ML20153B006
| ML20153B006 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/17/1988 |
| From: | Jordan E Committee To Review Generic Requirements |
| To: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML20153B009 | List: |
| References | |
| NUDOCS 8807120677 | |
| Download: ML20153B006 (28) | |
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UNITE 3 STATES N d 4"'
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NUCLEAR REGULATORY COMMCION 5
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June 17, 1988 MEMORANDUM FOR:
Victor Stello, Jr.
Er.ecutive Director for Operations FROM:
Edward L. Jordan Chairman Committee to Review Generic Requirements
SUBJECT:
MINUTES OF CRGR MEETING NUMER 137 The Committee to Review Generic Requirements (CRGR) met on Wednesday, June 8 V
1988 from 1-4 p.m.
A Ifst of ~ attendees for this meeting is attached p"
(Enclosure 1).
The following items were addressed at the meeting.
./ /
1.
M. Malsch (OGC) tnd S. Crockett (OGC) presented for CRGR review a ks proposed draft role, 10 CFR Part 52, on Site Permits, Design Certifications, and Combined Licenses.
The Committee did not complet 8
their review of this matter at this meeting, but planned to continue their review and provide final comments and recommendations during the week of June 13, 1988.
This matter is discussed in Enclosure 2.
2.
C. Rossi (NRR) and C. Berlinger (NRR) presented for CRGR review a proposed NRC Bulletin, Thermal Stresses in Piping Connected to f<eactor Coolant Systems.
The Comittee recoenended in favor of issuing the proposed bulletin, subject to several minor clarifying modifications (to be coordinated with the CRGR staff).
This matter is discussed in.
3.
A. Thadani (NRR) and V. Thomas (NRR) briefed the Comittee on the staff's proposed Safety Evaluation Report on generic report BAW 47115091-00, the B&W Owners Group response to the requirements in the ATWS rule (10 CFR 50.62).
The Cemittee concluded that the report did not require formal review by CRGR, because it only implements the legal requirements in the ATVS rule and previously-approved staff positions.
This matter is discussed in Enclosure 4.
In accordance with the ED0's July 18,1983 directive concerning "Feedback and Closure on CRGR Reviews," a written response is required from the cognizant office to report agreement or disagreement with the CRGR recommendattons in these minutes.
The response, which is required within five working days after receipt of these minutes, is to be forwarded to the CRGR Chairman and if there is disagreement with CRGR recoenendations, to the E00 for decisionsaking.
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Questions concerning these meeting minutes should be referred to Jim Conran (X29855).
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- t. L Jwenn Edward L. Jordan, Chairsan Committee to Review Generic Requirements
Enclosures:
As stated cc w/ enclosures:
Commission (5)
SECY Office Directors Regional Administrators CRGR Members W. Parler M. Malsch C. Rossi A. Thadant Distribution: w/o enc.
Central File POR(NRC/CRGR)
- 5. Treby W. Little M. Lesar B. Doolittle (w/ene.)
CRGRSF(w/ene.)
M. Taylor (w/ enc.)
E. Jordan (w/ene.)
J. Heltenes (w/ene.)
J. Conran (w/ene.)
C. Sekenes (w/ enc.)
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ATTENDANCE LIST CRGR MEETING NO 137 June 8, 1988
[
l CRGR E. Jordan
[
J. Snferek R. Bernero C. Paperiello J. Goldberg(for D. loss)
I G. Arlotto NRC STAFF J. Heltenes J. Conran J. Wilson T. Cox M. Taylor l
Z. Rosztoczy M. Malsch S. Goldberg l
R. Bangart S. Treby C. Y. Cheng l
l M. Hartzman l
N. Kadambi t
W. Lanning C. Berlinger i
E. Rossi l
B. Collins (NRC cvaultant) t V. Thomas l
R. Kendall S. Newberry A. Thadani i
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, to the Minutes of CRGR Meeting No.137 Proposed Rule on Standardization gr.e8.IR8 TOPIC M. Malsch (0GC) and 5. Crockett (0GC) presented for CRGR review a proposed draft rule,10 CFR Part 52, or, Site Permits Design Certifications, and Combined Licenses intended to implement both the Commission's revised Policy Statement on Nuclear Power Standardization and, to the extent permitted by present statutory law, the license reform legislation proposed by the Commission to Congress lost year.
