ML19269E993
| ML19269E993 | |
| Person / Time | |
|---|---|
| Issue date: | 05/31/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0458, NUREG-458, NUDOCS 7911130488 | |
| Download: ML19269E993 (62) | |
Text
NUREG-0458 SHORT TERM SAFETY ASSESSMENT ON THE ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT OF SEP OPERATING REACTORS
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2192 176 Office of Nuclear Renctor Regulation U. S. Nuclear Regulatory Commission 1911ygn gg
Available from National Technical Infnrmation Service Springfield, Vir:~nia 22161 Price: Printed CopySS.Zb ; Microfiche $3.00 The price of this document for requesters outside of the North American Continent can be obtained from the National Techc.ical Information Service.
2/9p 177
NUREG-0453 SHORT TERM SAFETY ASSESSMENT ON THE ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT OF SEP OPERATING REACTORS Manuscript Completedi May 1978 Date Published: May 1978 Division of Operating Reactors
- 7p Office of Nuclear Reactor Regulation I ' O U.S. Nuclear Regulatory Commission Washington, D.C. 20555
TABLE OF CONTENTS PAGE 1.0
SUMMARY
AND CONCLUSIONS 1
2.0 GENERIC CONSIDERATIONS OF ENVIRONMENTAL QUALIFICATION I< FORMATION 7
3.0 REVIEW 0F LICENSEE RESPONSES 12 4.0 TECHNICAL ASPECTS OF ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT 19 4.1 Environmental Conditions 19 4.1.1 Submergence and Chemical Spray 19 4.1.2 Design Basis Event Insije Containment 20 4.1.3 Pesign Basis Event Outside Containment 28 4.2 Aging,unsiderations 30 4.3 Margin and Sequence 34 4.4 Qualification Methodology 37 4.5 Radiation Considerations 37 Appendix A: NRC Letter to SEP Licensees dated 12/23/77 Appendix B: Main Steam Line Break Best Estimate Method and its Effect on Environmental Qualification Appendix C:
Essential Elements of Proposed Circular 2192 179
TABLE OF CONTENTS (CONTINUED)
Figure 1:
Best Estimate Containment Atmosphere Vapor and Saturation Temperature Response for Main Steam Line Break in Typical PWR Figure 2: Bounding Containment Temperature Response for Main Steam Line Break in BWR with Mark I Containment Table 1:
SEP Facilities Information Table 2:
Potential Environmental Conditions of Submergence and Spray for Safety-Related Electrical Equipment
- ii -
2192 180
. 1.0
SUMMARY
AND CONCLUSIONS The subject of environmental qualification of safety-related electrical equipment is addressed in the "NRC Staff Report on Union of Concerned Scientists' Petition for Emergency and Remedial Action". The report, which was submitted to the Commission on December 15, 1977, included a staff evaluation of concerns regarding the environmental qualification of specific electrical equipment (in particular, electrical connectors and penetrations). Overall considerations of environmental qualification of safety-related equipment were discussed in in AppendixE to that report.
Therein, the staff concluded that reasonable as surance existed that electrical connectors and penetrations would perform their required function in the environment of a loss-of-coolant accident (LOCA) and that no imediate remedial action was required for operating reactor facilities.
However, the staff also concluded tnat it would be prudent to determine, in a relatively short time, whether the scope of the environmental qualification review effort should be expanded from the two specific types of components to include all safety-related electrical equipment located both inside and outside of containment for all design basis events (DBEs).
E " Staff Report on Environmental Qualification of Safety-Related Electrical Equipment" was included in the report to the Commission as Appendix B, and was issued separately as NUREG-0413 in Fc'uruary 1978 under the same title.
2192 181
. To provide a more comprehensive basis for this conclusion, the staff determined that a preliminary review on such an expanded scope on a select group of operating reactors was appropriate. The findings for this representative group of plants would then be used to determine whether a need existed for any further action on all other operating reactors.
The subject of environmental qualification is also included in the recently instituted Systematic Evaluation Program (SEP) as one of the safety topics to be reviewed on older operating reactors. Therefore, the staff elected to use this program as the framework for the expanded review effort and decided that this topic should be the first issue to be evaluated as part of this program. The SEP includes eleven of the older operating reactor facilities as listed in Table 1.
The Systematic Evaluation Program is a staff evaluation to determine the degree to which these eleven facilities deviate from current licensing criteria for specific subjects, to assess the significance of these deviations, to develop an overall and balanced position concerning any potential need for backfitting, and to document the results of such an evaluation. The program for the eleven plants is expected to be completed within about three years. However, because of the potential generic safety implications of the environmental concern, this subject was initiated on an accelerated schedule.
2192 182 The expanded review of environmental qualification of safety-related electrical equipment would include a survey of all safety-related electrical equipment in the SEP facilities, and a short term safety assessment of those plants in order to determine, in sufficient detail, any generic safety implications and to decide whether any additional review of this subject is required for other operating reactors.
Subsequent to the report to the Commission of December 15, 1977, the 2
staff, by letters dated December 23, 1977 ], requested the licensees of the SEP facilities to provide additional infonnation on the environ-mental qualification of safety-related electrical equipment. On January 5,1978, the staff met with the licensees to discuss this topic and explain the purpose of and the need for the short term safety assessment and its relationship to the overall SEP effort. Subsequently, the staf" met with each of the licensees individually to discuss the progress of the licensee's effort in responding to the December 23, 1977 letters. During these meetings, which were held at the plant sites, the staff also viewed much of the safety-related electrical equipment as installed, reviewed with the licensee typical qualification records available at the facility, and discussed with the licensee the specific infonnation needed by the staff for this short term assessment. All A typical letter is included as Appendix A to this report.
2192 183 licensees have submitted their responses to the staff request and the staff has completed its hort term safety assessment.
On April 13, 1978, the Commission directed that this assessment be completed within one month and the results reported to the Comission.
The staff performed a completeness review of the infonnation submitted by the licensees with respect to specific safety-related electrical equipment, environmental conditions postulated and methods of qualifi-cation.
The staff included in its considerations its general knowledge of redundancy and location of components, the time when systems typically perform their safety function and the principal effects of such para-meters as steam, temperature, radiation, pressure and chemistry.
In judging the reasonableness of licensee statements, the staff relied on its best engineering judgment and past experience in evaluation of such matters.
Based on a preliminary review of the information provided by the licensees for the eleven SEP facilities, including previously docketed information, and based on observations by the staff of safety-related electrical equip-ment in the facilities, we have concluded that:
1.
No significant safety deficiencies requiring imediate remedial actions were identified in the assessment of the eleven SEP plants.
2192 184 2.
The results of the staff's safety assessment of the eleven SEP plants has not changed the staff's basis and conclusions in NUREG-0413.
3.
The changing and upgrading of older operating plants is a continual process. While this process is not part of a comprehensive program, it has provided and continue's to provide additional assurance that safety-related equipment will perform its function when required.
