ML20198A205
| ML20198A205 | |
| Person / Time | |
|---|---|
| Site: | 05000000, North Anna |
| Issue date: | 10/07/1985 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20198A170 | List: |
| References | |
| FOIA-86-203 TAC-57927, TAC-57928, NUDOCS 8510110151 | |
| Download: ML20198A205 (12) | |
Text
.,
OCT 0 71985 MEMORANDUM FOR: Gus C. Lainas, Assistant Director for Operating Reactors, DL FROM:
R. Wayne Houston, Assistant Director for Reactor Systems, DSI
SUBJECT:
NORTH ANNA POWER UPGRADE (TACS-57927 AND 57928)
Plant Name:
North Anna Power Station Units 1 and 2 Docket Nos.:
50-338, 339 Responsible Branch:
OR3 Project Manager:
Leon Engle Review Branch:
Reactor Systems Branch Review Status:
Additional Information Required The Reactor Systems Branch is reviewing the licensee's revised safety analysis for higher power level in accordance with the Standard Review Plan. We require the attached additional information in order to complete the review.
Of particular concern is the use by Westinghouse of an unapproved model to assess fuel failure following a locked rotor accident which is based on cladding temperature rather than DNBR less than 1.'.
Westinghouse has a
attempted to utilize this model in a number of licensing actions during the past several years but it has never been accepted by the staff.
Other problem areas involve inconsistencies between the Technical Specifications and the Safety Analysis, and the lack of LOCA analysis at shutdown as required by the SRP.
C -~
..-. 4 Oo. W D'= n N R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration cc:
R. Bernero DISTRIBUTION E. Butcher Docket File L. Engle RSB R/F L. Phillips RSB P/F:
North Anna 1/2 G. Hulman WJensen R/F H. Gilpin WJensen /
M. Chatterton NLauben CONTACT:
W. Jensen, RSB x29406
- BShero, AD/RS Rdg.
DocName: WJ-03, JF "0FFICIAL RECORD COPY" h
l l
y a s DSI:RSB DSI:RSB DSI:RSB:BC 091';' LAD-
'()i WJensen:jf NLauben BSheron RHeuston
/)I' 10/03/85*
10/03/85*
10/03/85*
10/]/85
- PREVIOUS CONCURRENCE SHEET ON FILE W/RSB i
r
e.,
ADDITIONAL INFORMATION REQUIREMENTS BY THE REACTOR SYSTEMS BRANCH FOR NORTH ANNA POWER UPGRADE 1
1.
(Section 3.1.3.3.12)
The ana' lysis of the postulated locked rotor accident predicted a peak cladding temperature of 2203 F, but concludes that fuel failure will not uccur based on a ciadding temperature less than 2700'f criterion.
It has been and continues to be the NRC staff position that cladding failure is assumed to occur when the fuel rod DNBR is less thaa 'l.3c' This position provides conservatism to cover analytical uncertainties in the core thermal hydraulics, geometry, and power peaking, in ad,dition to uncertaintles in experimental accuracy.
l Therefore, you should assume that all fuel which experiences a DNBR of less than 1.3 fails, and calculate the offsite dose consequences.
In the offsite dose analysis you should assume maximum technical specification pre-acci' dent coolant activity and steam generator leakage.
Single
)
failures should be considered, including a stuck open secondary relief valve.
Loss of offsite power should be assumed per GDC-17.
The effect of steam generator tube uncovery on the offsite dose conscouences caused by l
'O, 2
operator action to isolate the affected steam generator should also be considered in the analysis, a.
Assuming one secondary system relief valve fails to close, provide the total steam releases to the atmosphere at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and upon reaching cold shutdown.
These steam releases should be provided separately for the intact steam generators and for the affected steam generator.
b.
Describe the operator procedural response to a locked rotor accident with reactor trip-turbine trip, and the consequential failure to close a secondary system relief valve.
Include any actions to isolate steam flow and feedwater flow to and from the affected steam generator, c.
Provide the water level relative to the tube bundle as a function of time for the affected steam generator.
d.
Provide the calculated percent of failed fuel, e.
Provide the secondary side iodine decontamination factors assumed for the affected and unaffected steam generators in the offsite dose calculations.
f.
Provide the offsite dose consequences for the accident.
3 2.
(Section 3.2)
Analyses were not provided for postulated loss of coolant accidents during shutdown, following operation at the increased power level. We conclude that such analyses are required to demonstrate compliance with 10 CFR 50.46.
