ML20209B346
| ML20209B346 | |
| Person / Time | |
|---|---|
| Issue date: | 06/29/1999 |
| From: | Beckner W NRC (Affiliation Not Assigned) |
| To: | Jennifer Davis NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| References | |
| PROJECT-689 NUDOCS 9907060395 | |
| Download: ML20209B346 (7) | |
Text
June 29, 1999
' ' Mr. JImes Dsvis, Director-Opercti:ns Dep rtm:nt
- Nuclear Energy Institute 1776 i Street, N. W.
Suite 400 Washington, DC 20006-3708 ,
Dear Mr. Davis:
This is to inform you that disposition has been made on twenty-three travelers containing .
proposed changes to the Standard Technical Specification (STS) NUREGs made by the NEl Technical Specification Task Force (TSTF). Those travelers that were Approved are TSTFs
--111, R.6; -218; -237,- R.1; -258, R.4; -288; -290; -291; -292; -293; 315; -318; -325; and -326.
Please see the enclosure for additional information regarding the approval of TSTF-218. ,
. Those travelers that were Modified, after discussion with the respective Owners Group '
. Chairman, were TSTFs -197, R.1; -267, -270, and -324. Those travelers that were Rejected, after discussion with the respective Owners Group Chairman, were: TSTFs -113, R.4; -141, R.1; -213; -228; -251; and -304. Please see the enclosure for NRC comments with regard to the travelers that were modified and rejected. ,
For your information, the following travelers are pending evaluation by a technical branch: l
' TSTFs -051, R.1 (SPLB & SPSB); -068, R.1 ( SPLB & SPSB); -212, R.1 (EICB); -226 (SRXB); -264 (SRXB); -287, R.2 (SPLB); -295 (EICB); -312 (SPLB & SPSB); -313 (MCEB);
and -330 (SPLB).
In addition, we request a status of tae seven TSB and nine EDIT travelers that NRC records show are pending. Those traveler numbers are: TSB-002, -003, -007., -012, -016, -017, -018 and -020; and EDIT-006, -007, -008, -014, -015, -019, -022, -023, and -024. Should there be a disposition other than approve, please allow the staff a discussion. I Please contact me at (301) 415-1161 or e-mail wdb@nrc. gov, if you have any questions or need further information on these dispositions, y
Sincerely, l
Original signed by:
'; William D. Beckner, Chief g 7 g 9y g Technical Specifications Branch PDR Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation Project No. 689
Enclosures:
As stated (,[
cc:
N.' Clarkson, BWOG DISTRIBUTION:
E-mail:
[ 7
' H. Pontious, BWROG Hard copy:/
T. Weber, CEOG File Center DBMatthews D(Os p PUBLIC SFNewberry l D. Buschbaum, WOG-
. TSB Reading File WDBeckner D. Hoffman, EXCEL - JBirmingham, Project 689 TSB Staff DOCUMENT NAME: G:\RTSBUOHNSON\06990lS.WPD *See previous COncurlences g 4g9 OFFICE - RTSB/ DRIP /NRR NRR []Y /) SC:RN@A f@P/NRR C:RTSB/ DRIP /NRR NAME DLJohnSon#/L JBirmir[a6a'm" RLDeKnia/ f WDBecknerAM DATE 06/28/99- 0 06/2[7/99 06/F /99 06/ oY /99 ,
OFFICIAL RECORD COPY !
060026 R i n e, P ao n s 9
p f*'*%1
- UNITED STATES
- r NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30088-4001
,,, , June 29, 1999 q p Mr. James Davis, Director J l Operations Department Nuclear Energy Institute l
' 1778 l Street, N. W.
l Suite 400 l Washington, DC 20006-3708
Dear Mr. Davis:
l ' This is to inform you that disposition has been made on twenty-three travelers containing l' - proposed changes to the Standard Technical Specification (STS) NUREGa made by the NEl l ' Technical Specification Task Force (TSTF). Those travelers that were Approved are TSTFs l l
-111, R.6; -218; -237, R.1; -258 R.4; -288; -290; -291; -292; -293; 315; -318; -325; and -326. 1 Please see the enclosure for additional information regarding the approval of TSTF-218. !
Those travelers that were Modified, after discussion with the respective Owners Group !
Chairman, were TSTFs -197, R.1; -267, -270, and -324. Those travelers that were Rejected, after discussion with the respective Owners Group Chairman, were: TSTFs -113, R.4; -141, R.1; -213;-228; -251; and -304. Please see the enclosure for NRC comments with regard to the travelers that were modified and rejected.
