts if the applicant dsmonstrates: le ue.

(1) That the applicant has naade n (d) For the purposes of this section $"the radistu use production, hazards releueincident to the disposal, or j

reasonable effort to minimise the ra. the concentration limits in Appendix  ;

dioactitity contained in effluents to B,

• Table II of this put shall apply at gpmence of rad' oacun' materials or other sources of radjstion under a spe-i 1

unrestricted treu; and the boundary of the restricted area. e acille set of conditions. When approprl.

(2) nat it is not likely that radlote. The concentration of radjonctive ma-att such evaluation includes a physi-tive material discharged in the efDu. terial discharged through a stack, pipe ca] survey of the location of materials ent would result in the exposure of an or similar condult may be determined and equipment, and meuurements of individua.1 to concentratjens of radio- with respect to the point where the materialleaves the conduit. If the con, levels of radlauon or concentrations of 1 active material in air or a ater exceed.

ing the limits specified in Appendix B, duit discharges within the restricted _radioacth e materkt oresent.

Table H of this part. A trea, the concentration at the bound-  !

(c) As appucation for higher limits 3 try may be determined by applying pursuant to paragraph (b) of this sec ," appropriate factors for duution, dis- b Eac De m ofM;f e(us)ew to be made suc tion tha11 include information demon. s persion, or decay between ca the point strating that the appucant has made a 2 dischtige and the boundary.

reasonable effort to minim!24 the ra. (e) In addjtlon to limiting concentra- ::may be necessary for the licensee to dioactitity discharged tn effluents to tions in effluent stretms, the Comruis- ; comply alth the regulations in this unrestricted treu, and sh:33 include, sion may Itmit quantities of radioac- Spart, and (2) are reasontbie under the as pertinent: tjve materla]s releued in a.it or mater circumstances to evaluate the extent of radiation hazards that may be pres.

(1) Information u to flow rates. appears that the daily intake of radio-during a specified period of time tf it ent.

total volume of effluent, peat eencen. " ' 7 metive materis] from tir, water, or food tratJon of each radionucUde in the ef.

Duent, and concentrauen of each radi. by a suitable sample of an exposedf h.M2 Pannel m%ing.

onucilde in the effluent tveraged over popuistfon group. tveraged over a (t) Each Ucensee shall supply appro-a period of one year at the point period not exceeding one year, would there the effluent leaves a stack, otherwise exceed the dtUy intake re. priate personnel monitoring equip- l tube, pipe, or similar condult; sulting from continuous exposure to ment to, and shall require the use of j (2) A description of the properties of air or water containing one third the such equipment by- i

% the effluents, including. concentration of radioactive materials (1) Each individual who enters a re-3 (1) Chesnical compositjen; spectiled in Appendix B Table II of I stricted Area under such circumstances (ll) Physleal characteristics, includ. - this part. . that he receives, or is likely to receive, i E ins suspended solids content h Uguld a done in any cajendar quarter in l

• effluents, and nature of gas or aerosol , excess of 25 percent of the appucable

~

(f) The provisions of paragraphs (a) J value specified in paragraph (t) of for air efDuents;

( 111 ) The hydrogen lon coceentra. through (e) of this secuon do not i 20.101.

tions (p") of liquid ef fluents; and apply to disposal of radlosetive materi. (2) Each individut] under 18 years of (iv) The size range of particulates in al into sanitary ses erage systems. age who enters a restricted tres under efDuents released ir.to air,  ; which is governed by i 20 303 . .

such circumstanees that he receives, (3) A descr.ption of the anticipated - (g) In addition to other require. or is likely to receive, a dose in any crJ.

human occuptney in the unrestricted [ ments of this part, licensees engaged endar quarter in exeen of 5 percent of area where the highest concentration . In uranium fuel cycle operatJons sub. Ithe applicable value specified in para.

of rud.ioactive material from the efflu.

• ject to the provisions of 40 CFR Part ~ I sraph (a) of I 20.101.

ent is expected, and, in the case of a 190. "En ironmental Radiation Protec- e (3) Each indJvidual who enters a river or stream, a descripuen of water tion Standard for Nuclear Power Op-

• high rsdiatJon area.

uses downstream frorn the point of re. frations," shall comply with that part. 2 (b) As used in this part, lease of the efDuent. (1) "Personnel monitoring equip-

=

(4) Information u to the highest 1 20.107 Medical d.iagnosis and therspy.

rnent" means devices designed to be concentration of each radionuttde in g , g an unrestricted aret including antici- Nothing in the regulations in this the purpose of men.suring the dose re-pated concentrations averaged over a part shall be interpret 4d u limiting celved (e.g., flim badges, pocket cham.

period of one year: the intentions) exposure of patients to bers, pocket dosimeters, flim rings, (1) In air at any point of human oc. radiktion for the purpose of medica] e te,y, cupancy;or diagnosis or medical therapy. (2) "Radiation area" means ar. "

(11) In water at points of use dost. area, accessible to personnel, in whlch

. stream from the point of rthase of $ 20.lt3 Orders requiring furnisMng of exts s radiation, originating in the efDnen.. bio-atta) sen kes. g ,g gggg (5) The background concentration of necessary or desirable in rial, at such levels that a major por.

j or%'here

, der to aid in determin!ng the extent tion of the body could receive in any radjonuclides in the recehing rhet or stream prior tc the release of 11ged ef. E of an individuals exposure to concen. one hour a dose in excess of 5 roll.

Duent trations of radioactive materitJ. the g Commission 11 rem, or in a.ny 5 consecuthe days a (6) A description of the emironmen. may incorporate appro- dose in excess of 100 milltrems; tal monitoring equipraent, irduding priate provisions in any !! cense, direct. (3)"HJgh radiation aret" means any sensitJvity of the system, and proce, ing the licensee to tr.tke asallable to tres, artessible to personnel,in which the individual approp' late bio assay there exists radiation originating in dures and calculations to deurmine concentrations of radlonuclides k the ser ices and to furnish a copy of the whole or in part within licensed mate. '

unrestricted tres and possible recon. reports of such senices to the Corn. rial at such levels that a major portion mission- of the body could receive in any one hout a dose in excess of 100 mil]! rem.