Regarding the need for an expanded scope of review of environmental qualification information for other operating reactor facilities, we have concluded that:
1.
Equipment located inside containment is subjected to the most extreme environmental conditions for postulated accidents. For equipment located outside containment: the high energj line break (HELB) typically produces the most severe environmental conditions.
Postulated HELBs outside containment were previously considered for all SEP facilities and other operating reactors.
2.
The postulated accident which generally could produce the most severe environmental conditions, is the loss-of-coolant accident (LOCA). The main steam line break (MSLB) can produce higher temper-atures than a LOCA; however, for all operating reactors, except a few of the older SEP facilities, the duration of this temperature
" spike" is so short that the thennal capacity of the equipment damps the higher environmental temperatures.
2192 185 3.
Any questions regarding the environmental qualification of safety-related electrical equipment do not appear to apply to all electrical equipment but seem to be limited to certain component types. Further-more, such problems appear to be more prevalent in small "of-the-shelf" components.
In reaching its recommendation regarding the need for additional review of this subject for other operating reactors, the staff also considered the generic implications of licensee responses and actions to staff con-cerns on the environmental qualification of electrical connections and penetrations previously reported to the Commission, expecially on those concerns identified after the staff's - G et of December 15, 1977.
Based on these considerations, and those discussed in Sections 2 and 3 of this Report, the staff concludes that, while no immediate remedial action is necessary, it is appropriate to devote some additional staff effort, consistent with available resources and other important activities, to examine the installation and environmental qualification documentation of specific electrical ' equipment inside containment of all operating reactors.
We are, therefore, recommending the issuance of an IE Circular to all licensees. Such a circular will provide for the effective feedback of the significant " lessons learned" regarding environmental qualification to all licensees.
It will also request additional licensee review in this area for selected safety-related electrical equipment inside containment. NRR will participate with I&E in followup inspections to resolve any questionable design practices. The essential elements of this Circular are presented in Appendix C.
g The staff will continue-to evaluate, on the long term basis and, as part of the overall SEP effort, the detailed infonnation provided in response to the request for information from the eleven plants. The evaluation will also consider the extensive information referenced in the licensees' responses which has previously been decketed, information included in support of Topical Reports, and other generic equipment qualification and test reports for specific types of electrical equipment that have not been previously docketed.
Should a potential significant safety deficiency be identified during this evaluation, the staff will follow its SEP procedure which provides for taxing such an issue out of the SEP for imediate evaluation and assessing the relationship of such an issue to non-SEP operating reactors.
2.0 GENERIC CONSIDERATIONS OF ENVIRONMENTAL QUALIFICATION INFORMATION The staff report of December 15,1977 (NUREG-0413) discussed the environmental qualifir,ation of electrical cable connectors and pene-trations. Since that time, a substantial amount of additional information has been collected on the status of environmental aualification of electrical equipment. Much of the information was included in responses from all licensees to bulletins issued by the Office of Inspection and Enforcement (I&E) which showed that for certain components documentation of environmental qualification, as currently required, could not be imediately identified. Other information was gained from qualification 2192 187 tests that were performed within the past four months for specific electrical components. The staff has reported to the Commission on each of these developments and on the disposition of each item on a regular basis, including a status report on all environmental qualifi-cation activities.O The questions about environmental qualification of safety-related electrical equipment are, in part, the result of the evolution of the regulatory and industry effort b the development of the current criteria, regulations and standards.O The design basis for functionability of safety-related equipment under accident conditions, including the adverse environment caused by the accident, has been recognized for many years and was implicitly included in the design of even the oldest nuclear power plant. However, the specific requirements and appropriate guidance (10 CFR 50 - Appendix B, IEEE-323-1974 and other appropriate standards, Regulatory Guide 1.89, Standard Review Plan 3.11) for the E Memo to the Commission from E. Case and E. Volgeneau, March 23, 1978.
N NUREG-0413, Appendix A, " Report on the Historical Evolution of Environmental Qualification Requirements for Safety-Related Electrical Equipment".
2 02 188
_g-development, conduct, documentation, and evaluation of specific environmental qualification programs to satisfy the design basis have been continually updated. For example, in response to I&E Bulletin 77-0E regarding the qualification of electrical penetrations, the licensee for Dresden Unit 1, which is the oldest operating facility, was able (albeit with great difficulty however) to trace and locate data from environmental qualification tests perfonned in the 1950's to provide a design basis for qualification. Thus while the general design basis was reasonably well adopted in the nuclear industry, specific requirements had not been estab-lished and considerable variation in licensee approaches existed for the older operating reactors. Therefore, the staff capability to perform meaningful inspections in this area, in particular qualification tests and evaluations and installation and mainter.ance procedures, was limited.
Based on the general considerations of the evolution of the requirements for environmental qualification and based on the specific results of the staff's and the licensees' activities in the past several months, 2192 189 we have made the following observations. First, many licensees have difficulties in tracing and obtaining information on the subject of environmental qualification.
In many instances such information is not available in the licensees' cagineering evaluations but is indicated in some correspondence between licensee and vendor or in purchase speci-fications. Secondly, the design of specific equipment, as evidenced by recent testing or evaluation, reflects a general sngineering recog-nition and requirement that the equipment be able to perform its function under accident conditions. Thirdly, where in recent instances equip-ment was replaced or modified by licensees, such action was, in some cases, not mand;ted by unsuitable design application but was determined to be more expeditious than demonstrating by test or analysis the quali-fication of the installed equipment.
Considering the infonnation on the environmental qualification of safety-related electrical equipment, including the preliminary review of the SEP facilities, the staff believes it prudent to assure that all licensees are familiar with the recent developments in the area of environmental qualification through issuance of an I&E Circular.
In addition, recog-nizing that a lack of specific design requirements in the past resulted in a reduced capability of the staff to conduct meaningful inspections in this area, an enhanced inspection effort, with input from the Office of Nuclear Reactor Regulation (NRR), should be initiated as the subject 2192 190 of environmental qualification. These staff reviews will be concentrated on electrical equipment where problems have been identified in the.past, i.e., small, "off-the-shelf" components such as tenninal blocks, connec-tors and splices, located inside containment and required to function under LOCA conditions. This approach should provide additional assurance that all operating plants have adequately addressed this issue.
There are already indications that some licensees have initiated such reviews. For example, on April 21, 1978, the licensee for D. C. Cook Units 1 and 2 informed the staff of changes that were made to electrical equipment such as tenninal blocks and cable and to reconnect safety-related instruments in flood-up tubes to protect them from direct exposure to caustic sprays and high temoeratures. On April 24, 1978, the licensee for Monticello indicated that questionable splices inside of containment would be replaced with qualified splices.
