ML19340D521
| ML19340D521 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1979 |
| From: | Szukiewicz A Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| REF-GTECI-A-24, REF-GTECI-IS, TASK-A-24, TASK-OR NUREG-0588, NUREG-0588-FC, NUREG-588, NUREG-588-FC, NUDOCS 8012310024 | |
| Download: ML19340D521 (70) | |
Text
NUREG-0588 For Comment Interim Staff Position on Environmental Qualification of l Safety-Related Electrical Equipment i
Resolution of Generic Technical Activity A-24 Manuscript Completed: August 1979 Date Published: December 1979 1. J. Szukiewicz, Task Manager A
iDivision of Systems Safety
'Offico of Nuclear Reactor Regulation iU.S. Nuclear Regulatory Commission iW:shington, D.C. 20555 7"%,,
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ABSTRACT This report provides the NRC staff positions regarding selected areas of environmental qualification of safety-related electrical equipment in the resolution of Generic Technice.1 Activity A-24, " Qualification of Class 1E Safety-Related Equipment". The positions herein are applicable to plants that are or will be in the construction permit (CP) or operating license (OL) review process and that are required to satisfy the requirements set forth in either the 1971 or the 1974 version of IEEE Standard 323. These positions were developed prior to the Three Mile Island Unit 2 event. Any recommendations resulting from the staff's review of that event will be provided later. The seismic qualification requirements are addressed elsewere and are not included in the scope of this document.
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TABLE OF CONTENTS Page ABSTRACT...............................................................
iii ACKNOWLEDGMENT.........................................................
vii INTRODUCTION...........................................................
1 DISCUSSION.............................................................
2 1.
Establishment of the Qualification Parameters for Design Basis Event Conditions.....................................
5 1.1 Temperature and Pressure Conditions Inside Containment -
Loss-of-Coolant Accident (L0CA)..............................
5 1.2 Temperature and Pressure Conditiona Inside Containment -
Main Steam Line Break (MSLB).................................
6 1.3 Effects of Chemical Spray....................................
7 1.4 Radiation Conditior.s Inside and Outside Containment..........
'7 1.5 Environmental Conditions Outside Containment.................
10 2.
Qualification Methods.............................................
11 2.1 Selection of Methods.........................................
11 2.2 Qualificatisn by Test........................................
12 2.3 Test Sequence................................................
13 2.4 Other Qualification Methods..................................
14 14 3.
Margins...........................................................
15 4.
Aging.............................................................
5.
Qualification Documentation.......................................
16 APPENDIX A - METHODS FOR CALCULATING MASS AND ENERGY RELEASE...........
A-1 APPENDIX B - MODEL FOR ENVIRONMENTAL QUALIFICATION FOR LOSS-OF-COOLANT ACCIDENT AND MAIN STEAM LINE BREAK INSIDE PWR AND BWR DRY TYPES OF CONTAINMENT.........
B-1 APPENDIX C - QUALIFICATION PROFILES FOR BWR AND ICE CONDENSER C0NTAINMENTS....................................
C-1 APPENDIX D - SAMPLE CALCULATION AND TYPE METHODOLOGY FOR RADIATION QUALIFICATION D0SE..............................
D-1 APPENDIX E - STANDARD QUESTION ON ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT.......................
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5 ACKNOWLEDGMENT Many NRC individuals participated in the development of the positions on environmental qualification of safety-related electrical equipment presented herein. The contributions of the-following individuals were particularly helpful in developing this report and are acknowledged:
A. J. Szukiewicz J. A. Zwolinski F. M. Akstulewicz L. Soffer H. E. Krug R. M. Satterfield E. Butcher T. G. Dunning C. F. Miller i
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INTERIM STAFF POSITION ON ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT INTRODUCTION Equipment that is used to perform a necessary safety function must be capable of maintaining functional operability under all service conditions postulated to occur during the installed life for the time it is required to operate.
This requirement, which is embodied in D aral Design Criteria 1, 2, 4 and 23 of Appendix A and Sections III and XI of Appendix B to 10 CFR Part 50, is applicable to equipment located inside as well as outside containment. More detailed guidance related to the methods, procedures and guidelines for demon-strating this capability has been set forth in IEEE Std. 323 and ancillary daughter standards (e.g., IEEE Stds. 317, 334, 382, 383) and has been endorsed with supplementary material as noted in NRC Regulatory Guides.
As part of the operating license review for each plant, the staff evaluates the applicant's equipment qualification program by reviewing the qualification documentation on selected safety-related equipment. The objective of this review is to provide reasonable assurance that the equipment can perform its intended function in the most limiting environment in which it is expected to function.
The staff review of the documentation submitted by both equipment suppliers and license applicants indicate that some have developed generally acceptable qualification programs. The efforts of others, as compared with the " state of the art," need improvements. This is due in part to the fact that the qualifica-tion requirements contained in national standards and other guidance related to equipment qualification have been evolutionary in nature and subject to diverse interpretation.
To promote more orderly and systematic implementation of equipment qualification programs in industry and to provide guidance to be used by the NRC staff for use in the ongoing licensing reviews, the staff has developed a number of positions on selected areas of the qualification issue. These positions, which are presented in this report, provide guidance on the establishment of service conditions, methods for qualifying equipment, and other related matters.
They do not address in detail all areas of qualification, since certain areas are not yet well understood and are the subjects of research studies conducted by the NRC and by the industry. For example, the effects of aging, sequential versus simultaneous testing, including synergistic effects, and the potential combustible gas and chloride formation in equipment containing organic materials are being evaluated.
It is expected that these studies will lead to the development of more detailed guidance in the future, and.may require changes to these positions.
These positions were developed prior to the staff completion of the TMI-2 event evaluation, and any additional requirements or modifications to these-positions as a result of this evaluation will be identified later.
In addition, seismic qualification is being pursued on a case-by-case basis by the Seismic Qualification Review Team (SQRT) and is outside the scope of this document.
These positions are applicable only to plants that are or will be in the construction permit or operating license review process. These positions do not apply to operating plants. Operating plant licensees have been required by the NRC Office of Inspection and Enforcement to reassess the qualification of safety-related equipment used in those facilities (see IE Bulletin 79-01).
Licensee responses are to be evaluated using guidelines being developed specifically for that effort.
DISCUSSION As part of the staff reviews of operating license applications, a number of positions have been developed on the methods and procedures used to environ-mentally qualify safety-related electrical equipment. These positions, which are described in the following sections of this report, supplement the require-ments found in the 1971 and the 1974 version of IEEE Standard 323*.
While alternatives to these positions may be proposed, the positions will be used, together with the standards, as the basis for reviewing all license applications.
The positions are divided into two categories.
Category I positions apply to equipment qualified in compliance with IEEE Std. 323-1974 and Category II positions apply to equipment qualified in compliance with IEEE Std. 323-1971.
Section 1 of the the following table contains positions related to the establish-ment of the service conditions for areas inside and outside containment to which equipment should be qualified.
It includes guidancef for calculating the pressure and temperature conditions that result from a high energy line break (LOCA and/or MSLB), and also provides guidance for determining the chemical spray and the radiation environments expected to occur during a design basis event condition. Section 2 provides guidance on the selection of qualification methods (that is, testing, analysis, etc.) to be used for equipment located-inside and outside containment. Sections 3, 4, and 5 provide guidance on the selection of margins, aging and the preparation of qualification documentation.
The appendices supplement the positions and identify specific codes, sample calculations, and procedures that should be used when qualifying equipment.
The term " equipment" referred to in the following secti,ons applies to safety-related electrical equipment required for accident mitigation, post-incident monitoring, and safe shutdown.
It should be noted that, although the intent of these positions is to define criteria related to electrical equipment, it is necessary to recognize.and address equipment interfaces (e.g., mounting, seals, terminations) in the qualification process to which these positions apply. Also, qualification programs for specific equipment, such as cables, valves, motors, and electrical penetrations, that are designed to conform with the requirements of the daughter standards of IEEE Std. 323-1974 (as endorsed by the NRC Regulatory Guides) are acceptable for demonstrating compliance with the objectives of IEEE Std. 323.
The daughter standards include standards such as IEEE Std. 383 for cables, IEEE Std. 323-1974, "IEEE_ Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations."
IEEE Std. 323-1971, "LEEE Trial Use Standard: General Guide for Qualifying Class lE Equipment for Nuclear Power Generating Stations."
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IEEE Std. 382 for valves, IEEE Std. 334 for motors, and T.EEE Std. 317 for electrical penetrations. These standards are endorsed by Regulatory Guides 1.131, 1.73, 1.40, and 1.63 respectively.
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s INTERIM ST AFF POSITION ON ENVIRONMENTAL QUALIFICXTTdN~DI TAFETY3ILATED ELECTRICXElQUIPMENT CATEGORYI CATEGORY II Applicable to Equipment Qualified in Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance with IEEE Std. 323-19T1 1.
ESTABLIS}NEh'T OF THE QUALIFICATION 1.
ESTABLISINEAT OF TPE gbALIFICATION PARAMETERS FOR DESIGN BASIS EVENTS PARAMETERS FOR DESIGN BASIS EVENTS 1.1 Terperature and Pressure Conditions Inside 1.1 Temperature and Pressure Conditions Inside Contairment - Loss-of-Coolant Accident (LOCA)
Containment - Loss-of-Coolant Accident (LOCA)
(1) The time-dependent temperature and pressure, (1) Same as Category I.
established for the design of the containment structure and found acceptable by the staf f, may be used for_ environmental qualification of equipment.
(2) Acceptable methods for calculating and (2) Same as Category 1.
establishicg the containment p essure and temperature envelopes to which i
equipment should be qualified are j
sunnarized below. Acceptable methods for calculating mass and energy release i
rates are suanarized in Appendix A.
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Pressurized Water Reactors (Phis)
Pressurized Vater Reactors (FVRs)
Dry Containment - Calculate LOCA con-Dry Containment - Use the same containment tatraent environment using CONTEMPT-LT models as in Category I.
The assumption or equivalent industry codes.
of partial revaporization will be allowed.
f Additional guidance is provided in Other assumptions that reduce the temperature Standard Review Plan (SRP) response of the containment will be evaluated Section 6.2.1.1. A NLTdG-75/087.
on a case-1,f-case basis.
Ice Cendenser Contairment - Calculate Ice Condenser Containment - Same as LOCA containment environment using LOTIC Category I.
or equivalent industry codes.