Copies of the briefino sildes used by the staff to guide their presentations and the discussions w' th the Committee on the proposed draft rule at this meeting are enclosed (see Attachment 1 to this enclosure).
This draft rule is related to, but is broader in scope, than two Commissien papers reviewed recently by the Committee (i.e., "Standardization of DOE-Sponsored Advanced Reactor Designs," and "Key Licensing Issues for DOE Sponsored Advanced Reactor Designs"). The most recent version of the i
(Part 52) draft rule is intended to reflect the Cosaittee's comments and recommendations on those two Commission papers; see enclosed excerpts from the minutes of CRGR Meeting Nos.135 and 136 (Attachments 2 and 3 to this enclosure).
BACKGROUND 1.
The Part 52 draft rule package was transmitted initially to CRGR for review by remorandum dated April 19, 1988, W. C. Parler to E. L. Jordan; i
that initial review package included the following:
a.
Policy Statement on Nuclear Power Plant Standardization, dated September 15, 1967 b.
Draft Federal Register Notice of Proposed Rulemaking, undated (70 pages) 2.
Subsequently, a revised draft Federal Register Notice, reflecting CRGR coments and recommendations on two related Advanced Reactor Commission papers (and coordination ef forts by the EDO of fice), was transmitted by memorandum dated June 7,1988, W. C. Parler to E. L. Jordan; the revised draft Federal Register Notice was broken into two separate parts, as follows:
a.
Supplementary Inforsation Section, undated (33 pages) b.
Draft Rule and lackfit Analysis Section, undated (39 pages)
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CONCLUSIONS /REcomDeATICu$
As a result of their reefsw of this satter, including the discussions at this meeting and earlier related discussions at. Meeting Nos.135 and 136, the Committee recommended that the proposed draf t rule be issued for comment, subject to the following modifications:
1.
At p. 21 of the draft rule, under Section 52.45(a), change the wording to read:
'Any person de holds o* has applied for a final design approval..."
2.
At p. 21 of the draft rule, under Section 52.45(c), replace the word "isolated" with the ters "appropriately sited".
3.
At p. 22 of the draft rule, under Section 52.45(d):
a.
Change the wording to read:
"Designs shall be complete in scope.
Only site-specific elements, such as... heat sink, may be excluded from the scope of the design.
(Excluded si ' specific elements that can affect safe operation sust be addressed in the specificaticn of the site envelope or interface requirements. )
b.
Also, delete the phrase "... essential to the safe operation of the plant' at the end of the second sentence there; and substitute the folicwing:
"... components that can affect the safe operation of the plant but are not fixed by site specific considerations or parameters."
4.
At pp. 22 23 of the draf t rule, under Section 52.47:
a.
Change the secsad sentence of the opening paragraph to read as follows:
"The infeenstion submitted for a _ design certification sust include performance requirements and Spei:ifications suf ficient to permit the drawing g of procurereent specifications.
b.
At the tap of p. 23 of the draf t rule, under Section 52.47, to avoid confusion regarding intent, dele *:e the word "unconditional" in the third line, and sedify the wording of that senten:e to convey the same thought but also take into iccount the fact that certification of a standard design will necess.irily include "conditions" (e.g., in the specification of interface requirements).
5.
At p. 29 of the draft rule, under Sec:f on 52.62(a), modify the wording of that entire subsectian to more clearly convey the intent that (a) during the initial or rensaal periods in which a design certification is in effect, no changes will be irposed by the Cownission except changes needed to ecgly with replattens in effect at the time of certification approval /
a
renewal, or to provide adequate safety; but (b) in the rulemaking for renewal of certification, changes that substantially improve safety in a cost beneficial sanner can only be imposed prospectively.
6.
Make clearer in the trestment of mechanises for p@lic participation in the regulatory process for certified designs (e.g., at pp.19-20 of the Supplementary Information package) the important distinction between (a) the rulemaking proceedings for Initial approval or renewal of certification, and (b) the rulemaking reoceedings for amerwhent of a i
certified design, i.e., the former includes provision for petitimrs to request informal hearings as a preliminary step, but the latter does l
not. This should be very clear because it may be a principal area of l
pubile comment.
i 7.