In recent reviews of plants for operating licenses, we have found the need for similar actions by licensees. Just as was the case for the SEP plants, certain environmen'.a1 parameters were not included in the original environ-mental testing for some of the current OL applications, and complete environmental qualification data may not always be available. For example, the reviews of North Anna Unit 1 and D. C. Cook Unit 2 revealed that certain Barton pressure transmitters will need to be requalified to in-clude sequential environmental effects. The staff is also reviewing the 2192 19i adequacy of the qualification testing performed on several models of Foxboro instrumentation transmitters and has indicated to the licensees that a more complete and adequate qualification basis is required. These findings will be included in the information being provided to all licensees (Appendix C) to ensure timely dissemination of information gained in the ongoing staff reviews and to ensure appropriate responses.
3.0 REVIEW 0F LICENSEE RESPONSES The responses from the licensees of the eleven SEP facilities consisted, in general, of a tabular list of electrical equipment including com-ponent identification and function, component location (e.g., inside/out-side containment, room, elevation), and the service environment and qualification environment (i.e., environmental parameters and values).
Extensive additional information was provided by references to previously submitted information. Although the extent of information provided varied among the licensees, the information was limited to safety-related electrical equipment in most responses.
In some cases, non-safety related equipment was included and to an extent, it complicated the review for this short term assessment. One licensee did not include a listing as such of specific safety-related electrical equipment but did discuss the environmental qualification aspect in the submittal. The staff utilized, for this review, other information referenced and/or previously submitted by the licensee.
~
2192 192 In general, the NRC short time constraints on licensees limited their efforts to provide detailed information on the documentation of the qualification for all safety-related equipment.
In most instances, the licensees advised the staff that their search for adc'itional quali-fication documentation is continuing. As a result, the submittals made extensive use of references to earlier submittals,(e.g., high energy line break reports) and to additional information within the licensees' records and files. These included communications betwee-the licensee and the equipment vendor, manufacturers' specifications, and test reports on specific equipment which were not normally submitted to the staff.
For some facilities, the licensees also identified specific safety-related electrical equipment or environmental considerations which will be further evaluated as the SEP effort progresses. The facilities and the concerns identified by the licensees are:
1.
Big Rock Point The licensee stated that investigation of the following environ-mental considerations will continue:
a.
The potential for an inadvertent release of boric acid from the liquid poison system, due to LOCA conditions, which could result in a chemically adverse environment not previously considered for safety-related electrical equipment.
2192 193 b.
The potential for the rupture of a low energy steam line (space heating) in the electrical equipment room and the effects on safety-related electrical equipment in the room, c.
The potential for the destructive failure of a non-safety motor control center (MCC) inside containment due to con-tainment spray, which could subsequently damage a safety-related reactor vessel low water level sensor located in the vicinity of the MCC.
The staff has reviewed the likelihood of these events and their consequences.
In the review, the staff has considered the availa-bility of redundant equipment and the time relationsnip between the time the safety-related equipment is required to perform its safety function and the time when it is postulated to fail.
Based on these considerations, the staff concludes no inmediate action is needed pending the completion of the review by the licensee and the staff.
In addition, two solenoid valves with an insufficiently qualified component were identified (nylon disc with qualification tempera-ture of 150 F). One valve (ASCO Model 831620) controls three isolation valvec in the resin sluice system (non-safety syste.m) and the other (ASCO Model 831622) is associated with scram valves for one control rod drive. The two solenoid valves are de-energized to perform their safety function and they are administratively controlled in this position. They will remain so until such time 2192 194 that an evaluation regarding their acceptability is completed or a modification made.
2.
Palisades The licensee stated in his submittal that further evaluation will be conducted on the following components':
a.
A transmitter for pressurizer pressure (PT 103), which is required for the determination of the long term cooling status, was found to have insufficient documentation of qualification for the radiation environment. The licensee advised us that an additional transmitter (PT 104), qualified for a LOCA environment and for a total dose of 5 x 106 rads, has been
- added, b.
Solenoid valves (Model ASCO 831614) for the cooling water valves to the containment air coolers, have inconclusive radiation qualification information. The licensee advised the staff that the valves will be replaced with qualified models.
In the interim, a failure of the existing solenoid valves (i.e., valve becomes de-energized) will result in the venting of the control valve and the valve will remain open, which is the required safety position.
2192 195 3.
Ginna The licensee has identified the following two areas for which further evaluation and modifications are currently in progress:
a.
The potential for the failure of a steam line (space heating) in the diesel generator rooms and in the screenhouse and the effect of the resulting advarse environment on safety-related electrical equipment.
b.
The potential for a high energy line break in the turbine building which could create an adverse environment for safety-related electrical equipment in the control building and in the diesel generator rooms.
The licensee has proposed to make modifications to the facility to protect safety-realted equipment from the postulated failures above.
The staff review of those modifications is scheduled to be completed in May, 1978.
In the interim, plant inspections are performed during each shift to detect leaks and to isolate affected pip;ng if necessary.
4.
Dresden Unit 1 The licensee committed to install new equipment, relocate some equipment from inside to outside the containment and perform quali-fication tests or analyses as part of a previous backfitting Order 2192 196 issued by the staff in 1976. A more recent staff safety evaluation discussing the situation was issued on January 6,1978. The plant is scheduled to be shutdown later in 1978 to make extensive plant moaifications. Until that time, surridient protection is provided by non-safety equipment located outside containment and by the qualification and redundancy of existing equipment.
The staff reviewed the above issues identified by the licensees and concludes that no immediate remedial action need to be initiated on these issues while the licensees are pursuing their qualification evaluation effort. We will include our conclusions on these ongoing efforts in our final evaluation of the Systematic Evaluation Program.
As requested by the staff, the licensees for each of the eleven SEP facilities has evaluated the information presented in their submittals.
Based on their review of that information, each licensee has concluded that there is a reasonable assurance that each facility can continue to operate safely and without endangering the health and safety of the public.
Based on our pr c'iminary review of the licensees' submittals and other referenced information, we also conclude that the specific iten:s identi-fied above are probably not generic in nature but are plant specific.
On the other hand, similar, but minor items would probably not be limited tc the SEP facilities b.ut may be considered to be symptomatic for other operating reactor facilities. These items were detennined to exist, in most cases, as the result of the licensees' systematic 2192 197 review of safety-related equipment with respect to environmental conditions and qualifications. Therefore, while major new safety prob-lems were not uncovered, it appears to be prudent for other licensees to review their operating reactor facilities to determine the existence of any similar problems.
2\\92 198 4.0 TECHNICAL ASPECTS OF ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT 4.1 Environmental Conditions The limiting environmental conditions considered by the staff in the preliminary assessment of electrical equipment quaiit ratior f or the eleven SEP facilities are discussed below. This includes Design Basis Events (DBEs), occurring both inside and outside the containment s tructure. Parameters considered were submergence, chemical spray, pressure, temperature, humidity, and radiation. These factors are addressed in the following sections.
4.1.1 Submergence and Chemical Spray In some of the SEP facilities, safety-related electrical equipment inside containment could be subjected to wetting by containment spray and to total submergence by water or aqueous chemical solutions.