Additional guidance is provided in SRP Section 6.2.1.1.B. NL7EG-75/087.
Boiling kater Reactors (BbTs)
Boiling Water Reactors (BVRs)
Mark I, 11 and III Containment -
Same as Category I.
Calculate LOCA environment using methods of GESSAR Appendix 3B or equivalent industry codes. Additional guidance is provided la SPJ Section 6.2.1.1.C, NUREG-75/OS7.
(3) In lieu of using the plant-specific (3) Same as Category I.
contairment temperature and pressure design profiles for PVR and ice condenser types of plants, the generic envelope shown in Appendix C may be used for qualification testing.
(4) The test profiles included in Appendix A (4) Same as Category I.
to IEEE Std. 323-1974 should not be con-sidered an acceptable alternative in lieu of using plant-specific contain-ment temperature and pressure design profiles unless plant-specific analysis is provided to verify the adequacy of those profiles.
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CATEGORYI CATEGORY II Applicable to Equipment Qualified in Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance with IEEE Std. 323-1971 1.2 Temperature and Pressure Conditions 1.2 Temperature and Pressure Conditions Inside inside Containment - Main $ team Line Containment - Main Steam Line Break _{MSLB)
Break (MSLB)
(1) Where qualification has not been con-(1) The environmental parameters used for pleted, the environmental parameters equipment qualification should be cal-used for equipment qualification should culated with a plant-specific model be calculated using a plant-specific reviewed and approved by the staff.
model based on the staff-approved assumptions di:. cussed in item 1 of Appendix B.
(2) Models that are acceptable for calculat-(2) Other models that are acceptable for ing containment parameters are listed calculating containment parameters in Section 1.1(2).
are listed in Section 1.l(2).
(3) In lieu of using the plant-specific (3) Same as Category I.
containment temperature and pressure design profiles for BWR and ice condenser plants, the generic envelepe shown in Appendix C may be used.
(4) The test profiles included in Appendix A (4) Same as Category 1.
to IEEE Std. 323-1974 should not be con-sidered an acceptable alternative in lieu of using plant-specific containment temperature and pressure design profiles unless plant-spec 2fic analysis is pro-vided to verify the adequacy of those profiles.
(S) Where qualification has been completed (3) Where qualification has been completed but only LOCA conditions were considered, but only LOCA conditions were considered, it must be demonstrated that the LOCA then it must be demonstrated that the qualification conditions exceed or are LOCA qualification conditions exceed or equivalent to the maximum calculated are equivalent to the maximum calculated MSLB conditions. The following tech-MSLB conditions. The following tech-nique is acceptable:
nique is acceptable:
(a) Calculate the peak temperature (a) Calculate the peak temperature from envelope from an MSLB using a an MSLB using a model based on the model based on the staff's staff's approved assumptions dis-approved assumptions defined in cussed in item I of Appendix B.
Section 1.l(2).
(b) Show that the peak surface (b) Same as Category I section 1.2(5)(b).
temperature of the component to be qualified does not exceed the LOCA qualification temperature by the method discussed in item 2 of Appendix B.
(c) If the calculated surface tempera-(c) If the calculated surface temperature ture exceeds the qualification exceeds the qualification temperature, temperature, the staff requires the staff requires that (i) additional that (1) requalification testing justification be provided to demonstrate be perfurmed with appropriate that the equipment can maintain its margins, or (ii) qualified physical required functional operability if its protection be provided to assure surface temperature reaches the calculated that the surface temperature will value or (ii) requalification testing be not exceed the actual qualifica-pe-formed with appropriate margins, tion temperature. For plants that or (iii) qualified physical protec-6
CATEGORYI CATEGORY 11 Applicable to Equipment Qualified in Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance wids IEEE Std. 323-1971 are currently being reviewed, or tion be provided to assure that the will be submitted for an operating surface temperature will not exceed license review within six months the actual qualificatior temperature.
f rom issue date of this report, compliance with items (i) or (ii) above may represent a sub-stantial impact. For those plants, the staff will consider additional internation submitted by the applicant to demonstrate that the equipment can maintain its functional operability if its surface temperature rises to the value calculated.
1.3 Effects of Chemical Spray 1.3 Effects of Chemical Spray The effects of caustic spray should be Same as Category I.
addressed for the equipment qualification.
The concentration of caustics used for qualification should be equivalent to or more severe than those used in the plant containment spray system. If the chemical composition of the caustic spray can be affected by equipment malfunctions, the most severe caustic spray envircament that results from a single failure in the spray system should be assumed. See SRP Section 6.5.2 (NUREG-75/087), paragraph II, ites (e) for caustic spray soluti.n guidelines.
1.4 Radiation Conditions Inside and Outside 1.4 Radiation Conditions Inside and Outside Containment Containment The radiation environment for qualification Same a s Cat ego ry I.
of equipment should be based on the normally expected radiation environment over the equipment qualified life, plus that asso-ciated with the most severe design basis accident (DBA) during or following which that equipment must remain functional. It should be assumed that the DBA related environmental conditions occur at the end of the equipment qualified life.
The sample calculations in Appendix D and the following positions provide an accept-able approach for establishing radiation limits for qualification. Additional radiation margins identified in Section 6.3.1.5 of IEEE Std. 323-1974 for qualification type testing are not required if these methods are used.
(1) The source term to be used in determining the radiation environment associated with the design basis LOCA should be taken as an instantaneous release from the fuel to the atmosphere of 100 percent of the noble gases, 50 percent of the iodines, and 1 percent of the remaining fission products. For all other non-LOCA design 7
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'ATEGORY I CATEGORY II Applicable to Equipment Qualified in Apphcable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance with IEEE Std. 323-1971 basis accident conditions, a source term involving an instantaneous release from the fuel to the atmosphere of 10 percent of the noble gases (except Kr-85 for which a release of 30 percent should be assumed) and 10 percent of the iodines is acceptable.
(2) The calculation of the radiation environment associated with design basis accidents should take into account the time-dependent transport of released fission products within various regions of containment and auxiliary s t ruct ures.
(3) The initial distribution of activity within the containment should be based en a mechanistically rational assumption. Hence, for compartmented containments, such as in a B'.1, a large portion of the source should be assumed to be initially contained in the drywell.
The assumption of uniform distribution of activity throughout the containment at time zero is not appropriate.
(4) Eff2 cts of ESF systems, such as containment sprays and containment ventilation and filtration systems, which act to remove airborne activity and redistribute activity within con-
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tainment, should be calculated using l
the same assumptions used in the cal-culation of offsite dose. See SRP Section 15.6.5 (NUREG-75/087) and the related sections referenced in the Appendices to that section.
(5) Natural deposition (i.e., plate-out) of airborne activity should be determined using a mechanistic model and best estimates for the model parameters.
The assumption of 50 percent instan-taneous plate-out of the iodine released from the core should not be made. Removal of iodine from surfaces by steam condensate flow or washoff by the containment spray may be assumed if such effects can be justi-fled and quantified by analysis or experiment.
(6) For unshielded equipment located in the containment, the gamma dose and dose rate should be equal to the dose and dose rate at the centerpoint of the I
containment plus the contribution from location dependent sources such as the sump water and plate-out, unless it can be shown by analyses that location and 8
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Applicable to Equipment Qualified in Applicable to Equipment Qualified in Accordance mdui IEEE Std. 323-1974 Accordance widh IEEE Std. 3 23-1971 shielding of the equipment reduces the dose and dose rate.
(7) For unshielded equipment, the beta doses at the surface of the equipment should be the sum of the airborne and plate-out sources. The airborne beta dose should be taken as the beta dose calculated for a point at the containment center.
(8) Shielded components need be qualified only to the gamma radiation levels required, provided an analysis or test shows that the sensitive portions of the component or equipment are not exposed to beta radiation or that the ef fects of beta radiation heating and ionization have no deleterious effects on component performance.
(9) Cables arranged in cable trays in the containment should be assumed to be exposed to half the beta radiation dose calculated for a point at the center of the containment plus the gamma ray dose calculated in accordance with Section 1.4(6).
This reduction in beta dose is allowed because of the localized shielding by other cables plus the cable tray itself.
(10) Paints and coatings should be assumed to be exposed to both beta and gamma rays in assessing their resistance to radiation. Plate-out activity should be assumed to remain on the equipment surface unless the effects of the removal mechanisms, such as spray wash-f off or steam condensate flow, can be justified and quantified by analysis or experiment.
(11) Components of the emergency core cool-Ing system (ECCS) located outside con-tainment (e.g., pumps, valves, seals and electrical equipment) should be qualified to withstand the radiation equivalent to that penetrating the con-tainment, plus the exposure from the sump fluid using assumptions consistent with the requirements stated in Appendix K to 10 CFR Part 50.
(12) Equipment that may be exposed to radia-4 tion doses below 10 rads should not be considered to be exempt from radiation qualification, unless analysis supported by test data is provided to verify that these levels will not degrade the operability of the equipment below acceptable values.
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CATEGORY!
CATEGORY II Applicable to Equipment Quahfied in Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance with IEEE Std. 323-1971 (13) The staff will accept a given component to be qualified provided it can be shown that the component has been qualified to integrated beta and gamma doses which are equal to or higher than those levels resulting f rom ar 4. lysis similar in nature and scope that included in Appendix D (wht.
Laer the source term given in item (1) abos=),
and that the componen* incorporatt.
I apprepriate factors pertinent to the plant desirn and operating character-istics, as given in these general guidelines.
(14) When a conservative analysis has not been provided by the applicant for staff review, the staff wil! use the radiation environment guidelines contained in Appendix D, suitably corrected for the differences in reactos power level, type, containment size, and other appropriate factors.
1.5 Environmental Conditions for Outside Containment 1.5 Environmental Conditions for Outside Containment (1) Equipment located outside contain-ment that could be subjected to high-(1) Equipment located outside containment energy pipe breaks should be qualified that could be subjected to high-energy to the conditions resulting from the pipe breaks -hould be qualified to the accident for the duration required.
e nditions t ting from the accident The techniques to calculate the f r the duration required. The tech-i environmental parameters described niques to calculate the environmental in Sections 1.1 through 1.4 above Parameters described in Sections 1.1 l
should be applied.
through 1.4 (Category II) above should be applied.