At p.13 of the draf t rule, under Section 52.17(c), modify the wording to read as follows:
"The application sust show, and the Commission mu;t make a findino, t
that the area surrounding...at the site."
The rule should bs modified, wherever necessary, to assure, to the extent practical, that emergency planning issues are resolved at the initial hearing, i.e., at an early site or combined hearing.
8.
Throughout the draft Part 52 package, delete the ters "... structures, systems, and components that are essential to the safe operation of the plant..." and substitute other wording that is not 50 likely to be misunderstood by applicants as synony'mous with the current safety classification ters safety-related.
The phrase "... structures, systetes, and coeponents that can significantly affect safe operation of the f acility..." (or something similar and equivalent) would be satisfactory and less likely to result in confusion on this important point.
9.
At p. 22 of the Supplement Information package, delete the third sentence on that page.
- 10. At pp. 29-30 of the Supplementary Information package, under Question 7 add the following:
"Should the standards for obtaining exemptions for a certified design be stricter to avoid continual regression from true standardization?"
- 11. The Coanittee understands from discussions with staff at this meeting that the proposed rule is intended to apply to existing as well ss advanced designs, e.g., to the AlWs and to (applicants who hold or have applied for) final design approvals for existing designs.
The wording at p.14 of the Supplementary Inforsation, however, package states (in a way that suggests exclusiveness) that the rule applies to "... procedural aspects of the certification of advanced reactor designs..." Clarify the intended scope of the rule, perhaps by deleting this wording.
1
The recommendations in the preceding for the most part are focused on changes in the wording of just e specific section of the total package submitted for reviev.
It is intended at the staff should carefully review, and make c :nforming che.nges to, other parts of the package as appropriate for ktency throaghout.
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l PART 52 IN A NUTSHELL (Sections changed f rom 4/19 draf t are marked
"*")
EARLY SITE PERMITS Annlicant is anyone who may apply for CP (15)
Contents of application (17)
Types of f acilities auitable to site Projected population profile Redress plan Good f aith efforts to get local cooperation on emergency planning EARA (19)
No application fee Review fees paid by permittee but deferred Hearinza are mandatory, and adjudicatory (21)
ACES (23) reviews any safety issues LWA-1 activities ok without separate authority (25)
Duration of permit is 10 years (27)
Renewals (29-33) 5-10 years Unlimited number Granted. if application meets current regulations Hearing opportunity ACRS review Permit remains valid during senewal proceeding or if cited in CP applica lon before expiration Use_of site for other purposes ok, with NRR review (35)
Einalitz (39)
If permit in ef fect, backfits only for undue risk CP/0L applicant may request variance under 50.92 Attachment I to Enclosure 2
DESIGN CERTIFICATIONS Ann 11eant is any person (45)
Advanced reactors can be certified (45)
Prototpye test is presumed Presumption can be overcome Inconolata desiana can be certified (45)
If everything "essential to safety" is in design Contenti of applications (47)
Level of detail FDA level Enough to draw up procurement specs, etc.
Enough so DC can be unconditional Technical information Required by applicable portions of Parts 20, 50, 73, & 100 Staff will advise applicant on what is needed Site parameters PRA Proposed tests, analyses, inspections, etc.
Modular design information Options for configuration PRA should take account of options Interface requirements with uncertified BOP, with Showing of verifiability of requirements Representative design for BOP l
Ecss (49)
No application fee l
Review fees paid by holder but deferred Ctztification oroceeding (51)
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Rulemaking l
Notice and Comment Informal hearing before ASLB ACBS need not consider issues it reviewed earlier (53)
Duration of certification is 10 years (55)
Reneuals (57-61) 5-10 years Unlimited number Granted if application meets current regulations Proceeding is rulemaking with informal hearing DC remains valid during renewal proceeding or if cited in CP application before expiration Einalitz (63)
If DC in effect, backfits only foi andue risk Bolder may apply for amendment to design Granted if it complies with regulations Backfitted if adequate protection requires Applicant or licensee may request 50.12 exemption Backfitted if adequate protection requires Licensee may make plant changes without prior NRC approval if change in outside design
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i COMBINED CP AND CONDITIONAL OL Anolicant is anyone who may apply for CP (75)
Esta are those for CP/0Ls in Part 170 (75)
Contents of Applications -- Technical information (79)
As for DCs, If cites DC and ESP, show them compatible Tech specs Emergency plans and good faith efforts to get local cooperation 4
Hearinas are mandatory and adjudicatory (85)
ACBS need not consider issues it reviewed earlier (87)
LWA-l activities ok without separate authority (91)
If early site permit cited Redress required in application denied and site permit expires without being cited Exeactions & variances (93)
Applicant may seek 50.12 exemption from DC Applicant may seek variance from permit Conversion to full operating license (103)
Each module the subject of a separat.e conversion Opportunity for hearing on grounds of Nonconformance of construction with DC, etc.
