ML20209E117
| ML20209E117 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Diablo Canyon |
| Issue date: | 03/20/1985 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082410749 | List:
|
| References | |
| FOIA-86-197 NUDOCS 8503280297 | |
| Download: ML20209E117 (4) | |
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A UNITED STATES yg g
NUCLEAR REGULATORY COMMISSION
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March 20, 1985 i
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Docket No. STN 50:ctet d
i MEMORANDUM FOR:
FROM:
Dennis M. Crutchfield, Assistant Director for Safety Assessment Division of Licensing
SUBJECT:
ADDITIONAL INFORMATION DIABLO CANYON UNIT 2 TECHNICAL SPECIFICATIONS l
Please request the additional information identified in the enclosure 1
from the applicant. This information is needed for completion of the review of the Diablo Canyon Unit 2 Technical Specifications. The items in the enclosure are numbered as additions to the items in the enclosure to our memorandum of March 6,1985.
1 Der nis tch i is nt Director forSafetyAssessment/
Division of Licensing Enclosure l
Diablo Canyon Unit 2 Information Request Technical Specifications rm lxA L-l i TsT032 reg)
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ENCLOSURE DIABLO CANYON UNIT 2 INFORMATION REQUEST 6.
Table 3.3-1, Reactor Trip System Instrumentation (Page 3/4 3-2)
Item 6.c on this table specifies that only one Source Range Monitor (SRM) channel is required to be operable during Modes 3, 4, and 5.
During these modes, the SRM does not provide a reactor trip function. However, it provides a boron dilution mitigation function by sensing the neutron flux increase and actuating alarms alerting the operator. One operable SRM represents a single point of vulnerability for the boron dilution mitigation system (BDMS). During the FSAR review stage, the staff reviewed and the BDMS on the premise that it is single failure proof. Therefore, this item represents an apparent deviation from the boron dilution analysis assumptions as approved by the staff.
We note that the BDMS as reflected in the Diablo Canyon Unit 2 Technical Specifications does not meet the singic failure criterion.
Either (a) propose appropriate changes to rendering the BDMS single failure proof, or (b) provide justification for not meeting the single failure criterion.
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Table 3.8-2, Reactor Trip System Instrumentation Response Times (Page 3/4-3 and Page B2-4)
In table 3.3-2, it is stated that the need to specify a response time for both the Intennediate Range and Source Range Neutron Flux Trip is "not applicable." Based on previous reviews, we understand that this is
- because they are claimed not to be taken credit for in safety analyses.
However, on page B2-4 'it is stated that they are relied upon. Therefore, provide response times, consistent with the need for a power range neutron flux response time.
8.
Table 3.3-3, Engineered Safety Feateres Actuation Systeris Instrumentation (Pages 3/4-15 to 3/4 3-17) a). We understand that Item 1.C, Automatic Safety Injection is required in Mode 4 on high containment pressure in order to protect the core in 1
the event of a LOCA. The same coninent applies to table 4.3-2, item 1.C.
b).
Item 4.a. manual steam line isolation capability should be required in Mode 4 to enable isolation of the faulted steam generator in case of a steam line break or a steam generator tube rupture.
9.
Section 3.4.4, Relief Valves (Page 3/4 4-10)
We understand that Diablo Canyon Unit 2 relies on the PORVs to be operable j
and available in order to meet the 10 CFR 100 guideline values. However, the proposed technical specifications would allow them to be taken out of k
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n service and put in an inoperable mode.
It is unclear if the action state-ments (a) through (e) of this section ensure that a PORV relief path will always be operable assuming a single failure.
In particular, if a plant q
lost 2 PORVs, one can not be taken out of service and rendered inoperable, since a single failure of a PORY would result in no pressure relief path and a violation of the FSAR assumptions for the postulated steam generator tube rupture event. Clarify the action statements to ensure that licensing bases are met or otherwise provide a basis for a conclusion that the licensing bases will be met.
- 10. Section 3/410.3, Special Test Exceptions, Reactor Coolant Loops
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(Page 3/4 10-4)
This technical specification permits plant operation without any reactor coolant pumps operating up to 10% thermal power on fission heat for startup or physics tests. The staff is unaware of any safety analysis that demonstrates that this operating condition would be acceptable.
Provide a basis for the acceptability of steady state operation at 10%
thermal power on fission heat that includes an assessment of reactor j
coolant system temperature profiles, margins to saturation, and core -
DNBR.
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