ML20125D398
| ML20125D398 | |
| Person / Time | |
|---|---|
| Issue date: | 12/20/1977 |
| From: | Mattson R Office of Nuclear Reactor Regulation |
| To: | Case E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20125D401 | List: |
| References | |
| REF-GTECI-A-26, REF-GTECI-RV, TASK-A-26, TASK-OR NUDOCS 8001140278 | |
| Download: ML20125D398 (20) | |
Text
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, / o UNITED STATES f (,
n NUCLEAR REGULATORY COMMISslON r> /
[ WASHINGTON, D. C. 20555 r., -l 8'g'
%*****/ DEC 2 01977
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MEMORANDUM FOR: E. G. Case, Chairman, Regulatory Requirements Review Committee FROM: R. J. Mattson, Director, Division of Systems Safety, NRP
SUBJECT:
REQUEST FOR RRRC REVIEW 0F PROPOSED REACTOR SYSTEMS BRANCH TECHNICAL POSITION ON REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION ,
I request that RRRC consider a proposed Branch Technical Position on Reactor Coolant System Overpressure Protection of Pressurized Water Reactors While Operating at Low Temperatures.
The following material is provided for review:
- 1. Working Paper on Overpressure Protection;
- 2. Proposed Branch Technical Position on Overpressure Protection:
- 3. Estimate of earthquake contribution to overpressure events;
- 4. Summary of disposition of comments received; and,
The Divisions of Operating Reactors and Project Management have concurred in the Proposed Branch Technical Position; however, DPM recomendations on the implementation section have not been incorporated. DPM's recomendations have merit, and the attention of RRRC is invited to the November 17, 1977 memorandum from R. Boyd to R. Mattson which is included '
as the last item of Enclosure 5.
l The submittal and approval of this Branch Technical Position represents a milestone on the D0R Task Action Plan for Reactor Vessel Pressure Transient Protection (Overpressure Protection) for Category A Technical Activity A-26.
r ,3 GR ,
j Roger J. Mapson, hirector l
Division ofdSystemd Safety l
Enclosures:
As stated cc: See next page 90017126 Contact NRR:S. Newberry 8001140 Ext. 27591
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E. G. Case DEC 2 01977 cc: ONRR Division Directors ONRR Assistant Directors S. Hanauer F. Schroeder T. Ippolito S. Pawlicki D. Eisenhut R. Baer C. Berlinger G. Lanik T. Marsh W. Minners G. Zech T. Novak S. Israel G. Mazetis S. Newberry R. Ireland 90017127 i
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i Enclosure 1 WORKING PAPER ON RCS OVERPRESSURE PROTECTION DURING STARTUP AND SHUTDOWN FOR PRESSURIZED WATER REACTORS I.
SUMMARY
OF PROPOSAL A Reactor Systems Branch Technical Position on Overpressure Protection is proposed to define design and functional requirements for overpressure protection systems in pressurized water reactors to prevent the reactor coolant system (RCS) pressure from exceeding the pressure-temperature limitations on the reactor vessel for protection against brittle fracture, particularly when the RCS is water-solid during startup or shutdown conditions.
It is proposed that the Branch Technical Position be appended to Standard Review Plan (SRP) 5.2.2 on Overpressure Protection as soon as it is approved. Modifications to the text of SRP 5.2.2, as appropriate to delineation of corresponding review responsibilities of the Reactor Systems and Materials Engineering Branches, would be made at a later time in connection with a general updatina of SRP 5.2.2.
II. BACKGROUND The general design criteria (Appendix A to 10 CFR 50) specify in General Design Criterion (GDC) 15 regarding the reactor coolant system design that:
"The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences."
Appendix A of 10 CFR 50 contains the following definition:
" Anticipated operational occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power."
The pressure-temperature limits imposed on the reactor coolant pressure boundary (RCPB) during operation and tests are provided to assure adequate safety margins against nonductile behavior or rapidly propagating failure of components of the RCPB, as required by GDC 31.
Appendix G of 10 CFR 50 describes the conditions that require pressure-temperature limits and provides the bases for these limits. Appendix G specifies minimum fracture toughness requirements for ferritic mate ialsary.
of pressure-retaining components of the reactor coolant pressure bo 90017128 .
