ML20196E165
| ML20196E165 | |
| Person / Time | |
|---|---|
| Issue date: | 06/11/1999 |
| From: | Stewart Magruder NRC (Affiliation Not Assigned) |
| To: | Carpenter C NRC (Affiliation Not Assigned) |
| References | |
| PROJECT-689 NUDOCS 9906280095 | |
| Download: ML20196E165 (28) | |
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June 11,1999 MEMORANDUM TO: Cynthia A. Carpenter, Chief Generic Issues, Environmental, Financial and Rulemaking Branch Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation FROM:
Stewart L. Magruder, Project Manager Original Signed By:
Generic issues, Environmental, Financial and Rulemaking Branch 4
Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation
SUBJECT:
SUMMARY
OF JUNE 2,1999, MEETING WITH THE NUCLEAR ENERGY INSTITUTE (NEI) REGARDING CONTROL ROOM HABITABILITY On June 2,1999, representatives of the Nuclear Energy institute (NEI) met with the staff of the Nuclear Regulatory Commission (NRC) at the NRC's offices in Rockville, Maryland. provides a list of attendees. Attachment 2 provides the meeting agenda.
The meeting was held so that representatives of the NEl Task Force on control room habitability could present a discussion on the uncertainty and conservatism associated with the control room habitability assessments. This discussion was presented in light cf the guidance provided in NEl 99-03," Control Room Habitability Assessment Guidance." The discussion focused on whether a more rigorous presentation of uncertainties and conservatism would be beneficial to the staff and should be included in NEl 99-03. Attachment 3 is a copy of the NEl handout which was utilized by NEl in the discussion.
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WASHINGTON, D.C. 20655-0001 49 * * * * *,o June 11, 1999 MEMORANDUM TO: Cynthia A. Carpenter, Chief Generic issues, Environmental, Financial and Rulemaking Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation FROM:
Stewart L. Magruder, Project Manager %M % %= b b
Generic issues, Environmental, Financial and Rulemaking Branch Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation
SUBJECT:
SUMMARY
OF JUNE 2,1999, MEETING WITH THE NUCLEAR ENERGY INSTITUTE (NEI) REGARDING CONTROL ROOM HABITABILITY On June 2,1909, representatives of the Nuclear Energy Institute (NEI) met with the staff of the Nuclear Regulatory Commission (NRC) at the NRC's offices in Rockville, Maryland. provides a list of attendees. Attachment 2 provides the meeting agenda.
The meeting was held so that representatives of the NEl Task Force on control room habitability could present a discussion on the uncertainty and conservatism associated with the control room habitability assessments. This discussion was presented in light of the guidance provided in NEl 9E-03, " Control Room Habitability Assessment Guidance." The discussion focused on whether a more rigorous presentation of uncertainties and conservatism would be beneficial to the staff and should be included in NEl 99-03. Attachment 3 is a copy of the NEl handout which was utilized by NEl in the discussion.
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k NEl/NRC CONTROL ROOM HABITABILITY MEETING June 2,1999 List of Attendees NAME ORGANIZATION Kurt Cozens NEl l
Jim Metcalf Polestar Gopal Patel Nucore John Duny PSE&G Steve LaVie NRC'NRR Mark Blumberg NRC/NRR Michelle Hart NRC/NRR Mark Reinhart NRC/NRR Jay Lee NRC/NRR Jack Hayes NRC/NRR Millard Wohl NRC/NRR I
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NEl/NRC MEETING ON CONTROL ROOM HABITABILITY ISSUES AGENDA June 2,1999 1:45 Introductory Remarks l
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Preliminary Information - for Illustration Only Conservatism in Design Inputs & Assumptions Used in L.icensing Basis Control Room IIabitability Analysis for Loss of Coolant Accident (LOCA)I No.
DESIGN INPUT Wily PARAMETER IS POTENTI AL MAGNITUDE PARAMETER PWR BWR CONSERVATIVE OF CONSERVATISM 1
Source Term Magnitu<le:
X X
NRC proposed revised source The overall impact of implementation Post-LOCA Containment term (NUREG-1465) to be of the revised source term in the Airborne Source Term voluntarily implernented by the majority of cases is to produce lower Noble Gas 100%
operating plants (SECY-98-158) calculated doses, ranging from a lodine 25%
which demonstrates that the use slight reduction up to an order of of TID-14844 & RGs 1.3 & l.4 magnitude decrease (SECY-98-154, Waterborne:
source term in the current page 4) lodine 50%
licensing basis analyses is conservative (SECY-98-154).
2 Source Term Timing:
X The revised source term treats The dose contribution from the un-The release of fission products the release of fission products as filtered ground level release during (source term) is instantaneous a time-dependent release; thus, drawdown time becomes relatively inside the containment.
reduction of dose will be inconsequential using revised source strongly influenced by safety term (SECY-98-154, page 5).
features which are timing sensitive.