Failure of riectrical equipment may result from the admission of water or electrolytic solutions to the equipment internals or from direct chemicai attack on the equipmeint.
Table 2 summarizes the submergence and chemical spray environments for the SEP facilities.
Based on our preliminary review of the infonnation submitted, we conclude that the adverse environment due to chemical spray and to submergence has been considered by the licensees and does not constitute a significant new safety concern that requires irmediate remedial action.
2192 199 The staff also determined that its conclusions for submergence and exposure to chemical spray, including the type of spray, can equally be applied to other operating reactor facilities since these areas were typically reviewed as part of past staff ECCS evaluations.
4.1.2 Design Basis Event Inside Containment The environmental conditions of pressure, temperature, and humidity used in the assessment of the qualification of safety-related eiectri-cal equipment inside containment are discussed below. The main steam line break (MSLB) results in the highest temperature condition inside containment, whereas the loss-of-coolant accident (LOCA) usually results in the highest pressure condition.
One hundred percent relative humidity was assumed inside containment for both the MSLB and LOCA environments.
High relative humidity conditions result in moisture condensation on and within electrical equipment which could cause electrical short circuits.
Since the plant type and the containment design (dry, or pressure suppression) detennine to a large measure the environmental conditions, the eleven SEP plants hav@ been grouped accordingly.
PWR and BWR Facilities With Dry Containments For pressurized water reactors (PWR) licensed prior to 1976, the environ-mental conditions, for which safety-related electrical equipment inside containment was qualified, were established by the loss-of-coolant 2192 200 accident (LOCA) analyses. Typically, the large LOCA (double-ended, instantaneous guillotine break of the largest reactor coolant system line) would result in a peak calculated containment pressure and temperature on the order of 40 - 50 psig and 260 - 280 F, respectively.
Analyses by the NRC staff and by several applicants have shown that the calculated temperature inside containment associated ictith a 0
main steam line break (MSLB) could be as much as 100 - 150 F higher than the predicted LOCA temperature for a short period (60 - 100 seconds) early in the MSLB accident.
The staff recognizes that the analytical methods used to predict the containment temperature following a MSLB contain significant conservatisms. Specifically, the analyses were performed for an instantaneous large steam line rupture assuming dry steam blowdown, conservative containment heat transfer coefficients, and conservative treatment of the thermodynamics of condensate behavior.
In view of these conservatisms, the staff developed a more realistic, yet still conservative, "best es!.imate" evaluation method for predicting the containment temperature and pressure and the thermal temperature peakN.
The "best estimate" evaluation method was developed for a typical PWR dry containment design.
The temperature response for a typical E
Memorandum R. Tedesco to R. Nattson, V. Stello, and R. Boyd, "Best Estimate Evaluation for Environmental Qualification of Equipment Inside Containment Following a Main Steam Line Break", February 24, 1978.
2192 201 Westinghouse PWR is shown in Figure 1.
The thermal capability of typical safety-related electrical components was evaluated with respect to the "best estimate" MSLB containment temperature transient. A compari-son of typical LOCA test envelope temperatures indicates that the cal-culated MSLB thennal response of such typical components will remain within the actual LOCA qualification temperature envelope. The method and its effect on the environmental qualification are described in neuw '% B.
The staff has considered the application of "best estimate" calculations to the eleven SEP facilities. The effort is still in progress; however, certain preliminary conclusions have been reached as detailed below.
The "best estimate" results and conclusions can be directly applied to the Ginna, Palisades, and San Onofre facilities, which are PWR facilities with typical dry containments.
Due to similarity of dry containment designs, this information can also be extrapolated to other operating plants of such typical design and thus provide confidence that safety-related electrical equipment would survive the environmental conditions associated with a main steam line break. However, the "best estimate" method can be applied only indirectly to the Haddam Neck and Yankee Rowe facilities (dry containment PWRs) and to the LaCrasse, Dresden 1, and Big Rock Point facilities (dry containment BWRs). The facilities do not have a containment spray system which responds automatically to terminate the high temperature transient in the same manner as assumed 2192 202 for the "best estimate". method.
Therefore, the "best estimate" treat-ment of electrical equipment heat transfer effects cannot be used for these facilities to demonstrate a temperature difference between the containment vapor and the component under cons.ideration.
The spray systems for Haddam Neck, Lacrosse, and Dresden 1 ara initiated by the operator at some time af ter a LOCA or MSLB. The containment sprays at Big Rock Point are initiated automatically, but af ter a 15 minute delay, and Yankee Rowe does not have a containment spray system. Due to these considerations, the staff has not yet made a final determination as to the potential temperature resulting from a MSLB inside these containments.
However, based on our experience from the reviews of typical electrical components, we have found that equipment qualified for a LOCA environment typically have margin available to sustain a higher temperature.
The staff will continue to evaluate, on a priority basis, the potential for the high temperature conditions for these five facilitie' and the avail-s able thermal response margin for affected safety-related electrical components.
If at any time specific components are identified which do not have the requisite environmental capability, the staff will take immediate remedial action.
2192 203 Facilities with Mark I Pressure Suppression Containments The boiling water reactor (BWR) Mark I containment is a pressure suppression system consisting of a drywell, which houses the reactor and most of the associated high energy steam and liquid systems, and a wetwell, which is a toroidal vessel encircling the drywell. A number of vent pipes connect the drywell to the wetwell.
In the event of a LOCA or MSLB inside containment, the drywell is designed to contain the energy as it is released and direct it through the vent pipes to the wetwell. The wetwell contains sufficient water to provide a heat sink for the released LOCA energy.
Typically, the environmental conditions for which containment structures and safety-related equipment inside containment have been designed are the conditions resulting from the large LOCA (double-ended instantaneous guillotine break of the largest reactor recirculation line). The large LOCA results in the highest calculated pressure inside containment. The temperatures associated with the LOCA are determined by assuming the containment environment consists of saturated steam at the calculated containment pressure.
For BWR Mark I facilities, typical LOCA con-tainment design temperatures and pressures are 280 F and 60 psig, respectively.
2192 204 In the case of the MSLB, the rupture location and size can be such that saturated steam, but no liquid, escapes to the drywell.
For this condition, the drywell temperature can be higher than the predicted LOCA temperature, because the throttling of high pressure reactor steam to the low pressure drywell results in superheating of the steam above the saturation temperature corresponding to the drywell pressure.
An example of the superheat phenomenon was experienced at an operating BWR (Dresden 2) when a transient resulted in reactor system safety valves becoming stuck in a partially open position.
In the first few seconds of this incident, the safety valves relieved high pressure saturated liquid to the drywell, and the containment temperature was calculated to rise as the saturation temperature corresponding to the rising drywell pressure. When steam, saturated at reactor pressure, began to flow through the safety valves it expanded and experienced superheating.
The highest drywell temperature and pressure during this incident were calculated to be 320 F and 20 psig.
The effect of superheating is evident because the saturation temperature of steam at 20 psig is 260 F.