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It requires that pressure-temperature limits must provide safety margins at least as great as those recommended in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G, " Protection Against Non-ductile Failure," during heatup, cooldown, and test conditions. Appendix G of the ASME Code provides a basis for determining allowable pressure-temperature relationships.
The proposed Branch Technical _ Position makes it clear that the above criterion and definition apply not only to power operation but also to startup or shutdown conditions where inadvertent overpressurization might lead to violation of pressure limits imposed on the reactor coolant boundary for protection against brittle fracture at low temperatures. It also requires that a pressure relief system be provided that will automatically give overpressure protection for such conditions.
Our review of recent operating experience indicates that there have been a number of reported incidents of reactor vessel overpressurization in pressurized water reactors in which the technical specification pressure-temperature limits have been exceeded. The majority of cases have occurred during reactor startup or shutdown when the reactor coolant system has been in a water-solid condition. These over-pressurization incidents have been initiated by a variety of causes, including the following:
(1) Isolation of the residual heat removal system or the letdown system while continuing to charge to the reactor coolant system.
(2) Thennal expansion, followinq the starting of a reactor coolant pump, due to transfer of thermal energy stored in the secondary system, through the steam generators, to the reactor coolant system.
(3) Initiation of operation of a reactor coolant pump or a high pressure safety injection pump. ,
(4) Inadvertent actuation of safety injection accumulators.
In essentially all of the events reported, a single personnel error, equipment malfunction or procedural deficiency has been sufficient to cause the event. The frequency of these events has been considerably higher than anticipated at the time the affected plants were reviewed.
A summary of the reported overpressurization incidents can be found in NUREG-0138 " Staff Discussion of Fifteen Technical Issues. . ."
(Issue 15).
III. PROPOSED POSITION The objective of the proposed Branch Technical Position is to establish definitive design and functional requirements for over-pze -- r protection systems which are installed in pressurized water reactors for the express purpose of assuring that 10 CFR 50 Appendix G reactor 90017129
coolant system pressure limits are not exceeded for all plant conditions under which Appendix G limits apply. The full text of the proposed Branch Technical Position is included in Enclosure 2. The elements of the proposed position are summarized below:
(1) A system should be installed which will be capable of relieving pressure at a rate sufficient to satisfy Technical Specification limits derived from requirements of Appendix G, particularly when the reactor coolant system is in a water-solid condition.
(2) The relief system must be able to perform its function following the initiation of a potential overpressure condition, assumino any single active failure of fluid system components which must operate in conjunction with providing overpressure relief.
Failures which may initiate an overpressure condition will not be considered as the single active failure, and potential initiating failures may not be ruled out by technical specifications or other administrative controls. Analyses must demonstrate that adequate relief capacity is provided for the worst case event.
(3) The system must operate automatically - without dependence on manual actions to enable the system to provide its pressure relief function.
(4) To assure system readiness:
- Test electronics prior to each shutdown;
- Test valves as specified in ASME Code Section XI: and,
- Test affected portions following maintenance.
(5) The system must meet the design requirements of IEEE-279, Regulatory Guide 1.26 and Section III of the ASME Code.
(6) The system need not be designed to meet Seismic Category I requirements if it can be demonstrated that an earthquake of a magnitudeyquivalenttotheSSEwouldnotinitiateanoverpressure In any case, the system design must not compromise transient.
the design of any safety-grade systems with which it interfaces and ;
the requirements of Regulatory Guide 1.29 must be satisfied. !
l (7) The system must not depend on the availability of offsite power to carry out its function, l
I The practical effect of the proposed position will likely be a requirement to design to Seismic Category I requirements. Opinion is divided on whether Seismic Category I requirements should be mandatory in view of the apparent small contribution seismic events would have on the overall frequency of overpressure events. Enclosure 3 provides further discussion.
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' (8) The effects of inadvertent system actuation should be analyzed, unless it can be shown that such effects are bounded by other analyses.
IV. VALUE ASSESSMENT The pressure-temperature limits imposed on the reactor coolant system assure adequate safety margins against nonductile or rapidly propagating failure of the reactor coolant system. The inclusion of system desian provisions to mitigate the potential inadvertent overpressurization transient provides reasonable assurance that nonductile or rapidly propagating failure will not occur. Since operating experience has shown that these overpressurization incidents have occurred'with a greater frequency than originally anticipated, it is important to
- safety to require that steps be taken to minimize ocer rences of overpressurization.