X There is little inherer.' dvantage of a time-dependent release in PWR containment if spray removal rate is large compared to the duration of i
release time (EPRI TR-105909, page A-9). There may still an advantage iri delaying the release.
3 Chemical Form of lodine:
X X
NRC proposed in revised source Treatment ofiodme primarily as an Liementaliodine 91%
term:
aerosolin revised source term Particulate iodine 5%
Elementaliodine 4.85%
resulted in a substantial reduction of Organic iodine 4%
Particulate iodine 95%
the thyroid dose, which is the most Organic iodine 0.15%
limiting dose for the operating plants.
4 Containment Leak Rate:
X X
The post-LOCA containment MELCOR analysis varies the Primary containment is leak rate used in DBA analysis is containment leakage in accordance assumed to leak at the based on the peak containment with the predicted post-LOCA Technical Specification leak pressure. The post-accident containment peak pressure (SECY-rate (weight % per day).
containment pressure varies98-154, Figure 6-12). The Type A substantially and eventually tests demonstrate that the become atmospheric or sub-containment overallintegrated atmospheric.Therefore, the leakage rate is substantially less than containment leakage is reduced half of the Technical Specification to a fraction of the peak leak rate limit. A margin of a factor of two pithin a few hours of accident, exists in the conteinment leakage.
5 Atmospheric Dispersion X
X lThe Murphy-Campe model does The NRC-accepted ARCON-96 I Magnituda of conservatisms given in the above Table represent generally observed margin in the design inputs and assumptions used in current Licensing basis analyses of the operating plants. The plant-specific information shot.d be used to obtain more accurate results.
Preliminary Information - for Illustration Only tmospheric dispersion models not predict the variations of the a
Factors:
concentrations near buildings predict that the dispersion factors Most operating plants use Murphy-Campe methodology well, particularly at low wind may be at least a factor of five lower for calculating the atmospheric speeds.
than the Murphy-Campe model predictions.
dispersion factors. (X/Qs).
6 Dose Conversion Factors:
Y X
The ICRP 2 DCFs are higher The use of FGR 11 & 12 (ICRP 30)
Regulatory Guide 1.3 & L4 than ICRP 30 DCFs. Federal DCFs alone will produce reduction in mandate use ofICRP 2 dose Guidance Report (FGR) No 12, the dose up to 40% (SECY-93154, conversion factors (DCFs).
which uses ICRP 30 DCFs is page 5).
However, some Licensees intended for use by Federal have used RG 1.109 DCFs.
agencies having regulatory responsibilities for protection of general public and'or werkers, such as the EPA, the NRC, and the Occupational llealth &
Safety Administration.
7 Allowable Dose Limits:
X X
GDC 19 requirer. that the CR Since 5 Rem TEDE is the NRC's Control Room (CR):
operator exposure should not latest definition of equivalency to the exceed 5 Rem whole body or its "5 Rem whole body" referred to in Thyroid 30 Rem Beta Skin 30/75 Rem equivalent to any part of body, GDC 19, it could be argued that a Whol,: Body 5 Rem for the duration of the accident. minimum margin of a factor of three Standard Review Plan 6.4 exists in the allowable dose limit of 5 provides acceptable CR operator Rem TEDE.
dose limits.
X The plant-specific Type C test The Type C tests (10 CFR 50, Bypass Leakages:
controls the leakages through Appendix J test) performed at vamu Various olant-specific bypass containment isolation valves plants demonstrate that the total leakages (including the MSIV (CIVs). The bypass leakage is bypass leakages are normally well bypass leakage) are considered categorized as a CIV leakage.
within the allow ible TS leak rate in the Licensing Basis Type C test controls the limits (a factor three or more).
analpis.
individual bypass leakage to be Therefore, a margin of a factor of within its allowable TS leak rate three exists in the BWR CR dose limit. The Bypass leakages resulting from the use of realistic contribute as much as 75% of bypass leakages (averaged over last the total CR thyroid dose in a few year Type C test results).
BWR F. ant.
X X
During the long-term re-Normally, the post-accident sump Most plants use minimum 10%
circulation phase the temperature water iodine concentration is low.
ofiodine in the ESF or ECCS of the sump or suppression pool Therefore, for the given sump water leakage is assumed to become
' water is much less than 212' F pH of 7 and temperrure of 180"F, the f
airborne during re-circulation (normally less than 180* F),
fiashing fraction c..odine would be i
phase (SRP 15.6.5, Appendix which is not sufficient to cause approximately 1% based on Table 7 flashing of the iodine in the ESF of EPRI Report EPRI NP-1271. The B).
leakage. In a PWR plant ESF partition factor used in the analysis leakages contribute as much as (10%)is at least conservative by a j
50% of total CR dose, factor five or greater.
j i
10 Containment Spray X
The spray removal rates are The use of updated spray modo l
Removal:
conservative comp.:relto the provides the higher values of spray Spray. removal rates for NRC accepted updatfmodel of, removal rates for elemental (r) and
]
1
)
Preliminary Information - for Illustration Only elemental and particulate are aerosol removal by containment particulate (,) iodines which reduce calculated based on guidance spray described in NUREG/CR-the CR thyroid dose by a factor of
{
provided in LRP 6.5.2 5966 for sump water pH 7 or two. This credit is not available for greater.
the plant having a very high (r).