(2) Equipment located in general plant areas (2) Same as Category I.
outside containment where equipment is not subjected to a design basis accident environment should be qualified to the normal and abnormal range of environ-mental conditions postulated to occur at the equipment location.
(3) Equipment not served by Class lE environ-(3) Same as Category I; or, there may be mental support systems, or ser'ad by designs where a loss of the environ-Class lE support systems tha ~ be mental support system may expose some se.ured during plant oper tion or shut-equipment to environments that exceed down, should be qualified to the limiting the qualified limits. For these designs, environmrntal conditions that are postulated appropriate monitoring devices should be for that location, assuming a loss of the provided to alert the operator that environmental support system.
abnormal conditions exist and to permit an assessment of the conditions that occurred in order to determine if cor-rective action, such as replacing any affected equipment, is warranted.
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CATEGORY I CATEGORYII Applicable to Equipment Qualified in Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance with IEEE Std. 323-1971 2.
QUALIFICATION METHODS 2.
QUALIFICATION METHODS 2.1 Selection of Methods 2.1 Selection of Methods (1) Qualification methods should conform (1) Qualification methods should conform to the requirements defined in IEEE to the requirements defined in IEEE Std. 323-1974.
Std. 323-1971.
(2) The cbtice of the methods selected (2) Same as Category 1.
1s largely a matter of technical judgment and availability of infor-mation that supports the conclusions reached. Experience has shown that qualification of equipment subjected to an accident environment without test data is not adequate to demonstrate functional operability. In general, the staff will not accept analysis in l
lieu of test data unless (a) testing i
of the component is impractical due to size limitations, and (b) partial I
I type test data is provided to support the analytical assumptions I
and conclusions reached.
l (3) The environmental qualification of (3) Same as Category 1.
l equipment exposed to DBA environ-ments should conform to the following positions. The bases should be provided for the time interval required for operability of this equipment. The operability and failure criteria should be specified and the safety margins defined.
(a) Equipment that must function in order to mitigate any acci-dent should be qualified by test to demonstrate its operabil-ity for the time required in the environmental conditions resulting from that accident.
(b) Any equipment (safety-related or non-safety-related) that need not function in order to mitigate any accident, but that must not fail in a manne r detrimental to plant safety should be qualified by test to demonstrate its capability to with-stand any accident environment for the time during which it must not fail.
(c) Equipment that need not function in order to mitigate any accident and whose failure in any mode in any accident environment is not detrimental to plant safety need only be qualified for its non-accident service environment.
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CATEGORY I CATEGORY II Applicable to Equipment Quahfied in Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance udth IEEE Std. 3 23-1971 Although actual type testing is preferred, other methods when justified may be found acceptable.
The bases should be provided for concluding that such equipment is not required to f unction in order to mitigate any accident, and that its failure in any mode in any accident environment is not detri-mental to plant safety.
(4) For environmental qualification of (4) Same as Category I.
equipment subject to events other than a CBA, which result in abnormal environmental conditions, actual type testing is prefer:ed. However, analysis or operating history, or any applicable combination thereof, coupled with partial type test data may be f ound acceptable, subject to the applicability and detail of information provided.
2.2 Qualificatico by Test 2.2 Qualification by Test (1) The failure criteria should be established (1) Same as Category I.
prior to testing.
(2) Test results should demonstrate that (2) Same as Category I.
the equipment can perform its required function for all service conditions postulated (with margin) during its installed life (3) The items described in Section 6.3 of (3) The items described in Section 5.2 of IEEE Std. 323-1974 supplemented by lEEE Std. 323-1971 supplemented by itent items (4) through (12) below constitute (4) through (12) below constitute acceptable guidelines for establishing acceptable guidelines for establishing test procedures.
test procedures.
(4) When establishing the simulated (4) Same as Category I.
environmental profile for qualifying equipment located inside containment, it is preferred that a single profile be used that envelopes the environmental conditions resulting from any design basis event during any mode of plant operation (e.g., a profile that envelopes the conditions produced by the main steamline break and loss-of-coolant accidents).
(5) Equipment should be located above flood (5) Same as Category I.
level or protected against submergence by locating the equipment in qualified watertight enclosures. Where equipment is located in watertight enclosures, qualification by test or analysis should be used to demonstrate the adequacy of such protection. Where equipment could be submerged, it should be identified and demonstrated to be qualified by test for the duration required.
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CATEGORYI CATEGORY 11 Applicable to Equipment Qualified in Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance with IEEE Std. 323+1971 (6) The temperature to which equipment is (6) Same as Category I.
If there were qualified, when exposed to the simulated no thermocouples located near the accident environment, should be defined equipment during the tests, heat by thermocouple readings on or as close transfer analysis should be used as practical to the surface of the com-to determine the temperature at the ponent being qualified.
component. (Acceptable heat transfer analysis methods are provided in Appendix B.)
(7) pe rformance characteristics of (7) Same as Category I.
equipment should be verified before, after, and periodically during testing throughout its range of required operability.
(8) Caustic spray should be incorporated (8) Same as Category I.
during simulated event testing at the maximum pressure and at the temperature conditions that would occur when the onsite spray systems actuate.
(9) The operability status of equipment (9) Same as Category I.
should be monitored continuously during testing. For long-term testing, how-ever, monitoring at discrete intervals should be justified if used.
(10) Expected extremes in power supply (10) Same as Category I.
voltage range and frequency should be applied during simulated event environ-mental testing.
(11) Dust environments should be addressed (11) Same as Category I.
when establishing qualification service conditions.
(12) Cobalt-60 is an acceptable gamma radia-(12) Same as Category I.
tion source for environmental qu.lification.
2.3 Test Sequence 2.3 Test Sequence 2
(1) The test sequence should conform fully (1) Justification of the adequacy of the to the guidelines established in test sequence selected should be Section 6.3.2 of IEEE Std. 323-1974, provided.
The test procedures should insure that the same piece of equipment is used throu hout the test sequence, and that the test simulates as closely as prac-ticable the postulated accident environment.
(2) The test should simulate as closely as practicable the postulated environment.
(3) The test procedures should conform to the guidelines described in Section 5 of IEEE Std. 323-1971.
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i CATECORY I CATEGORY!!
Appheable to Equipcient Qualified in Appbcable to Equilscent Quahfied in Accordance with IEEE Std. 323-1374 Accordance with IEEE Std. 323-1971 i
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(4) The staff considers that, for vital electrical equipment such as penetrations, cotnectors, cables, valves and motors, and transmitters located icaide containment or exposed to hostile steam envirennents outside containment, separate ef f ects I
testing for the most part is not an accept-able qualification cethod. The testir.g 1
of such equipment should be ccaducted in a r.anner tnat subjects the same piece of equiptent to radiatien and the hestile steam enviren ent sequent ially.
I 2.4 Other Qualification Methods 2.4 Other Qualification Metheds Qualification by analysts or cperating Sase as Category 1 (except that IEEE experience implemented, as described in Std. 323-1971 and ancillary standards I
IEEE Std. 323-1974 and other ancillary endorsed at the time the CP SER was issued standards, may be fcund acceptable. The may be used).
adegancy of these methods will be evalu-i ated on the basis of the quality and detail of the information subettted in support of the assumptiens made and the specific func-3 tien and location of the equipoent. These methods are ecst suitable for equipment where testirig is precluded by physical size of the equipment being qualified. It is required 1
that, shen these methods are employed, some partial type tests on vital components of the equipment be prcvided in suppcrt of these methods.
3.
MAEGINS 3.
MAFGINS
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(1) Quantified margins should be applied to (1) Same as Category I.
the design parameters discussed in i
Section I to assure that the postulated I
accident conditions have been enveloted l
daricg testicg. These margins should be applied in addition to any margins (con-I servatism) applied during the deriva-tzen of the specified plant parameters.
(2) In lieu of other proposed sargins that (2) The margins provided in the design will may be fcund acceptable, the suggested be evaluated on a case-by-ca_c basis.
values indicated in IEEE Std. 323-1974, Factors that should be considered in Section 6.3.1.5, should be used as a guide.
quantifying margins are (a) the environ-(Nete exceptiens stated in Section 1.4.)
mental stress levels induced duricg test-l ing, (b) the durat2cn of the stress, (c) the number of itans tested and the i
number of tests performed in the hostile environment, (d) the performance character-istics of the equipment while subjected to the envirennental stresses, and (e) the specified function of the equipment.
(3) When the qualification envelope in (3) Same as Category I.
Appendix C is used, the only required margins are those accounting for the inaccuracies in the test equipment.
Saf ficient conservatism has already been included to account for uncer-L 14
CATEGORYI CA"2 GORY 11 Applicable to Equipment Quahfied in Applicable to Equipment Quahtied in Accordance with IEEE Std. 323-1974 Accordance with IEEE Std. 323-1971 tainties such as production errors and errors associated with defining satisf actory perfo rmance (e.g., when only a small number of units are tested).
(4) Some equipment may be required by the (4) Same as Category 1.
design to only perform its safety func-tion within a short time period into the event (i.e., within seconds or minutes), and, once its function is complete, subsequent failures are shown not to be detrimental to plant safety.
Other equipment may not be required to perform a safety function but must not fail within a short time period into the event, and subsequent failures are also shown not to be det rirent al to plant safety. Equipment in tuese categories is required to remain functional in the accident environment for a period of at least 1 bour in excess of the time assumed in the accident analysis. For all other equipment (e.g., post-accident monitoring, recombiners, etc.), the 10 percent time margin identified in Section 6.3.1.5 of IEEE Std. 323-1974 may be used.
4.
AGING 4.
AGING (I) Aging effects on all equipment, regard-(1) Qualification programs that are com-less of its location in the plant, mitted to conform to the requirements should be considered and included in of IEEE Std. 382-1972 (for valve the qualification program.
operators) and IEEE Std. 334-1971 (for motors) should consider the effects of aging. For this equipment, the Category I positions of Section 4 are applicable.
(2) The degrading influences discussed (2) For other equipment, the qualification in Sections 6.3.3, 6.3.4 and 6.3.5 programs should address aging only of IEEE Std. 323-1974 and the electri-to the extent that equipment that cal and cechanical stresses associated is composed, in part, of materials with cyclic operation of equipment susceptible to aging effects should should be considered and included as be identified, and a schedule for part of the aging programs.
periodically replacing the equipment and/or materials should be established.
During individual case reviews, the staff will require that the ef fects of aging be accounted for on selected equipment if operating experience or testing indicates that the equipment may exhibit deleterious aging mechanisms.