The proposed Branch Technical Position provides definitive guidance with respect to design and functional requirements for pressure relief systems needed to prevent Appendix G limits from beina exceeded.
It has already been determined that some pressure relief capability must be provided in pressurized water reactors to minimize the potential for reactor coolant system damage from inadvertent overpressure in temperature ranges where Appendix G limits apply. Promulgation of the Branch Position will remove the ad hoc nature of case-by-case review now taking place on the basis of other criteria.
V. IMPACT ASSESSMENT The proposed position applies only to pressurized water reactors since the design configurations of PWRs during normal plant operating modes during startup and shutdown are vulnerable to those types of failures which can lead to overpressurization incidents. BWRs are not brought into a water-solid condition.
Plants which receive an operating license prior to promulgation of the proposed Branch Position, will be required to commit to installation of a " satisfactory" overpressure protection system. These plants will not require back-fit to meet the Branch Position, but must meet criteria l J
which have been agreed to on a case-by-case basis. Thus, these plants are not affected by the Branch Position.
Plants which are expected to receive an operating license in the 12-month period following promulgation of the Branch Position will be required to commit to installation of a pressure relief system meeting the requirements of the Branch Position (with the exception of reasonable deviations from IEEE-279) prior to startup following the i first refueling outage.
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Schedule impact on these plants should be negligible except for the possibility of a somewhat prolonged outage during the first refueling.
Plants receiving an OL beyond this 12-month period will be required to install the pressure relief system prior to plant startup. For these plants there should be no schedule impact, since ample time would be available for design and installation of a satisfactory system.
Because of the variety of approaches available to designers in meeting the proposed Branch Position, the dollar cost of design and installation is difficult to judge. However, Consolidated Edison is installing systems in Indian Point, Units 2 and 3 which essentially meet the proposed requirements. According to Con-Ed, the total engineering and installation cost will be about one million dollars per unit. How much of this cost is attributable to "back fit" is not known; however, it should bound the cost for a typical plant still under construction.
VI. PROPOSED IMPLEMENTATION PLANS It is proposed that the Branch Technical Position be implemented as follows:
(1) All aspects of the Branch Technical Position will be applicable to CP and PDA applications docketed after the date of promulgation.
(2) CP and OL applications docketed before the date of promulgation and for which licensing action is pending will be required to meet the requirements of the Branch Technical Position with the exception of reasonable deviations from IEEE-279.
(3) Applicants receiving an OL in the 12-month period followina promulgation will be required to commit to installation of a satisfactory system no later than the first refueling outane.
For these plants DSS will obtain the commitment and provide assurance that the reactor can be operated throuah the first cycle without endangering the integrity of the reactor coolant system if pressure reaches the safety relief settings while the coolant is at ambient temperature. In addition, DSS will examine administrative controls to assure that the likelihood of an overpressure transient is minimized.
(4) Applicants receiving an OL 12 months or later from the date of promulgation of the Branch Technical Position will be required to install a satisfactory pressure relief system prior to plant startup.
(.5) The Branch Technical Position will not be applied to plants which have received an OL prior to its promulgation. DSS believes that the criteria used by DSS prior to promulgation of this position and those used by D0R on operating reactors are satisfactory.
Therefore, backfitting of these plants is not intended. The acceptance criteria for operating reactors are addressed as part of
,, the Task Action Plan for Reactor Vessel Pressure Transient
! Protection (Overpressure Protection) for Category A Technical Activity A-26, l 90017132 o
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-- . The Reactor Systems Branch will take lead responsibility in incorporatinn ,
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the proposed Branch Technical Position into the review process. The j Reactor Systems Branch will .also take steps to assure that proper coordination occurs with the instrumentation and Control Systems, Mechanic.a1 Engineering and Materials Engineerina Branches pending update of the review procedures in SRP 5.2.2.
VII. COORDINATION 1
The Reactor Systems Branch solicited staff comments in September 1976, prior to drafting the Branch Technical Position on Overpressure Protection. In April 1977 a " final" draft was circulated to DSS and I DOR for comment. The disposition of the comments and recommendations received in response to these requests is summarized in Enclosure 4. l Conrnents by Robert L. Baer in his October 14, 1977 memorandum to Darrel G. Eisenhut, DOR have been incorporated in the workinn paper and in Section B.2 of the proposed Branch Technical Position; D0R l has concurred in the proposed Branch Technical Position.