Some change necessary for adequate protection 4
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Enclosun 2 to the Minutes of CRGR Meetina No.135 4
DOE Sponsored 1.1vanced Reactor Desferis - Key I.' censing Issues
~
April 17. 198_8 TOPIC The Committee continued at this meeting their review of the proposed Commis-sion Paper, "Key Licensing Issues Associated With DOE Sponsored Advanced Reactor Designs" which as begun at Meeting No'.133 - (see Pnutes dated 5/6/88).
W. Morris (RE5) and T. King RES) were the princ@l staff representatives presentia0 and discuss (ing at this seating the proposed d criteria for, and approach to the staff's review and certification of, three advanced reacter designs that are being sponsored by D0E. Attachment 1 to this enclosure is a copy of a briefing slide used by the staff at thi: see'6fng to clarify the relationship between Advanced Reactor Certification arkt Standard-ization issues (as reflected in the two draft Commission Papers subtitted by RES for CRGR rerview), and the draf t rulemakirv; package (Part 50) subeltted to CRGR for review recently by OGC.
BACKGROUIC In addition to the doctments subeltted initially by RES for CRGR review in this matter (see listing of those doctments in Minutes of CRGR Meeting No.133), the Commi8 tee was provided the following supplemental infonnation subsequent to Meeting No.133:
Memorandum dated April 21, 1988, J. H. Conran to E. L Jordan et al., with attachments as fo11cws:
1.
Comments by Individual CRGR Membe; (Ross), undated, "Major Points to Make on Commission Paper on Adycnced Reactors."
2.
Note dated April 12, 1988, S. Treby to T. King, subject: "SECY Paper on Key Licenslig Issues Asrociated with DOE-Sponsored Advanced Reactor Designs."
3.
Revised Pages (pp. 8, 9, 16, 17, & 18) for the Draf t (Key Licensing issues) Commission Paper, dated February 9,1988, that was stheitted initially by RES for CRGR review in this matter.
(The documents listed above are included as a part of these Minutes - see to this Enclosure.)
CONCLUSIONS / REC 0fmE)SATIIIIS As a result of their review of this matter, including the discussions with the staff at Meeting Mos.133 and 135, the Coenittee recommended in favor of send-ing the proposed "Key Licensing Issues' Commission Paper fomard for E00 and Consission consideratten, subject to the following comments and modifications:
Attachtnent 2 to Enclosure 2
1.
Tim Committee strongly believes that, as a principal policy objective, NRC should require that advanced reactor designs provide an improved level of safety compared to currently operating LWRs, at least with respect to the degree of confidence la the level of safety achieved.
2.
lhe Committee believes that the proposed approach to advanced reactor design certification and the preliminary design criteria set forth in the draft Commission Paper provide an acceptable basis for approving the construction and testing of prototypes of the advanced reactor designs involved. Information, experience, and test results obtained in the construction and testing of prototypes should be factored into the development of the final design criteria and guidance that will serve as the basis for certification of advanced designs. As a specific censider-ation, the Commission should address explicitly the timely development of appropriate codes and standards to support design, construction and review of the a6anced designs.
3.
The Cornittet noted that the proposed advanced reactor designs appear to involve significant safeguards issues, because their fuel designs espicy plutonium or uranits acre highly enriched than in current LWR designs; but these are not addressed in the draf t Commission Paper.
Ihe Committee was informed that the staff intends to address these issues separately and later in the process.
Notwithstanding such future plans, the Commit-tee recommended that the draf t ("Xey Issues") Commission Paper be revised to include identification, and at least some discussion, of the signifi-cant safeguards issues associated with the use of plutonium and the more highly enriched uranits.