~ -

Septembw 1,1982 20-6

l l

'b'401(a)

• 20.401(c)

PART 20

• STANDARDS FOR PROTECTION AGAINST RADIATION ,

1 l

I tt.sel Reewds of smeys, redlauee amen!tering, and disposal ( )

1 (a) Each ljeensee shall maintain re. I cords showing the radiation exposIru 1 of all indhiduals for thom personnel '

monitoring is required under i 20.202 l of the regulatjons in this part. Such

{ '

, records shall be kept on Fbrm NRC-8,

= in accordance uith the instructions A contained in that form or on clear and l

, legible records containing all the in. '

formatJon required by Form NRC-5 The doses entered on the forms or re-cords shall be for periods of time not exceedine one calendar quarter.

(b) Each licensee sha!! malatein

. records in the same units need la thle

part, showing the reeslts of seveys

required by l 30J01(b). maahering g required by 1130J06(b) and 3DJos(c).

~ and esposals made ander ll stLaat, l

' atkaan,termond l 30J04,'and Part #1 of '

this chaptar,

(

l (cXII Records of l'ndhidual exposure to rahtico and to radioactive materi.

al shich must be ma!ntained pursuant to the provisions of paragraph (a) of this section and records of t'ossaars,  ;

including results of whole body count-ing exarninstions, made pursuant to i i 20.108, shall be preserved until the l g

, Commission authortses dispositjon. j (2) Records of the results of sitrveys '

f and monitoring which must be main-

. tained pursuant to paragraph (b) of

' this section shall be preserved for two years after completion of the survey except that the following records shall be maintained until the Comtnission authortzes their dispositjore (1) Re.

cords of the results of surveys to de-termine complJance tith l 20.103(ar, (11) in the absence of personnel monl-toring data, records of the results of surveys to determine externaj rsdl.

ation dose; and (!!!) records of the re-sults of surveys used to evaluate the release of radionethe effluents to the environment.

(

l December 30,1962 twi u,. a te a ll l

34.32 APP.A(V) l 34.32 Opersting and emergency pro- and *orn by only one individual. I 34 44 % = of reeogreeere'

(

• • • • "8"**-

ced ures. (b) Pocket dosimeters shall be read and exposures .ecorded daily. Whenever a radiographer's sulstant The hcensee's operating and emer- uses red >ographic exposure dev6ces, (c) Pocket desimetera shall be gency procedures shall include instruc* steched at penode not to eaceed one uses sealed sources of related source tions in s't least the followmg i year for correct response to radiation. handhas tools, or conducts radiauon (a) The handhng and use of Icensed .Aseptable dos 6, *rs shall read within , surveys required by I 34 43(b) to sesled sources and radiogrsphg eaposure :phis or minus 30 percent of the true  : determine that the sealed source has devices to be employed such List no *8'diation exposure. 3 returned to the shielded position after e (d)If an individual's Pocket dosimeter a an exposure. he shall be ander the person is hiely to be enposed to radsatiop

• in discharged beyond its ra nge. his film

. ' penonal supervision of a radiographer, doses in encess of the hmits estabhshed in : bdge or T1D shau be immediately sent he personal supervision shall include Part 20 of this chapter "St ndards for (sj the radiographer'o personal presence

{h processing Protection Against Radtation ; et the site where the4ealed sources are (b) Methods end occasions for con- (e) Reports receive'd from the flht being used. (b) the abdity of the

ducting ndistion surveys- undge or UD proceesor abaD be kept for rad ographer to give immediate

~ (c) Methods foe controlhr: scress to 4spection until the Commisalon a nistance if required. and (c) the a radiographic areas, e authorisse their disposal. Records of radiegrapher's watching the assistant's

' (d) Methods and occasions for locking

• da0y pocket dosimeter tesdings shad be F*rformance of the operstionii referred and secutir:g rsdiographg eaposure de.Japt for two part, j b M ***

vices, storage cont ainers and ne ste d EXEM PTION6 sources; itECAUTION ARY PROCEDURES IN R ADIOG R APHIC OPE R ATIONS $ 34.51, Applications for enernptions.

(e) Personnel monitones and the use of personnel monitonng equipment. $ 14.41 Security. O The Commission may, upon applica-(f) Transporting sealed sources t k mh ndm@c opernion 2 tion by any licenses or upon its own field locations. iricluding packing of rad" initiative, grant such esemptions from the ographic esposure derxes and saorage h ndWhr M ndiwhu's e K requirerne.nts of the regulations an this containers in the vehicles, posting of W shad mh e Met survehce of

.g ' pa1 as it deterrnines are authorized by vehicles and control of the sealed sources law arid wal not result in undue hazard to during transportation; ued em Wo a @ ndiHion m life of property.

em defined in Part 20 of this chapter, _

I (3) Minimaing eaposure of perions in .mcrpt (a) where the high radiation area is bthe event of an accident; spipped with a control device or an ,

1. W "^ [

• I rundame.cols of Aaserson sofety (h) The proce dure for nouf yin g at a t m system ae described in \

• proper persons in the event of as acce 520.203(c)(2) of this chapter, or (b) A Charactenstica of samune redataan.

Ident;and where the high radiation eres is locked Io a Units ed rsdistaan dose lanm) and 1 (i) Maintenance of records pote ct against u nauthoriz ed or a cci- Th n

"[  %

I b (j) The inspection and snainteaance ef **jb of ndiate W bcensed i 34.4 2 Posting.

f ndiographic enposure devges and morage L Wethods of centronmg redation do.e, .

m containert Notv ithstanding any provisions in 1. Worbag tune-l

120.204(c) of this chapter, areas in 1 * ** '8 ^(**

$ shct rsdiography is being performed o i

(k) Steps that must be taken est! be c,nspicuously posted as required 3 II A***'" M***a la8ha*88868a T*

l

immediately by radiography personnel s in the event a pocket doeuneter is found ;is$20.203(b) and (cXI) of this chapter.

• A Use of redation survey instrwmenta

• • to be off.ecale. SK43 Redecen surveys . 1. Curstmoew h is) Atleast one calibrated and t Cawsixart i inatrument 1 I"* tmana-  !