In a memorandum from Roger S. Boyd to Roger J. Mattson, dated November 17, 1977, DPM generally concurred in the proposed Branch Technical Position. However, DPM recommended specific changes in the implementation section to cover applications using the replication and reference design options of the standardization programs and to provide more explicit guidance on how the position can be imposed on applicants with pending near term licensing decisions. These recommendations have not been incorporated in the proposed Branch Technical Position. The DPM memorandum is included in Enclosure 5.
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I ENCLOSURE 2 1
l PROPOSED REACTOR SYSTEMS BRANCH TECHNICAL POSITION OVERPRESSURIZATION PROTECTION OF PRESSURIZED WATER REACTORS WHILE OPERATING AT LOW TEMPERATURES l
A. Background General Design Criterion 15 of Appendix A,10 CFR 30, requires that "the Reactor Conlant System and associated auxiliarv, control, and ?rotection systems shC1 be designed with su"icient maroin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of *cmial operation, includinc anticipated operational occurrences."
Anticipated operational oCCJfrdnCe5 85 defined in Appendix A of 10 CFF 50 1 are "those conditions of nc + ~ 'Deration which are expected to occur one or more tiMc . Gring the i %(r4o ruclear cowr unit 7 d include but ar" not limited to loss of powe s at i recirce W on >u- t, t ipp'rq of the turbine generator set. isola ica af the 6 (qdensee, n! loss of 6 o'? site power."
ippenuix G of 10 CFR 50 pr:,s . des t e v. ui
- out.oness requirements for reactor pressure vessels under all conc lions. Tn assure t!.at the Appendix G limits of the re-tor coolant are:' are bound +.r / are not c>ceeded during any anticipted 00" '-
'nal accurrences. lechnn al Specificatior aressure-temperttu- an ts arc ?rreinec for t erating the pl ant .
The primary concern is that dur u, Ltartup vd shutdcwn ct " .tions at low temperature, especially i. a water-solia condition, the reactor c ool ar ' system pressure might evceed the reactor essel pressure- ;
teu 'erature limitations in the Technical Specificat'c w established for !
protection against brittle ir,ctere This ina6nrter.t nerpressurizatio" could Se gm. crated by any ona f a .ariety of malf unctions or operator err ~ r. Many incidents have eccerid in operatina plants as described in Reference 1.
This position uses Appendix G .s the limit for all postulated initiating cuents, regardless of the probability of that event or combination of events occurring. The probability of a safe shutdown earthquake witn a resultant overpressurization event is less than that of an overpressurization event caused by an anticipated operational occurrence and therefore the application of the criteria of Appendix G to these less probable events may represent a high degree of conservatism. However, there is currently no less conservative vessel brittle fracture limit within the NRC; therefore, this position does not attempt to define any additional criteria less limiting than Appendix G.
Additional discussion on the background of this position is contained in Reference 1.
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l B. . Branch Posit' ion j I
- 1. A system should be designed and installed which will prevent exceeding the applicable Technical Specifications and Appendix G {
-limits for the reactor coolant system while operating at low temperatures. The system should be capable of relieving pressure during all potential overpressurization events at a rate sufficient to satisfy the Technical Specification limits,. particularly while ]
the Reactor Coolant System is in a water-solid condition.
- 2. The system must be able to nerform its function assumino any sinale l active component faihere. An:1"ses using appropriate calculational techniques must be provided which demonstrate that the system will provide the required pressure relief capacity assuming the most !
limiting single active failure. The cause for initiation of the event; e.g. , operator error, armonent malfunction, will not be l considered as the single active sailure. The analysis should assume -i the most limiting allowabic 'oerating conditione and systems j configuration at the tine of the postulated cause of the ovarpressure i event. All potential :verpressurization everte must be corsidered when establishing tne wors* m se event. Pote ial event r,ay not be eliminated from coisiner m on in overpressure protectian system de ign analyses mere'y by tse imposit on of technical spedfications or other administrative controls (e.n , prchit,itions on sa'ety injection pump operation).
- 5. The system must operate autora+ + y, providino a completely independent backup prote.tive " ature for he "operetor. Thc. design must not require manual activas to enabic or turn on the system or to mitigate the consequences f a potential overpressure event.