4.
The Coccittee recoemended, and the staff agreed to make, a number of revisions to the wording of the proposed Comission Paper in the specific areas indicated in the following:
a.
Page 2, middle of page:
The stiff should revise the paper with regard to use of the term "current generation LWRs", in this section and throughout the paper.
That ters should not be applied to "paper" reactor designs that have not yet been built.
Also, usage of that ters in this draft Commis-sion Paper does not appear consistent with its usage in the related (Part 52) rulemaking package provided by OGC (e.g., see under definition of 'Mwanced reactor," at pp. 40-41 of the OGC paper),
b.
Pages 2 & 3, under "General Criteria":
The wording of this section is too vague or ambiguous; it should be revised / clarified along the following lines:
1.
... existing rules and regulations, as interpreted for advanced reactor concepts... "
(This is sluply too vague to u.derstated for review purposes, as written. Revise this section here, and deretar else these or
sieller words appear, to reflect the explanation to the Commit-tee at the meeting, i.e., that the designers will propose, and the staff will review and determine finally, which parts of the existing rules, SWs, and other guidance are applicable.)
11.
... fission product retention capability at least equivalent to
. Usts... "
(Indicate more clearly what is intended here, e.g., retention la the fuel, in the reactar, in the containment, etc., and how this is to be measured / evaluated for review purposes.)
111. "... defense in depth..."
(The intended meaning of this ters appears a be different in the advanced design context than is commonly understood for current Lidt designs.
Discuss / explain these differences and their safety implications, e.g., is this a less (or more) striagent criterion for the advanced designs than for existing LWR designs with regard to number of barriers and overall effect on safety.)
iv.
... demonstrate enhanced safety / margin via testing..."
(Indicate more clearly how this will be done, e.g., which systaes or features will need to be tested for certification of an advanced design, and how that will be determined.
Note Itse 4.e. below in this context. )
v.
"Advanced reactor designers should ensure..."
(Use of the terms "shall" or "must" is more appropriate for specifying design criteria, i.e., design recuirements.
'Shosid" is used in connection with non-mancatory guidance.)
The draft paper should be reviewed carefully throughout for consist-ency in the use of terminologies, and further revised as necessary, in accordance with the preceding.)
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c.
Page 3. under "Specific Criteria:
I Change Un term "release limits" to "dose limits," and revise to
(
delete the ters "ad hoc" in discussions of Emergency Planning (here i
l and throghout the paper).
Clarify what is intended with regard to l
Emergency Plenning for advanced designs, in accordance with the discusslee with the Committee in this meeting (e.g.Ing reduced, no same onsite I
planning as for current plants, prompt offsite plann f
drills /precedures required, etc., similar to fuel cycle facilities).
RES sgreed to work with OGC in revising this section, and the more detailed discussion of EP matters in Section II.B.4 (at pp.
19-21 of the draf t Commission Paper).
4 d.
Page 10, under Subsection iv, first bullet:
Clarify what is intended by the term "shutdown" fcr the advanced reactor designs involved, in accordance with the discussions at this meeting (e.g., one cold shutdown mechanisa/ system independent of offsite power; one systes/sechanisa capable of holding the reacter safely at hot shutdown for "X" Murs/ days / weeks).
e.
Page 10, under Subsection !!.#fs.v and Page 11, under Subsection II.A.2.iii:
Revise wording to specify that the results of the PRA should be used in determining what tests are to be performed for certification.
f.
Page 11, under second indented paragraph:
Clarify the wording of this section with regard to QA requirements for advanced designs.
For example, as indicated to the Committee at this meeting, the QA required for inherent / enhanced safety features say be even more stringent than existing Appendix B, certainly not less stringent as was suggested to some by the existing wording).
g.
Paaes 13-14, under "Event Categories III & IV:
The staff should specifically call to the Commission's attention, and get explicit Commission agreement on, the proposed requirement that the advanced reactors be designed for very severe accidents, clearly beyond what are currently recognized as the Design Basis Events.
Also, there should also be explicit acknowledgement of, and appropri-ate cautionary guidance regarding, the large uncertainties associated with event frequencies as low as E-6 and E-7, as referred to in this section.