(3) Synergistic effects should be considered in the accelerated aging programs.
Investigation should be performed to assure that nc known synergistic ef fects have been identified on materials that are included in the equipment being I
15 j
JATEGOIY
- C ATEGC'AY !!
App;2 d:e to Escp: rent Q.:ah'rd =
A;;2.ath 1: Eq=; rect @c" +-d =
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r::arams. iefer te G M 3 -;2't 15NC 75-C'H) ami U E '3-2; (58JC ~!-!.52), "Q221tf.catt:: Test-1:4 Ivalune: 0 arterly Eerc ris,'
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an acceptatie mett:d c f a f fres sitt at:elerated agiq.
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be c: s dered 2 the agt:g rf electrical j
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- te palified lif e of the ew2;eest l
f f
{st-fler c: v ert as applicatle) and j
the tas:s f:r its selectac sh:211 te deft:cf.
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(10) ? alif ted life stnl* te estathshed i
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the c;eratx4 itstery. 2:* ---'er j
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(;) !ie staf f enfarses tie regirenetts
'll S ame a s Ca te pry I-4 stated ic IIE 5t2. 3 3-!S74 that. "Tre qsalif a catice de-nsnert ati:: stall verify that eact type cf electrical epipn+:t I
is waltf ref for its a;;hcatt:c as!
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I CATEGORYl CATEGORYTI Applicable to Equipment Qualified in Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1974 Accordance with IEEE Std. 323-1971 requirements. The basis of qualificar'on shall be explained to show the relat.
ship of all facets of proof needed t support adequacy of the complete equip-ment. Data used to demonstrate the qualification of the equipment shall be pertinent to the application and organized in an auditable form."
(2) The guidelines for documentation in (2) Same as Category 1, except the guide-IEEE Std. 323-1974 when fully imple-lines of IEEE Std. 323-1971 may be used.
mented are acceptable. The documenta-tion should include sufficient informa-tion to address the required information I
identified in Appendix E.
A certificate of conformance by itself is not acceptabic l
unless it is accompanied by test data and information on the qualification program, i
l l
l s
i I
d 6
17
APPEhTIX A METHODS FOR CALCULATING MASS AND ENERGY RELEASE k
APPENDIX A 4
METHODS FOR CALCULATING MASS AND ENERGY RELEASE Acceptable methods for calculating the mass and energy release to determine the loss-of-coolant accident (LOCA) environment for PWR and BWR plants are described in the following:
(1) Tct. cal Report WCAP-8312A for Westinghouse plants.
(2) Section 6.2.1 of CESSAR System 80 PSAR for Combustion Engineering plants.
(3) Appendix 6A of B-SAF-205 for Babcock & Wilcox plants.
(4) NEDO-10320 and Supplements 1 & 2 for General Electric plants.
t Acceptable methods for calculating the mass and energy release to determine the main steam line break (MSLB) environment are described in the following:
(1) Appendix 6B of CESSAR System 80 PSAR for Combustion Engineering plants.
(2) Section 15.1.14 of B-SAR-205 for Babcock & Wilcox plants.
(3) Same as item (4) above for General Electric plants.
(4) Topical Report WCAP-8822 for Westinghouse plants.
(Although this Topical Report is currently under review, the use of this method is acceptiole in the interim if 'o entrainment is assumed. Reanalysis may be required following i e NRC staff review of the entrainment model as presently describ
.)
A-1 e
w._..
I APPEhTIX B i
MODEL FOR ENVIRONMENTAL QUALIFICATION FOR LOSS-OF-COOLANT ACCIDENT AND MAIN STEAM LINE BREAK INSIDE PkR AND BWR DRY TYPE OF CONTAINMENT l
APPENDIX B l
MODEL FOR ENVIRONMENTAL QUALIFICATION FOR LOSS-OF-C001 ANT ACCIDENT AND MAIN STEAM LINE BREAK INSIDE PW AND BWR DRY TYPE OF CONTA1hENT
- 1.. Methodology to Determine the Containment Environmental Response a.
Heat Transfer Coefficient for heat transfer cuafficient tu the heat sinks, the Tagami con-densing heat transfer correlation should be used for a LOCA with the maximum heat transfer rate determined at the time of peak pressure or the end of primary system blowdown. A rapid transition to a natural convection, condensing heat transfer correlation should follow. The Uchida heat transfer correlation should be used for MSLB accidents while in the condensing mode. A natural convection heat transfer coefficient should be used at all other times when not in the condensing heat tranfer mode for both LOCAs and MSLB accidents.
The applicatica of these correlations should be as follows:
(1) Condensing heat transfer q/A = h
- (s~
w}
cond where q/A = the surface heat flux h
the condensing heat transfer coefficient
=
c g,3, T, = the steam saturation (dew point) temperature T =
surf ace temperature of the heat sink y
(2) Convective heat transfer q/A = h (v~ w}
e where h
= c nvective heat transfer coefficient e
the bulk vapor temperature T
=
y All other parameters are the same as for the condensing mode.
b.
Heat Sink Condensate Treatment When the containment atmosphere is at or below the saturation tempeca-ture, all condensate formed on the heat sinks should be cansferre<1 directly to the sump. When the atmosphere is superheated, a maximum of 8 percent of the condensate may be assumed to remain in the vapor region. The condensed mass should be calculated as follows:
Mcond = X q / (h -h) y t
B-1
l where M
= mass e ndensation rate cond
{
X = mass condenaation fraction (0.92) q = surface heat transfer rate h = enthalphy of the superheated steam y
h = enthalphy of the liquid condensate entering the sump region (i.e., average enthalpy of the heat sink condensate boundary layer) c.
Heat Sink - Surface Area s
The surface area of the heat sinks should correspond to that used for the containment design pressure evaluation.
d.
Single Active Failure Evtluation i
Single active teilures should be evaluated for those containment '
safety systems and components relied upon to limit the containment i
temperature / pressure respoi a to a LOCA or MSLB accident. This i
evaluation should include, out not necessarily be limited to, the loss or availability of offsite power (whichever is worse), diesel-generator failure when loss of offsite power is evaluated, and loss of containment heat removal systems (either partial or total, whichever is worse).
i e.
Containment Heat Removal System Actuation i
The time determined at which active containment heat removal systems become effective should include consideration of actuation sensors and setpoints, actuation delay time, and s'istem delay time (i.e., time i
required to come into operation).
i f.
Identification of Most Severe Environment The worst case for environmental qualification should be selected considering time duration at elevated temperatures as well as the maximum temperature.
In particular, consider the spettrum of break sizes analyzed and single failures evaluated.
2.
Acceptable Methodology for Safety-Related Component Thermal Analysis l
Component thermal analyses may be performed to justify environmental qualification test conditions that are found to be less than those calculated during the containment environmental response calculation.
The heat transfer rate to component should be calculated as follows:
a.
Condensing Heat Transfer Rate g/A = h
- (T - T) cond s
w I
ew>,,
,m.
m, nn.~
w - m. a w e..e nen-n
--~w-,c-v+<a-.
where q/A = component surface heat flux hcond = condensing heat transfer coefficient is equal to the larger of 4x Tagami correlation or 4x Uchida correlation T = saturation temperature (dew point) 3 T = component surface temperature b.
Convective Heat Transfer A convective heat transfer coefficient should be used when the condensing heat flux is calculated to be less than the convective heat flux. During the blowdown period, a forced convection heat transfer correlation should be used. For example:
\\
NU = C (Re)"
where Nu = Nusselt number Re = Reynolds number C,n = empirical constants dependent on geometry and Reynolds number The velocity used in the evaluation of Reynolds number may be determined as follows:
V = 25 " D CON 7 where V = velocity in ft/sec M
= the blowdown rate in lbs/hr 2
BD V
C "
I""*"' " "** '" '
CONT After the blowdown has ceased or reduced to a negligibly low value, a natural convection heat transfer correlation is acceptable.
However, use of a natural convection heat transfer coefficient must be fully justified whenever used.
B-3 m
APPENDIX C QUALIFICATION PROFILES FOR BWR AND ICE CONDENSER CONTAINMENTS
Legend:
For IEEE Std. 3231974 (all Category I plants), use the double " spike" as shown.
BWR Containments
~
==== fce Condenser Containments 3231971 {all Category 11 plants), ti s envelope starts at T.
For IEEE Std.
o T
T T
T
?
?
?
?
I 10 sec
? 'ir 6 hr 24 hr l
l 3%
l 340
\\
\\
\\
k 300 t
y j
p Decay rate can im varied between the 3
required end points.
o e
\\
h g
250 N
No staf f N N requirement g
for rise time Qualification pressure N
200 or decay rate should be equal to or N \\
for the first greater than the design 16 t
transient-pressure or equal to the Dwell at peak is l
saturation pressure,1t the g
ed to &
l-temperatures indicated.
Rate of decrease 150
' nd duration af ter 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -
a should be identified and Not justified on a case-by-case
~
~
basis.
g gf;g T
Time (as indicated) o Figure C-1. Qualification Profiles for BWR and Ice Condenser Containments.
l l
t i
i APPENDIX D SAMPLE CALCULATION AND TYPE METIIODOLOGY FOR RADIATION QUALIFICATION DOSE k
f APPENDIX D SAMPLE CALCULATION AND TYPE METHODOLOGY FOR RADIATION QUALIFICATION DOSE
{
This appendix illustrates the proposed staff model for calculating dose rates and integrated doses for equipment qualification purposes. The example doses shown below include contributions from several dose point locations in the
[
containment and cover a period of only thirty days following the postulated fission product release. The values shown are not intended for use as appropriate equipment qualification levels. The dose levels intended for l
qualification purposes should be determined using the maximum time the equipment is intended to function which, for the design basis LOCA event, may well exceed thirty days.
The beta and gamma integrated doses presented in the tables below have'been estimated using models and assumptions consistent with those of Regulatory Guides 1.7 and 1.89.
This analysis is conservative, but it does not ignore important time-dependent phenomena related to the action of engineered safety features (ESFs) and natural phenomena, such as plate-out, as done in previous staff analyses.
Doses were calculated for points within the containment atmosphere, at the containment surface (taking sprays and plate-out mechanisms into account), and near the sump water.