- 4. To assure operational readiness, the overpressure protection systero must be tested in the foll, wing manner:
- a. A test must be performed to assure operability of the system elcctronics prior to each shutdown,
- b. A test for valve operability must, as a minimun be conducted as specified in the ASME Code Section XI.
- c. Subsequent to system, valve, or electronics maintenance, a test on that portion (s) of the system must be performed prior to declaring the system operational.
- 5. The system must meet the desian requirements of IEEE-279, Reaulatory Guide 1.26, and Section III of the ASME Code.
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- 6. The-protection system need not meet Seismic Category I requirements if it can be shown that an earthquake of a magnitude equivalent to the SSE would not initiate an overpressure transient. If the earthquake can initiate an overpressure transient, the protec. tion system ipould be' designed to Seismic Category I requirements. Should the applicant show that a postulated earthquake could not cause an overpressure event, l
the overpressure protection system design must not compromise the l design criteria _ of any other safety-grade system with which it would interface. The requirements of Regulatory Guide 1.29 must be satisfied.
- 7. The overpressure protection systen must not depend on the availability of offsite power to perform its function.
- 8. Overpressure protection systeme which take credit for an active component (s) to mitigate the consequences of an overpressurization event must include additional analyses considering inadvertent system initiation / actuation or provide justification to show that existing analyses bound such an event.
C. Implerentation The Branch Position in paracraph 8 will be uted in the review of all PDA, CP, and OL applications fc cressurized mtec reactors. All CP and PDA applications docketed af ter (date of promigation) must meet the requirements of IEEE-279. Applicants for CPL and OLs docketed before (date of promulgation) will be Clowed to justify reasonable deviations from the requirements of IEEE-279. Those apolicarts issued an operatina license during the twelve-montL per.cJ followina (date of prom,lgation) must commit to installation of all equipment no later than the first refueling outage. Those applicants receiving an operatino license beyond this twelve-month period must install all equipment prior to plant startup.
D. References
- 1. NUREG-0138, Staff Discussion of Fifteen Technical Issues listed in Attachment to November 3,1976 memorandum from Director, NRR, to NRR staff.
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'hhD hD Enclosure 3 j BOUNDING ESTIMATE OF EARTHQUAKE CONTRIBUTION i
TO OVERPRESSURE EVENTS POTENTIALLY VIO'ATING APPENDIX G LIMITS
- The probability of a seismic event generating an overpressure event potentially causing a violation of Appendix G while shutdown at low temperatures can be calculated as follows:
- 1. The probability of an earthquake greater than.0.lg as sta, i in WASH-1400 for an average sitn in the area East of the Rocky Mountains is about 4x10-3/yr.
- 2. The extremely conservative assumption can be made that every earthquake greater than 0.19 will cau" an overpressure ev W e ile shutdown at low temperatures; i .e. , P (overpressure) = 1.0 if t.e p' ant is in cold shutdown at the time of the event.
- 3. The time a plant is ic.utdown a' low temperstures, wt 'ch % tne condition
~." corsa~ r'isely _
required for such overpruturt evente to o c c ..
assumed to be ten day 't.ach :/cer, or a projability ( " c. ci 3xio "
tnat the plant will be vulnerabia to overp ms /re.
On this basis the frequency of overpres e e : vents induced by earth-quakes is conservatively estimated to bt 4x10
-3 3, 1,3 x xO u ~ - ._. f erents .
yr.
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yr.
To date thirty-one events violating the '\ppendix .i it have Seen reported. From this operating plant data it can be conse events have occurred at the rate of 2x10 yvativeij thown that the overp
/ year.
From these frequencies, the percentage contribution cf seimically induced overpressure events to the total occurrence rate woulo be:
1.2x10-4 x 100 - 0.06%
(2x10-I) + 1.2x10-4)
It can be postulated that for a seismic event greater than 0.19, the plant would have to be shutdown for inspection / repair for a prolonged period and that it would be continuously vulnerable to overpressure during this time. This presumes that a relief system, if provided, has been damaged, is not repaired and the coolant system is exceptionally vulnerable to being put in a water-solid condition during the entire period. In this case a bounding estimate of the percentane contribution of seismic events to the total can be made by assuming that every seismic event greater than 0.1g has a probability of 1.0 of causina an overpressure event; i .e. ,
4x10-3 x 100 ; 2.0%
(2x10-l) + (4x10~3) 1 B001
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2 j Therefore, it can be concluded that the overall contribution to the frequency of-overpressure events from a postulated seismic event is not significant and i need not be considered.
