ML20038A889
| ML20038A889 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/17/1981 |
| From: | Adler R PENNSYLVANIA, COMMONWEALTH OF |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8111240398 | |
| Download: ML20038A889 (11) | |
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UNITED STATES OF AMERICA DOLMETED hUCLEAR REGUIAIORY CDtMSSION U%RC BEFORE THE ATOMIC SAFE 1Y AND LICEtBING BOARD
'81 NOV 19 P2:42 In the Matter of:
, _ mm.gf J,. :EimCE PENNSYLVANIA PORER & LIGdT (X).
C and F1EGHEIE ELECIRIC COOPERATIVE, Docket tbs. 50-387 j
, -,, 'a y)d INC. (Susquehanna Steam Electric 50-388
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COFIDWEAlmi 0F PEh15YLVANIA'S FDTION TO P
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TAKE OFFICIAL IDTICE OF CERTAIN DOCUMENIS S i, "% " ISS7 f
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The Correnwealth respectfully taoves the Board, pursuant to o
C.F.R. 52.743(h) and (i), to take official notice of the following doctInents :
(1)
U.S. thclear Regulatory Conmission, Standard Reviea Plan, Section 15.6.2, Fadiological Consequences of the Failure of Scull Lines Carrying Primary Coolant Outside Containment (Rev. 2 - July 1981); and (2)
U.S. Nuclear Regulatory Cocmission, Standard Technical Specificac'.cns for General Electric Boiling Water Reactors (%R/5),
hTREG-0123, Rev. 3, p. B 3/4 4-3 (specific activity in reactor coolant system).
Both of these documents are relevant to Contention 21 in tids proceeding, as explained belcw.
Copies of the docucents are attached and are being served en all parties.
The IRC Staff, in response to Contention 21, introduced into evidence as an exhibit hTREG-0803, Generic Safety Evaluation Report Regarding Integrity of ER Scram System Piping.
Staff Ex. 5.
hu EG-0803 p54 S 8111240398 811117 PDR ADOCK 05000287 SO /
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O cites both Section 15.6.2 of the Standard Review Plan and the Standard Technical Specifications for BWR plants as key reference documents underlying the analysis:
The iodine concentration in the primary coolant is expected to undergo a teporary increase above its equilibrium level following a scram and prompt depressurization (" iodine spiking").
The model used to account for this phenomenon is the one identified in Section 15.6.2 of the SRP (Ref.
18)--a temporary increase of 500 tires the equilibriun release rate from the fuel. which in turn is determined from the equilibrium coolant concentration.
This equilibrium concentration may vary substantially during the lifetime of the plant.
In all cases, however, it is limited by the technical specifications for coolant activity.
The Standard Technical Specifications (SIS) for BWR plants, therefore. were used as the base case for this analysis.
Because the technical specifications for coolant activity substantially deviate from the SIS for some BWRs, the calculations also were performed for a plant with typical nonstandard technical specifications, that is, those for the Broms Ferry plant.
Staff Ex. 5, at 4-18 (sphasis added).
The Coamorseealth requests the Board only to take official notice of the existence and content of these documents.
It is understood that the validity of the information contained therein could be subject to dispute, and wculd not properly be the subject of official notice. Therefore, the purpose for which official notice is requested is limited.
The documents identified above are official agency records within the meaning of 10 C.F.R. 52.743(h). Moreover, the authenticity of the attached documents is a fact that can be readily and accurately ascertained.
Therefore, the Coccorseealth respectfully requests that this nution be granted.
Respectfully submitted, f
. p20 ROBERT W. ADLER
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Assistant Counsel Conronwealth of Pennsylvania s
NU REG-0800 (Formerly NUREG 75/087) 00'KETED
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USNRC U.S. NUCLEAR REGULATORY COMM:SS!GN
!4,3$,,%)STANDARD REVEW PLAN e2:42
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NO f OFFICE OF NUCLEAR REACTOR REGULATION ECRE TI RY
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ERV:CE 15.6.2 RADIOLOGICAL CONSEQUENCES OF THE FAILURE OF SMALL LINES CARRYING' Q~
PRIMARY COOLANT OUTSIDE CONTAINMENT REVIEW RESPONSIBILITIES Primary - Accident Evaluation Branch (AEB)
Secondary - None l
I.