Although some equipment damage did occur during the event, the ECCS and containment systems functioned properly.
Further investigation of this incident resulted in the conclusion that, under certain conditions, the peak temperature that could be achieved is approximately 340 F for BWRs with Mark I containments.
2192 205 Upon examining the spectrum of BWR main steam line break locations and sizes, it is concluded that:
1.
A discharge of saturated steam produces higher containment temperature than a discharge of saturated water because of s uperheating.
2.
The most severe temperature transient in the drywell is caused by a small break, steam only, which does not cause reactor system depressurization or automatic operation of emergency core cooling systems.
It is assumed that following a break of this size, the reactor operators will initiate an orderly shutdown and cooldown of the plant.b 3.
The large break, steam only, produces the most rapid rise of con-tainment temperature to the superheat temperature but the duration of the high temperature condition is less than that of the small break because the large break rapidly depressurizes the reactor coolant system and causes the automatic initiation of the emergency core cooling systems which cool the reactor system and ultimately the containment.
N Preliminary Safety Analysis Report, Section 6.2, Grand Gulf Nuclear Station Units 1 and 2, Docket Nos. 50-416 and -417.
2192 206 Based on the above discussions, a qualification temperature for the typical Mark I BWR was developed (Figure 2). The profile conservatively combines both the rapid rise to 340 F characteristic of the large MSLB and the extended duration with subsequent decrease in temperature corre-sponding to a small MSLB for a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> r.ormal reactor plant cooldown.
The SEP facilities with Mark I containments are the Dresden 2, Millstone 1, and Oyster Creek facilities. The type and quantity of safety-related electrical equipment inside containment of these facilities is typical of that used in other Mark I BWRs and has been essentially standardized by the NSSS vendor,
.neral Electric.
Typical safety-related equipment inside the drywall includes inboard containment isolation valves, pressure relief valves, the high pressure coolant injection (HPCI) pump steam supply valve, recirculation pump discharge and discharge bypass valves, and electrical cabling, penetrations, and connections associated with this equipment.
Consideration has been given to the time required to perform its safety function within a few seconds of a MSLB or LOCA and before a prolonged containment high temperature condition exists that can cause degradation of the equipment. The effect of high containment temper-ature over long periods is mitigated by the redundancy available to the containment isolation, HPCI, and pressure relief systems and the location of much of this equipment, outside of containment.
In addition, the high temperature condition may be further mitigated by operation of the containment spray system, normally manually initiated.
2192 207
~
The staff is continuing its assessment of the Mark I SEP facilities to take into account plart specific environmental qualification factors.
For example, the autimatically initiated containment spray system at Oyster Creek would ra. 'dly reduce the containment temperature and Figure 2 may be too severe. Based on the experience gained at the Dresden 2 facility and based on the typicality of safety-related electrical equipment used in nuclear reactor facilities, the staff concludes that no immediate remedial action need to be taken for the BWR Mark I facilities.
4.1.3 Design Basis Event Outside Containment In December of 1972, the staff sent letters to all licensees with operating reactors requesting an analysis of postulated failures of high energy lines outside of containment.
High energy lines are defined as pipes carrying a fluid at a temperature in excess of U
200 F and at a pressure in excess of 275 psig.
The staff's letter included the criteria and requirements for evaluating existing designs for postulated high energy line breaks (HELB).
The consequences of such failures were evaluated to determine the impact of such failures on systems required to shutdown and maintain the reactor in a safe condition.
The effects associateo with pressure buildup, pipe whip, jet impingement, and environmental conditions of temperature, pressure, and humidity subsequent to a postulated failure were considered in the evaluation.
2192 208 Meetings with the licensees were held in January of 1973 to discuss the potential problems at their facilities and to quantitatively assess the potential for damage. The analyses presented by the licensees provided information concerning areas containing high energy lines, equipment such as instrumentation in close proximity of such lines, structural loading anticipated from pipe whip, and environmental conditions resulting from the postulated HELB. The licensees further proposed corrective measures to prevent or mitigate the consequences of the event.
The proposed modifications varied in degree from installing barriers and pipe restraints to relocation of equipment or piping.
The staff required an augmented inservice inspection pro-gram in areas of high energy lines for those facilities which required extended time to complete the proposed modification.
A review and evaluation of HELBs outside of containment was incorporated in the licensing review for all facilities receiving operating licenses after 1972.
The staff review of high energy line breaks outside of containment for the SEP facilities has been completed and a Safety Evaluation Report (SER) was issued for these plants except for the Ginna plant. The review for Ginna is scheduled for completion in May, 1978.
2192 209 The environment from this event has been postulated to result in a temperature of approximately 210 F in most areas (in others as high U
as 250 F)
'00% relative humidity, and a pressure increase up to 15 psig.
The modif edtions proposed by the licensees were to relocate or pro-tect temperature and humidity sensitive equipment and instrumentation, and to restrain or isolate pipes containing high energy fluids so as to preclude subsequent damage from pipe whip. As indicated previously an augmented inservice inspection program has been implemented for facilities which have not completed the accepted modifications.
This program will remain effective until all modifications are completed.
The staff has performed a preliminary reassessment of high energy line breaks outside of containment with respect to environmental qualification of safety-related electrical equipment for the SEP plants. This reassessment indicates that the corrective measures taken to prevent or mitigate the consequences of the event are acceptable.
We have concluded that no immediate action is required with respect to the environmental qualification.
4.2 Aging Considerations After publication of IEEE-323-1974, the staff undertook to develop Regulatory Guide 1.89 to officially endorse it for use in the licensing 2192 210 review process. One of the significant changes in this standard relative to IEEE-323-1971 was the consideration for establishing a quali-fied life for safety-related equipnient.
In 1974, during the delibera-tions. of the NRC's Regulatory Requirements Review Committee on the imple-mentation of this guide, consideration was given to the incremental ir.travements t:; safety it afforded in comparison to the then current staff revie'. prcctice. The Comittee recommended that the guide be appliM only to future CP applicaticns; i.e., it should not be back-fitted. That decision was based on the Consideration that the incremental improvements were not significant to safety and that full implementation of IEEE-323-1974 required the further development of other ancillary standards to provide guidance on specific safety-related equipment and components. Subsequent public comments and review by the ACRS did not alter the recommendation concerning implementation of Regulatory Guide 1.89.
1.'e iecognize frc, oer current licensing experience in implementing Regulctory Guide 1.89 and our participation in the development of IEEE-323-1974, that additional guidance is needed in the area of accelerated aging techniques used to establish a qualified life for electrical equipment and assemblies.
Our Category A technical activity on equipnent qualification (Task Action Plan A-24) and the NRC research program are intended to provide guidance for the development 2192 211 of test methods and licensing review procedures on aging. These efforts are described in our December 15, 1977 report.
