ML20056E526
| ML20056E526 | |
| Person / Time | |
|---|---|
| Issue date: | 06/25/1993 |
| From: | Remick NRC COMMISSION (OCM) |
| To: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| References | |
| NUDOCS 9308240190 | |
| Download: ML20056E526 (13) | |
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2n:1920 TO THE PDP NOTATION V0TEj f// /93 ci__,
5 RESPONSE SHEET
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SAMUEL J. CHILK, SECRETARY OF THE COMMISSION FROM:
C0114ISSIONER REMICK
SUBJECT:
SECY-93-087 - POLICY, TECHNICAL, AND LICENSING ISSUES PERTAINING TO EVOLUTIONARY AND ADVANCED LIGHT-WATER REACTOR (ALWR)
DESIGNS In N J zn A /
APPROVED Y DISAPPROVED X ABSTAIN NOT PARTICIPATING REQUEST DISCUSSION COMMENTS:
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Commissioner Remick's Comments on SECY 93-087:
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My vote, as indicated below addresses only the twenty items in SECY-93-087 for which the staff indicates it needs Commission i
decision at this time.
Comments on other items in SECY-93-087 will be provided separately.
l I approve, in part, and disapprove, in part, the staff's recommendations outlined in SECY 93-087, as supplemented in the Commission meeting on the subject paper, and as supplemented in the staff's letter to the ACRS dated May 19, 1993.
I also want i
to take this opportunity to commend the staff for a job well done on these highly complex technical issues.
1.
I.E.
Fire Protection:
I approve the staff's position that the passive plants should also be reviewed against the enhanced fire protection criteria approved in the Commission's SRM of June 26, 1990.
l I would like to be kept informed of the staff's resolution of the i
issue related to common-mode failures through common ventilation systems.
2.
I.F.
Intersystem Loss-of-Coolant Accident:
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I approve the staff's position that the passive plants should also be reviewed for compliance with the intersystem LOCA criteria approved in the Commission's SRM of June 26, 1990.
The staff should clarify the intent of the phrase "could not practically be designed to meet such a criterion."
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3.
I.G.
Hydrogen Control:
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I approve the staff position that the passive plants should be
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designed, as a minimum, to the same requirements applied to evolutionary designs.
Specifically, passive plants must:
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accommodate hydrogen generation equivalent to a 100% metal-water reaction of the fuel eladding; o
limit containment hydrogen concentration to no greater than t
10%; and provide containment-wide hydrogen control (such as igniters o
or inerting) for severe accidents.
1 i
I approve the staff's clarification, as expressed at the Commission briefing, that the possible use of passive autocatalytic hydrogen recombiners should not be precluded from
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consideration a priori.
The staff's resolution of the issue of the placement of hydrogen control devices should be provided to the ACRS and the Commission for infermation.
i 4.
I.H.
Core Debris Coolability:
I approve the staff's position that both the evolutionary and passive ALWR designs meet the following criteria:
o Provide reactor cavity floor space to enhance debris spreading.
o Provide a means to flood the reactor cavity to assist in the cooling process.
o Protect the containment liner and other structural members with concrete, if necessary.
o Ensure that the best estimate environmental conditions (pressure and temperature) resulting from core-concrete interactions do not exceed Service Level C for steel containments or Reactor Load Category for concrete containments, for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Ensure that the containment capability has margin to accommodate uncertainties in the environmental conditions from core-concrete interactions.
2 With regard to the 0.02m /Mwt reactor vessel cavity floor area, the staff should continue its research activities and supporting analyses, as documented in its May 19, 1993 letter to the ACRS.
With respect to the containment response to ex-reactor vessel core debris, the staff should not limit licensees to only one method for addressing containment responses to severe accident events but also permit other technically justified means for demonstrating adequate containment response.
5.
I.I.
High Pressure Core Melt Ejection:
I approve the staff's position for the general criteria that the evolutionary and passive LWR designs:
o provide a reliable depressurization system; and 2
i provide cavity design features to decrease the amount o
of ejected core debris that reaches the upper.
i containment.
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6.
I.J.
Containment Performance:
The recommendations on containment performance, as outlined in SECY 93-087, could be read to imply that the staff is no longer proposing to use the concept of conditional containment-failure probabilities (CCFP).