AREAS OF REVIEW This SRP section covers the radiological consequences of failures outside the containment of small lines connected to the primary coolant pressure boundary, such as instrument lines and sample lines.
The review includes the following:
1.
The identification of small lines postulated to fail and the isolation provisions for these lines, including the applicability of General Design Criterion 55 (Ref. 1), which requires isolation capability of the line inside and outside containment, and Regulatory Guide 1.11 (Ref. 2), which requires isolation capability outside containment for those lines that are exempt from*GDC 55.
The implementation of these regulatory positions and guidelines is reviewed by the Containment Systems Branch (CSB) under SRP Section 6.2.4.
2.
The failure scenario, as described by the applicant, to assure that the most severe radioactive releases have been considered.
3.
The models ad assumptions used by the applicant for the calculation of the thyroid and whole-body doses for the postulated failure.
4.
An evaluation of the primary coolant iodine activity, including the effects of a concurrent iodine spike, and the technical specifications for the reactor coolant iodine activity.
5.
An independent calculation by the staff of the thyroid and whole-body doses for the small line failure, including an evaluation of the isolation times and maximum leak rates of the isolation valves.
Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN Star.dard rwew plans are prepared f cr the guidance of the office of Nuclear Reactor Regafation staf f responsible for the review of applications to construct and operate nuctear power pfants. These documents are made available to the pt.blic as part of the Commission s pohcy to inf orm :he nuclear industry and the general public of regulatory procedures and policies. Standard review clans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required The standard review plan sactions are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
Not all sections of the standard Format have a carresoondmg review plan.
Pubteshed standard review plans will be revised periodically, as appropriate. to accommodate comments and to reflect new inf orma-tion and experience.
Con ments and suquestions f 3r impeavement wel be consioared and should be sent to the U S. Nuclear Regulatory Commission.
Of f:ce of Nuclear Reactor Regulation. Washingten o C. 20555
y 6.
A comparisen of the doses calt' lated by the applicant and by the staff with appropriate exposure gui elines of 10 CFR Part 100, 9100.11 (Ref. 3),
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as stated in subsection II below.
t In addition, AEB will coordinate its review with other branches that interface in the overall review of the break analysis.
The Reactor Systems Branch (RSB) i upon request by the AE8 will confirm the value used by the applicant for the mass of coolant released in the accident and to determine if this accident will cause fuel failures.
The Containment Systems Branch (CSB) upon request by the AEB, will verify Mat secondary containment integrity and leaktightness are maintained during the course of the accident.
4 i
II.
ACCEPTANCE CRITERIA u
The acceptance criteria for this SRP section are based on the relevant i
requirements of the following regulations:
l 1.
General Design Criterion 55 (Ref.1) as it relates to the idantification of small diameter lines connected to the primary system that are exempted from the isolation requirements of GDC 55 and that are acceptable on the I
basis of meeting item (2) below, t
2.
10 CFR Part 100, 9100.11 (Ref. 3) as it relates to the radiological consequences of a small line break carr;ing primary coolant cutside I
containment.
The plant site and the dose mitigating engineerei safety feature (ESF) systems are acceptable with respect to the radiological consequences of a postulated failure outside the containment of a small line carrying reactor coolant if the calculated whole-body and thyroid doses at the exclusion area and the low population zone outer boundaries do not exceed a small fraction of the exposure guideline values of 10 CFR Part 100. g100.11 (Ref 3) as stated in position C.1.b of Regulatory Guide 1.11 (Ref. 2).
A "small fraction" of 10 CFR Part 100 means 10 percent of these exposure guideline values,.that is, 2.5 rem and 30 rem for the whole-bcdy and thyroid doses, respectively.
A plant-specific technical specification is required for the iodine activity in the primary coolant system.
