ML20058P828
| ML20058P828 | |
| Person / Time | |
|---|---|
| Issue date: | 08/15/1990 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Kintner E ALWR UTILITY STEERING COMMITTEE |
| References | |
| PROJECT-669A NUDOCS 9008200128 | |
| Download: ML20058P828 (8) | |
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O Project No. 669 August 15, 1990 lir. E. E. Kintner, Chairman ALWR Utility Steering Committee GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054
Dear Mr. Kintner:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION AND C0fEENTS ON EPRI'S ALWR REQUIREMENTS DOCUMENT As a result of our review of the EPRI ALWR Requirements Document, the staff has determined that it needs additional information in order to complete our review of the design criteria. The additional information is needed on EPRI's proposed resolutions for Generic Safety Issues and Unresolved Safety Issues. Our coments and concerns are discussed in the enclosure to this letter.
Please respond to this request within 60 days of the date of this letter.
If you have any questions regarding this matter, call me at (301) 492-1120.
Sincerely, Original signed by T. Kenyon Thomas J. Kenyon, Project Manager Standardization Project Directorate Division of Reactor Projects - III, IV, V, and Special Projects Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
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ENCLOSURE Connents Regarding USIs and G!s Addressed in Appendix 3 to Chapter 10 3-47
" Safety Implications of Control Systems" -
(Resolved by GL-89-19, 9/20/89)
The resolution of USI A-47 is stated in Generic Letter (GL) 89-19, which should be acknowledged and referenced by the EPRI discussion.
We note that many of the requirements discussed in the EPRI-ALWR Requirements Document's Chapter 10, Section B.1 (Control System Tailures) are not directly related to GL-89-19, and are therefore not appropriate for discussion in this section.
The EPRI discussion should be re-directed to address GL-89-19, which contains dif ferent requirements for dif ferent plant designs.
- Thus, not all of the letter's requirements will necessarily be applicable to the IPRI design, and so EPRI should state which parts are applicable and how their design is in conformance with those parts of the generic letter.
Absent such a discussion, we find that Section B.1 is too general for us to reach a finding regarding the EPRI design's conformance to NRC's requirements resulting from the resolution of USI A-47.
GI-10I
The resolution of GI-101-is stated in Generic Letter 89-11, which should be acknowledged and referenced by the GI-101 discussion given in the EPRI-ALWR Requirements Document's Chapter 10, Section B.4 (Design of ABNR Water Level Instrumentation).
The EPRI discussion should be re-directed to address GL-89-11, which concludes that all present designs are acceptable, although some systems are preferable to others.
EPRI should identify which type of the more preferable designs they intend to use, and show how their system is in conformance to the generic letter's description of that j.
system's design features.
In addition, EPRI should state that it is their intention that licensees will provide and continee to monitor the appropriate procedures as described in GL-89-11, and will provide an acceptable l
training program for its operators regarding those procedures.
Absent such a discussion, we find that Section B.4 does not provide sufficient detail for us to reach a finding regarding the EPRI I
design's conformance to NRC's resolution of GI-101.
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2 GI-75 "Gener!.c Implications of ATWS Events at Salem Nuclear Plant" (SchedJ18d for Resolution 11/90)
Although resolution of this issue has not been finalized, the close-out memorandum is in concurrence.
The review results stated below assume the memorandam will rective concurrence without significant change.
The close-out memorandam states that GI-75 is resolved with respect to existing reactors on the basis of the requirements stated in Generic Letter (GL) 83-28.
The memorandum further states that GI-75 is closed with respect to future reactors on the basis that additional guidance may be issued whenever such plants have been proposed in sufficient numbers to make the effort worthwhile.
That effort has not been made; however, future plants should at a minimum meet the requirements of
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existing plants.
Therefore, our initial review of the ALWR design with respect to GI-75 is by comparison with the requirements for existing plar.ts stated in GL-83-28.
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We agree with EPhl's statement that GL-83-28-required licensee actions generally concern programmatic efforts by an operating plant that are not totally within the scope of the ALWR Requirements Document.
However, we also agree with EPRI's acknowledgment that there are requirements on the ALWR design which will impact how these operations will be carried out.
This review concerned itself primarily with such design requirements.
Item-by-item comments follow.
The numbering corresponds to that in GL-83-28, 1.1 Post-Trip Review (Program Description and Procedure)
EPRI does not discuss this item, and we agree that it's too early in the process to expect such a discussion.
1.2 Post-Trip Review (Data and Information Capability)
We believe a thorouch discussion of this subject should be included.
- Diagno, e data capability should be " designed in" from the start.
At very minimum, EPRI should state that their design will meet the requirements in GL-83-28 and TER SAIC-85/1513.
If EPRI has not made this commitment because they believe it's too early in the process, then it's too early in the process for us.to be reviewing this part of their design.
