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APR 0 5-1985 G
MEMORANDUM FOR:
Thomas M. Novak, Assistant Director g
for Licensing, DL 4
F. ROM:
R. Wayne Houston, Assistant Director for Reactor Safety, DSI
SUBJECT:
- DIABLO CANYON UNIT 2 LICENSE
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Plant Name:
Diablo Canyon Unit 2 y
Docket No.:
50-323 (j
Responsible Branch:
LB-3 C
Project Manager:
H. Schierling Review Stage:
OL q
Review Status:
Continuing
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Per your request we have reviewed the draft License for Dielo Canyon Unit 2. to this memorandum provides an ~SSER input on "The Thermal Mechanical N,,
Report, NUREG-0737, II.K.2.13." The staff finds the applicant's response
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acceptable and concludes that this issue is resolved for Unit 2.
9 k provides proposed license conditions for resolution of the Steam Generator Tube Rupture design basis event (This issue is further discussied in a til!
memo _ from B. Sheron, Chief, RSB, to G. Knighton, Chief, LB-3), and the Diablo Canyon Unit 2 Technical Specifications.
The RSB SSER input for NUREG-Oi37, Item II.D.1, " Safety and Relief Valve Testing" is being forwarded to you separately.
Stisinal signed by*
k:
R. Wayne Houston, Assistant Director r
for Reactor Safety Division of Systems Integration cc:
R. Bernero DISTRIBUTION H. Schierling DOCKET FILE E. Buckley RSB R/F RSB Section Leaders-RSB P/F:
Diablo Canyon 2 SDiab R/F WJensen R/F LMarsh NLauben BSheron AD/RS Rdg.
N iab WJensen CONTACT:
S. Diab, RSB X-29440 l
"0FFICIAL RECORD COPY" r
- PREVIOUS CONCURRENCE SHEET ON FILE W/RSB t
- DSI:RSB
- DSI:RSB
- DSI: RSB
- DSI:RSB DS B
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' MEMORANDUM FOR:
Thomas M. Novak,' Assistant Director
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for Li, censing, DL FROM:
R. Wayne Houston, Assistant Director for Reactor Safety, DSI
SUBJECT:
,DIABLO CANYON UNIT 2 LICENSE Plant Name:
Diablo Canyon Unit 2
/
Docket No.:
50-323 Responsible Branch:
LB-3 Project. Manager:
H. Schierling
/,
Review Stage:
OL Review Status:
Continuing Per your request we have reviewed the draft License' for 'Diablo Canyon Unit 2.. to this memorandum provides an SSER input on "The Thermal Mechanical
- Report, NUREG-0737, II.K.2.13."
The staff finds the applicant's response acceptable and concludes that this issue is resolved for Unit 2.
2 provides proposed license conditions for resolution of the Steam
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Generator Tube Rupture design bas.is event (This issue is further discussed in a memo from B. Sheron, Chief, RSB, to G. Knighton, Chief, LB-3), and the Diablo Canyon Unit 2 Technical Specifications.
The RSB SSER input for NUREG-0737, Item II.D.1, " Safety and Relief Valve Testing" is being forwarded to you separately.
J R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration i
cc:
R. Bernero DISTRIBUTION H. Schierling DOCKET FILE B. Buckity RSB R/F RSB Section Leaders RSB P/F:
Diablo Canyon 2 SDiab R/F WJensen R/F LMarsh NLauben BSheron AD/RS Rdg.
SDiab WJensen CONTACT:
S. Diab, RSB X-29440 "0FFICIAL RECORD COPY" I
DSI:RSB D}l. B DSI:RSBl DSI:RSB DSI:AD:RS DSI:RSp[y NLauben g
SDiab:jf WJ. en LMarsh BSheron RHouston 04/ 3/85 04n /85 04/J/85 04/y/85 04/ /85 04/ /85 O
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ENCLOSURE 1 THEFRAL-MECHANICAL REPORT,- NUREG-0737, II.K 2.13 In resolution of NUREG-0737, Item II.K 2.13 the applicant referenced the i
Westinghouse Owners Groupi-rep'ortNCAP-10n19, December 1981, and the NRC staff evaluation of this ' issue', dated Junc '29,1984. The staff safety evaluation concluded that the in'ddittyYe'ipdn'sei tiiI.V.I.13., coupled with that e.xperience gained through the PTS frogram and with changes in requiramr-nts concording tipi operation, are judged by iNe s'taff to be adequate in demonstruting vessel integrity.