The above probabilistic approach can be at odds with General Design Criteria 2 1 which states that. 1 i
" structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes..."
Since it is possible to postulate that a seismic event could generate an overpressure event while shutdown at low temperatures, it can be aroued that the system (s) designed to mitigate the consequences of such an-event should be designed to withstand the effects of the seismic event, no matter how low ,
the probability. Unfortunately, since at present we are judging the acceptability ~ !
of all overpressure transients regardless of their anticipated frequence of arrival against Appendix G limits, these very low probability events are also required to meet'the same limits as anticipated transients.
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. Enclosure 4
SUMMARY
oOfDISPOSITIONOFCOMMENTS The Reactor Systems. Branch' solicited staff' comments prior to the draftinc of a Branch Technical Position'on Overpressure Protection During Startup and Shutdown (Reference 1) and, subsepently, on the final draft version [
(Reference 5). The' comments received are discu<*d below.
Reactor Safety Franch DnB (Memo dtd 10_Q Clfd, The positions in the memorar.dum were inaorporated into tt, araft.wir.h the following exceptions.
- 1. Operator Action /Autamatic Enabie: The Rea t m Safet/ 9 r a n a'1, DDR, allows operator actior, an ".
- o n : e7@li n 7 Sineo spr"oxi'1a t elv two-thirds of the overio ." J.s .i f E m >uarred era-caused by operator error, Reactor '3ysttu Branch tesirea to remove allowance for operat..' n.3tta) after 10 miqutae, from the system design. In addition, Renator Syste13 3rech halieven
+,o it prudeqt for the overpressure protection systea he folly automatic. The requirerieit to ex lade the oper3 tar from the proteati m sequ mre .'.3 furthea substantiated by in ever t 3.1 an operating plant in which the operator forgot t > eoele th, installed low temperature overpru aure protection systro. !
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- 2. Testing: Reactor Systeas Branch ooncurs with the recomianndativi of MEB (below) that the safety /relier valves used in the overpressuen !
protectian syst<rs be test ai a s ne picei Sy the ASME Code. A full system functional test cach refueling oJtage in not a Code re.pice i- t )
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Mechanical Encineerine Branch. DSS (Memo dtd 9/29/76)
- 1. Any pressure relief valve installed for the specific purpose of providing protection against exceeding Appendix G limit be designed and fabricated to ASME Code Section III Class 1 "equirements.
This. requirement is included 1r the BTP in that the applicant must comply with Section III of the ASME Code.
Since the ASME Code '.a now in + ' e process of being changed to include this requiremer.t, we agree with MEB that protection systems 5hould be Class 1. )
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n reac. , f ot w' fr-qu icv any
- 2. 11EB roes not fec1 t here ocau-izer different from what dc r ?cified in AS"a. S 2t ion XI fc- t: :
safety valves. 10 CFR 50.55a(-) r muires that e.1h.- sr .he na valve types be tested in accordanec win.Section XI.
This recommendation has been incorporar."1 tr'.9 .ha Branch Position.
Valve functional requirement 9: Detailed desien raquirements or 3
the ASME Code for liquid relief valves are net intended to be part of the Branch Technical Position. Reactor Systems Br4nch recognizes the importance of the ASME Code requirements and will therefore review each design on plant-specific design submittals.
l Auxiliary Systems Branch. DSS (Memo dtd 10/14/7f.)
- 1. System modifications or combinations thereof to prevent overpres-surization transients should be designed to Seismic Category I
. requirements.
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' Reactor Systems Branch does not agree with this. requirement
.as' stated in paragraph 6 of-the final BTP and Reference 5.
The BTP does' require that (he system meet the requirements of Reg. Guide 1.29 should the system be a part of the reactor coolant pressure boundary.
- 2. Three ' cases of system modifications and the associated quality r
group requirements of these desigre 3re discussed in Reference 3..
Reactor Systems Branch cont.;rs with the requirement of this recommendation for Case (1) and (2). Howeve , the ASME C-de will require that the protaction sy.mtem be'da91rned to ASMF Section III, Class 1 requirtments, ever _r art of a ..as? II system. These requirements wern included in the bTP. .