2.1 & 2.2 Equipment Classification and Vendor Interface (Reactor Trip System Components) & (Programs for All Safety-Related Components)
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3 The action with respect to these items is reportedly covered in the Configuration Management Section (Chapter 1, Section 2.2.C).
With the exception of requiring a standard identification system which might lead to proper classification of reactor trip and other safety related components, there is nothing more to review.
The design phase (i.e., D2w) is the time when safety re* ated systems, A
components etc. should be identified and tracked.
Now is
,L also the time to think about what kind of vendor information systems should be built into the purchase orders.
Such discussions and commitments should be added.
3.1 and 3.2 Post-Maintenance Testing m
Section 2.2.4 requires pre-installation test procedu;es and then states that they should serve as the basis for post-installation validation tests and the long term surveillance and maintenance which confirm installed system operability.
We assume that this means that during design they will prepare test procedures and provide the required testability so that post maintenance testing can be performed.
Sections 3.1.3.6.1 and 3.7.7.1 require post-maintenance testing to be included in the test plans prepared during design.
We believe those commitments are appropriate for this item during the definition-of-design-requirements phase.
In addition, EPRI adopts'h taquirement that testing should be possible without ti.e use of jumpers and lif ting of leads, which we find commendable.
However, the stated minimum-mean-time-between-failures-requiring-corrective-maintenance of 14 days for the
-protection system, plant control system and the plant information and monitoring system appears to be low by a significant amount.
EPRI should reconsider, and should clearly state the bases they used to justify the time specified.
4.2 Reactor Trip System Reliability (Preventive Maintenance and Surveillance Program for Reactor Trip Dreakers)
EPRI has not addressed the underlying requirement.
Certainly trip system reliability would be covered under the generic. requirement for reliability analyses of important systems.
And certainly EPRI will select trip system components which will be free from the defects found in the l'
Salem ATWS-era breakers.
And if they find breakers which will last forever.without maintenance, then their design is acceptable.
Sections 8.3.4.1.1 & 2 seem to imply that EPRI thinks that's achievable.
Otherwise, what we require here is a commitment to the design phase activities associated
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4-4 with - (1) definition of periodic maintenance, and (2) a designed-
'in' capability.for life testing and post-maintenance testing.
We
-believe that a short-term, high-cycle-frequency factory testing program cannot be substituted for an in-plant testing-and-
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trending program which leads to periodic replacement.
The design f*
must support the testing program, and such commitmerds should be f
added in the discussion.
4.5 Reactor Trip System Reliability (System Functional Testing) li EPRI has not addressed the underlying requirement.
Perhaps EPRI feels that breaker testing is unnecessary because (at discussed above) they'll be able to find " perfect" breakers.
However, this section covers a requirement for testing of diverse trip features.which are not part cf the breaker.
We believe that a specific commitment should be added to require testing 'of the diverse -trip features.
j, B-17
- 0riteria for Safetv-Jalated Operator Actions" -
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(Scheduled for Resolution 1/92)
We approve EPRI resolution of this item, 2H llr1 assumption th31 EPRI intends to comply with provisions of any revis e and apdated versions of ANSI 58.8, such as those now being considere: hy NRC-sponsored 4vL studies at LLL and ANS.
(The' present ANSI 58.8 Sas " time tests" that are used to determine when an action should be 4stomated, and when it Those tests are based or simulator data taken k,.neednot-beautomated.
from studies using AIRDi-based procedures.
The industry is now using gymptom-based procedures.
It is likely that the studies now underway it aT will result in reconnendation that an updated ANSI 58.8 be adopted that is based on the new symptom-based procedures.)
%n addition to the above consnants regarding USIs and GIs addressed by EPRI' in Appendix B to Chapter 10, we believe EPRI should also address the icilowing four Generic Issues in that Appendix:
5 RJ 115
" Reliability of W Solid State Protection Systems" -
(Resolved 4/89'by.NUREG-1341)
The resolution of this generic issue dqes not involve any regulatory requirements'for existing plants.
Certain insights, however. were
- gained in the course of evaluating this issue which we believe should
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Lbe considered in the design.cf the ALWR.
Consistent with.ACRS comments in a memorandum from Forrest J. Remick to Victor Stello, Jr.,
dated April 11, 1989, we recommend that EPRI evaluate innovative design changes to enhance both the reactor trip reliability and the m
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reactor core and thermal-hydraulic response to an ATWS.
For example,
- a part of such consideration could include the possibility of-relaxing the test frequency of the reactor trip breakers (RTBs)'in conjunction with the addition of an automatic trip. function to the breakers supplying the field current to the M/G sets, or,_ alternately, to the input or output breakers of the M/G sets' power supplies.