5.
The changes recomended above should be coordinated with the CRGR staff.
In addition, RES should obtain NRR concurrence on the revised package.
If this results in further substantive modifications to the package, it l
should be resubmitted for review by the Comittee.
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. to the Minutes of CRGR Meetino No.136 Standardization of DOE-5ponsoreed Advanced Reactor Desig May 19. 1988
. TOPIC The Cosmittee continued at this meeting their review of the proposed Commission Paper, "Standardization of Advanced Reactor Designs," begun at Meetime)
- (see Minutes dated May 6,1988).
W. Morris (RES) and J. Wilson (RES were the principal staff representatives at this meeting who presented the staff's pmposed plans for review of three D0E-sponsored advanced reactor designs and the proposed (preliminary) design criteria that those advanced designs will be reviewed against. Copies of the briefing slides used by the staff at this meeting are enclosad - (see Attachment I to this Enclosure).
BACKGROUND In addition to the documents submitted initially by RES for CRGR review in this matter (see listing in Minutes of CRGR Meeting No.133), the Committee was provided supplemental inforsation by the staff subsequent to Meeting No.133, as follows:
l 1.
Revised (draft) Coemission Paper, dated May 10, 1988 i
[See enclosed - Attachment 2 to this Enclosure) 2.
Memorandum dated May 19, 1988, S. A. Treby to E. L. Jordan, "Standardization of Advanced Reactor Designs" (See enclosed - Attachment 3 to this Enclosure)
CONCLUSIONS / REC 0 MEN 0ATIONS As a result of their review of this matter, including the discussions with the staff at Meetings 133 and 13(, the Committee recommended in favor of sending the proposed (Standardization of Advar.ced Reactor Designs) Commission Paper forward for E00 and Commission consideration, subject to the following caveats:
l 1.
The scope of the staff's review should be an entire prototypical advanced reactor plant (not just the safety-related envelope); and the staff should require a level of detail sufficient to do a complete review of both balance-of plant and safety related systees, and to understand fully any interactions between them.
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2.
The staff should require prototype testing at an isolated site.(i.e., not l
a site that could be approved in accordance with Reg. Guide 1.47).
The test configuration should be closely representative of a whole plant (e.g., a single module with a heat sink other than a turbine should not be i
considered acceptable).
The purpose of this prototype testing would be to demonstrate the inherent / enhanced safety features of the advanced design, to Enclosure 2
.g.
and to verify the intended absence of interactions between safety equipment and other plant equipment / systems.
The results of the prototype testing should be used to determine the scope of certification of the final standard design; and this would be aJdressed in the SER accepting the advanced design for certification.
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. to the Minutes of CRGR Meeting No.137 Proposed Bulletin on Thermal Stresses in Piping June 3, 1988 TOPIC C. Rossi (MAR) and C. Berlinger (NRR) presented for CRGR review a proposed NRC Bulletin requesting that licensees (1) review their reactor coolant systems (RCS) to identify any connected, unisolable piping that could be subjected to temperature distributions that would result in unacceptable thermal stresses, and (2) take action, where such piping is identified, to ensure that the piping will not be subjected to unacceptable thermal stresses.
Copies of the briefing slides used by the staff to guide their presentation and their discussions with the Consittee on the proposed bulletin at this meeting are enclosed (see attachment to this enclosure).
BACKGROUND The documents submitted for review by CRGR in this matter were transmitted by memorandum dated May 16, 1988, J. H. Sniezek to E. L. Jordan; the review package included the following:
- 1. - Draft !!RC Bulletin, undated, "Thermal Stresses in Piping Connected to Reactor Coolant Systems," and attachments as follows:
a.
Figure 1 "Farley 2 Temperature Data" b.
Figure 2 "Farley ECCS"
- 2. - Infomation Required by Section IV.B of the CRGR Charter for Review of the Proposed Bulletin l
CONCLUSIONS / REC 0leEN0ATIONS As a result of their review of the proposed bulletin, including the I
discussions with the sponsoring office staff in this meeting, the Committee i
recommended in favor of issuing the proposed bulletin, subject to the l
following minor revisions (to be coordinated with the CRGR staff):
(
1.