Determiniitic fYadt'ur's sechinici analyses have demonstrated no loss of vessel integrity'at' erid ~5f-life' c5n~dilion 'for a II.K.2.33 event. A pro-
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babilistic assessment iddicited'thdt the conditional probability of through-wall r-j cracking, given a II.K.'i.15 ev'idt,2 is Tess than one in one hundred occurrences.
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This probability is sufficiently low within the context of the proposed PTS rule. That is, the probabil'it'y df a II.K.2.13 event occurring which ultimately leads to a through-wall crack is on the order of one in one million reactor years.
A through wall crack does not necessarily lead to loss of vessel integrity (for example, the crack size may be small enough to allow the safety injectionsystemtd'adidtdincord'dooling).
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Rosever,.the above mentioned staff safety evaluation was specifically directed to the operating plants including Diablo Canyon Unit 1.
In letters dated January 28, 1985, and February 27, 1985 the, applicant stated that the Diablo
- aryon Unit 2 material properties at the celtline weld location and the base f
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plate are more favorable than those of Unit 1 from a thermal shock standpoint.
Therefore, the applicant stated, the analyses performed.for Unit 1 in the WCAP-10019areb'5undingforUnit2.
The' staff finds this to be art acceptable resolution to the II.K.2.13 requirements for the Diabl62anyon Unit 2.
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ENCLOSURE 2
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PROPOSED'LICEN'SE CONDITIONS
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1.
Steam Generator Tube Rupture Issue (Section 15)
Prior to restart following the first refueling outage, PG&E shall demon-strate to the satisfaction of the NRC that the steam generator tube rupture (SGTR) analysis presented in the FSAR is the most; severe case with.
respect to the release of fission products and calculated doses.
In addition, if resolution-of this issue demonstrates the need for certain equipment for mitigation (e.g. pressurizer PORVs and/or steam generator atmospheric dump valves),' suitable Technical Specifications shall be developed for these equipment to assure their operability.
2.
Reactor Trip Instrumentation (T.S. Section 3.3.1.1, Pe 3/4 3-2)
Within one hundred and eighty days of receiving the Operating License for Unit 2, PG&E shall develop technical specifications for the source range and intermediate range neutron flux monitors response times which are consistent with the FSAR safety analyses for control rod withdrawal accidents from all shutdown modes of operation.
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Reactor Coolant System Hot Shutdown (T.S. Section 3.4.1.3, Page 3/4 4-3)
Prior to restart following the first refueling outage, PG&E shall develop technical specifications for reactor coolant pump'6perability at shutdown which.are consistent with the FSAR safety analysis for control rod
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withdrawal accidents.
In addition prior to energizing the control rods before the first criticality, PG&E shall submit informention to the NRC sufficient to demonstrate that, until the above technical specifications are developed and implemented, appropriate actions such as administrative procedures have been implemented to ensdre that fuel failure will not result from a centrol rod withdrawal ' accident while at hot or cold shutdown.
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4.
Plant System, Main Steam Isolation Valves (T.S. Section 3.7.1_5 Page 3/4 7-9)
Prior to restart following the first refueling outage, PG&E shall develop technical specifications for the main steam isolation valves that are consistent with the FSAR safety analysis for steam generator tube rupture accidents while operating in a hot shutdown condition (mode 4).
5.
Reactor Trip Instrumentation (T.S. Table 3.3.1 Page 3/4 3-2)
Prior to energizing the control rods before the first criticality, PG&E shall provide technical specifications that are consistent with the single f ailure criteria of 10 CFR 50 A::per::ix A and the FSkR safety analysis f c-baron cilution accicents f rom a'l shutdown modes of operation.
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3 6.
Engineered Safety Features Actuation Systems Instrumentation (T.S. Table 3.3-3 Page 3/4 3-15)
Prior to restart foklowing the first refueling outage PG&E shall develop technical specifications for actuation of safety injection that are consistent with the FSAR safety analysis for loss of coolant accidents while operating in a hot shutdown condition.
7.
Relief Valves (T.S. Section 3.4.4, Page 3/4 4-10)'
i Prior to power operation PG&E shall develop technical specifications for operability of the Pressurizer PORVs or equivalent safety related equipment for reactor system depressurization consistent with the FSAR Safety Analysis for steam generator tube rupture and the single failure criteria of Appendix A to'10 CFR 50.
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Special Test Exemption, Reactor Coolant Loops (T.S. Section 3/410.3)
Prior to preforming natural circulation using fission heat PG&E shall i
provide technical specifications that are consistent with the FSAR safety analysis for loss of forced reactor coolant flow.
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