In our current review of applications for operating licenses (0L) we do not require that the equipment qualification program include establishment of a qualified life for electrical equipment. However, applications for construction permits, filed since issuance of Regulatory Guide 1.89 in 1974, have been made subject to a requirement for documentation, at the OL stage of review, which shows that a qualified life for safety-related equipment has been established in accordance with IEEE-323-1974.
No. OLs have yet been issued for plants falling in this category. There continues to be a need for development of test methods in this area, hence the high priority of this subject in the NRC research program.
The staff will take into consideration the available information per-taining to the effects of age on the functional capability of safety-related equipment based on operating experience both in this initial safety assessment of the eleven SEP facilities and during our evaluation effort of these facilities. The operating experience of these eleven plants will be helpful in providing information in. support of the functional capability of the safety-related equipment in operating plants.
The licensee of the Yankee Rowe facility, for example, has indicated that the limiting environmental condition for equipment in controlled environments would be a loss of offsite power which results in a loss of ventilation.
This is because loss of ventilation typically increases 2192 212 the environmental temperature.
However, in 17 years of operation of the Yankee Rowe facility, this has occurred only once and lasted for only 20 minutes.
In addition, the loss of offsite power also trips all non-safety loads thus eliminating the primary heat sources and limiting the adverse environmental conditions. As of July, 1977, there have been a total of 44 loss-of-offsite power incidents, 18 of which have been of one hour or longer duration. The longest incident lasted one day at Dresden 1 following a tornado. The loss-of-offsite power incidents have occurred at approximately one-half of the reactor sites.
Based on the number and duration of loss-of-ventilation incidents experienced in con-trolled environments should not have been adversely affected by the higher temperature environment for the time periods involved.
In addition, the staff will assess the surveillance and maintenance records of the eleven SEP plants for equipment inside and outside of containment that has been effectively " aged" and could be subjected to extreme environmental conditions during an accident contion.
In addition to the initial qualification tests or analyses, operating experience with failures during surveillance and frequency of maintenance interval will provide an indication of the capability of the equipment to perform its function. This consideratior has not been factored into the 1nitial assessment due to the limited time available for both the licensees and staff. However, it will be pursued during our continuing evaluation. The 2192 20 continuing review will also include specific considerations of environmental qualifications with respect to the aging of cabling. This was recently identifiedU as a potential mechanism for degradation of cable relia-bility due to the aging effects of temperature and radiation level.
In addition, licensees of operating plants have replaced and added new safety-related equipment as the result of the continual development of nuclear power plant designs, modifications and associated licensing requirements to comply with NRC regulations.
Such replacement equip-ment was typically purchased to current industry standards and is evalu-ated in accordance with current NRC criteria and regulations to determine their acceptability.
4.3 Margin and Sequence Margin is an appropriate consideration in evaluating the qualification of electrical equipment because uncertainties and variations exist in the manufacturing and quality control of safety-related electrical equipment.
IEEE-323-1974 indicates that margin is the difference between U Memo from E. Case and E. Volgenau to S. Levine, " Aging Effects on Polyethelene Cabling", March 15, 1978.
2192 214 the most severe specified service conditions of the plant and the conditions used in type tests or analysis to account for normal variations in commercial production of equipment and reasonable errors in defining satisfactory performance.
IEEE-323-1974 further provides a specific sequence of testing, including aging, which defines the most severe sequence for most equipment; however, the sequence used shall be justified as the most severe for the item tested.
As indicated above, these items were considered during the development of Regulatory Guide 1.89 concerning the implementation requirements.
Both of these req 0irements are to provide additional assurance that the type test of a single unit or analysis with justification performed is adequate.
In evaluating the capability of the safety-related equipment to perform this safety function, the staff considered not only the specific information provided for each plant, but also the dacketed information available for the eleven SEP plants.
In addition, many of the components have been generically type-tested for a group of facilities, encompassing the worst case conditions, which provided margin for most plants.
2i92 215 The test envelope for specific type tests identified by the eleven SEP plants as the basis of their qualification was usually exceeded by the testing facility, thus providing margins. Generally, this margin results from the requirement to maintain a specific rise time, tempera-ture and pressure profile'. The licensee for the Ginna facility, for E to simulate a LOCA profile tempera-example, referenced type tests U
ture of 286 F at a pressure of 52 psi.
The actual peak temperat'ures, pressures and duration identified in the test data were:
329 F 90 psig 20 minutes (motor operated valves) 0 296 F 60 psig 20 minutes-(instruments and cables)
The type tests identified as the basis for qualification of equipment for the eleven SEP plants did not include protective mounting boxes or racks for most components and equipment tested due to test chamber size limitations.
In addition, for both type tests and analysis no credit was taken for the spacial location and protection afforded by walls, floors and other structures that exist within the containment.
Based on our initial safety assessment and based on the considerations identified in NUREG-0413, it is the staff's judgement that margin exists for such uncertainties and variations for the 8f Westinghouse Topical Report WCAP 7744 7410-L 2192 216 safety-related equipment for the eleven SEP plants a 1d other operating plants. We will continue to take into account the odequacy of the margin provided for the specific equipment in the eleven plants as part of our overall SEP effort.
- 4. 4 Qualification Methodology IEEE-323-1974 as augmented by Regulatory Guide 1.89 provides the current criteria, used as guidelines, for evaluating the adequacy of the quali-fication of safety-related electrical equipment.
IEEE-323-1974 indicates that qualification may be accomplished in several ways, such as type testing, operating experience, or by analyses based on appropriate test data. These may be used individually or in any combination depending upon the particular application.
In NUREG-0413, the staff identified the process and considerations that are typically' utilized in the environmental qualification. The report indicates the historical evolution of the environmental qualifi-cation requirements and identifies the procedures that are being used for the evaluation as part of the SEP in addition to the examples and details provided in this report.
4.5 Radiation Considerations The SEP licensees, with the exception of the licensees for Oyster Creek and Millstone, have addressed the subject of equipment qualifica-tion for a radiation environment. These two licensees are being requested to supplement their submittals to address this environmental parameter during the continuation of our evaluation.
2192 217 A number of assumptions need to be made to calculate the radiation environmental conditions inside the containment following a design basis event.E The conservative positions taken in Regulatory Guide 1.89 assume an ins +3ntaneous release and uniform distribution in the contain-ment atmosphere of 100% of the noble gases, 25% of the iodines, and 1%
of the solids in the reactor core. An additional 25% of the iodines are assumed to plate out on the containment walls. Using these assump-tions and conservatively assuming no removal of radioactivity by sprays or filters, the integrated 60 day LOCA dose to electrical equipment due 0
8 to gamma radiation ranges from 10 to 10 rads for the 11 SEP plants.
The dose due to beta radiation is of comparable magnitude. These estimates represent core degradation substantially beyond that which ECC systems are designed to prevent. With a minimum operation of the core cooling systems, as required by 10 CFR 50.46 and Appendix K, it is expected that the fission products released from such an event would be significantly lower than the assumed values (a factor of 10 to 100).