At p. 2 of the proposed bulletin, in the first full paragraph, change the l
wording of the second sentence to read as follows:
"Because valves often leak, an unanalyzed condition may exist for those reactors that can be sut,jected to this unanticipated phenomenon."
2.
At p. 2 of the proposed bulletin, in the first full paragraph, preface the last sentence in that paragraph with the following phrase:
"Although failure is unlikely to occur..."
2-Then reexamine the remaining wording of that sentence (referring to double-ended failure), in comparison to the wording of the response to Question (viii) on p. 6 of Background Item 2 (referring to small-break LOCA), and revise as necessary for consistency.
3.
At p. 2 of the proposed bulletin, under Action No. 2, delete the words
"... safety injection..." in the first line of that paragraph.
4.
At p. 2 of the proposed bulletin, under Action No. 2, add a sentence as follows:
"If affected welds are found in the unisolable section being examined, extend the examination to other likely affected portions l
of that section (e.g., to any adjacent elbows).
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l'.'lt THERMAL STRESSES IN PIPING CONNECTED TO REACTOR COOLANT SYSTEM BACKGROUND:
FAPLEY UNIT 2 EVENT (12/9/87)
THERMAL FATIGUE FAILURE OF UNISOLABLE PIPING CONNECTED TO RCS.
BULLETIN REQUESTED ACTIONS:
1.
REVIEW PIPING SYSTEMS CONNECTED To RCS - UNISOLABLE AND POSSIBLY SUBJECTED TO UNANALYZED THERMAL STRESSES.
2.
FOR SUCH P! PING - EXAMINE NONDESTRUCTIVELY WELDS AND HEAT AFFECTED ZONES.
3.
PLAN / IMPLEMENT CONTINUING PROGRAMS TO:
REDESIGh/ MODIFY PIPING TO RESIST CYCLIC THERMAL STRESSES.
INSTALL T/C TO MONITOR TEMPERATURE AND MAINTAIN WITHIN ACCEPTABLE TEMPERATURE DISTRIBUTION.
MONITOR PRESSURE UPSTREAM OF BLOCK VALVES AND KEEP BELOW RCS PRESSURE.
SCHEDULE:
OPERATING PLANTS HQT IN EXTENDED OUTAGE:
ACTION 1:
60 DAYS ACTION 2/3:
BEFORE END OF NEXT REFUELING IF RESTART SCHEDULED FOR3> 90 DAYS, OR BEFORF RESTART FROM NEXT REFUELING IF CURRENT STARTUP SCHEDULED WITHIN 90 DAvs.
OPERATING PLANTS IN EXTENDED OUTAGE AND CP HOLDERS:
C ACTION 1:
LATER OF 60 DAYS OR INITIAL CRITICALITY ACTION 7/3:
BEFORE CRITICALITY IF CRITICALITY SCHEDULED FOR 3> 90 DAYS, OR BEFORE RESTART FROM NEXT REFUELING IF CURRENT STARTUP SCHEDULED WITHIN 90 DAYS.
ll REPORTING REQUIREMENTS - CONFIRMATORY LETTERS - WITHIN 30 DAYS OF COMPLETION OF ACTION 1 AND ACTIONS 2/3.
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. to the Minutes of CRGR Meetina No.137 Proposed SER for B&W Owners Group's Response to the ATWS Rule June 8, 1988 TOPIC A. Thadani (NRR) and V. Thomas (NRR) briefed CRGR on the proposed Safety Evaluation (SER) for generic report BAW 47-115091-00, the B&W Owners Group's (BWOG's) response to the ATWS Rule (10 CFR 50.62).
The proposed SER accepts the BWOG response, subject to implementation of certain contended recommenda-tions that will raquire modification of the affected plants.
The issue to be determined in the the discussion with the staff of this matter was whether the recommended modifications involved new or previously-approved staff positions, and therefore whether or not this proposed SER required CRGR review prior to issuance.
Copies of the briefing slides used by the staff to guide their presentation and their discussion with the Committee at this meeting are enclosed (see attachment to this enclosure).
BACKGROUND 1.
The material provided initially for CRGR consideration in connection with this item was transmitted by memorandum dated May 9,1988, J. H. Sniezek to E. L. Jordan; that package included the following:
a.