I1 Suppression Pool (SP)
X The Reactor Safety Study Calculations for a BWR Mark I used Decontamination Factor (WASII-1400) assumed a DF in NUREG-1150 indicate that DFs (DF):
of 100 for subcooled SPs. The ranged from 1.2 to about 4000 with a The SP decontamination DFs are conservative compared median valuc of about 80. The PWR factors for iodines are to the NRC accepted updated never credited the spray system, calculated bned on guidance model for decontamination by which can be combined with the provided in SRP 6.5.5.
BWR SPs described in smpression poot additional NUREG/CR-6153. Considering sc ubbing. The overall conservatisni the fission products which of a factor five or greater is available.
bypass the pool, the effective DF will be less than 100, regardless of the pool's ability to screb fission products (NUREG-1465, Section 5.2).
12 Control Room Occupancy:
X X
Normally, the control room Parametric study shows that rotating 0-24 hrs 100 %
operators work in a rotating shift the CR operator in the normal 12 24-96 hrs 60%
of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> / day. Majority of the hours shift or retiring the maximum 96-720 hrs 40%
post-LOCA CR thyroid dose exposed operator from future CR occurs in first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the operation can reduce the CR thyroid accident. Twelve hours are dose by a factor of two.
enough to identify the cause of j
(
accident, initiate the safe shutdown action, and dose mitigation function. Enough trained operators are available to continue the safe shutdown functions.
13 CR Charcoal Filter X
X Most of the plants use lower The control room charcoal filtei is Efficiencies:
charcoal filter efficiencies than capable of removing more iodme Normally, the smaller than typically tested to meet TS than is typically credited.
tested values of CR charcoal acceptance criteria (ASTM filter efficiencies are used in D3803-89).
the analysis.
14 CR HEPA Filter Efficiency:
X X
The llEPA filter is capable of The use of average tested llEPA filter Normally, the smaller than removing more iodine than is efficiency will reduce the CR thyroid tested value of CR llEPA filter typically credited.The use of
- dose, efficiency is used in the average tested HEPA filter analysis.
efficiency will reduce the CR thyroid dose sub.tantially with the application of revised source term where 95% of the iodine is released m the aerosol form.
e Preliminary Information - for Illustration Only l
Explanation of Conservatism in Dose Limit Pilot Plant Control Room Dose (Rem)
Identification Thyroid Whole TEDE l
Thyroid Remaining Body CEDE TEDE A
B C
D = A x 0.03 C-D Surry - Phase I 19.4 0.I 1 0.82 0.582 0.238 Grand Gulf-Phase I 50.3 0.57 3.89 1.509 2.381 Surry-Phase 11 23 0.12 0.935 0.69 0.245 Grand Gulf-Phase II 3.94 0.392 0.532 0.118 0.414 Surry-Phase III 17.50 0.082 0.69 0.525 0.165 0.116 0.413 Grand Gulf-Phase 111 3.85 0.39 0.529 i
0.295 Average TEDE from Noble Gas and Particulates 0.207 Average TEDE from Noble Gas (see Note)
Note: Grand Gulf-Phase I dose is excluded from the average because thyroid dose exceeds 30 Rem limit. The TEDE doses in Table 4.1-1 are calculated based on the revised source term, which includes the airborne particulates. The particulates contribute over 30% of the inhaled dose. In the current Licensing Basis, the particulates are assumed to be in either suppression pool or sump water. Only noble gas and iodine become airborne and contribute to the inhaled dow. Therefore, the TEDE contribution from noble gas would be 0.207 Rem (0.295 x 0.70 = 0.207).