In addition, other assuruptions, such as uniform distribution and exposure calculations at the center of the containment with no shielding, represent a signifi-cant conservatism.
Therefore, a large safety margin exists in the level E
The LOCA is the DBE of primary concern from the radiation standpoint.
While the MSLB inside containment in PWRs may produce high temperatures and pressures, this event does not result in a significant radiation field. We have considered several other events and have concluded that the associated radiation levels are not likely to be more severe.
2192 218 of environment qualification for radiation due to the source term assumed in staff licensing reviews.
These conservative radiation dose levels should be put in context by relating them to potential material damage of most elastomers (rubber-like materials) which begin to exhibit damage when exposed to doses on the order of 10 rads or higher. Minor damage is initiated for certain materials at somewhat lower doses but not to the extent of functional impairment.
For most thermoplastic resins (e.g., polymers, cellullosics, 6
vinyls, etc.) mild damage becomes evident starting at about 10 rads.
Radiation damage to metals is not a concern at the levels considered here. As a result, it is expected that most of the equipment needed only during the initial phases following the LOCA would be able to perform its functions before any impairment by the radiation level. E Therefore, the primary concern is with equipment needed for long term accident mitigation (e.g., containment integrity and residual heat removal).
For this equipment, operators can often take appropriate action to assure adequate performance of such safety equipment.
E The one-hour integrated dose is conservatively estimated to be on 6
the order of 10 rads assuming a Regulatory Guide 1.89 release.
2192 219 The estimates of the radiation level and the data on equipment qualifi-cation provided in the licensees' submittals refer only to gama radiation. The effects of beta radiation have not been consistently addressed in past licensing reviews, but recent studiesb ndicate i
that the beta dose may be of comparable magnitude to the gamma dose.
Historically, beta radiation has not been explicitly considered in environmental qualification due to the inherent protection much of the equipment has to beta radiation. Specifically, any equipment in metal enclosures would not be directly exposed to the containment atmosphere and would not be affected due to the limited penetration ability of beta radiation. Similarly, plastic jacketing on cabling provides some pro-tection against beta radiation.
For example, a thickness of 0.2 cm of PVC outer jacketing will reduce the beta dose for 1 Mev betas by a factor of 100 and protect the inner insulation of indivudual wires.
Based on these facts, it appears that any equipment required to function only during the first hour following the LOCA will not be likely to 6
be exposed to doses in excess of 10 rads and will not have its safety function impaired.
For equipment needed for long-term accident miti-7 8
gation, a qualification to radiation doses ranging from 10 to 10 rads, Bonzon, L.
L., " Radiation Signature Following the Hypothesized LOCA", Sandia Laboratories, September,1977. SAND 76-0740. NUREG 76-6521.
2192 220 depending on the plant size, is acceptable.
For this equipment, however, the potential incremental effect of beta radiation on organic materials will be considered during the long term SEP evaluation.
Should this review indicate the need to alter our present licensing review practices, appropriate action will be recommended for the licensing process.
2192 221
APPENDIX A
.... ge UNITED STATES NUCLEAR REGULATORY COMMisslON s
WASHINCTON, D. C. 20555 o
f
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December 23, 1977 TO ALL SEP LICENSEES Our letter of December 1,1977, discussed your participation in the Systenatic Evaluation Program (SEP) far operating reactors and provided a report describing the pror}ran.
The progran includes the topic
" Environmental Qualification of Safety-Related Equipment" as one of the topics of safety significance.
The progran further pi vides that.opics considered to te of special safety significance would continue to be-evaluated on a case-by-case basis in advance of completing th. oprall program. The staff has determined that it is aporopriate to omolete its review of this subject as the first topic of the SEP.
- he enclosed " Staff Report on the Environmental Qualification of Safety-Related Electrical Equipment" (NUREG 413) supports tb staff position that no immediate action for your facility is required, but that it is appropriate te initiate the review of this topic. This document was prepared as part of the NRC staff's response of December 15,1977 to a Union of Concerned Scientists petition dated November 4,1977.
The report also provides the staff reviaw procedure, a oistussion of the historical evolution of regulatory criteria, current ceneric staff effort, a sumnary of relevant operating experience and other infomation relevant to this topic.
Therefore, pursuant to 50.54(f), please deliver to the Office cf Nuclear Reactor Regulation, Nuclear Regulatory Commission within 60 days from the date of this letter three signed oriainals and 40 copies of the following infomation:
2192 222
A-2 December 23, 1977 1.
Identification of safety-related electrical equipment located both inside and outside containment which are required to perform a safety function under the envirorcental condition resulting from each Design Basis Event (DBE) for your facility. The DBE is defined as any single event which could potentially result in greater than routine release of radioactivity from the site.
Briefly describe the safety function provided by each item of the equipment identified.
Describe the location of the equipment.
Identify any non-safety system, equipment or components, which, if subjected to the environnental conditions associated with a DBE, could affect the safety function of any safety-related system.
Identify non-safety systens which could perform the function of a safety system by ameliorating the consequences of a DBE and specify electrical equipment required to assure function of such non-safety systems.
2.
Definition of the limiting service environmental condicions for operating of the eeuf pnent and components identified above. The environmental parameters should include pressure, temperature, hunidity, subnergence, steam, raoiation, chemicals, vibration or any conbination of the above (seismic conditions are not to be included but will be considered elsewhere in the SEP).
These environnental conditions should be presented as a function of time and the DBE producing the conditions should be identified. The time period during which each item of equipment would be required to operate in a DBE environment should also be identified.
3.
Determination of the current status of environnental qualification for safety-related electrical equipment and identification of the supporting documentation. Any evidence by tests and analyses of environmental qualification for any environmental condition should be considered and provided.
Any information previously provided to the NRC that is still approoriate may be provided by reference.
We will be contacting you shortly to arrange a meeting to provide you with additional background information.
In addition, a visit of your facility will be scheduled in the near future to obtain information on safety-related electrical equipnent through direct onsite observation of these systens.
2192 223 9
A-3 December 23, 1977 Our safety assessment of this review topic will consider:
(1) ability of the facility to adequately respond tc all design basis events, (2) importance of the safety function and alternate ways of performing the safety function such as the possible use of non-safety system, (3) all available testing and qualification data, (4) any existing protection of the equipment f rom the environnent, or (5) any other bcsis for acceptability that you might identify.
Sincerely, Victor Stell{o, J[., Director
(
Division of Operating Reactors Office of Nuclear Reactor Regulation
Enclosure:
Staff Report cc w/ enclosure:
See next page 2192 224
B-1 APPENDIX B Main Steam Line Break Best Estimate Evaluation Method And Its Effect on Environmental Qualification The "best estimate" containment temperature calculation involved the reevaluation of three processes which affect the heat balance in the containment atmosphere during a MSLB. The processes are entrainment of liquid in the break effluent, condensing heat transfer, and heat sink condensate revaporization.