I (1) The procedure (s) for identJying agerable red 2stion and teportir;g defeets and duc be available at the survefocation of E 5"'"7 h l***

E neograPhic oP*rationa whenever C U*' P"'**^'I "**"**4 'M** *'

noncomplia nce, a s required by Part !! of 1. 7%m beden and thermolummeceos

. this chapter. . Edognphac opersuons an betag deemeten innd 7'I*""d 1 Poc.he dosi.meten.

__' $) A survey with a radiation survey a hulrument shall be made after each //l. Aod>cgraphac Equipment To Se Used

)3433 Personnet monsorW9

• neog+sphic exposute to deIermine thet (a) The licensee sheD nos permit any A Remese handhng epipment it.dividual to act ee a red rspher or e  ? le sealed source hae been teturned to E Radu:paphse upoem devicet

. radiographer's assistant u1n ess, at aD hahielded position.ne entire C Storese containers.

• tirnes dunng radiographic operations. siscumference of the radiographic IV /M pd MeWew pufM each such individual wears a direct egeture device shsU be sumyed. If the by rAe AodermpAm 8 rta ding pocket desirneter and either a udegraphic esposure device has a

', Glm badge or a thermoluramescent suecs gWde tube, the survey shau V Ccse Nissonas of Ao6cympAy Accidears

• dosimeter (TID). Pocket doometers huisde the gWde tube.

I shall have a range froen sero to atleast M A reccrd of the survey required in

} 2C0 milliroentgens and shaR be pra6raph (b) shaU be maintained for recharged at the stari of each shJt.Each to years when the survey la the last (Note removed 49 FR 19623)

Glm bedge end T1D shD be sesigned to ervey pnoe to loding the radiographic supeure device and ending direct arveillance of the opervion. (

Mey 31,1964 344

-e

~ -

= w,.

NUREG/CR PNL-5809 Vol. 2 Review of Light Water Reactor Regulatory Requirements i

Assessment of Selected Regulatory Reauirements That May Have Marginal Importance To Risk

- Reactor Containment L'eakage Rates

- Main Steam Isolation Valve Leakage Control Systems

- Fuel Design Safety Reviews

' Prepared by M. F. Mutten, W. J. Bailey, C. E. Beyer, G. J. Konzek, P. J. Pelto, W. B. Scott Pacific Northwest Laboratory Operated by Battede MemorialInstitute Prepared for U.S. Nuclear Regulatory Commission

-- I, b '

SUMMARY

BACKGROUND AND OBJECTIVE The U.S. Nuclear Regulatory Comission (NRC) has initiated a program to  ;

review current light water reactor (LWR) regulatory requirements to see if some could be relaxed or eliminated to reduce regulatory burdens without compromising public health and safety (Federal Register, October 3, 1984).

Pacific Northwest Laboratory (PNL), which is operated for the Department of Energy (DOE) by Battelle Memorial Institute, is conducting a series of studies in support of this NRC program. This report covers a portion of PNL's work.

The purpose of the report is to present information on the risks benefits of streamlining regulatory requirements in three areas:, costs and reactor containment leakage rates e

main steam isolation valve (MSIV) leakage control systems (LCS)

NRC licensing review of fuel design information.

These three areas of regulation were se'lected by NRC staff for analysis in the initial (pilot) phase of the regulatory review program.

CONCLUSIONS Analyses were performed to assess the effects of streamlining regula-tory requirements in the three selected areas. The basic framework for the analyses was that presented in the Regulatory Analysis Guidelines, NUREG/-

BR-0058 (NRC 1984a) and in the Handbook for Value-Igact Assessment NUREG/CR-3568 (Heaberlin et al.1983). Probabilistic risk assessment, supplemented by other considerations where appropriate, was used to evaluate the risk

~ significance of streamlining the requirements. Various measures of risk l-- were examined including population dose, expected early fatalities and injuries, and individual dose. Sensitivity studies were also performed to explore the effects of such fet. tors as accident source terus.

The results of the analyses are summarized in Table S.1. Several coments and observations concerning the table are provided here.

The three areas of regulation cover a range of different types of regula-tory requirements, and the analyses considered a range of different degrees of regulatory relaxation. In the case of the containment leakage rate limit the regulatory modification considered was to relax the numerical leakage rate limit. In the case of main steam isolation i

valve leakage control systems, the regulatory modification considered

{

was complete elimination of the requirement, and disabling of the systems I currently in place. In the case of fuel system safety reviews, the  !

modification considered was the selective elimination of items in the  !

current reytew procedure that may have marginal risk significance.

i y  !

' 'W9.

TABLE S.I. Sumary of Risk Impacts, Benefits, and Benefit-Risk Comparisons -- Total for All Affected Reactors Area of ReQulation Main Steam Isolation Valve Reactor Containment Leakage Control Fuel System LeakageRate(a) System (b) Safety Reviews (c)

Effect on public risk Marginal Marginal Marginal if requirements were (On the order of a (Less than one few percent, or (Notquantified) streamlined (d) percent of less of overall overall risk) risk .

Benefits of Greater than $10 7 6 Greater than $10 Marginal streamlining requirements

(,) (Notquantified)

Benefit-Risk NA II)

In the range of Greater than $10 4

comparison, if 3 4 per person-rem requirements were $10 -$10 streamlined p" p"son-rem saved per per(dollars son-rem j ofrisk)  !

-(a) Increase allowable leakage rate to 10% per day. Currently, typical allowable leakage rates are 0.1% for PWRs and 1% for BWRs.

(ti). Eliminate the requirement for MSIV leakage control systems and disable I

.~

the systems in plants that currently have (or will have) them.

(c) Eliminate selected items from the current procedures for fuel system safety reviews, which are set forth in Section 4.2 of the Standard Review Plan, NUREG-0800 (NRC 1981a).

(d) Various measures of public risk were considered, including population and individual dose, early fatalities and injuries, and latent cancers.

For each of these measures the effect of streamlining the requirements was marginal.

(e) Costs and cost savings in this table are sumed over the remaining lifetimes of all affected plants and discounted to present value at a 10% real discount rate as suggested b Guidelines, NUREG/BR-0058 (NRC 1984a)y the Regulatory Analysis (f) Not applicable. It is assumed that the benefit-risk conparison is not of interest when the benefits are marginal.