AVERAGE NON-THYROID TOTAL EFFECIIVE DOSE EQUIVALENT (TEDE) 0.207 Rem x 1.40 = 0.3 Rem (TID-14844 DCFs)
(remaining organs excluding the thyroid organ =
= 0.207 Rem using ICRP 30 DCFs)
Assumption:
Allow 1.7 Rem for the external WB exposure, resulting from the containment, external cloud, and CR filter shine doses NON-THYROID TOTAL El FECTIVE DOSE EQt'lVALENT = 0.3 + 1.7 = 2.0 Rem TEDE TEDE MARGIN = 5.0 - 2.0 - 3.0 Rem TEDE THYROID CEDE FOR MAXIMUM ALLOWABLE
=
LIMIT OF 30 REM 30 Rem x 0.03 = 0.90 Rem CEDE M ARGIN IN Al 1,0W ABI E THYROID DOSE LIMIT = 3.0/0.9 = 3.3 (i.e.," effective" thyroid lunit for comparison to 5.0 rem TEDE using current Licensing Basis source term and DCFs = 3.3 x 30 rem = 100 rem)
r Preliminary Information - for Illustration Only 1
t Combining ofIndividual Conservatisms to Obtain an Overall Conservatism 1
j Design input Parameter PWR BWR Magnitude of (Item Nos. In Conservatism Table)
Plant Plant Conservatism Containment Leak Rate (4)
X X
2 Atmospheric Dispersion factor (5)
X X
5 Dose Conversion Factor (6)
X X
1.67 Allowable Dose Limit (7)
X X
3 Bypass Leakage (8)
X 3
ESF/ECCS Leakages (9)
X X
5 Containment Spray Removal Rate (10)
X 2
1 Suppression Pool Decontamination Factor (11)
X 5
i Control Room Occupancy (12)
X X
2 Typical Conservatism in PWR Plant I icensina Bases Assumptions:
- 1. Contamment leakage contributes approximately 50%
- 2. ESF leakage contributes approximately 30%
- 3. Bypass leakage contmutes approximately 20% of the total CR thyroid dose. However, small variations are expected in the percentage contributions based on the plant-specific designs.
Reduction in the CR thyroid dose (containment leakage contribution / magnitude of (due to containment + ESF + bypass leakages and conservatism is the containment leakage / magnitude of j
PY conservatism in the containment spray removal rate) +
1 (ESF leakage contnbution/ rnagnitude of conservatism in the ESF leakage) + (bypass leakage contribution /
magnitude of conservatism in the bypass leakage /
magmtude of conservatism in the containment spray j
removal rate) i (0.50/2/2 + 0.30/5 + 0.20/3/2) = (O.125 + 0.06
+ 0.033) = 0.218 t
Overall conservatism in the CR thyroid dose (due to containment + ESF + bypass leakages and 1/0.218 = 4.59.
spray removal)
The remaining conservatisms in annospheric dispersion factors (X/Qs), dose conversion factors (DCFs), allowable dose limit, and control room occupancy apply equally to all sources of radiation.
Total consert stism in the Licensing basis analysis = 4.59 (containment + ESF + bypass leakages) x 5 (X/Qs) x I
1.67 (DCFs) x 3.0 (allowable dose limit) x 2.0 (CR occupancy) = 229 Maintaining a margin of a factor of 3 for uncertainty associated with design input parameters and perfonnance of ESF components:
1 THE CONSERVATIVE MARGIN IN Tile LICENSING BASIS AN AINSIS = 229/3 = 76
==
Conclusions:==
' Die existing conservative margia of 76 in the LOCA licensing basis analyses can be utilized to compensate for an increase in the CR doses due to additional unfiltered in-leakage. The normal in leakage of 10 cfm
I Preliminary Information - for Illustration Only 7,
(used in most CR habitability analyses) can be increased to at least 760 cfm without exceeding the allowable 19 limits.
r A
Nuclear Energy Institute Project No. 689 cc:
Mr. Ralph Beedle Ms. Lynnette Hendricks, Director Senior Vice President Plant Support and Chief Nuclear Officer Nuclear Energy Institute Nuclear Energy Institute Suite 400 Suite 400 1776 l Street, NW 1776 i Street, NW Washington, DC 20006-3708 Washington, DC 20006-3708 4
Mr. Alex Marion, Director Mr. Charles B. Brinkman, Director Programs Washington Operations Nuclear Energy Institute ABB-Combustion Engineering, Inc.
Suite 400 12300 Twinbrook Parkway, Suite 330 1776 I Street, NW Rockville, Maryland 20852 Washington, DC 20006-3708 Mr. David Modeen, Director Mr. Robert R. Campbell. President Engineering Nuclear HVAC Utilities Group Nuclear Energy Institute Tennessee Valley Authority Suite 400 1101 Market Street, LP4J-C 4
1776 i Street, NW Chattanooga, TN 37402-2801 Washington, DC 20006-3708 Mr. Anthony Pietrangelo, Director Mr. Dennis Adams Licensing Nuclear HVAC Utilities Group Nuclear Energy Institute Comed Suite 400 1400 Opus Place 1776 l Street, NW Downers Grove, IL 60515 l
Washington, DC 20006-3708
)
Mr. Jim Davis, Director Operations i
Nuclear Energy Institute l
Suite 400 l
1776 i Street, NW Washington, DC 20006 ?708 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Activities Nuchar and Advanccd Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230