Provide analyses of a large break loss of coolant accident during hot a.
standby at a reactor system pressure of 1000 psig.
The accumulators should be assumed to be isolated in accordance with the Technical Specifications.
b.
Provide analyses of a large break loss of coolant accident during hot shutdown. Automatic actuation of safety injection should not be assumed unless required by the Technical Specifications. Justify operator response times for manual operation.
Provideanalysesofsmallbreaklossofcoolantaccidentswhenthe c.
reactor is at 1000 psig and the accumulators are isolated in accordance with the Technical Specifications. Justify operator response time if required to restore the isolated accumulators.
Justify that operator training and procedures contain instructions
~
to restore isolated accumulators if required to mitigate small break LOCA.
4 3.
(Section 3.1.3.3.1)
The minimum DNBR from uncontrolled control rod withdrawal from subcritical was shown to be acceptable for one or two reactor coolant pumps in operation.
The analyses included the effect of a positive moderator coefficient which would be permitted by the Technical Specifications in modes 1 and 2 below 70% power. The analyses were not demonstrated to be bounding for all shutdown conditions.
Therefore, provide the following information:
Justify that the analyses which were performed at the reactor a.
temperature for hot zero power would conservatively bound events at lower temperatures.
b.
The analyses take credit for reactor trip from the power range high neutron flux channels. The power range high neutron flux channels are not required to be operable during shutdown by the Technical Specifications. The source range and intermediate range channels are required to be operable; however, the Technical Specification Basis (Page B2-4) states that no credit was taken for these trips.
Technical Specification Table 3.3-2 states that delay times for the source and intermediate trip functions are not applicable.
Correct this apparent inconsistency between the Technical Specifications and the safety analysis as required by 10 CFR 50.36.
w g_
5 c.
Analyses of inadvertent control rod withdrawal were performed for reactor pump operation, but were not performed for Residual Heat Removal Operation such as would be the case in modes 4 and 5.
Provide analyses of inadvertent control rod withdrawal in modes 4 and 5 or provide evidence that the event cannot occur.
Preventive measures at other plants include requirements that either (1) the control rods will be deenergized, (2) the reactor coolant pumps will be operating, or (3) the reactor will be sufficiently borated so that criticality cannot occur through control rod movement.
The preventive measures should be included in the Technical Specifications.
+
e O
OCT 0 71985 MEMORANDUM FOR: Gus C. Lainas, Assistant Director for Operating Reactors, DL FROM:
R. Wayne Houston, Assistant Director for Reactor Systems, DSI
SUBJECT:
NORTH ANNA POWER UPGRADE (TACS-57927 AND 57928)
Plant Name:
North Anna Power Station Units 1 and 2 Docket Nos.:
50-338, 339 Responsible Branch:
OR3 Project Manager:
Leon Engle Review Branch:
Reactor Systems Branch Review Status:
Additional Information Required The Reactor Systems Branch is reviewing the licensee's revised safety analysis for higher power level in accordance with the Standard Review Plan. We require the attached additional information in order to complete the review.
Of particular concern is the use by Westinghouse of an unapproved model to assess fuel failure following a locked rotor accident which ir based on cladding temperature rather than DNBR less than 1.3.
Westinghouse has attempted to utilize this model in a number of licensing actions during the past several years but it has never been accepted by the staff.
Other problem areas involve inconsistencies between the Technical Specifications and the Safety Analysis, and the lack of LOCA analysis at shutdown as required by the SRP.
- M--
..... m
&.Dfhnf R. Wayne Houston. Assistant Director for Reactor Safety Division of Systems Integration 3
cc:
R. Bernero DISTRIBUTION E. Butcher Docket File L. Engle RSB R/F L. Phillips RSB P/F:
North Anna 1/2 G. Hulman WJensen R/F H. Gilpin WJensen /
M. Chatterton NLauben CONTACT:
W. Jensen, RSB x29406 BSheron AD/RS Rdg.
DocName: WJ-03, JF "0FFICIAL RECORD COPY"
~
q %1 &
DSI:RSB DSI:RSB DSI:RSB:BC DS11'pA, dPG WJensen:jf NLauben BSheron Rijetiston 10/03/85*
10/03/85*
10/03/85*
10/]/85
- PREVIOUS CONCURRENCE SHEET ON FILE W/RSB
?