However, based on discussions held during the Commission meeting on this subject, the staff informed the Commission that it intends to continue to apply the 0.1 CCFP in implementing the Commission's defense in depth regulatory j
philosophy and the Commission's policy on Safety' Goals.
t Therefore, I approve the staff's position to use the following i
deterministic containment performance goal in the evaluation of the passive ALWRs as a complement to the CCFP approach approved by the Commission in its SRM of June 26, 1990:
"The containment should maintain its role as a reliable, i
leak-tight barrier (for example, by ensuring that containments stresses do not exceed ASME Service Level C i
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limits for metal containments, or Factored Load Category for concrete containments) approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of core damage under the more likely severe accident challenges and, following this period, the containment should continue to provide a barrier against the uncontrolled release of fission products."
I approve the staff's interim approach subject to the staff's review and recommendations resulting from public comments on the l
" Advance Notice of Proposed Rulemaking on Severe' Accident Plant Performance Criteria for Future ALWRs."
l 7.
I.K.
Dedicated Containment Vent Penetration:
i I approve the staff's position that the need for a containment vent for the passive plant designs should be evaluated on a I
design-specific basis.
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8.
I.L.
Eauipment Survivability:
I approve the' staff's position that the passive plant design features provided only for severe-accident mitigation need not be subject to the environmental qualification requirements of 10 CFR Section 50.49; quality assurance requirements-of 10 CFR Part 50, Appendix B; and redundancy / diversity requirements of 10 CFR Part 50, Appendix A.
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9.
I.M.
Elimination of Operatinc-Basis Earthquake (OBELL 4
I approve the staff's recommendation to account for earthquake cycles in the fatigue analyses of piping systems performed until the new guidance is issued, using two SSE events with 10 maximum stress cycles per event (20 full cycles of the maximum SSE stress range).
Alternatively, the number of fractional vibratory cycles equivalent to that of 20 full SSE vibratory cycles may be used l
(but with an amplitude not less than one-third of the maximum SSE amplitude) when derived in accordance with Appendix D of IEEE Standard 344-1987.
I approve the staff's recommendation that the effects of anchor i
i displacements in the piping caused by an SSE be considered with the Service Level D limit.
I i
I approve the staff recommendation to eliminate the OBE from the c
design of systems, structures, and components.
When the OBE is eliminated from the design, no replacement earthquake loading should be used to establish the postulated pipe rupture and leakage crack locations.
i I approve the staff's recommendation that the mechanistic pipe break and high-energy leakage crack locations determined by the d
l piping high stress (without the OBE) and fatigue locations may be l
j used for equipment environmental qualification and compartment pressurization purposes.
I approve that with the elimination of the OBE, two alternatives exist that will essentially maintain the requirements provided in IEEE Standard 344-1987 to qualify equipment with the equivalent i
a of five OBE events followed by one SSE event (with 10 maximum i
stress cycles per event).
Of these alternatives, the equipment should be qualified with five on-half SSE events followed by one full SSE event.
Alternatively, a number of fractional peak i
cycles equivalent to the maximum peak cycles for five one-half SSE events may be used in accordance with Appendix D_of IEEE 1
Standard 344-1987 when followed by one full SSE.
I approve that the above requirements should also apply to passive ALWRs.
It is my understanding that the OBE will continue to be used as a 1
threshold criterion for conducting inspections following an earthquake event.
The staff should keep the Commission and the ACRS informed as the staff's further analysis and review proceed.
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10.
II.A. Industry Codes and Standards:
I approve the staff's position that consistent with past practice, that staff will review both evolutionary and passive plant design applications using the newest codes and standards that have been endorsed by the NRC.
Unapproved revisions to codes and standards will be re/iewed on a case-by-case basis.
11.
II.D.
Leak Before Break:
I approve the staff's recommendation that the leak before break approach should be applied to both the evolutionary and the passive ALWRs seeking design certification under 10 CFR Part 52.
This approval should be limited to instances in which appropriate bounding limits are established using preliminary analysis results during the design certification phase and verified during the COL phase by performing the appropriate ITAAC.
12.
II.E.
Classification of Main Steamlines in Boiling Water Re_ actors.
I approve the staff's position that neither the main steam drain and bypass lines from the first valve up to the condenser inlet, nor the piping between the turbine stop valve and the turbine inlet should be classified as safety-related or as seismic Category I.
Rather, these lines should be analyzed using a dynamic seismic analysis to demot'trate structural integrity under SSE loading conditions.
The turbine stop, control, and bypass valves and the main steam lines from the turbine control valves to the turbine shall meet all of the quality group'and quality assurance guidelines specified in SRP Section 3.2.2, Appendix A.