Liquid entrainment in the break effluent affects the containment tempera-ture because the expansion of entrained high pressure saturated water into contarment results in partial flashing of the water to steam and both water and steam will be in a saturated condition at the saturation temperature corresponding to the containment
- pressure.
But the expan-sion of high pressure dry saturated steam into containment results in superheating the steam since an essentially constant enthalpy throttling of the steam occurs.
Thus, for a given amount of energy removed via the break, the release of dry saturated steam develops a higher containment temperature than 'the release of a saturated liquid or a mixture of saturated steam and liquid.
2192 225
B-2 By examining the amount of liquid entrainment in the MSLB effluent as a function of break size, it was shown that the break size producing the greatest amount of superheating is in many cases significantly less than the largest possible break size.
The condensation of steam from the containment atmosphere on the passive heat sinks inside the containment removes energy from the containment.
This is identified as condensing heat transfer which is also affected by the MSLB size.
Large breaks result in turbulent convection heat transfer inside containment. Turbulent convection heat transfer is more effective in removing heat from the containment atmosphere than the natural convection heat transfer which dominates for the small breaks.
Condensate revaporization is the process whereby superheated steam is condensed on a heat sink and is then transported back into the super-heated environment where the liquid is revaporized.
The net effect of this process is to cool the containment atmosphere by re-moving the superheat energy needed to change the phase of the liquid to vapor. By considering experimental data, the staff determined that the turbulent flow characteristic of large MSLBs would result in a greater amount of condensate revaporization.
2192 226
B-3 By examining the sensitivity of containment atmosphere temperature to the three processes, it is possible to maximize the "best estimate" containment temperature response.
Figure 1 shows the temperature trans-ient for a typical Westinghouse PWR.
The sudden decrease in temperature at apprcximately 100 seconds shown in Figure 1 is caused by the initiation of containment spray with the resulting condensation of the superheated steam. Thus, the high temperature conditions in containment are termini.ted by the containment spray system.
The "best estimate" containment temperature was then used to perform thermal analyses of various typical safety-related electrical components.
The containment atmosphere temperature transient and heat transfer coefficients were varied to evaluate a best estimate component thermal response.
Since evaluation of component failure modes was not rigorously included in the "best estimate" study, the primary para-meter of interest was the component surface temperature.
The surface temperature was selected so that it could be compared with the actual thermal environment for which the component was qualified.
Past environmental qualification testing programs have typically maintained the peak test chamber temperature for a sufficient period of time to reach thennal equilibrium. Thus, the comparison of the analytical surface temperature peak with the test envelope provides an indication of component capability to funct'on in the containment environment predicted by the "best estimate" analysis.
2192 227
B-4 In evaluating the thermal response of electrical equipment, the staff also considered the effects of equipment location relative to the MSLB location.
It was concluded that it is reasonable to assume a 25 F temperature difference, as a minimum, between the area where the break is located and those locations separated by walls, floors, or large distances.
Temperature transients similar to a Westinghouse PWR shown in Figure 1 are predicted for Babcock & Wilcox and Combustion Engineering plants by an extension of the "best estimate" approach.
Since steam blowdown analyses for these plants predict significantly larger break sizes which result in pure steam release than that predicted by Westinghouse analyses, higher calculated containment atmosphere temperatures and, therefore, higher component temperatures are calculated for Babcock & Wilcox and Combustion Engineering.
This was considered in the "best estimate" evaluation.
Based on this. evaluation, it is concluded that:
1.
The maximum containment atmosphere temperature predicted by a realistic, but still conservative, method is, in many cases, significantly less than that predicted by previous methods (e.g.,
a decrease of approximately 80 F for a typical Westinghouse PWR).
2192 228
B-5 2.
Using "best estimate" methods, a heat transfer analysis indicates that the thermal response of electrical equipment will remain within the temperature range for which it has been qualified.
Although the "best estimate" method has been developed for PWRs and using PWR data, the method can be extrapolated to BWRs with dry con-tainments. This is possible because the method is based on a realistic treatment of generic thennodynamic processes to derive the containment temperature response and electrical equipment response to the high temperature short duration " spike". The method derives mainly from an assessment of liquid entrainment in the break effluent, heat trans-fer coefficients for containment heat sink, and condensate revapori-zation, whereas plant specific factors, such as reactor power rating, containment volume, and heat sink surface area, have a smaller effect on the temperature transient.
The initiation of the containment spray system is of major iniportance in tenninating the high temperature condition.
2192 229
C-1 APPENDIX C ESSENTIAL ELEMENTS OF PROPOSED CIRCULAR INTRODUCTION - PURPOSE 1.
Feedback of identified problems to licensees.
2.
Emphasize licensee responsibility to assure that electrical equipment in safety systems will function under predicted accident environment.
FOLLOWUP ACTIVITIES 1.
Licensee review of past problems outlined below to assure that an adequate basis for qualification of similar equipment exists and that corrective action is taken, if necessary.
2.
Augmented NRC Inspection program will focus on environmental qualification to assure that Class IE equipment is qualified for accident conditions.
LESSONS LEARNED 1.
In general, licensees did not have adequate documentation of the qualification of certain types of electrical equipment and initial responses of some licensees indicated a lack of detailed knowledge of the quality of installed equipment.
2.
Many problems resulted from inadequate consideration of "as installed" characteristics of equipment assemblies and their interfaces with other equipment. This was especially true for smaller, "off-the-2 shelf" components.
C-2 EXAMPLES OF DEFICIENCIES IDENTIFIED 1.
Connectors - IE Bulletin and responses 2.
Penetrations - IE Bulletin and responses 3.
Terminal Blocks - Deficiencies identified at Connecticut Yankee, Ginna, Rancho Seco, D. C. Cook and Palisades 4.
Limit Switches - IE Bulletin and responses 5.
Splices - Deficiencies identified at Monticello 6.
Other issues under staff review:
a) Solenoid Valves - (Big Rock Point and Palisades) b) Transmitters - (Palisades, North Anna and D. C. Look) c) Cables - (D. C. Cook) 2192 23I
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Table 2: P0TENTIAL ENVIRONMENTAL CONDITIONS OF SUBMERGENCE AND SPRAY FOR SAFETY RELATED ELECTRICAL EQUIPMENT FACILITY SUBMERGENCE CONTAINMENT SPRAY Big Rock Point yes/ water yes/ water Dresden 1 no yes/ water Dresden 2 no yes/ water Ginna yes/ sodium hydroxide, yes/ sodium hydroxide, boric acid boric acid Haddam Neck no yes/ boric acid Lacrosse yes/ sodium pentaborate yes/ sodium pentaborate Millstone 1 no yes/ water Oyster Creek no yes/ potassium chromate Palisades yes/ boric acid, sodium yes/ boric acid, sodium hydn xide, hydrazine hydroxide, hydrazine N
San Onofre yes/hydrazine, trisodium yes/hydrazine, trisodium
]
phosphate, boric acid phosphate, boric acid Yankee Rowe no no ro U
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J 2192 236