The specification is acceptable with respect to the postulated failure if the calculated doses resulting from the failure are within the above exposure guidelines.
III.
REVIEW PROCEDURES The reviewer selects and emphasizes specific aspects of this SRP section as are appropriate for a particular plant.
The areas to be given attention and emphasis are determined by the similarity of the information provided in the applicant's Safety Analysis Report (SAR) to that recently reviewed on other The plants and whether items of special safety significance are involved.
review consists of the following steps:
1.
Review of the applicant's description of the small line failures to determine the appropriateness and conservatism of the assumptions used in the analysis.
i 2.
Identification of the small lines connected to the primary reactor coolant s
system and penetrating the containment.
The isolation provisions a:e 15.6.2-2 Rev. 2 - July 1981 w-w*
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identificd with respect to the applicability of GDC 55 (i.e., isolation capability inside and outsiae containment) and Regulatory Guide 1.11 (i.e., isolation capability outside containment for lines exempt from GDC 55).
The implementation of these guidelines is reviewed by the Containment Systems Branch under SRP Section 6.2.4 The AEB reviewer will ccordinate his review with CSB if additional clarification is needed.
3.
Performance of an independent analysis by the staff.
The reviewer selects for a failure analysis those small lines that most likely will result in the highest offsite radiological consequences.
The selection is largely based on the ' analysis performed on recently reviewed plants' but should include, if appropriate, the letdown line of the chemical volume and control system (CVCS) and the. largest instrument and sample line.
The I
following conservative assumptions are made for the analysis:
a.
For small lines that meet GDC 55, such as the CVCS letdown line, the failure is assumed to occur downstream of.the outboard containment isolation valve in' conjunction with a single failure of one of the two containment isolation valves.
The amount of primary coolant i
released outside the containment is determined by considering the method. capability and time required to detect such failure and the time required to isolate the failure (i.e., time to close the operable isolation valve).
b.
For small lines exempt from GDC 55, such as instrument lines, but which meet the isolation guidelines of Regulatory Guide 1.-11 (i.e.,
containment isolation valve outside containment), the failure is postulated to occur downstream of the valve in conjunction with a single failure (i.e., valve does not close).
Unless other isolation or flow reduction capabilities are provided (e.g., orifice in line) which will be evaluated on a case-by-case basis, it is assumed that this line failure cannot be isolated and the primary coolant release will continue until the primary system is depressurized.
c.
The amount of primary coolant released is conservatively estimated by assuming critical flow at the small line break location with the reactor coolant fluid enthalpy corresponding to normal reactor coerating conditions.
The reviewer evaluates the reactor coolant release rates provided by the applicant, taking into consideration l
similar information for plants recently reviewed.
The reviewer-should verify the release rates and the total amount of coolant released with the RSB in a coordinating review effort.
I d.
The initial fission product concentrations in the primary coolant are assumed to be the maximum equilibrium values permitted by the standard technical specification for the NSSS vendor or those pro-l vided by the applicant.
In addition, it is assumed that an iodine spike occurs as a result of the reactor shutdown or depressurization of the primary system.
The spike is modelled by increasing the equilibrium fission product activity release rate from the fuel by a factor of 500.
4 The reviewer consults with the RSB regarding the potential for and g
extent of camage to the fuel as a result of the line failure.
If i
appropriata, the additional fission product activity in the primary coolant activity will be included in the analysis.
i 1
15.6.2-3 Rev. 2 - July 1981
The fraction of the iodine assumed to beccme airborne and available for release to the atmosphere, without credit for plateout,.is equal to the fraction of the coolant flashing into steam in the depressuri-zation process.
The flash fraction is determined by assuming the discharge to be a constant enthalpy process, e.
For a plant with a dual containment system, it is assumed that the small line failure occurs outside the secondary containment if the line penetrates or bypasses the secondary containment.
The release is assumed to occur within the secondary containment if the line terminates ins'ide the secondary containment.
The reviewer verifies, in a coordinating review effort with the CSB the integrity and leak-tightness of the secondary containment during the pressure transient associated with the postulated small line failure within its boundaries.