~
ADDITIONAL INFORMATION REQUIREMENTS BY THE REACTOR SYSTEMS BRANCH FOR NORTH ANNA POWER UPGRADE 1.
(Section 3.1.3.3.12)
The analysis of the postulated locked rotor accident predicted a peak cladding temperature of 2203*F, but concludes that fuel failure will not occur based on a cladding temperature less than 2700*F criterion.
It has been and continues to be the NRC staff position that cladding failure is assumed to occur when the fuel rod DNBR is less than 1.3.
This position E
provides conservatism to cover analytical uncertainties in the core thermal hydraulics, geometry, and power peaking, in addition to uncertaintles in experimental accuracy.
Therefore, you should assume that all fuel which experiences a DNBR of less than 1.3 fails, and calculate the offsite dose consequences.
In the offsite dose analysis you should assume maximum technical specification pre-accident coolant activity and steam generator leakage.
Single failures should be considered, including a stuck open secondary relief valve.
Loss of offsite power should be assumed per GDC-17.
The effect of steam generator-tube uncovery on the offsite dose consequences caused by -
2 operator action to isolate the affected steam generator should also be considered in the analysis.
a.
Assuming one secondary system relief valve fails to close, provide the total steam releases to the atmosphere at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and upon reaching cold shutdown.
These steam releases should be provided separately for the intact steam generators and for the affected steam generator.
b.
Describe the operator procedural response to a locked rotor accident with reactor trip-turbine trip, and the consequential failure to close a secondary system relief valve.
Include any actions to isolate steam flow and feedwater flow to and from the affected steam generator.
c.
Provide the water level relative to the tube bundle as a function of time for the affected steam generator.
d.
Provide the calculated percent of failed fuel.
e.
Provide the secondary side ' iodine decontamination factors assumed for the affected and unaffected steam generators in the offsite dose calculations.
f.
Provide the offsite dose consequences for the accident.
~.
3 2.
(Section 3.2)
Analyses were not provided for postulated loss of coolant accidents during shutdown, following operation at the increased power level. We conclude that such analyses are re' quired to demonstrate compliance with 10 CFR 50.46.
Provide analyses of a large break loss of coolant accident during hot a.
standby at a reactor system pressure of 1000 psig.
The accumulators should be assumed to be isolated in accordance with the Technical
-Specifications.
b.
Provide analyses of a large break loss of coolant accident during hot shutdown.
Automatic actuation of safety injection should net Fe assumed unless required by the Technical Specificatisns.
Justify operator response times for manual operation.
Provideanalysesofsmallbreaklossofcoolantaccidentswhenthe c.
reactor is at 1000 psig and the accumulators are isolated in accordance with the Technical Specifications.
Justify operator response time if required to restore the isolated accumulators.
Justify that operator training and procedures contain instructions to' restore isolated accumulators if required to mitigate small break LOCA.
4 3.
(Section 3.1.3.3.1)
The minimum DNBR from uncontrolled control rod withdrawal from subcritical was shown to be acceptable for one or two reactor coolant pumps in operation.
The analyses included the effect of a positive moderator coefficient which would be permitted by the Technical Specifications in modes 1 and 2 below 70% power.
The analyses were not demonstrated to be bounding for all shutdown conditions.
Therefore, provide the following information:
Justify that the analyses which were performed at the reactor a.
temperature for hot zero power would conservatively bound events at lower temperatures.
b.
The analyses take credit for reactor trip from the power range high
?
neutron flux channels.
The power range high neutron flux channels are not required to be operable during shutdown by the Technical Specifications. The source range and intermediate range channels are required to be operable; however, the Technical Specification Basis (Page B2-4) states that no credit was taken for these trips.
Technical Specification Table 3.3-2 states that delay times for the source and intermediate trip functions are not applicable.
Correct this apparent inconsistency between the Technical Specifications and the safety analysis as required by 10 CFR 50.36.
5 c.
Analyses of inadvertent control rod withdrawal were performed for reactor pump operation, but were not performed for Residual Heat Removal Operation such as would be the case in modes 4 and 5.
Provide analyses of inadvertent control rod withdrawal in modes 4 and l
5 or provide evidence that the event cannot occur.
Preventive measures at other plants include requirements that either (1) the control rods will be deenergized, (2) the reactor coolant pumps will be operating, or (3) the reactor will be sufficiently borated so that criticality cannot occur through control rod movement.
The preventive measures should be included in the Technical Specifications.
3 e
o e'
_. _ _ _ _ _ _ _ _