11t CTY-DAY INTEGRATED DOSES Integrated Dose (Rad)
Location Beta Gamma 7
Containment Atmosphere 1.4 x 108 1.5 x 10 Containment Surface 1.1 x 107 9.1 x 106 Near Sump Water 7.2 x 10 4.4 x 106 7
l l
1.
General Summary of the LOCA Scenario l
The accident considered in this report for determining the radiation environ-ment for qualification of safety-related equipment is a design basis LOCA.
l The following is a description of the events that are postulated to occur. At the time t=0, the pipe break occurs and results in rapid blowdown of the l
reactor coolant system (RCS). The blowdown of the RCS ends approximately 20 l
to 40 seconds after the break. Flashing and escape of the coolant during
(
blowdown remtves heat rapidly from the primary system and causes the fuel rod l
cladding temperature to drop.
Consequently, only a few fuel rods are expected l
to fail during the blowdown period.
l l
Following the end of blowdown, the fuel rods are uncovered and the stored heat in the fuel and the decay heat are transferred to the cladding, thus raising the cladding temperature. Some fuel rods may experience cladding failure during this period. The ECCS refills the lower reactor vessel and then l
D-1 t
refloods the core region within 100 to 300 seconds, causing cladding temperature turnaround. During the initial blowdown, only the radioactive naterial contained in the coolant from steady-state operation would be released to the containment. During reflood/ refill when fuel rod cladding failure may occur, the noble gases would be transported out of the primary system by steam flow and would become airborne within the primary containment of a PWR (or within the drywell of a BWR).
Some fraction of the iodines and less volatile fission products that are released as a result of fuel rod failure would also be transported out of the primary system by the steam flow and become airborne, and some fraction would remain in solution in the sump water or would be deposited on surfaces within the primary system. The amount that becomes airborne outside the primary system would be strongly dependent on the time of fuel rod fatlure and the transport phenomenon for each species within the primary system.
Following the release from the primary system, the fission prcducts would be distributed within the containment. For a PWR containment, the released airborne activity would rapidly disperse and become uniformly distributed l
within the primary containment. For a BWR, the released activity would be airborne within the drywell. Following initial release to the containment atmosphere, the action of natural convection currents and ESF equipment, such as cooling fans, will cause time-dependent redistribution of the activity within the containment. Natural removal processes, such as deposition on containment surfaces and washout from the containment atmosphere by the containment spray systems, would reduce the airbarne activity concentration and would redistribute this activity to the containment surfaces and to the containment sump water.
1 During the same period of time, leakage of radiocctivity from the containment to the atmosphere could take place. This would te processed to some extent, j
by ESF filters if present, causing a buildup of activity on these filters.
In l
addition, there could be some deposition and platecut of radioactivity (iodine and daughters of noble gases) on surfaces of ductwork or on the walls of secondary containment.
During the longer term, contaminated primary coolant could be circulated through pipes outside of containment (PWR residual heat removal model). The staf f usually assumes a failure of a seal in the ECCS equipment, such that significant quantities of coolant could leak into compartments outside of containment. The leaked fluid is either retained in a sealed room or transported to the radwaste system.
Some portion of this leaked fluid is volatilized and also transported in the air of these compartments. These sources would be processed to some extent by ESF filtcrs.
2.
Basic Assumptions Used in the Analysis Gamma and beta doses and dose rates were determined for three types of radioactive source distributions:
isotopes suspended in the containment atmosphere, plated-out on containment surfaces, or mixed in the containment sump water. Thus, a given piece of equipment may receive a dose contribution from any or all of these sources. The amount of dose contributed by each of these sources is determined by the location of the equipment, the time-dependent and location-dependent distribution of the source, and effects of shielding.
D-2
Previous guidance issued by the staff regarding the source term for equipment qualification was general in nature. Recognizing tt.at implementation of that guidance required a number of assumptions to be made regarding the time-dependent behavior of material within and outside of containment, the staff, in this report, has performed an analysis of the radiation environment tnat is associated with the source term of position C.2 of Regulation Guide 1.89, using assumptions and methods which were intended to be consistent with staff practices in analyzing the radiological consequences of a design basis LOCA.
Position C.2 of Regulatory Guide 1.89 assumes a source term condition associated with a core meltdown. To get a feel for the degree of conservatism in this assumption, calculations using the RELAP-EM (Evaluation Model) program, which uses the conservative assumptions given in Appendix K to 10 CFR Part 50, predict that the peak cladding temperature attained by the hottest fuel rod will be less than 2200 F.
Based on the predicted distribu ion of cladding temperature throughout the core, it is estimated that between 20 and 80 percent of the fuel rods could experience cladding failure for a PWR with a lesser fraction for a BWR.
Calculations performed using the 7 ore realistic RELAP-BE (Eest Estimate) program predicted much lower cladding temperatures than RELAP-EM. Based on the RELAP-BE predictions, the number of fuel rod cladding failures is estimated to be less than 10 percent.
A Sandia Laboratories report (SAND 76-0740, " Radiation Signature Follcwing the Hypothesized LOCA") also analyzed the radiation environment associated with the conditions of position C.2 of Regulatory Guide 1.89.
But as noted in the text of that report (ct. Table 1.1, for example), those analyses are based upon calculational assumptions that are n<c consistent (are overly conservative) with respect to staff recommended practices. Therefore, the results in that report should not be directly applied.
Table D-1 compares the source terms of position C.2 of Regulatory Guide 1.89 to source terms used for other design basis events.
3.
Analysis of the Concentration of Fission Products in Air 4
This section discusses the physical model used to simulate the PWR containment and to determine the time-dependent and location-dependent distribution of noble gases and iodines airborne within the containment atmosphere and plated-out on containment surfaces.
The staff has developed a computer program (TACT) to be published that is used to model the time-dependent behavior of iodine and noble gases within a nuclear power plant. The TACT code is used routinely by the staff for the calculation of the offsite radiological consequences of a LOCA, and is an acceptable method for modeling the transfer of activity from one containment region to another and in modeling the reduction of activity due to the action of ESFs.
Another staff code, SPIRT (Ref. 1), is used to estimate the removal rates of elemental iodine by plate-out and sprays, and is a needed input to TACT.
These codes were used to develop the source te nn estimates.
The source terms used in the analysis assumed that 50 percent of the core iodines and 100 percent of the core noble gases were released instantaneously to the containment atmosphere. The following assumptions were also used to calculate the distribution of radioactivity within the containment:
D-3
a.
The representative containment free volume was taken as 2.52 x 106 ft. 3 Of this volume, 74 percent or 1.86 x 10 ft3 6
is assumed to be directly covered by the containment sprays.
b.
6.6 x 105 ft3 of the containment free volume is assumed unsprayed, which includes regions within the main containment room near the containment dome and compartments below the cperating floor level.
Good mixing of the containment activity between the sprayed and unsprayed regions is assured by natural convection currents and ESF fans.
The ESF fans are assumed to have a design flow rate of 220,000 cfm c.
in the post-LOCA environment.
Since mixing between all major un-sprayed regions and compartments and the main sprayed region will occur, the containment was modeled with TACT nodes.
d.
Air exchange between the sprayed and uns, ayed region was taken as one-half of the design flow rate of ESF fans plus the effect of natural convection.
The containment spray system was assumed to have two equal capacity e.
trains, each designed to inject 3000 gpm of boric acid solution into the containment.
f.
Trace levels of hydrazine was assumed added to enhance the removal of iodine.
g.
The spray removal rate constant (lambda) was calculated using the staff's SPIRT program, conservatively assuming only one spray train operation and an elemental iodine instantaneous partition coefficient (H) of 5000. The calculated value of the elemental iodine spray removal constant was 27.2 hr 1, which represents an elemental iodine asidence half-life in the sprayed region of approximately 1.5 minutes.
h.
Plate-out of iodine on containment internal surfaces was modeled as a first-order rate removal process and best estimates for model param-eters were assumed.
Based on an assumed total surface area within 5 ft2, the calculated value for containment of approximately 5.0 x 10 the overall elemental iodine plate-out constant was 1.23 hr 1 i.
The spray removal and plate-out process were modeled as competing iod'ne removal mechanisms.
4.
Departure from Past Practices i
Computing the radiological consequences at the exclusion radius and the low population zone, the staff usually assumed that an instantaneous release of f
100 percent of the noble gases and 25 percent of the core iodines is available j
for leakage from the containment. Recognizing that it would take some time before a release of this magnitude could occur, even assuming degraded emergency core cooling system (ECCS) operation, the staff has also assumed, for purposes of estimating offsite dose consequences, that the source is uniformly distributed and that containment sprays activate at the time the large source is available for release (both of which would also take time to D-4
j 1
occur). Also implicit in the 25 percent release of iodines was the assumption i
that 50 percent of a 50 percent release of iodine from the fuel is plated-out in a very short period of time.
l The staff usually limits credit for element iodine spray removal to no more l
than 10 hr 1, for an assumed release of 25 percent of the balogens to compensate j
for the articial assumption of instantaneous plate-out.
If a release of 50 percent were assumed (as is implied by Regulatory Guide 1.7 and TID-14844),
i the actual conservatively calculated spray lambdas would be appropriate.
In any event, removal of elemental iodine from the conta.nment atmosphere by i
spray and plate-out is assumed to cease when the concentration in the sprayed region is reduced by a factor of 200 (when the initial concentration of iodine in the containment is calculated assuming 50 percent of the core i
l inventory of iodines is initially airborne). This reduction factor is obtained by doubling the reduction factor used in the LOCA dose analysis. The intent is to achieve an equilibrium airborne iodine concentration that is consistent with the LOCA analysis. Since the initial (t=0) concentration is assumed to be twice that of the LOCA analysis (50 versus 25 percent), the reduction factor has been doubled.
The staff assumes that more than one species of iodine is present, or will be formed, in a design basis LOCA (see Regulatory Guides 1.3 and 1.4).
For our analysis, it is assumed that 2.5 percent of the core inventory of iodine released is associated with airborne particulate material and 2 percent of the core inventory of iodine released forms organic compounds. Even though these values would not be obtained until several hours after the LOCA, it is the staff assumption that the aforementioned composition is present at t=0.
A removal rate constant for particulate iodine was calculated to be 0.43 hr 1 The organic iodine concentration in the containment atmosphere was assumed to be unaffected by containment sprays or plate-out. The action of sprays would not commence at t=0 (e.g., some time would elapse between the onset of the LOCA and the delivery of spray solution to the spray nozzles). Similarly, the assumed large source would not be immediately released from the fuel, and some time would pass before any airborne iodine would be distributed throughout containment.