For your information, th9 following travelers are pending evaluation by a technical branch:
TSTFs -051, R.1 (SPLB & SPSB); -068, R.1 ( SPLB & SPSB); -212, R.1 (EICB); -226 (SRXB); -264 (SRXB); -287, R.2 (SPLB); -295 (EICB); -312 (SPLB & SPSB); -313 (MCEB);
j and -330 (SPLB).
l In addition, we request a status of the seven TSB and nine EDIT travelers that NRC records I
show are pending. Those traveler numbers are: TSB-002, -003, -007, -012, -016, -017, -018 I
and -020; and EDIT-006, -007, -008, -014, -015, -019, -022, -023, and -024. Should there be a disposition other than approve, please allow the staff a discussion.
Please contact me at (301) 415-1161 or e-mail wdb@nrc. gov, if you have any questions or need further information on these dispositions.
Sincerely, j' FJ, ly '
William D. Beckner, Chief Technical Specifications Branch Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation l .
l Project No. 689
Enclosures:
As stated -
cc- N. Clarkson, BWOG D. Buschbaum, WOG H. Pontious, BWROG D. Hoffman, EXCEL l T. Weber, CEOG L _
~
,, DISPOSITION
SUMMARY
TSTF-218: Anorove with Chance There is an apparent discrepancy in the Shutdown Bypass section of BWOG Bases 3.3.1.
The third paragraph refers to a new High Pressure 1850 psig trip, rather than a 1720 psig l Shutdown Bypass High Pressure trip, as given in Table 3.3.1-1 (see attachment). This l discrepancy was discussed and agreed to by Noel Clarkson, BWOG Chairman and William Beckner, NRC, on June 17,1999. The solution, discussed and agreed to on June 23,1999, by Debbie Johnson, NRC, Noel Clarkson, and Vince Gilbert, NEl, is for NRC to make this change to the BWOG Bases 3.3.1 paragraph to change 1850 psig trip to 1720 psig trip at the same time the other changes are being incorporated with this TSTF.
1 TSTF-197. R.1: Modify l
1 Changes to containment penetrations as modified are acceptable. Changing the words !
"outside atmosphere" to " environment" is rejected. The term "outside atmosphere" in LCO 3.9.4 and related specifications should remain.
TSTF-267: Modify i
The staff agrees that surveillance frequencies such as illustrated in the example are not easily explained using the existing frequency examples in Section 1.4. However, such frequencies are worded in a variety of ways, and it is not certain that the proposed example would completely solve the problem. Before adopting the proposed example, a complete search of all SR Frequencies in the five NUREGs should be made to ensure that all frequencies may be readily understood using the examples of Section 1.4 as augmented by the proposed example and also the example proposed by TSTF-270 to the PWR NUREGs.
It would be better to resolve the issue of missing examples once and for all; not in an iterative fashion. The TSTF should modify TSTF-267 by confirming this analysis is complete and by adding needed changes, if any, arising from the analysis results. Also, the numbering of the proposed example should be 1,4-6 accounting for TSTF-270.
TSTF-270: Modifv Language changes are needed for consistency between NUREGs: l
" Performed" example:
(1) Example 1.4-4 should be numbered 1.4-5 in all NUREGs.
(2) Last sentence of second paragraph: change " event" to "even".
(3) First sentence of third paragraph, third line: change "have" to "had."
(4) Second sentence of third paragraph, fifth line: Insert " entering" before " MODE 1".
TSTF-324: Modify -
1
- 1) Revise Insert B to replace [500] with (400] for the pressurizer pressure to maintain consistency with the last sentence in Insert B and with NUREG-1431.
- 2) Make a global change to inserts A, B, C and D by rep! acing " bypass" with " trip bypass." This change is needed because as used in Notes A, B, C, and D ' bypass"
r
. DISPOSITION
SUMMARY
(Continued) l Is jargon referring to the reactor protection system design feature for placing a reactor trip channel in a condition which removes the channel from the specified function trip logic.
j l l TSTF-113f R.4: Relect l The proposed changes are rejected for the following reasons:
i
- 1) The statement regarding Mode 3 with the RCS average temperature <500*F that indicates that the leak flow through ruptured tube will stop after the SG is filled to capacity is not valid. Since the RCS temperature <500*F may/can be associated with a higher RCS pressure (around 2000 psia), following a SGTR, the leak flow will not stop because of a high delta P across the ruptured tube, caused by pressurization of the secondary side and will release the reactor coolant to atmosphere via lifted main steam safety valves. Even if the RCS pressure is not higher than the main steam safety valve lift setpoint at the SGTR initiation point, j having the main steam line filled with liquid water following a SGTR is still not !
acceptable.
- 2) There is margin between the temperature at Mode 4 (350'F) and the enable temperature of the LTOP system. The current STS will not cause any operational problems.
- 3) The proposed changes could lead to the loss of mitigation capability during a SGTR l at a temperature less than 500*F.
Also, see the attached Required Safety Analysis to Support Proposed TSTF-113, Rev. 4.
TSTF-141. R.1: Reject l WOG and BWR OGs indicate that the safety analyses do not support a reduction in LCO Applicability to MODE 1 only; DBAs can be initiated with reactivity anomalies from l
MODE 2 that could threaten core thermallimits. The staff does not understand how
- vendor differences would change this conclusion. Therefore, the change is rejected l pending adequate explanation of vendor differences or agreement that the change is acceptable.