Further, that seismic analyses be performed to ensure that the condenser anchorages and the piping inlet nozzle to the condenser are capable of maintaining their structural integrity during and after the SSE.
I approve that the above-described approach to resolve the main steamline classification for both evolutionary and passive ALWRs.
13.
II.F.
Tornado Design Basis:
I approve the staff's position that a maximum tornado wind speed of 482 km/hr (300 mph) be used in the design-basis tornado employed in the design of evolutionary and passive ALWRs.
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i 14.
II.H.
Containment Leak Rate Testing:
l I approve the staff's position that until the rule change proceedings for Appendix J of 10 CFR Part 50 are completed, the maximum interval between. Type C leakage rate tests for both evolutionary and passive plant designs should be 30 months, s
rather than the 24 months maximum interval currently required in Appendix J to 10 CFR Part 50.
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l 15.
II.I.
Post-Accident Sampling System:.
I approve the staff's position as modified below.
My approval is conditioned on my understanding of the staff's position, as described to me in meetings with the staff.
i I approve the staff's recommendation that the post-accident sampling systems for evolutionary and passive ALWRs of the pressurized water reactor type be required to have the capability to analyee determine the cross amount of dissolved gases and eh4er4de in cccordance aith-the requirements as an acceptable means of satisfying the intent of 10 CFR 50.34 (f) (2) (viii) and Item II.B.3 of NUREG-0737.
I approve that the time for taking these samples can be extended j
to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident.
i I approve that for evolutionary and passive ALWRs of the boiling water reactor type, there would be no need for the post-accident sampling system to analyze dissolved gases.
l I approve the deviation from the requirements of Item II.B.3. of l
NUREG-0737 with regard to requirements for sampling reactor coolant for boron concentration and radioactivity measurements using the post-accident sampling system in evolutionary and j
passive ALWRs.
The modified requirement would. require the capability to'take boron concentration samples and radioactivity measurements 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, respectively, following the accident.
i As a result of staff and industry briefings on the PASS system, I j
conclude that there is much confusion and miscommunication on j
this matter, especially as to the position that the staff is recommending.
My comments, explalation of modifications and the l
conditions to my approval are as follows.
The PASS system is a current regulatory requirement of 10 CFR
A PASS system has never been called upon for use.
It is most likely that a PASS system could not be justified under the " substantial increase" standard of the current backfit rule of 10 CFR 50.109.
Based on industry 6
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experience with existing PASS systems, the staff is recommending modifications to the current requirement as follows:
t 1.
The staff has previously exempted performing a hydrogen i
analysis of containment atmosphere samples and is recommending excluding this requirement for the advanced reactors.
However, hydrogen as a component of the gross amount of dissolved gases in the primary system would continue to be included as described below.
2.
The staff recommends excluding the determination of chloride concentration as this is a long-term corrosion consideration, not an accident mitigation consideration.
However, an error was made in including chlorides in the staff recommended position.
I have i
therefore modified the recommendation by removing the l
reference to chlorides.
3.
Some licensees have indicated that they have had up to three individuals on site for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day to meet the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> sampling requirement for boron concentration and radioactive isotope analyses.
The staff's recommendation would relax that requirement by
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requiring boron samples be taken 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after a severe accident event and that radioactivity measurements be i
taken 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after such an event.
4.
The staff believes that for certain accident scenarios, operators may be reluctant to depressurize PWR primary i
systems due to concern over contaminating the plant inside containment (the staff does not specifically i
identify such scenarios).
The staff therefore thinks i
that it would be prudent to be able to take samples of the pressurized primary system, especially to determine i'
whether dissolved non-condensible gases (including hydrogen) are contained in the primary coolant, which might interfere with subsequent heat removal capability.
It is my understanding that the staff is recommending that the PASS system for PWRs (only) be capable of taking a sample from the pressurized primary 1
system (not necessarily a pressurized sample) in a manner such that the gross amount of dissolved gases (including hydrogen) can be determined.
It is not intended that the sample be taken to a laboratory to i
analyze for specific dissolved gas components.
I have i
therefore modified the staff's recommendation l
consistent with this understanding of the staff's intent.
Resolution of this issue is further confounded by indications l
that the staff is considering eliminating some of the 7
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I requirements at issue.
For example, in NUREG/CR-4330, Review of Light Water Reactor Regulatory Requirements: Assessment of t
Selected Regulatory Requirements That May Have Marginal Importance To Risk, it is stated that i
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" Elimination of the Dissolved Gas Samples:
Item 2c of NUREG-0737 requires analysis of the reactor coolant sample l
for dissolved gases, in particular, hydrogen.