Draft Letter (undated), J. A. Calvo to Multiple Intended Addressees), subject:
"NRC Evaluation of BWOG Generic Report -
Design Requirements for DSS and AMSAC."
b.
Safety Evaluation, dated February 1988, of Topical Report (B&W Document 47-115091-00), "Design Requirements for DSS (Diverse Scram System) and AMSAC (ATVS Hitigation System Actuation Circuitry).
2.
Subsequently, at the request of the CRGR staff, the staff provided supplemental documents for consideration by the Committee, as follows:
a.
Letter dated October 9,1985, J. T. Enos to H. L. Thompson, subject: "B&W Owners Group ATVS Design Basis," and attachment entitled:
"Design Requirements for DSS (Diverse Scram Sy' stem) and AMSAC i
l (ATVS Mitigation System Actuation Circuitry),
dated September 1985.
b.
Letter dated December 1,1987, J. T. Enos to F. J. Miraglia, subject: "8W Response to ATWS SER Items," and attached specific l
responses (nine p q d l
l
. CONCLUSIONS /RECOMENDATIONS As a result of their discussions with the staff at this meeting, the Committee agreed that the subject proposed SER is based only on previously-approved positions and legal requirements in the ATWS rule.
Accordingly, this item does not require formal review by CRGR prior to issuance by the NRC staff.
s l
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F4 BRIEF ATWS HISTORY OUTLINE B&W o
THREE PROPOSED ATWS RULES DEVELOPED (STAFF RULE /HENDRIE RULE / UTILITY RULE) 1981 o
PUBLIC COMMENT ON PROPOSED ATWS RULES JUNE 1982 o
STEERING GROUP FORMED TO DEVELOP FINAL RULE JULY 1982 o
SALEM - EVENT (FEB. 1983) 00 RESULT:
NUREG-1000, VOL 1 - 1983 GL 83-28 (NUREG-1000, VOL. 2 - AUG. 1983 o
SECY-83-293 (JULY 1983) o COMMISSION V0TE o
FINAL RULE - NOV. 1983 10 CFR 50.62 - JUNE 1984 F
Attachment to Enclosure 4 i
i I
B8WOG o
B8W DOCUMENT - GENERIC REPORT (DSS /AMSAC DESIGN REQUIREMENTS)
OCTOBER 1985 o
MEETING:
BWOG/NRC STAFF DRAFT SER/BaWOG GENERIC REPORT OCTOBER 1987 o
BWOG SUPPLEMENTAL LETTER RESPONSES TO OPEN ITEMS DECEMBER 1987
.. o SER/IN HOUSE REVIEW CURRENT
OPEN ITEMS (DISCUSSION) o NO FORESEEABLE PROBLEMS WITH THE IDENTIFIED AREAS OF CONCERN TO THE STAFF WITH ONE EXCEPTION:
o SAFETY-RELATED (IE) POWER SUPPLY -
NOT REQUIRED, BUT MUST DE CAPABLE OF PERFORMING SAFETY FUNCTIONS WITH LOSS OF 0FFSITE POWER.
LOGIC POWER MUST BE FROM AN INSTRUMENT POWER SUPPLY INDEPENDENT FROM THE POWER SUPPLIES FOR THE EXISTING REACTOR TRIP SYSTEM.
EXISTING RTS SENSOR AND INSTRUMENT CHANNEL POWER SUPPLIES MAY BE USED PROVIDED THE POSSIBILITY OF COMMON MODE FAILURE IS PREVENTED.
00 DESIGN REQUIREMENTS FOR THE USE OF POWER SUPPLIES AS DESCRIBED IN THE BWOG GENERIC DOCUMENT AND SUPPLEMENTAL LETTER 12/87 IS NOT ACCEPTABLE TO THE STAFF.
00 AN ACCEPTABLE METHOD IS PRESENTED IN SECTION 5.6 l
OF THE SER.
I e
CONCLUSIONS o
IF INDEPENDENT SEPARATE POWER SUPPLIES ARE USED, GENERIC APPROVAL CAN BE GRANTED o
OTHERWISE, ISSUE BECOMES PLANT SPECIFIC AND WILL BE REVIEWED AS SUCH o
OGC AGREES WITH THE STAFF'S POSITION CONCERNING INTERPRETATION OF POWER SUPPLIES e
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