Peactor Safety Brar;h d OF (Meetine - 5/10/77.1
- 1. Allrwing no operator action to enable the protection system (para. 3) yet allowing applicants to prevent initiating eve er th ouch administrative controls is not consistent (para. 9).
Paragraph 9 was deleted from the branch position such that each applicant must consider ard fully analyze each initiating event.
2._ Would the staff allow Appendix G to be exceeded for short periods due to pressure overshoots caused by relief / safety valve setpoints
.and initial operating conditions 7 i
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Paragraph 1 of the position does not allow exceeding Appendix G for short periods. Realizing the constraint of the conservatisms designed into Appendix G, no guideline presently exists which allows exceeding this limit under any circumstances.
3 How much of the protection system will be tested? At what frequency?
Paragraph 4 of the BTP has been revised in response to this comment. The testing requirements are consistent 1
with the ASME Code Section XI.
)
- 4. The IEEE-279 requirements are not specific.
Paragraph 5 of the BTP has been revised in resnonse to this comment.
l 5 Provide further discussion with respect to the sei?cic requirem nts J
of the system.
This position states that the overpressure protee' ion system must be Seismic Category I unless the applicant can show .
that an overpressure event will not be caused by an SSE.
RSB realizer,.the low probabilities of such events and the associated difficulty with the application of the Appendix G criteria; however, no other limit has been established. RSB recommended in Reference 5 that applicants be permitted to propose an appropriate pressure-temperature limit curve for'these lower probability events. The acceptability of this limit would be reviewed by MTEB.
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Instrumentation'and Control Systems Brangh. DSS (Meetinz - 5/11/77)
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- 1. The position should require the identification of sytems/ components that could cause an overpressure event assuming an~ operator error-or single failure unless the overpressure protection system precludes this event from exceeding acceptable limits 'no matter what failure or error initiates the event.
Paragraphs 1 and 2 of.the position require that the protection system be designed tr. litigate the ecnsequences ,
of all potential overpressure events assuming the most. ,
i limiting single' failure and allowable operating conditions.
The applicant will be required to justify tnat such failures and operating conditior.a are the wor?t case.
- 2. The E, I & C should meet all IEEE-279 requirements ecpecially seismic and environnental qualifientions.
The implementation requirements of the position were revised to require all CP applications docketed after January 1, 1978 be designed to neet IEEE-27').
The requirements for. seismic systems are discussed in previous comments by RSB, DOR, paragraph 6 of the BTP, and the cover letter forwarding this BTP.
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-3 Paragraph 9, which gives;the applicant the. opportunity to rule out l
certain overpressure events through the use of administrative controls.,
is not consistent with the requirement for automatic enabling.
RSB concurs with this comment. Paragraph 9 was deleted from the position.
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- 4. A problem exists with interfacing a potentially nonsafety-grade overpressure protection system with nonsafety-grade systems.
The change in IEEE-279 requirements (above) will. minimize ,
i the effects of this problem. The review of OL plant over-pressure protection systems will require that'no safety-grade system is compromised by an interfacing cystem of ,
lower quality. !
- 5. The BTP is not clear in stating tha' 5 ,verrre'sure protection system must operate independently from the initiating event.
The position has been restated to reinforce this ?nnect>t.
- 6. ICSB desires the automatic enable pet tic'. of t he sols t em .
This portion of the position has aean restated to state this requirement more clearly.
Mechanical Engineering Branch (Memo dte 5/9/771
- 1. The " UPSET" definition in the A3ME Code has been deleted and the terminology changed to " Service Limit B". ,
This position was changed to incorporate this comment.
- 2. "High quality equipment" in paragraph 5 should read " Quality Group A equipment".
This matter is addressed in MEB comment #1 (9/26/76). The quality requirements for the overpressure protection system must be such that the design of the connecting system (s) is not down graded.
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References
- 1. Memorandum from T. M. Novak to Distribution dated September 7, 1976,
- 2. Memorandum from R. Baer to T. Novak dated October 7, 1976.
3 Memorandum from Robert Kirkwood to Thomas. M. Novak dated October 14, 1976.
- 4. Memorandum from R. J. Bosnak to T. N'ovak dated September 29, 2976.
- 5. Memorandum from T. M. Novak to Distriaution dated April 29, 1977.
~; . Mencrar.dum from R. J. Bosnak to T. M. Novak dv i av 9, ' T 7.
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