The assumption of a large release, uniformly distributed in containment (or in the sump water as will be discussed later) is a convenient simplification for I
purpose of the dose assessment in a PWR containment, and is conservative in terms of specifying the time-dependent radiation environment. Accurate coupling of the various time sequences is beyond the scope of this analysis.
l The calculated values of noble gas and airborne iodine activity in the con-tainment as a function of time following the LOCA are presented in Table D-2.
5.
Analysis of the Concentration of Fission Products on Surface The air dose model assumed that only one spray train and one ventilation system train were operable.
If both trains of both systems were operable, spray washout would progress more rapidly in the sprayed regions and the
" equilibrium" of concentrations between sprayed and c, sprayed regions would be reached more quickly. The result would be lower dose rates due to plate-out D-5
a activity on surfaces or suspended in the air in sprayed regions, and in unsprayed regions during the early phases of the accident.
It has been suggested that the plate-out source used in estimating the radiation environment should assume that 50 percent of the released elemental iodine is instantaneously plated-out on containment and equipment surfaces.
This assumption is inconsistent with the time-dependent model used to characterize the concentration ot' iodines in the air.
It is the staff's view that the estimates should be mechanistically consistent. A large margin of 3
conservatism already exists by virtue of the assumed large source term.
In any event, the subsequent removal of deposited material by washoff (by sprays or condensate flow) may be important.
Ignoring this factor (as was done for this short-term effort) introduces conservatism.
Current staff guidelines do not include an acceptable method for estimating this effect.
In the absence f
of such methods, it has been assumed that all plated-out material is retained by the containment surfaces. Table D-3 gives the values calculated for the iodine activity buildup on the plate-out surfaces of the containment.
6.
Analysis of the Concentration of Fission Products in the Sump Regulatory Guide 1.7 (Table D-1) recommends that 50 percent of the iodines and 1 percent of the remaining fission products present in the core are assumed to be intimately mixed with the coolant water. These values stem directly from TID-14844 (and we presume that the 1 percent solids refer to fission products other than halogens and noble gases). No specificatica of the time dependencies for this source are given. However, for a PWR with containment sprays, the elemental iodine (constituting about 95 percent of released iodine) is rapidly washed out of the containment atmosphere and transported to the containment sump (over 90 percent in less than 15 minutes is a typical result).
Table D-4 presents an estimate of buildup of iodine in the sump uid.
There is little difference in the estimated integrated dose from the susp water between these values and values resulting from an assumed instantaneous release of 50 percent of'the core iodines into the sump.
The inclusion of solid fission products in the sump source seems to be an artifact from the source of TID-14844. Although it may have applicability to the estimates of hydrogen production per Regulatory Guide 1.7, its applicability to radiation dose estimates has not been fully resolved.
Pending this resolu-tion, it should be assumed that the sump fluid contains 1 percent of the solid fission products and that the solid fission products are released and uniformly distributed in the sump fluid at t=0.
7.
Estimates of the Radiation Environment Dose and Dose Rates Previous staff estimates did not take into account the important time-dependent and spatially dependent phenomena. The calculated radiation environment was
)
generally taken as a point on a surface or in the center of containment.
l l
The activities within the containment regions were used as input to calculate i
the beta and gamma dose rates and integrated doses. One typical location was assumed to be a point located in the center of the main containment region. A second location was assumed to be a point on a containment inner surface. A third location would be adjacent to the sump water. Doses for representative D-6 l
l
l points outside containment were taken from Reference 2 and are also listed for completeness.
The gamma transport calculations were performed in cylindrical geometry.
Containment internal geometry was not modeled because this was considered to involve a degree of complexity beyond the scope of the present work. The calculations of both References 3 and 4 indicate that the specific internal shielding and structure would be expected to reduce the gamma doses and dose rates by factors of two or more, depending upon the specific location and geometry.
The beta doses were calculated using the infinite medium approximation.
Because of the short range of the betas, this was shown in Reference 5 to result in only small error. The beta doses are not expected to be significantly reduced by the presence of containment internal structures.
Finally, tne doses were multiplied by a correction factor of 1.3 as suggested by Reference 5 to account for the neglect of the decay chains with subsequent growing-in of additional daughter products.
a.
Containment Atmosphere Doses and Integrated Dose The beta and gamma dose rates and integrated doses for a point detector on the containment centerline exposed to the airborne activity within the containment atmosphere was calculated. The containment was modeled as an air-filled cylinder whose height equaled the diameter.
Containment internal structure and shielding were neglected. The gamma dose rate contribution for the plate-cut iodine on containment surfaces to the detector was also modeled and included as a contributor. The gamma dose rates and integrated doses are shown in Table D-5, whereas the beta dose rates and integrated doses are shown in Table D-6.
The increased pressure effects in a post-LOCA containment have little shielding importance and therefore was not considered. This results in a small conservatism in the calculated dose.
b.
Surface Dose and Dose Rates The beta and gamma dose rates and integrated doses were computed for containment coatings on which iodine fisssion products were presumed to be plated-out. The containment coatings were assume 3 to have a 3
thickness of 10 mils (0.0254 cm) with an average density of 2 gms/cm.
Removal of plated-out activity with time is expected to be a complex phenomenon dependent upon such conditions as whether the surface is exposed to the sprays and whether moisture condensation and runoff can be expected to remove surface activity. Assuming complete retention of plate-out activity, half of the beta energy from plated-out iodine is assumed directed toward the coated surface.
The airborne contribution was added to the plate-out contribution, and all the betas directed toward the coating were assumed to be absorbed in the coating. This is conservative since the maximum range for betas is greater than the coating thickness.
Hence, this assumption may overestimate the beta dose for a specific coating, D-7
but may be appropriate for a cable insulation layer. The airborne contribution was taken to be one-half the dose rate from an infinite cloud.
The gamma done rate at the plated-out surface exposed to airborne activity was calculated to be one-half of the dose rate for a detector at the contairiment centerline. Although half of the gamma energy from plated-out iodine is also directed toward the coating, the coating is calculated to be relatively permeable to gammas with only about 1 percent of the plated-out gammas absorbed by the coating, and this contribution is considered negligible.
The gamma dose rates and integrated doses are therefore half of the centerpoint values for an airborne detector. The gamma dose rates are not significantly affected by the radioactive decay of plated-out activity with time.
The beta dose rates and integrated doses for "well-washed" and
" unwashed" surface, respectively, are shewn in Table D-7.
Note that a plate-out "washoff" model was not used for the "well-washed" example, the plate-out dose rate component was set equal to zero.
c.
Dose Near Sump Water The activity in the sump water was assumed to vary with time, and to be initially free of any iodine fission products. Ultimately, essentially all of the iodine released appears in the sump water.
Table D-4 gives the iodine activity in the sump as a function of time. Note that the maximum is reached in about 0.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with radioactive decay reducing the activity afterwards. The beta and gamma dose rates and integrated doses were computed for a detector located at the surface of a large pool of sump water contaminated by iodine and solid fission products. There was 44,200 cubic feet of water that was assumed to cover the bottom of the containment. The containment geometry was simplified to assume a uniform depth of water of about 2.5 feet, and the dose rates were calculated at the sump water surface excluding the effects of buildup. The gamma dose rate and integrated dose from the sump water source are given in Table D-8.
d.
Equipment Outside Containment Although not specifically calculated in this study, several values of dose rates and doses at points outside of containment were taken from Reference 2 for completeness. The method used in this report in arriving at these results are acceptable for plant-specific determination.
The gamma dose rates and integrated doses at a point outside of containment are shown in Table D-9 (taken from Reference 2).
The containment source was assumed to be a Regulatory Guide 1.4 source (with a power level of 4000 MWt) and was shielded by 3 feet of concrete. The dose rates at the beginning of recirculation near a pipe containing water contaminated by iodine fission products was D-8
i also calculated in Reference 2 and the dose rates are shown in Table D-10.
l 8.
Comparison of a PWR and a BWR A detailed model for a BWR equivalent to the PWR model is not prese:.ced in this report. Doses to equipment inside a BWR containment (primarily considering a BWR with a MARK III type of containment structure) would not be expected to differ greatly from the doses calculated for PWR equipment. However, some differences in equipment doses will result due to the compartmented design of l
BWR containments, and the fact that most BWRs do not have containment sprays designed for rapid iodine removal.
i l
Several of the models and assumptions used in the PWR analysis would not be appropriate for an equivalent analysis for a BWR.
Specifically:
a.
The assumption of an initial uniformly distributed airborne con-centration of activity throughout the containment is not appropriate for a BWR containment.
b.
Following the blowdown portion of the LOCA, the air exchange rates i
between the drywell region and the remainder of the containment free volume will be relatively small, Since any major releases of activity would be initially into the c.
drywell and would occur following the blowdown period, only relatively slow transport would occur to the main containment volume.
Consequently, an appropriate model for a BWR containment should consider that all (or most) of the activity is initially released into the drywell region.
d.
It is important to correctly estimate the atmospheric mixing rates between the drywell and the main containment regions (including l
sprayed and unsprayed regions) to adequately estimate the time-dependent and location-dependent distribution of activity. This l
should include an estimate of the flow between the drywell and the l
main containment that bypassea the suppression pool. This suggests I
a relatively detailed multi-node containment model, if overly conservative estimates of the radiation environment are to be avoided.
e.
Removal of iodines from the main containment region and from the j
drywell, by operation of ESF systems such as containment sprays, should be modeled in a manner similar to that used in calculating offsite doses (i.e., single failure etc.),
l f.
Time-dependent deposition of iodines on surfaces by natural processes should be evaluated using mechanistic models and best estimates for model parameters; this wi.1 require a relatively detailed evaluation of potential deposition surfaces within the main containment and drywell.
l g.
Capture of iodines in the suppression pool, although not currently assumed, may be important and should be evaluated.
D-9
i P
i Table D-1.
Source Terms: Activity Released from the Fuel as a Percentage of the Total Core Inventory f
Activity Released (percent)
Source Terms Noble Gases Iodines Solids 1.
Source term based on TID-14844 required by Reg.
Guides 1.3 and 1.4) 100 50 0
1 2.
Source term as required by Regulatory Guides 1.7 and
.i 1.89 Rev. 0 (base case)*
100 50 1
l 3.