! TSTF-213: Reject
! No actual problem has occurred, and the wording is unlikely to find its way into additional conversions. The wording appears clear enough.
The proposed change would remove the phrase "that are used to shut down the reactor and ,
l control power level during maneuvering operations," from the STS definition of full-length control rods (CONTROL RODS). The motivation for the proposed change is Oconee's concem that this phrase could be misinterpreted to limit the use of the full-length control rods. That is, these control rods could not be used for purposes other than shutting down the reactor and controlling power during maneuvering operations.- Other uses of full-length i
DISPOSITION
SUMMARY
(Continued) control rods include startup, control of xenon oscillations, control reactor imbalance, etc. I Since adopting the STS, including the STS definition CONTROL RODS, Crystal River 3 and j Oconee 1,2, & 3 have experienced no technical specification (TS) interpretation problems involving this definition.; in addition, the TSs for the B&W plants that have not yet adopted the STS, ANO-1, Davis-Besse and Three Mile Island 1, have no definition for control rods.
TSTF-228: Relect The information given in the traveler is not sufficient to make a judgment. Despite repeated requests to the Og Chairperson for additional information, none has been received. Therefore, the proposed change is rejected.
TSTF-251: Re* met LThe proposed change deletes Technical Specification 5.6.9, Tendon Surveillance Report, from the Standard Technical Specifications (STS). The justification states "TS 5.6.9 requires only reporting of ' abnormal degradation' of containment structure. However, such reporting of ' abnormal degradation' is duplicative of the 10 CFR 50.73 reporting
< requirements if the condition of the principal safety barrier is sufficient to be considered
' seriously degraded, or it otherwise meets the criteria of 10 CFR 50.73(a)(2)(ii). If the 10 CFR 50.73 reporting criteria are not met, the condition of the containment would not be of value to the exercise of the Commission's responsibilities and would not, therefore, warrant such additional reporting requirements."
Section 5.6.9 of the Babcock and Wilcox Owner's Group, the Westinghouse Owner's ,
Group, the Combustion Engineering Owner's Group, and the Boiling Water Reactor /6 STS !
all state "Any abnormal degradation of the containment structure detected during the ,
tests required by the Pre-stressed Concrete Contain' ment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shallinclude a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken."
The abnormal degradation reporting requirement of the STS is a lower threshold than the seriously degraded reporting requirement of 10 CFR 50.73. The containment is the final barrier to the release of fission products. Abnormal degradation can cause a reduction in safety margins. It is certainly important that the NRC be informed of abnormal degradation so that it can make an independent evaluation to determine if there is still reasonable assurance that the containment can perform its safety function, rather than having to wait until a licensee considers the degradation to be serious enough to report.
Recently, a licensee reported to the NRC that, during the containment tendon surveillance testing of one of their units, degradation of vertical tendons due to corrosion was discovered. This was not considered serious enough to not operate the unit.
Further inspections of that unit and the plant's two other units found similar degradation of vertical tendons, if the Technical Specification had not required a report, the licensee, relying on the seriously degraded criterion in the regulation, might not have reported the degradation to the NRC. The staff is currently in the process of preparing an Information Notice to warn all licensees of plants with similar tendon configurations of the degradation potential.
3-
DISPOSITION
SUMMARY
(C:ntinued)
It l's imperative that the NRC be informed of abnormal degradation to all safety structures, systems, and components in a timely manner. Therefore, the request to eliminate TS section 5.6.9 is rejected.
ISTF-304: Relect it is acceptable to suspend certain LCOs for Physics Tests if fuel damage criteria are not exceeded. Fuel damage criteria are preserved during Physics Tests by maintaining limits on power distribution and shutdown capability. Since ASI must be maintained within limits to ensure that the core power distribution is limited to the initial values assumed in the accident analysis and CE deviation penalty factors generated by the CEACs for use in the Core Protection Calculator (CPC) trip functions are designed to accommodate individual CEA related AOOs to ensure that Specified Acceptable Fuel Design Limits (SAFDLs) are not exceeded. ASI and CEAC function must be retained during Physics Tests. Therefore, we do not approve the suspension of LCO 3.2.5 and LCO 3.3.3 during Physics Tests in LCO 3.1.10.
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,. TECHNICAL BRANCH NAMES AND ACRONYMS Division of Enoineerino (DE)
Materials and Chemical Engineenng Branch (EMCB)
Mechanical and Civil Engineering Branch (EMCB)
Electrical and Instrumentation Controls Branch (EICB)
Division of Systems Safety and Analysis (DSSA)
Plant Systems Branch (SPLB)
Reactor Systems Branch (SRXB)
Probabilistic Safety Assessment Branch (SPSB) 9 9
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