This change should also be accompanied by assuring that the undervoltage driver cards of the universal trip logic circuitry comply with the recommendations of Westinghouse Technical Bulletin " Solid State
' Protection System Undervoltage Output Driver Card", NSID-TB-85-16, July 31,1985.
Tesues Dealina with Human Factors similar to B-17, which was addressed in thd s Chapter 10.
Thus, these issues should also be addressed:
HF 4.4
" Guidelines for Upgrading Other Procedures" HP 5.1
" Local Control Stations" HP 3.2
- Rev. Crit. for Human Factors Aspects of ADV Controls and Instrument." (involves Annunciators) 4 6
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CbmmentsRegardingOtherUSIsandGlsinALWR.RequirementsDocument' r+
I US1 A 49-
" Pressurized Thermal ShoIk (PTS): EPR1'did not specifically L
address the PTS issue. EPRI should statt that all ALWR vessels t
will be designed to stay well below the 10CFR50.61-specified RT screening limits throughout their' intended life.
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~G1 B-55:
' Target Rock S/R Valves". EPRI did not address because they claimed the issue falls in " category 3" (not applicable to standard plant designs, per NUREG 1197). EPR1 should fimly state in Section 5.2 that Target Rock $/R valves will not be used, or EPR1 should discuss this issue in the " Requirements" document.
El'C-8:
'MSL Leakage Control". EPRI did not address because they claimec the issue falls in " category 6" (current requirements are acceptable, thus no need to consider, per NUREG-1197). The concern is with EPRI's not having a safety grade leakage control system. RES understandTThat discussions are currently J!
(6/897/89) underway _between NRR and EPRI regarding this matter; EPRI and other advanced designs should all conform to whatever i
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- requirements are agreed upon as a result of these discussions.
61-15:
' Rad. Effects on RV Supports". EPRI plans are not adequate: pp.
4.6-3 (for PWRs) should say fluence at the vess F supports should be low, or the material must not be sensitive to radiation damage, or the supports must be shielded; in contrast, the pg.
4.3-4 requirements (for SWRs) T10w fluence location).are acceptable (a " sk f
will be used, which will be in
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61-43:
" Air System Reliability". EPRI didn't address this issue because it wasn't priorittud previous to July 1,1986, tb date EPR) nad agreed to. EPRI should comit to meet the requirements of-a Generic Letter 88-14, which resolved this issue for operating J
. plants, and should recomend a testing program that individual applicants should be required to commit to perform. Air system acceptability is highly de>endent on a good testing program. The required testing program siould include a series o' slow air system depressurization tests (the problem is %at valves could orperhapsevenbefluttering)partiallyopenAlso, EPRI inculd propose a be in an infinity of fully or closed positions.
testing program that individual applicants should be required to comit to perform similar (or identical) to Regulatory Guide 1.80, "Preoperational Te ting of Instrument Air Systems".
GI-57:
" Fire Syst. Inadvert. Actuation'. EPRI did not address, but should address in Ch. 9 including:, a rehense to NRR Questions 3 and 7 involving qualification of fire barriers and coa 6ustible gas storage separation criteria (Me=c to Kintner from Long, draf',
j sent to Woods 6/12/89); show compliance with GDC-3; and describe requirements for use of CO, in the D.G. rooms, including the pqtential for CO, ingression into diesel air intakes.
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1 GI-87:'
'HPCI Turbi $L Break w/o 1 solation". EPRI did not address i
because they claimed the issue falls'in " category 3' (not applicable.tostandardplantdeigns,perNUREG-1197). RES recomends~ that the EPRI document should require all valves be demonstrated capable of performing their DBA functions by prototype valve test data 9 DBA conditions, before the valves are specified and/or instiTTid for the application.
'BWR H,0 Level Redundancy". EPL addressed in Ch. 4 pg. 4.314, G1-101:
although the subject was not identified as "GI-101".
Contrary to requirementssoontobeissuedforexistingdesigns(generic L
letter to be issued 6/89 or 7/89), for new designs RES expects L
that independence between the water levi Tinstrumentation of the
' feedwater control systems and the protection system will be cost effective, and should be provided, including independent dedicated sensing lines for each system. Procedure and training l
aspects of 61-101 should be listed in thi'EPRI document for' future consideration and comitment by individual applicants.
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GI-113:
"Dyn. Testing of Large Hyd. Snubbers". EPRI did not address because they concluded the issue is in " category 1" (superseded by other issues). RES recomends that for new plant designs large bore snubbers (50 kips and greater load capacity) s30uld not be considered as " rigid" restraints as they pretently are for operating plants. Rather, they should be considered as active components and be subjected to the single failure criteria.
Additionally, testing and environmental and dynamic qualification requirements should be recomended by EPRI and listed in the document for future consideration and comitment by the individual applicants.
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