Source term based on conser-vative gap release (Reg.
10 Guide 1.25)
(30 of 10 0
Kr-185) j 4.
Best estimates of total j
activity gap:
f WASH-1400 3
5 j
NUREG/CR-0091**
1.27 2.79 f
- Case 2 was used in the calculations presented in this appendix.
i
- Calculated for stable and long half-life isotopes (Ref. 8).
t t
l I
D-10 1
.-<g,g,-m---ny,-m,.,
a.-a,m,,w,,-
w,,,,we
,,w.w.,
,,.ew,--
,,vv..
,,-e,-,,,,yyym,,-,-,,-
.g,,.
,. -.. ~,..,,
Table D-2.
PWR Airborne Activity Distribution Within Containment Versus Time - Base Case, Ci Time Noble Elemental Organic Particulate Total Total (hours)
Gases Iodine Iodine Iodine Iodine Airborne l
0.0 1.31 + 9 4.37 + 8 9.15 + 6 1.14 + 7 4.58 + 0 1.77 + 9 i
0.03 1.19 + 9 4.17 + 8 9.07 + 6 1.13 + 7 4.37 + 8 1.63 + 9 0.50 7.36 + 8 3.56 + 6 7.98 + 6 8.58 + 6 2.01 + 7 7.56 + 8 0.75 6.80 + 8 3.35 + 6 7.51 + 6 7.46 + 6 1.83 + 7 6.98 + 8 1.00 6.41 + 8 3.17 + 6 7.11 + 6 6.52 + 6 1.68 + 7 6.58 + 8 2.00 5.54 + 8 2.66 + 6 5.95 + 6 3.96 + 6 1.26 + 7 5.67 + 8 8.00 3.62 + 8 1.62 + 6 3.62 + 6 3.56 + 5 5.60 + 6 3.68 + 8 24.00 2.33 + 8 9.11 + 5 2.04 + 6 1.21 + 3 2.95 + 6 2.36 + 8 1.57 + 6 1.66 + 8 60.00 1.64 + 8 4.84 + 5 1.09 + 6 1.13 + 6 1.34 + 8 96.00 1.33 + 8 3.47 + 5 7.78 + 5 7.11 + 5 7.91 + 7 192.00 7.84 + 7 2.19 + 5 4.92 + 5 4.82 + 5 4.54 + 7 298.00 4.49 + 7 1.48 + 5 3.34 + 5 3.42 + 5 2.76 + 7 394.00 2.73 + 7 1.05 + 5 2.37 + 5 1.89 + 5 1.22 + 7 560.00 1.20 + 7 5.76 + 4 1.31 + 5 720.00 6.01 + 6 3.23 + 4 7.36 + 4 1.06 + 5 6.12 + 6 1
I D-11 l
4 -
Table D-3.
Total Plate-out Surface Activity in the 2
Containment Versus Time for the Base Case Iodine Activity Time Deposited on (hours)
Surfaces,_Ct 0.0 0.0 0.03 1.57 + 7 0.07 2.96 + 7 0.14 3.92 + 7 0.2e 4.23 + 7 0.40 0.50 4.23 + 7 0.75 3.98 + 7 1.00 3.77 + 7 2.00 3.15 + 7 8.00 1.92 + 7 24.00 1.08 + 7 60.00 5.76 + 6 96.00 4.13 + 6 192.00 2.61 + 6 298.00 1.77 + 6 394.00 1.25 + 6 560.00 6.91 + 5 720.00 3.90 + 5 D-12
Table D-4.
Iodine Activity in Containment Sump Versus Time Iodine Activity in Containment Sump, Ci Time Elemental Particulate
- Total Iodine (hours)
Iodine Iodine in Sump 0.0 0.0 0.0 0.0 j
O.03 0.0 0.0 0.0 0.07 2.04 + 8 2.04 + 8 3.04 + 8 0.14 3.04 + 8 0.20 3.35 + 8 3.35 + 8 0.25 3.44 + 8 3.44 + 8 0.50 3.34 + 8 1.39 + 6 3.35 + 8 0.75 3.15 + 8 1.93 + 6 3.17 + 8 1.00 2.98 + 8 2.36 + 6 3.00 + 8 2.00 2.49 + 8 3.48 + 6 2.52 + 8 8.00 1.52 + 8 4.18 + 6 1.56 + 8 24.00 8.58 + 7 2.54 + 6 8.83 + 7 60.00 4.56 + 7 1.36 + 6 4.70 + 7 96.00 3.27 + 7 9.75 + 6 3.37 + 7 192.00 2.06 + 7 6.15 + 5 2.12 + 7 l
298.00 1.40 + 7 4.18 + 5 1.44 + 7 394.00 9.43 + 6 2.96 + 5 9.73 + 6 560.00 5.48 + 6 1.63 + 5 5.64 + 6 720.00 3.09 + 6 9.30 + 4 3.18 + 6 l
- Particulate iodine activity in the containment sump for times less than 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is small and, when added to the elemental iodine activity, does not
(
significantly affect the tctal magnitude of the iodine activity in the sump.
l l
l D-13
Table D-5.
Total Gamma Dose Rates and Integrated Doses at the Containment Center in Air Versus Time - Base Case Unwashed Gamma Dnee Gamma Dose Rate in Air Total Gamma Total Integrated Time Rate From From Plate-out Dose Rate Gamma Dose in the (hours)
Airborne (R/hr)
Source (R/hr) in Air (R/hr)
Containment Air (R) 0.0 4.92 + 6 1.56 + 4 4.92 + 6 0.03 4,43 + 6 5.59 + 4 4.49 + 6 2.06 + 5 0.50 1.33 + 6 1.44 + 5 1.47 + 6 1.18 + 6 0.75 1.16 + 6 1.33 + 5 1.29 + 6 1.55 + 6 1.00 1.05 + 6 1.23 + 5 1.17 + 6 1.82 + 6 2.00 7.75 + 5 9.44 + 4 8.69 + 5 2.80 + 6 8.00 2.37 + 5 4.14 + 4 2.78 + 5 6.0
+6 24.00 5.19 + 4 1.58 + 4 6.77 + 4 7.1
+6 60.00 1.70 + 4 6.36 + 3 2.34 + 4 9.2
+6 96.00 1.30 + 4 4.36 + 3 1.74 + 4 1.0
+7 192.00 7 66 + 3 2.66 + 3 1.03 + 4 1.15 + 7 298.00 4.38 + 3 1.80 + 3 6.18 + 3 1.20 + 7 396.00 2.67 + 3 1.28 + 3 3.95 + 3 1.25 + 7 560.00 1.14 + 3 7.04 + 2 1.84 + 3 1.30 + 7 720.00 5.14 + 2 3.98 + 2 9.12 + 2 1.36 + 7 D-14
Table D-6.
Beta Dose Rates and Integrated Doses at the Containment Center Versus Time in Air
(
Time Dose Rate in Integrated Dose in (hours)
Containment Air (R/hr)
Containment Air (R) 0.0 2.373 + 7 l
l 0.03 1.951 + 7 8.89 + 5 0.25 5.856 + 6 3.55 + 6 0.5 4.198 + 6 4.93 + 6 0.75 3.671 + 6 6.0
+6 1.0 3.369 + 6 7.13 + 6 2.0 2.758 + 6 1.03 + 7 8.0 1.538 + 6 2.21 + 7 24.0 7.068 + 5 4.1
+7 60.0 3.919 + 5 6.1
+7 96.0 3.117 + 5 7.2
+7 192.0 1.871 + 5 8.9
+7 298.0 1.083 + 5 1.03 + 8 394.0 6.807 + 4 1.08 + 8 560.0 3.278 + 4 1.17 + 8 720.0 1.901 + 4 1.26 + 8 l
l I
l l
{
l l
D-15
Table D-7.
Beta Dose Rates and Integrated Doses for Paint on Containment Wall - Washed and Unwashed Cases a
Dose Rate
- Dose Rate **
Dose Dose Time Unwashed Washed Unwashed Washed (hours)
(R/hr)
(R/hr)
(R)
(R) 0.0 1.19 + 7 1.19 + 7 0.0 0.0 0.03 1.01 + 7 9.76 + 6 4.99 + 5 6.46 + 5 0.25 3.79 + 6 2.93 + 6 1.81 + 6 1.69 + 6 0.5 2.92 + 6 2.10 + 6 2.70 + 6 2.32 + 6 0.75 2.60 + 6 1.84 + 6 3.65 + 6 3.0
+6 1.0 2.39 + 6 1.68 + 6 4.20 + 6 3.25 + 6 2.0 1.94 + 6 1.38 + 6 6.39 + 6 4,77 + 6 8.0 1.07 + 6 7.69 + 5 1.42 + 7 9.9
+6 24.0 5.05 + 5 3.53 + 5 2.55 + 7 1.77 + 7 60.0 2.60 + 5 1.96 + 5 3.90 + 7 2.73 + 7 96.0 1.96 + 5 1.56 + 5 4.6
+7 3.3
+7 192.0 1.16 + 5 9.36 + 4 6.0
+7 4.4
+7 298.0 6.90 + 4 5.42 + 4 7.0
+7 5.2
+7 394.0 4.45 + 4 3.40 + 4 7.6
+7 5.6
+7 560.0 2.22 + 4 1.64 + 4 8.2
+7 6.1
+7 720.0 1.28 + 4 9.51 + 3 8.29 + 7 6.33 + 7
- Includes both the containment airborne and plate-out contributions.
- Includes only the containment airborne contribution.
D-16
i Table D-8.