1 vi

)

l In all three cases, judiciously streamlining the existing regulatory requirements is estimated to have marginal effect on public health and safety. Marginal, in this context, means that the effect is relatively small, on the order of a few percent, or less, of overall plant risk.

The benefits of streamlining the existing regulatory requirements vary.

In the case of reactor containment leakage rate limits, the estimated benefits are on the order of several tens of millions of dollars. In the case of fuel system safety reviews, the benefits are insignificant (no dollar estimate was computed). The case of the MSIV leakage control system occupies a middle ground, with benefits of several million  !

dollars.

Comparisons of benefits and risks also vary. In the case of the contain-ment (dollarsleakage rate limit}

per person-rem is the ratto oftodollars estimated be slightly savedhigher to risks incurred than the benchmark of $1000 per person-rem that has been used in some other contexts (e.g., the proposed safety goals, and 10 CFR 50, Appendix I).

In the case of the MSIV leakage control system, the ratio is estimated to be considerably higher than the $1000 per person-rem benchmark.

For the case of the fuel system safety reviews, no ratio was estimated; it was assumed that cogarison of benefits to risks is not of interest in this context because the benefits are insignificant.

The quantitative analyses on which these conclusions are based are highly uncertain, and should be interpreted cautiously. For this reason, i the results in the table are reported in tems of ranges and orders of magnitude. All of the usual caveats and uncertainties surrounding the use of quantitative risk-cost-benefit analysis apply. Specific areas of uncertainty and possible areas of conservatism in the analyses are discussed in the main report.

F

• It should be strcased that analyses of this kind, and especially the quantitative portions of principal, basis for regu.such latoryanalyses, decisions.are Rather,not thethe sole, or even the number of inputs. As noted by Heaberlin et al. (1983) yinare the one of a Handbook for Value-lapact Assessment, the real strengths of quantitative analysis are the discipline that it provides and its display of key infonnation and assumptions in understandable fom so that they can be scrutinized '

and, if appropriate, challenged by interested parties.

STRUCTURE OF THIS DOCUMENT l

The main report which follows this sumary, consists of four sections.

Section 1 covers the, background, objectives, and scope of this study. Sec-  !

tion 2 presents the analysis of containment leakage rate testing. Section 3 covers the MSIV lukage control system. Section 4 covers fuel system safety reviews. Two appendices contain supporting infonnation on the containment leakage rate analyeis.

vit

OVERVIEWS OF THE THREE ANALYSES that were perfomed in each of the three areas of regula the NRC staff for examination in this study.

Reactor Containment Leakage Testina Reactor containments constitute one of the principal lines of defense of light water power reactors.in the defense-in-depth design philosophy embod the consequences of accidents, containments are subject to a variet One element of the containment reguletory requirements th considerable 10 CFR 50, Appendixattention is the requirement for leakage testing set forth in J, "Primar Water-Cooled Power Reactors." y Reactor Containment Leakage Testing for ments for preoperational and periodic verification by tests of the leak-A i tight integrity of the primary reactor containment. ,

to assure that leakage through the containment will not exceed allowableThese ti leakage spect rate values, which are defined in each plant's technical fications. .

basis to meet the dose limits in 10 CFR 100, assuming accident.  !

(

100 limits is written into the plant's technical specifications.In Typical

! pr allowable leakage rates are 0.1% per day for a PWR and 1% per day fori a BWR Probabilistic risk assessments, beginning with the Reactor Safety Study WASH-1400 (NRC 1975), have consistently shown that containment leakage is relattvely minor contributor to overall plant risk.

ment-related contributions to risk stem from accidents in which theTh i

~

containment ruptures (due to steam explosions, overpressure, hydrogen i

-containment).

(e.g., an interfacing systems LOCA with resulting direct r significant role.In these dominant scenarios, containment leakage plays no the cost impact of containment leakage rate testing The is substa primary reason for this is that integrated leak rate tests (ILRTs) of th*

entire reactorcontainment outage critical(called path. Type A tests in Appendix J) are generally on the days of incremental plant downtime at an estimated cost of $1.3 to $2 million.

latory requirements without compromising public health and savings might be substantial.

Objective of the Con _tainment Leakage Rate Analysis Consistent with the overall objectives of the NRC program to review -

the effectiveness of current LWR regulatory requirements in limiting risk, v111

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

the purpose of this analysis is to provide information on the risks, costs, and benefits that would result if the current requirements for testing con-tainment leakage rates were mndified to reduce regulatory burdens without compromising public health and safety. For purposes of this analysis, the option under consideration is the following:

Increase the allowable leakage rate for both PWRs and BWRs to 10% per day. Sensitivity studies to show the effect of varying this numerical value are included in the analysis. The test frequency is not changed.

The underlying hypothesis is that in:reasing the allowable leakage rates might reduce regulatory burdens, perhaps substantially, without causing any significant adverse impact on safety.

Alternatives to Modifyino Containment Leakage Rate Requirements current regulatory requirements pertaining to containments are complex.

Furthermore, a host of technical issues involving containments and their role in reactor safety have been identified and are currently being studied in research programs worldwide. Numerous alternatives for modifying contain-ment requirements exist and are being considered. For exagle, a major revision and updating of 10 CFR So, Appendix J is pending at the NRC. A comprehensive reassessment of containment requirements, based en recent research on severe accidents, is planned. Other examples can be cited.

Although many of these other alternatives may lead to reduced regula- -

tory Durdens without adversely affecting safety, it is not possible within '

the scope of the present study to consider all of them. A much more extensive effort, with a different emphasis, would be needed to fully explore all of the options for modifying regulatory requirements for containment.

Consequences of Increasino the Allowable Containment leakage Rate A series of risk sensitivity calculations was performed under a range )

of conditions and assumptions to explore the effects on public health and safety of increasirg the allowable leakage rate. Consideration was given to the following factors, among others:

• population dose in person-rem e early fatalities and early injuries e

individual dose impacts (both whole body and thyroid doses) as a function of distance from the plant i

" 1

+ the effects of alternative source term assumptions.  !