This provides
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information to determine the corrosive potential of the j
coolant and to infer the potential for in-vessel gas bubbles.
The formation of non-condensible bubbles in the reactor vessel has the potential to uncover the core and decrease the heat removal capability of the engineered safety features (ESF) systems.
However, other plant systems adequately remove the potential for non-condensible gases to interfere with core cooling.
The PASS sample requirement is i
redundant with these requirements and therefore is considered a candidate for elimination."
i.
Further, the staff's concern about operators not depressurizing the system appears inconsistent with the plant procedures for PWRs, which direct the reactor operator to the emergency i
procedures whenever the in-core thermocouples register greater than 1200 degrees Fahrenheit (e.g.,
Emergency Guideline F-0.2 for AP-600).
That procedure then directs the operator to FR-C.1 on RED Path and addresses when to manually activate the j
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depressurization system.
Therefore, advanced PWRs are not expected to find themselves in a high pressure condition with I
degraded core conditions.
d With respect to the argument that because BWRs have a suppression l
pool they do not contaminate the containment building as much as 4
PWRG, it should be noted that the advanced PWRs also incorporate 1
a type of suppression pool within containment, which is called by a different name (
i.e.,
in-containment refueling water storage tank).
4 Therefore, arriving at the decision to approve the staff's position as modified is a close call.
The staff indicates that l
a 1
its position is justified on prudence and not on risk arguments.
l I approve as indicated on the basis that the PASS system is an 2
existing requirement, out of respect for staff judgement, and on the belief that a relatively simple system can be designed to
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meet the modified requirement.
(It is my understanding that a system can be designed which is simple, does not require chemical analysis of the gases in solution, and will provide the reactor operator information as to whether significant amounts of non-condensible gases exist in the reactor coolant.)
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16.
II.N. Site-Specific Probabilistic Risk Assessments and Analysis of External Events I approve, in part, and disapprove, in part, the staff's position i
on site-specific probabilistic risk assessment and analysis of i
external events, as listed below.
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I approve the position that the analyses submitted in accordance with 10 CFR 52.47 should include an assessment of internal and external events.
l I disapprove the staff's recommendation to use two times the Design Basis SSE for margins-type assessment of seismic events.
l I
I approve the use of 1.67 times the Design Basis SSE for a
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margin-type assessment of seismic events.
l I approve the following staff's recommendation, as modified:
PRA insights will be used to support a margins-type assessment of seismic events.
A PRA-based seismic margins analysis will consider sequence-level High Confidence, Low l
Probability of Failures (HCLPFs) and fragilities for all
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sequences leading to core damage or containment failures up i
to approximately one and two-thirds the ground motion i
acceleration of the Design Basis SSE.
e I approve the staff's position that the simplified probabilistic I
methods, such as but not limited to EPRI's FIVE methodology, will 4
be used to evaluate fires.
-l I approve the staff's position that traditional probabilistic techniques should be used to evaluate internal floods.
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l I approve the staff's position that the ALWR vendors should f
f perform bounding analyses of site-specific external events likely 4
to be a challenge to the plant (such as river flooding, storm i
surge, tsunami, volcanism, high winds, and hurricanes).
i s
I approve the staff's position that when a site is chosen, its characteristics should be compared to those assumed in the bounding analyses to ensure that the site is enveloped.
l t
I approve the staff's position that if the site is enveloped, the l
COL applicant need not perform further PRA evaluations for these external events.
The COL applicant should perform site-specific l
PRA evaluations to address any site-specific hazards for which a
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bounding analysis was not performed or which are not enveloped by j
t the bounding analyses to ensure that no vulnerabilities due to citing exist.
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l The above modifications to the staff's recommendations are for two purposes.
First, it is the ratio of the ground motion acceleration of the Seismic Margin Earthquake (SME) to that of the Design Basis SSE that should be referred to and not the ratio of the magnitude of the SME to that of the Design Basis SSE.
Therefore, ground motion acceleration has been substituted for magnitude.
Second, it is the ratio of the ground motion acceleration of the Seismic Margin Earthquake to that of the site-specific SSE that is of importance for a specific plant.
Inasmuch that it is most likely that the site-specific SSE for most sites selected in the U.S.
for the advanced designs will be less than 0.3g (the Design l
Basis SSE), the ratio of ground motion acceleration of the SME
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used for the seismic margin analysis of the design (i.e.,
0.5g) will be more than one and two-thirds and perhaps more than twice F
that of the site specific SSE.