Containment Sump Gamma Dose Rates and Integrated Doses Versus Time Dose Rate Dose Rate Total at the Sump at the Sump Total Dose Integrated Surface From Surface From Rate at Gamma Dose
)
i Time 5
Iodine in 1% Solids in the Sump at the (hours)
(Mev)
Sump (R/hr)
Sump (R/hr)
Surface (R/hr)
Surface (F) 0.0 0.887 0.0 5.90 + 4 5.90 + 4 0.03 0.887 0.0 3.09 + 4 3.09 + 4 4.65 + 2 0.07 0.886 1.18 + 5 0.14 0.884 1.79 + 5 2.21 + 4 2.01 + 5 1.23 + 4 0.20 0.882 1.94 + 5 0.25 0.880 1.99 + 5 1.90 + 4 2.18 + 5 2.82 + 4 0.50 0.873 1.83 + 5 1.59 + 4 1.99 + 5 7.89 + 4 0.75 0.866 1.71 + 5 1.00 0.860 1.56 + 5 1.25 + 4 1.68 + 5 1.68 + 5 2.00 0.839 1.19 + 5 1.01 + 4 1.29 + 5 3.00 + 5 8.00 0.763 5.08 + 4 24.00 0.569 1.61 + 4 4.99 + 3 2.11 + 4 1.15 + 6 60.00 0.401 6.04 + 3 96.00 0.357 3.81 + 3 3.09 + 3 6.90 + 3 1.95 + 6 192.00 0.332 2.20 + 3 298.00 0.330 1.50 + 3 2.14 + 3 3.64 + 3 2.95 + 6 394.00 0.330 1.06 + 3 560.00 0.330 5.86 + 2 1.61 + 3 2.20 + 3 3.65 + 6 720.0 0.330 3.30 + 2 1.42 + 3 1.75 + 3 3.96 + 6 r
D-17
Table D-9.
Gamma Dose Rates Outside Shielded Containment (3-foot Concrete Shield)
Time After Dose Rate Integrated Release (hours)
(R/hr)
Dose (Rads) 0 4.0 x 10 0
2 2
1 2.5 x 10 3.2 x 10 2
3 1.2 x 10 6.9 x 10' I
10 2.8 x 10 1.2 x 10 0
3 30 2.4 x 10 1.5 x 10
-2 3
100 2.8 x 10 1.6 x 10
=
t D-18
)
Table D-lb.
Gama Dose Rates at Beginning of Recirculation Near Pipe Containing Iodine Fission Products Distance Dose Rate (R/hr) 5 4 inches 1.6 x 10 5.3 x 10' 1 foot 3 feet 1.8 x 10
(
\\
D-19
REFERENCES 1.
A. K. Postma and P. T.m. " Technological Bases for Models of Spray Washout and Airborne Contamiaants in Containment Vessels," USNRC Report NUREG/CR-0009, November 1978. Available for purchase from National Technical Information Service, Springfield, Virginia 22161.
2.
A. K. Postma and R. Zavadoski, " Review of Organic I.^'ide Formation Under Accident Conditions in Water Cooled Reactors," WASH d233, October 1972, pp. 62-64.
Availabic for purchase from National Technical Information Service, Springfield, Virginia 22161.
3.
Memorandum from R. B. Minogue, NRC, to R. F. Fraley, October 25, 1975,
" Response to Request for Additional Information Concerning Regulatory Guide 1.89 on Qualification of Class IE Equipment." Available for inspection and copying for a fee from NRC PDR, Washington, DC 20555.
4.
E. A. Warman and 3.
4.
Boulette, " Engineering Evaluation of Radiation Environment in LWF; Containments," Vol. 23, pp. 604-605 in Transactions of the American Nuclear Society, 1976. Available from technical libraries.
5.
M. J. Kolar and N. C. Olson, " Calculation of Accident Doses to Equipment Inside Containment of Power Reactors," Vol. 22, pp. 808-809 in Transactions of the American Nuclear Society, 1975. Available from technical libraries.
6.
D. C. Kocher, ed., " Nuclear Decay Data for Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle Facilities," ORNL/NUREG/TM-102, August 1977. Available for purchase from National Technical Information Service, Springfield, Virginia 22161.
7.
E. Normand and W. R. Determan, "A Simple Algorithm to Calculate the Immersion Dose," Vol. 18, pp. 358-359 in Transaction of the American Nuclear Society, 1974. Available from technical libraries.
8.
R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, " Fission Product Source Terms for the LWR Loss-of-Coolant Accident: Summary Report,"
USNRC Report NUREG/CR-0091, May 1978. Available for purchase from National Technical Information Service, Springfield, Virginia 22161.
1 i
D-20
N APPENDIX E STANDARD QUESTION ON ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT
/
}
s APPENDIX E STANDARD QUESTION ON ENVIR0hMENTAL QUALIFICATION nF CLASS 1E EQUIPMENT In order to ensure that your environnental qualification program conforms with General Design Criteria 1, 2, 4 and 2 3 of Appendix A and Sections III and XI of Appendix B to 10 CFR Part 50, and to the national standards mentioned in part II " Acceptance Criteria" (which includes IEEE Std. 323) contained in Standard Review Plan Section 3.11, the following information on the qualification program is required for all Class 1E equipment.
1.
Identify all Class 1E equipment, and provide the following:
Type (functional designation) a.
b.
Manufacturer c.
Manufacturer's type number and model number d.
The equipment should include the following, as applicable:
(1) Switchgear (2) Motor control centers (3) Valve operators (4) Motors (5) Logic equipment (6) Cable (7) Diesel generator control equipment (8) Sensors (pressure, pressure differential, temperature and neutron)
(9) Limit switches (10) Heaters (11) Fans (12) Control boards (13) Ingtrument racks and panels (14) Cannectors (15) Electrical penetrations (16) Splices (17) Terminal blocks 2.
Categorize the equipment identified in item 1 above into one of the I
following categories:
Equipment that will experience the environmental conditions of a.
design basis accidents for-which it must function to mitigate said accidents, and that will be qualified to demonstrate operability in the accident environment for the time required for accident mitigation with safety margin to failure.
b.
Equipment that will experience environmental conditions of design basis accidents through which it need not function for mitigation of said accidents, but through which it must not fail in a manner detrimental to plant safety or accident mitigation, and that wi,ll be qualified to demonstrate the capability to withstand any accident environment for the time during which it must not fail with safety margin to failure.
E-1
l c.
Equipment that will experience environmental conditions of design basis accidents through which it need not function for mitigation of said accidents, and whose failure (in any mode) is deemed not detrimental to plant safety or accident mitigation, and need not be qualified for any accident environment, but will be qualified for its non-accident service environment.
d.
Equipment that will not experience environmental conditions of design basis accidents and that will be qualified to demonstrate operability under the expected extremes of its non-accident service environment. This equipment would normally be located outside the reactor containment.
3.
For each type of equipment in the categories of equipment listed in item 2 above, provide separately the equipment design specification requirements, including:
a The system safety function requirements.
b.
An e: iironmental envelope as a function of time that includes all extreme parameters, both maximum and minimum values, expected to occur during plant shutdown, normal operation, abnormal operation, and any desiga basis event (including LOCA and MSLB), including post-event conditions.
c.
Time required to fulfill its safety function when subjected to any of the extremes of the environment envelope specified above.
d.
Technical bases should be provided to justify the placement of each type equipment in the categories 2.b and 2.c listed above.
4.
Provide the qualification test plan, test setup, test procedures, and acceptance criteria for at least one of each group of equipment of item 1.d as appropriate to the category identified in item 2 above.
If any method other than type testing was used for qualification (operating experience, analysis, comb.'ned qualification, or ongoing qualification),
describe the method in sufficient detail to permit evaluation of its adequacy.
k 5.
For each category of equipment identified in item 2 above, state the actual qualification envelope simulated during testing (defining the duration of the hostile environment and the margin in excess of the design requirements).
If any method other than type testing was used for qualification, identify the method and define the equivalent
" qualification envelope" so derived.
- 6.
A summary of test results that demonstrates the adequacy of the qualification program.
If analysis is used for qualification, justification of all analysis assumptions must be provided.
- For applications for construction permits, it is acceptable to state that items 6 and 7 will be supplied in the initial application for an operating license.
E-2
\\
- 7.
Identification of the qualification.locuments which contain detailed supporting information, including test data, for items 4, 5 and 6.
In addition, in accordance with the requirements of Appendix B of 10 CFR 50, the staff requires a statement verifying that (1) all Class IE equipment has been qualified for an operating license (OL) or will be qualified for a construction permit (CP) to the program described above, and (2) the detailed qualification information and test results are (or will be) available for an
(
NRC audit.
- For applications for construction permits, it is acceptable to state
\\
that items 6 and 7 will be supplied in the initial application for an operating license.
E-3
'U" U S. NUCLE AR REGULATORY COMMISSION fiUREG-0588 BIBLIOGRAPritC DATA SHEET 4 TIT L E AN D SU BTI T LE (A dd Vosurne No, rt acormerare) 2 tt ese nys j Interin Staff Position on Environrental Qualification of Safety-Related Electrical Equiprent 3 nE clPIE N T'S ACCESSION NO 7 AUTHOR tSi
- 5. D ATE HE POHT COVPLE TE D vosTs
! naR A. J. Szukiewicz and others August
_ __ 379 9 PE uf ORVING ORGANIZATION N AVE AND M AILING ADDRESS f/nelude la Codel D AT E RE POR T 155UE D l n au Division of Systems Safety v em "
Office of Nuclear Reactor Regulation J4ovenber 1979 U. S. fluclear Regulatory Conmission c neue vrai Washingtion, D. C.
20555 8 l Leave bir k) 12 SPONSORtNG ORG ANIZ ATiON N AVE AND M AILING ADDRESS (lac 6de 2 0 Codel 10 PROJ5 CT T ASIVWOHK UNIT NO Sare as 9 above.
Regulatory Report il/A 15 SUPPLEVENT ARY NOTES
- 14. fleme ural This report dccments resolution of Generic Tech. Act. A-2L 16 ABSTH ACT (200 enecs or less)
This docurrent provides the fiRC staff positions regarding selected areas of environ-mental qualification of safety-related electrical equipment, in the resolution of Generic Technical Activity A-24 " Qualification of Class 1E Safety-Related Equipment."
The positions herein are applicable to plants that are or will be in the construction permit (CP) or operating license (OL) review process and that are required to satisfy the requirerents set forth in either the 1971 or the 1974 version of IEEE-323 standard. These positions were developed prior to the Three flile Island Unit 2 event. Any recommendations resulting from the event, pending staff's conpletion of the review of that event, will be provided in a supplemental report. The seismic qualification requirements are addressed elsewhere and are not included in the scope of this docunent.
I 17 K E Y WOR DS AND DOCUVE NT AN ALYSIS 17a. DESCRIPTORS t
17b IDENTIFIE RS'OPEN ENDE D TERMS 18 AV AILABILITY STATEMENT
- 19. SECURITY CLASS (Th,s report) 21 NO OF P AGE S N/A Unlimited availability
- 20. SE Cu RiTY CLASS (Th,s pap')
- 22. PRICE fi/A S
NRC F ORM 335 (7 77)