Full details are contained in the body of the report. Only the most salient points will be highlighted in this sumary section. l l

4 IX

$N' ' ' ' . E i E .__ i ' ._ _ _5_ 5 l. _ [

5 _ . _ _ _ _ _ _ _ _ _ , _ , , _

i l

Table S.2 shows the estimated sensitivity of risk (population dose in person-rem per plant year) to leakage rate for four different cases. In  ;

each case, the risk is not very sensitive to changes in leakage rates; increasing the leakage rate to 10% would increase the calculated risk by a few person-rem per plant year.

information and assumptions: These calculations were based on the following ,

Accident frequencies were obtained from the Reactor Safety Study (Surry 1 and Peach Bottom 2, NRC 1975) and two probabilistic risk assessments  ;

(Oconee 3 and Grand Gulf 1, NRC 1981b) performed as part of the Reactor Safety Study Methodology Applications Program (RSSMAP).

Risk sensitivity values were obtained from a study by Oak Ridge National Laboratory (Hermann and Burns 1984). Their analysis of containment leakage rate sensitivity used a set of generic source terms and fre- i quencies of occurrence developed as representative of the range of LVR accidents.

)

Population doses were calculated by the CRAC2 (Ritchie et al.1983 and 1984) program using a set of stande.~1 assumptions, including a uniform )

1 population density of 340 persons per square mile within 50 miles of the plant, which represents an average population density for all US  !

plants. Site-specific consequence analyses were not performed. The  :

l standard assumptions were those of the Handbook for Value-Impact Assessment (Heaberlin et al.1933).

TABLE S.2. Sensitivity of Risk to Containment Leakage Rate

, for Four Cases c Expected Population Dose, PWR person-rem / reactor-year Expected Population Dose, Leak Rate BWR person rem / reactor-year

%/ day Leak Rate ~

Surry 1 Oconee 3 _ %/ day Peach Bottom 2 Grand Gulf 1 1.0 71 207 0.5 151 250 10.0 72 210 5.0 153 254 100.0 82 238 50.0 174 288 i

An assessment was made of the cost impacts to the NRC and industry of increasing the allowable leakage rate. Impacts on occupational exposure were i

also considered. A sumary of the estimated cost impacts is presented in X

i Table 5.3. The over lion to $74 million. g impact is an estimated net cost savings of $40 mil-The largest component of this is a reduction in plant downtime. With an increase in the allowable leakage rate, plants )

1 would be less likely to fail their Type A Integrated Leak Rate Tests, (current i failure rates are in the neighborhood of forty to fifty percent), and the l

additional downtime due to test failures would be avoided, with resulting cost savings. It should be noted that this cost savings is subject to con-siderable uncertainty. The estimate is heay11y dependent on the assumed Type A test failure rate after the allowable leakage rate is increased.

This failure rate cannot be predicted precisely. Hence, the uncertainty range on the estimate is large. Other key assumptions are discussed in detail in the main report.

Calculations were also done to explore the sensitivity of individual dose, early fatalities, and early injuries to increases in leakage rate.

Again, the effect was found to be small.

_ TABLE S.3. Sumary of Cost Impacts of Increasing the Allowable Containment Leakage Rate --

Total for All Plants 1 CostCatetory(,) Qualitative EstimatedCostImpact,(b) s Effect Thousands of Dollars Industry Implementation Costs Cost Increase 800 Operation Costs Cost Savings 42,000 to 76,000 NRC Iglementation Costs Cost Increase 1,000  ;

Operation Costs Cost Savings 7 to 13  !

, ;-- - Total Cost Savings l

40,000 to 74,000 l

(c) Implementation costs are the one-time initial costs of iglementing the change. Operation costs are the recurring costs (or cost savings)  !

over the remaining life of the plants.

(b) Costs shown are not discounted. Discounting at a 10% real discount rate would reduce the total net cost savings by approximately a factor of 3. Discounting at 5% would reduce them by approximately a factor of 2.

(a) It should be noted that these estimates are not discounted. Discounting at a 10% real discount rate would reduce them by approximately a factor of 3. Discounting at 5% would reduce them by approximately a factor of 2.

xi h

1 Variations in assumed source terms and ott parameters were also con-sidered. Although the details of the calculations are, of course, affected by these variations, the basic conclusion is not altered.

Conclusions -- Containment Leakage Rate Requirements I

If the effects of increasing the allowable leakage rate are expressed on a dollars per person-ren basis, the ratio is on the order of several thousand dollars saved per person-rem of public exposure. This is deter-  :

mined as follows. A cost savings of $40 million to $74 million, discounted at a 10% real rate as suggested by the Regulatory Analysis Guidelines (MUREG/BR-0058) yields a present value of roughly $13 million to $24 mil- I lion. An increase of a few person-rem per plant year (range: 1 to 5) times 120 plants (operating and planned) times 30 years (nominal average remaining plant lifetime) yields 3,600 to 18,000 person-rem. The resulting ratio could range from about $700 per person-rem (i.e., $13 million/18,000 person-rem) to about $7,000 per person-rem (i.e., $25 million/3,600 person-rem).

These ratios can be compared to the benefit-cost guideline of $1000 per person-rem that has been used in certain other contexts (i.e., the pro-posed safety goals and 10 CFR 50, Appendix I). However, it should be stressed that o.uantitative calculations of this nature, even if they are assumed applicable in this instance, are never the sole or even the principal basis for regulatory decisions. Other regulatory considerations, such as defense-in-depth, must be factored into the process. Moreover, the numerical values i are highly uncertain and should be interpreted cautiously.

Main Steam isolation Valve (MSIV) Leakage Control System Most of the boiling water reactors (BWRs) that are currently operating I

and soon to be operating have been required to install leakage control systems  !

~

(LCS) to control leakage past the main steam isolation valves (MSIVs) in '

the event of an accident. The purpose of the LCS is to collect and process (filter) any leakage of fission products past the MSIVs and thereby ensure i that the radiological effects of certain postulated accidents do not exceed l the numerical limits set forth in 10 CFR 100, "Reactor Site Criteria." The NRC staff's regulatory position on the MSIV leakage control system is spelled l out in some detail in NRC Regulatory Guide 1.96, "Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants.' The rationale supporting the current requirements for leakage control systems is e antially deterministic. The systems are designed to ensure that the offsite dose limits in 10 CFR 100 are not exceeded under the following conditions: '

e design basis loss-of-coolant accident (LOCA) e missiles, dynaste effects (e.g., pipe whip) and environmental condi-tions (pressure, tesperature, steam) resulting from a design basis LOCA e an assumed single active failure concurrent with the LOCA (e.g., failure of one of the MSIVs to close) xii

1 l

l 1

i e

an assumed single failure in the leakage control system itself a

loss of all offsite power coincident with the LOCA  !

occurrence of a safe shutdown earthquake coincident with the LOCA (i.e., l the leakage control system is designed to seismic category I) l fission product source ters as defined in TID-14844, "Calculations and Distance Factors for Power and Test Reactor Sites," (DiNunno et al. l 1962.)