An approximate mixing volume is determined from the location of the assumed failure location and the proximity to the secondary containment ventilation system assumed to be operating (if any).
The release of the airborne radioactivity frem the secondary containment to the outside atmosphere is evaluated in accordance with the assumption of SRP Section 15.6.5, Appendix A, subsection III.3.
f.
The operation and effectiveness of an ESF grade filtration system for removal of airborne radiciodine will be reviewed on a case-by-case basis.
The reviewer verifies that all potential locations for a small ljne break are within ventilation zone of the system.
Depending on the type of air treatment system credited in the analysis, a ground-level or elevated (stack) release is assumed.
The appropriate atmospheric dispersion factors (X/Q values) are provided by the assigned meteorologist in accordance with SRP Section 2.3.4.
4.
Review of dose calculations.
The wnole-body ano thyroid doses calculated by the staff and by the applicant are compared with the acceptance criteria stated in subsection II of this SRP section.
If the doses calculated by the staff are not within the exposure guidelines (i.e., they are not less than 10 percent of 10 CFR Part 100, 9100.11), then the staff will pursue alternatives with the applicant to reduce the doses to within the guideline values.
IV.
EVALUATION FINDINGS The reviewer verifies in the Safety Evaluaton Report (SER) that sufficient information has been provided in the SAR.
The applicant's analysis and the staff's independent calculations are summarized.
The SER should identify the specific small line fa'ilure that was analyzed by the staff and the calculated doses, including the assumptions and unique system and operation provisions.
The evaluation snould support conclusions of the following type to be included in the SER:
The staff concludes that the distances to the exclusion area and to the low population zone outer boundaries for the (insert PLANT NAME) site, in conjunction with the operation of the dose mitigating ESF 15.6.2-4 Rev. 2 - July 1981
v systems, are sufficient to provide reasonable assurance that the calculated radiological. consequences of a postulated small line failure outside the containment, assuming the primary coolant equilibrium iodine concentrations permitted by the standard technical specifications, in combination with an accident generated iodine spike, do not exceed a small fraction of the exposure guidelines as set forth in 10 CFR Part 100, 6100.11.
The results of
- the staff's calculations are listed in Table 15.
The staff's conclusion is based on (1) the staff review of the applicant's classification and identification of small lines in accordance with General Design Criterion 55, " Reactor Coolant Pressure Boundary Penetrating Containment," and Regulatory Guide 1.11,
" Instrument Lines Penetrating Containment," (2) the staff review of the applicant's analysis of radiological consequences, (3) the independent dose calculation by the staff using regulatory position C.1.b of Regulatory Guide 1.11 and conservative atmospheric dispersion factors as discussed in Chapter 2 of this report,-cnd (4) the (insert NSSS VENDOR) standard technical specifications for the equilibrium iodine concentrations in the primary coolant system.
The staff will review the (PLANT NAME) specific technical specifications to assure that the dose guidelines stated above are not exceeded.
V.
IMPLEMENTATION The following provides guidance to applicants and licensees regarding the staff's plans 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, i
the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.
Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulatory guide.
VI.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 55, " Reactor Coolant Pressure Boundery Penetrating Containment."
2.
Regulatory Guide 1.11, " Instrument Lines Penetrating Primary Containment."
3.
10 CFR Part 100, S100.11, " Determination of Exclusion Area, Low Population Zone and Population Center Distance."
i
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t 15.6.2-5 Rev. 2 - July 1981
NUREG-0123 Revision 3 Standard Technical Speci3 cations for General Electric Boiling V/ater Reactors (BWR/5)
Revision Issued Fall 1980 Supercedes NUREG-0123, Rev. 2 issued by the U.S. Nuclear Regulatory Commission O
a Office of Nuclear Reactor Regulation
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REACTOR COOLANT SYSTEM BASES 3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100.
The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the (
) site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site.
This reevaluation may result in higher limits.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0 micro-
/ ~x curies per gram DOSE EQUIVALENT I-31, accommodates possible iodine spiking
(
)
phenomenon which may occur folicwing changes iri THERMAL POWER.