Further, the use of an SME ground h
motion acceleration of 0.5g will enable the designers to continue to use the safety margin equipment catalog compiled over a number of years.
For these reasons, I consider a ground motion acceleration of 0.5g for the SME adequate for seismic margin analyses 17.
II.O. Defense Against Common-Mode Failures in Digital Instrumentation and Control Systems:
I approve, in part, and disapprove, in part the staff's J
recommendation.
I approve the staff's revised position, as modified, that:
i 1.
The applicant shall assess the defense-in-depth and diversity of the proposed instrumentation and control j
system to demonstrate that vulnerabilities to common-
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node failures have adequately been addressed.
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2.
In performing the assessment, the vendor or applicant shall analyze each postulated common-mode failure for cach event that is evaluated in the accident analysis j
section of the safety analysis report (SAR) usina best-estimate probabilistic methods.
The vendor or applicant shall demonstrate adequate diversity within
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the design for each of these events.
3.
If a postulated common-mode failure could disable a
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safety function, then a diverse means, with a documented basis that the diverse means is unlikely to be subject to the same common-mode failure, shall be required to perform either the same function or a i
different function.
The diverse or different function 1
may be performed by a non-safety system if the system j
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is of sufficient auality to perform the necessary function under the associated event conditions.
4.
A set of safety--grede displays and controls located in the main control room shall be provided for manual, i
system-level actuation of critical safety functions and monitoring of parameters that support the safety functions.
The displays and controls shall be l
independent and diverse from the safety computer system
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identified in-items 1 and 3 above.
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I have modified the staff's position in essentially two respects:
First, inasmuch as common mode failures are beyond' design-basis I
events, I have clarified that the analysis of such events should be on a best-estimate probabilistic basis.
Second, the staff indicates in its discussion of the third part of its position that "The diverse or different function may be i
perferned by a non-safety system if the system is of sufficient quality to perform the necessary function under the associated
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event conditions."
Therefore, I have added this clarification to I
the fourth part of the staff's position (which refers to a subset of the safety functions referred to in the third part) by l
removing the safety grade requirement.
Further, the remainder of the discussion under the fourth part of the staff position is highly prescriptive and detail;d (e.g.,
"shall be evaluated,"
"shall be sufficient," shall be hardwired," etc.).
I approve only that such prescriptiveness be considered as general guidance, the practicality of which should be determined on a case-by-case basis.
18.
II.R. Steam Generator Tube Rupturtel II.R.1. Multiple Steam Generator Tube Ruptures:
I approve the staff's position to require that analysis of multiple steam generator tube ruptures (STGRs) involving two to five steam generator tubes be included in the application for design certification for the passive PWRs.
As discussed in the Commission meeting on this SECY paper,-'I approve the staff's recommendation that since the steam generator multi-tube rupture event is beyond the design basis requirements for FWRs, realistic or best-estimate analytical assumptions may be used to assess plant responses.
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II.R.2. Containment Evpass Potential Resultina From SGTRs:
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I approve the staff's recommendation that the applicant for design certification for a passive or evolutionary PWR assess design features to mitigate the amount of containment bypass leakage that could result from steam generator tube ruptures.
19.
II.T. Control Room Annunciator (Alarm) Reliability:
I approve the staff's recommendation that the alarm system for ALWRs should meet the applicable EPRI requirements for redundancy, independence, and separation.
In addition, alarms
-that are provided for manually controlled actions for which no automatic control is provided and that are required for the safety systems to accomplish their safety functions, shall meet t
the applicable requirements for Class 1E equipment and circuits.
20.
III.H.
Role of the Passive Plant Control Room Operator:
I approve the staff's recommendation that sufficient man-in-the-loop testing and evaluation must be performed.
In additica, a fully functional integrated control room prototype is likely to be necessary for passive plant control room designs to demonstrate that functions and tasks are properly integrated into j
the man / machine interface.
21.
Meeting with Industry:
j During the Commission meeting of 10 June 1993 with EPRI, the industry stated that several issues identified in the SECY paper i
required additional comment or clarification because they could be interpreted more than one way.
During my discussions with the i
staff, it became clear that in some cases, what the staff wanted and what the SECY paper stated were not consistent or not readily apparent.
I therefore recommend that the staff and industry i
representative meet to ensure a common understanding of the requirements such that industry's design activities are i
appropriately directed to comply with the requirements, i
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