Substantial elements of conservatism inherent in this deterministic approach I have been recognized for a long time. In recent years, improvements in the data and methods available for measuring risks have provided additional i insights into the benefits of MSIV leakage control systems, both in an l absolute sense and relative to other systems designed to protect the health and safety of the public. Estimates of the benefits of MSIV leakage control systems, based on probabilistic risk assessment techniques, indicate that the benefits are marginal at best, and that implementation of such systems is difficult to justify from the standpoint of cost-effectiveness. The basis for these conclusions 1. documented in the body of the report. The next sections of this sumary highlight the more salient points.

Objectives of the MSIV Leakaac Control Systems Analysis Consistent with the overall objectives of the NRC program to review the ef fectiveness of current LWR regulatory requirements in limiting risk, the purpose of. this analysis is to provide infonnation on the risks, costs, and benefits that would result from elimination or modification of current requirements for MSIV leakage control systems. For purposes of this analysis, the option under consideration is to:

c. . I

~ eliminate the requirement for MSIV leakage control systems (i.e., j eliminate NRC Regulatory Guide 1.96, Standard Review Plan Section 6.7  ;

(NRC 19814), and make conforming changes in other regulator

, such as technical specifications and 10 CFR 50, Appendix J)y documents e

disable the leakage control systems in plants that currently have them (or will have them), l i

i Alternatives to Modifying MSIV_LeGace Control System Requirements There are a number of complex technical issues surrounding the require-ments for MSIV leakage control systems. Each of the technical issues, in turn, gives rise to a number of regulatory alternatives. A comprehensive examination of the full range of issues has been conducted as part of a large-scale, multi-year effort to resolve Generic Safety Issue C-8, 'MSIV Leakage and LCS Failure." A re NUREG-1169 (portand Ridgely on the Wohl resolution of Generic 1986). Among Issue C-8 the issues )

has beeninprepared,G-1169 considered NURE are  ;

l l

l xili

l l

l l

. methods for reducing MSIV leakage (thus reducing the need for leakage  ;

controlsystems) j l

. allowable MSIV leakage rates ,

j i

. alternative methods / pathways for attigating the consequences of MSIV leakage

. analytict) methods for more accurately calculating the consequences of 1 MSIV leakage (taking account of such factors as fission product depost- i tionanddecay).

The scope of the present report is more liutted, since the intent is to  !

provide infomation on the effects of eliminating or relaxing current l requirements.

Constauences of Eliminatina Requirements for MSIV Leakage Control Systems j Risk sensitivity calculations were perfonned to estimate the change in I risk that could result if MSIV leakage control systems were eliminated.

With the Grand Gulf 1 BWR as the reftrence case (Hatch et al.1981) event ,

trees were constructed to model fission product leakage scenarios following core-melt accidents. Coiiservative (i.e., optimistic) assumptions were made for the effectiveness of the leakage control system vis-a-vis the alternative (no system). Even with these optimistic assumptions, the risk reduction attributable to the MSIV leakage control system was estimated as 0.3 person- I res per reactor year.

Two qualitative insights should be noted in order to place this estimate

. in perspective. First, even in the absence of a leakage control systes, MSIV leakage is a small contributor to overall plant risk (on the order of

~~

2 person-rem per reactor year in the calculations presented in this report);

_J so there is only limited risk reduction to be achieved even by a highly effective leakage control system. Second, the MSIV leakage control system is effective only to a limited degree. It eliminates about 15% of the risk contribution due to MSIV leakage (0.3 person-rem /2 person-rea). The reason for this limited effectiveness is that the LCS is effective only when the ,

leakage is less than about 100 standard cubic feet per hour (SCFH); the l system is not effective when large leakages (on the order of 1000 SCFH) l occur. These large leakage scenarios, although they have low probability, have relatively large consequences, and are the dominant contributor to the risk due to MSIY leakage. )

1 These findir,gs concerning the effectiveness of the LCS are consistent I l

with the results of Generic Issue C-8of(NUREG-1169 the recently). completed study Among other by Ridgely conclusions, theand key Wohl findings (1986) of the C-8 report are:

. At most plants there are alternative MSIV leakage paths that do not depend on the availability of offsite power and that are at least as effective as the LCS systems presently required.

l xiv

l l

l Alternative pathways for MSIV leakage control that take advantage of the condenser holdup volume are extremely effective in mitigating the of fsite radiological consequences of an MSIV failure to close; this is true even if offsite power is lost, i

e In the attegt to meet the current strict MSIV leakage requirements, l utilities have sometimes performed excessive maintenance on valves.  !

In some cases, this maintenance has damaged the valves (e.g., seat 1 refurbishment in situ has resulted in out-of-round seats) without pro-  !'

viding any substantial safety benefit.

• From the PRA analyses examined the requirement for a safety-grade LCS could not be defended on a valu,e-igact basis using a value of

$1000/ person-res saved.

On the cost side of the ledger, estimates were obtained for the industry and NRC cost impacts that could result if MSIV leakage control systems were eliminated. Because the risk reduction due to the LCS was found to be small, it was not necessary to quantify the costs with great precision. For in-dustry, the cost to procure and install an LCS was estimated as $500,000 per plant (initial cost). Operating costs for maintenance and surveillance were estimated at $20,000 per reactor year. Therefore, if the requirement for the LCS were eliminated, an operating cost savings of $20,000 per reactor year could be achieved. The $500,000 initial system cost, on the other hand, is a sunk cost and would not be affected in any way unless some BWRs still under construction have not yet acquired the systems.