Operation with
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specific activity levels exceeding 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 must be restricted to no mcre than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year, approximately 10 percent of the unit's yearly operating time, since these activity levels increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to (20) following a postulated (steam line rupture).
The reporting of cumulative operating time over 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any 6 month consecutive period with greater than 0.2 micro-curies per gram DCSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit.
Information obtained on iodine spiking will be used to assess the param-eters associated with spiking phenomena.
A reduction in frequency of isotopic analysis following powcr changes may be permissible if justified by the data obtained.
Closing the main steam line isolation valves prevents the release of activ-ity to the environs should a steam line rupture occur outside containment.
The surveillance requirerents provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
k i
'Qi GE-STS 8 3/4 4-3
s UNITED STATES OF AMERICA NUCLEAR REGLIATORY C1 MISSION BEFORE THE AIDMIC SAFETY AND LICEISING BOARD In the Matter of:
PENtEYLVANIA POWER & LIGIT C0.
and ALLEGiENY ELECIRIC (I)0 PERATIVE, Ibcket Nos. 50-387 INC. (Susquehama Steam Electric 50-388 Station, Units 1 and 2)
CERTIFICATE OF SERVICE I hereby certify that copies of the attached "Comomealth of Pemsylvania's Motion To Take Official Notice of Certain Documents",
with attacITc.ents, were served on the parties on the attached service list this 17th day of Nova 6er,1981, by deposit in the U.S. mail, first class, postage prepaid.
ROBERT W. ADIZR
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LUITED STATES OF AMERICA IED BUCLEAR PSGLIAIORY CCEESSION fypc BEFORE THE ATOMIC SAFETY AND LICENSIl0 BOARD
'81 NOV 19 P2:42 In the Matter of e ? SECRETARY
'"~;il G & SERV!CE PESBSYLVANIA POWER & LIGHT CO.
TRANCH and ALIJEdENY ELECIRIC COOPERATIVE, Docket Mos. 50-387 INC. (Susquehanna Steam Electric 50-388 Station, Units 1 and 2)
SERVICE LIST Secretary of the Ccarission Th =as J. Halligan U.S. Nuclear Regulatory Comrission Correspondent: CABO Washington, D.C.
20555 P.O. Box 5 Scranton, Pennsylvania 18501 Docketing and Service Section Office of the Secretary Ms. Colleen Marsh U.S. thclear Regulatory Conrission 558 A, R.D. #
Washington, D.C.
20555 Mt. Top, Pennsylvania 18707 Atomic Safety and Licensing Board Panel, Jessica H. Laverty, Esquire U.S. :hclear Regulatory Comission Office of the Executive Legal Director Washington, D.C.
20555 U.S. Nuclear Regulatory Conrission Washington, D.C.
20555 Adrnnistrative Judge Ja.es P. Gleason 513 Gilmoure Drive Jay Silberg, Esquire Silver Spring, Maryland 20901 Shaw, Pitt an, Potts & Trowbridge 1800 M Street, U.W.
Glenn O. Bright Washington, D.C.
20036 Atomic Safety & Licensing Board Panel U.S. :bclear Regt.latory Comission Bryan Snapp, Esquire Washington, D.C.
20555 Pennsylvania Pcwer & Light Company 2 North Ninth Street Dr. Paul W. Purdom Allentown, Pennsylvania 18101 245 Gulph Hills Road Radnor, Pennsylvania 19087 James BL Cutebin, IV Office of the Executive Irgal Director Dr. Judith H. Johnsrud U.S. Nuclear Regulatory Comission Co-Director Washington, D.C.
20555 Enviror:rantal Coalition on Nuclear Power 433 Orlando Avenue DeWitt C. Smith State College, Pennsylvania 16301 Director Pennsylvania Emergency Manage:ent Agency Susquehanna Envirotrental Advocates Transportation and Safety Building c/o Gerald Schultc, Esquire Harrisburg, Pennsylvania 17120 Post Office Building Wilkes-Barre, Pe.nsylvania 18703
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