Iglementing the change in regulatory requirements would entail some additional costs both for industry and for the NRC. Physically disabling the LCS could cost several thousand dollars (nominal estimates one man-

. week of effort or $2,000). Changing plant technical specifications and other documentation is estimated to cost industry about $10,000 per plant.

This would be a one-time cost. NRC costs for the technical specification changes would be about the same (noatnal estimate: $11,000perplant). .

The cost for the NRC to develop and iglement the revised staff position (i.e., eliminate Regniatory Guide 1.96, and conforming chan documents, such as the Standard Review Plan and Appendix J)ges in other was estimated at up to $500,000, although with the imminent completion of staff work on Generic Issca C-8, an estimate of 1/2 man-year, or $50,000, to prepare recom-mendations, a regulatory analysis and supporting documentation may be more i realistic.

Conclusions -- MSIV Leakage Control Systems If the M51V leakage control system were a new requirement to be evaluated under current procedures and policies such as the revised backfit rule or the pro >osed s tty goals it could not be justified on the basis of a quanti- ,

l tative sene% ' guh!eline of $1000 per person-rem; conservatively calcu-lated, the n...., for the LCS is on the order of $100,000 per person-rem.

If one considers eliminating the systems, the calculations are slightly I different. First, the large initial cost of the system is now a sunk cost xv r?.nw a . m n w r . n m u .-- - - n -- --

for the current generation of plants. Second i

plant for the industry (i.e.,g$300,000); initial 25 plants times costperto effec

$12,000 for the NRC, $50,000 for revising the staff positten plus $250,000 for technical specification changes (i.e.

$300,000). After this initial outlay of $600 000, savings would then accru,e to industry at the rate of $500,000 per year (,25 plants times $20,000 per reactor year is operating savings) over the remaining life of the current plants.

Adding all the costs and cost savings, and discounting future cost savings at a les real discount rate (as suggested in NUREG/BR-0058, NRC 1984a) over 38 years, the net monetary benefit is ,

$500,000 a 9.43 - $600,000 = $4.1 million.

The increase is risk to the public is estimated at 7.5 person-rem per year (0.3 person-rem per reactor year times 25 reactors). This works out to ,

$18,000 saved per person-rem of dose increase.

These quastitative calculations are provided for perspective. It should be stressed that quastitative analyses of this nature are not the sole or '

even the principal basis for regulatory decisions. Moreover, the numerical values are highly uncertain, and should be intenreted cautiously.  !

Fuel Systes safety heriews A fundanestal cancept in the design of nuclear power plants is the provision of altiple fission product barriers to protect the health and safety of the pelic from releases of radioactive material during normal i operations and under accident conditions.

barriers is provided by the fuel cladding in the fuel system.The Fuel system first of thes i components plugs, fill gas, inclade etc. the fuel rods (including pellets, cladding, springs, end burnable poison rods control rods and various associated hanhare su)c,h as spacer grids, spr,ings, end plat boxes.

c Because of its role as the first line of defense in the defense-in-depth design philosegby, the licensee's fuel system design is carefully reviewed by the RRC staff to ensure compliance with applicable regulatory requirements.

Procedures for fuel system safety reviews by NRC staff are set forth primarily in section 4.2 of the Standard Review Plan, NUREG-0800 (NRC 1981a).

The reviews identified overaddress a large number of complex technical issues that have been the years.

in the body of this report. An overview of these review procedures is given With about one thousand reactor operating years of experience in the United States, and on the order of four thousand reactor operating years worldwide, the technology of fuel design is mature.

Given the accumulated experience and the saphistication of current analytical models and design practices, it is concrivable that the NRC's reviews of fuel system design information sdudtted by licensees could now be streamlined and simplified without adversely affecting safety. To test this hypothesis, this report considers the cassequences of eliminating some steps from the current review procedures.

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'g UNITED STATES

! n l 1 NUCLEAR REGULATORY COMMISSION i E  : ,I f wAsmuorow, o. c. mu y)

\ **"* / SEP 20 686 MEMORANDUM FOR: Victor Stello, Jr.

Executive Director for Operations FROM: James H. Sniezek, Chairinan Comittee to Review Generic Requirements

SUBJECT:

MINUTES OF CRGR MEETING NUMBER 95 The Comittee to Review Generic Requirements (CRGR) met on Wednesday, August 27, 1986, from 1-3 p.m. A list of attendees for this meeting is enclosed (Enclosure 1).

1.

L. Shotkin and J. Reyes of RES continued their CRGR review presentation on g p, j k and changes to Appendix K to Part 50.the matter of the proposed ECCS r T / 9, CRGR of the actual rule language revisions now underway by RES, the CR u4 favorably supported the staff's proposed course of action toward resolution of the several open issues that resulted from CRGR meeting No. 91.

Enclosure 2 summarizes this matter (Category 2 item).

t 2..j R. Bernero (NRR) requested an opportunity to brief the CRGR members about p% >

activities being currently discussed with the BWR Owners' Group (BWROG) i concerning severe accident management strategies to enhance accident u'lk. l ./ mitigation capabilities of the Mark I containment.

issuance of a generic letter on this matter in the near future.HRR CRGR is considering provided no coments or recomendations on this matter. Enclosure 3 sumarizes this matter.

Enclosure 2 contains predecisional infonnation and, therefore, will not be released to the Public Document Room until the NRC has considered (in a public forum) or decided the matter addressed by the infonnation.

In accordance with the ED0's July 18, 1983 directive concerning "Feedback and Closure on CRGR Reviews," item 1 above requires written response from the cognizant office to report agreement or disagreement with CRGR recommendations in these minutes.

l The response, which is required within 5 working days after receipt of these meeting minutes, is to be forwarded to the CRGR Chainnan and if there is disagreement with the CRGR recomendations, to the EDO for decisionmaking, d M ^

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. :' . i' Enclosure 1 LIST OF ATTENDEES CRSR MEETING NO. 95 August 27, 1986 CRGR MEMBERS J.H. Sniezek R. Bernero R.Starostecki(E. Jordan)

J. Scinto J. Heltemes R. Cunningharn D. Ross OTHERS J. Zerbe M. Taylor W. Schwink C. Thomas N. Lauben C. Berlinger J. Stewart M. Jamgochian M. Lesar P. Boehnert M. Fleishman L. Shotkin B. Morris K. Kniel W. Beckner J. Reyes

-Enclosure 2 to the Minutes of CRGR Meeting No. 95 Proposed Revisions to 10 CFR 50.46 (ECC5 Rule),

Appendix K and Associated Regulatory Guide (draft) t.. Shotkin and J. Reyes (RES) continued from Meeting No. 91, the CRGR review presentation on the proposed ECCS rulemaking package. A copy of the vugraphs used in presenting this matter to CRGR is attached. (Refer to the Minutes of Meeting No. 91 for a more complete background on the proposed ECCS rulemaking package.) At Meeting No. 91, the CRGR concluded that the proposed ECCS ,

rulemaking package was in need of further language revisions and that there were three open issues requiring additional staff consideration and development of regulatory positions. Basically, these were:

1. The potential impacts on licensees of elimination of the Dougall-Rohsenow correlation as an acceptable modeling feature of Appendix K.

2. The appropriateness of the reporting requirements in the proposed rule.

This included the necessity for and promptness with which each and every calculational error should be reported regardless of magnitude, the in-clusion of reporting requirements for applicants as well as the licensees, and the need for the staff to impose upon itself a 60-day response re-quirement for addressing these ECCS calculational errors.

3. A determination by the Director of the Office of Research, in accord with CRGR Charter, whether the proposed rule actions would result in a decrease in the overall protection of cublic health and safety and whether the proposal would result in cost savings for NRC and/or the industry.

By memorandum from the Acting Director of Research (dated August 19,1986),

these issues were formally addressed. Regarding issue 1. above, concerning the Dougall-Rohsenow correlation and its proposed deletion from Appendix K, the staff expressed the view that with today's data and science available on ECCS performance and post-CHF correlations, it was not a good staff practice to continue to endorse by rule an outdated post-CHF correlation. While the staff noted that the Dougall-Rohsenow correlation could be made to give a better data fit by use of different correlating parameters, they would prefer its deletion from Appendix K. As indicated in the attached vugraph copy, the staff has considered four potential options having various impacts on licensees. For practical considerations, the staff has reached the conclusion that the option i (Option 2) which pennits existing Evaluation Models (EMS) to be "grandfathered" with the Dougall-Rohsenow correlation should be followed. At the same time this option would remove the Dougall-Rohsenow correlationas a generally i

< acceptable feature from Appendix K. It was the staff's overall judgment that  !

the current ECCS EMS are sufficiently conservative, on the whole, to pennit the l continued use of this correlation in existing EMS, For future modifications j to EMS, including error corrections, the impact of continued EMS acceptability i

! will be considered by the staff based on the overall conservatism of the eval-I uation model. If a new evaluation model is submitted, such as an EMS modified

2-because of errors or model revisions, or use of a new computer code, the ap- .

plicability of the Dougall-Rohsenow post-CHF correlation would have to be ad- i dressed. By this approach, the staff would intend to gradually phase out use of an outdated correlation while at the same time trying to minimize the impact to the licensees (and to NRC resources) by not requiring wholesale ECCS re-analyses at large costs. The staff recognized that this option would still leave an inconsistency in the rule, but such a gradual phase-out was the preferred approach for minimizing the potential analysis burdens. After further discussion on the various proposed options, CRGR generally endorsed the staff's plans to proceed with Option 2 as described above. CRGR further observed that it would be useful to develop a more specific standard of ac-ceptability to accompany Option 2 (e.g., an acceptable level of temperature increase to be allowed if continued use of the existing EMS with the Dougall-Rohsenow correlation were to be permitted) and reconinended the staff work toward this development.

Regarding the issue of reporting requirements, the staff expressed the view that the proposed ECCS rule revisions represent an improvement over the exist-ing ECCS rule that, as implemented, has resulted in the licensee to intnediately report and correct all errors --even those that may be very minor. The pro-posed ECCS rule revisions that were questioned by CRGR at Meeting No. 91 would reflect an NRR preference to be imediately informed of all errors, even minor, but would provide NRR with some flexibility on how quickly these errors are to be corrected. On the other hand, RES believes all minor errors should be noted and submitted to the agency on a regular basis (such as annually). Imediate  !

reporting of minor errors (less than a 50'F effect on calculated peak cladding temperature and not exceeding the 650.46(b) criteria) would not be a require-ment. CRGR expressed a generally favorable view toward the proposed RES ap- l proach to this matter of reporting the minor errors. However, CRGR noted that the RES staff would need to develop additional justification (or a rationale) to explain and support an NRC policy of allowing the knowledge of erroneous models to exist unreported for a period of up to a year. Also, the cununulative effects of such minor errors should be addressed. The RES staff advised CRGR that information was at hand that could be relied upon to develop this addi- ,

tional justification for periodic reporting of the minor errors. I Concerning the issue of whether the proposed rule would result in a decrease in  !

the overall protection of public health and safety, it was the determination of the Office of Research that the proposed action would not result in any de-  ;

crease in plant safety because the conservative safety limits of 150.46(b) l would not be reduced. However, it was made clear by RES that while this action  !

may not be considered a decrease in plant safety, it does represent a reduction l in margin introduced by the currently acceptable, yet overly conservative, cal-culational methods. On balance, RES expressed the view that using more realism in this matter should produce some safety benefit, albeit unquantifiable. Re-garding costs, RES was of the view that most of the Westinghouse plants could potentially benefit from the proposed ECCS rule revisions should this option

3 for use of more realistic ECCS performance calculations be exercised. An aver-age plant may be able to upgrade total power by an estimated 5% as a result of the proposed ECCS rule. This could result in lifetime energy replacement cost savings having a present value of between $70M to $100M. The estimated 3 to 4 staff years by NRC that would be needed to review the generically-based, rea-listic ECCS modeling subnitted by vendors is believed to reasonably reflect the NRC cost expenditures. The staff resources are estimated to be adequate for this future review purpose.

In the overall, CRGR endorsed the above courses of action proposed by the staff to resolve those open issues from Meeting No. 91. CRGR recommended that RES work y ' th OGC in developing the final proposed ECCS rule language. When ready, RES should provide the CRGR with the revised rule package for a determination as to whether further CRGR review is needed.

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