ML20236K242
| ML20236K242 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 07/31/1987 |
| From: | Stachew J EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20236K230 | List: |
| References | |
| CON-FIN-D-6023 EGG-NTA-7759, TAC-65396, NUDOCS 8708070043 | |
| Download: ML20236K242 (18) | |
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EGG-NTA-7759 July 1987 INFORMAL REPORT 7p
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2 TECHNICAL EVALUATION REPORT FOR REVIEW 0F RIVER BEND STATION UNIT 1 PROPOSED TECHNICAL SPECIFICATION AMENDMENT (RBG-25955) i.
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This boom was prepared as an account of wort sponsored by an agency of the United
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States Government. Neither the United States Government nor any agency thereof.
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nor any of their employees, makes any warranty, express or impled, or assumes any l
legal hacelity or responsibility for the accuracy, completeness, or usefulness of any information, ap;aratus, product or process d:sclosed, or iepresents that its use would not.ntnnge privately owned nghts. References herein to any specific commercial product, process, or service by trace name, tracemark, man facturer, or otherwise.
u does not necessanly constitute or imply its endorsement, recommendation, or favonng by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
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EGG-NTA-7759 r
TECHNICAL EVALUATION REPORT FOR REVIEW OF RIVER BEND STATION UNIT-I PROPOSED TECHNICAL SPECIFICATION AMENDMENT (RBG-25955)
Decket No. 50-458 J. C. Stachew Published July 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc.
Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. 06023
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ABSTRACT Tnis EG&G Idaho,.Inc., report evaluates the submittal (RBG-25955) provided by Gulf States Utilities Comoany (GSU) for River Bend Station (RBS)-Unit-1. The evaluation was to determine the acceptability of the proposed Technical Specification (TS) amendment for a one time extension of surveillance for' specific Reactor Protection System and
' Engineered Safety Feature instrumentation.
FOREWORO This report is supplied as part of the " Technical Assistance for Operating Reactors Licensing Actions" being concucted for the U.S. Nuclear Regulatory Commission, Washington D.C., by EG&G Idaho, Inc., NRR and I&E Support.
The U.S. Nuclear Regulatory Commission funced the work under authorization B&R 20-19-10-11-1, FIN No. 06023.
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Docket No 50-458 TAC No. 65396 ii
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CONTENTS
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ABSTRACT.........................................................
ii FOREWORD...........
11 1
1.
INTRODUCTION...................................................
1 2.
DISCUSSION AND EVALUATION......................................
2 0.1 Reactor Vessel Steam Dome Pressure-High....................
2 2.2 Main Steam Line Flow-High..
4 2.3 Low Reactor Vessel Water Level.............................
6 2.4 Reactor Protection Instrumentation System..................
9 3.
CONCLUSIONS 10 4
REFERENCES..................
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O TECHNICAL EVALUATION REPORT FOR REVIEW 0F RIVER BEND STATION UNIT-1 PROPOSED TECHNICAL SPECIFICATION AMENDMENT (RBG-25955) 1.
INTRODUCTION Gulf States Utilities Company (GSU) has submitted a proposed amendment (RBG-25955) to the Technical Specifications for River Bend Station (RBS)
Unit-1. This report provides an evaluation of the proposed one-time extension of surveillance for specific Reactor Protection System and Engineered Safety Feature instrumentation.
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1.
2.
DISCUSSION AND EVALUATION The Licensee provided four Attachments.to their Technical Specification Amendment application,1 Each of these Attachments is discussed and evaluated in an individual subsection below. Also, the Licensee'.s letter of May 11, 1987,4 requests changes to the Technical Specifications that affect pages that are common to the Licensee's changed f
pages that are the subject of this report.
The individual changes proposed by the Licensee's letter of May 11, 1987, and those in the four attachments to the subject Licensee's letter of this report, have been reviewed for their consistency when all of the individual changes are accumulated on the f
f affected common pages.
As the changes whicn affect common pages only involve adding superscript note' designations to specified surveillance for
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dif ferent instrunent types, accumulation of tne individual changes is acceptable (the different instruments do not have an interaction and the j
acceptability of the added superscript can be judged on an individual l
basis). One recurring carryover is the Licensee's reference to Note (d) on Technical Specification page 3/4 3-11.
This note (d) does not appear in the subject Table 4.3.2.1-1 in the subject letter of this report, but does appear in the Licensee's letter of May 11, 1987,4 and is, therefore, acceptable.
In each case, the requested extension of the surveillance already has accounted for the maximum extension of 25's allowed by Technical Specification 4.0.2 to the existing required 18-month surveillance interval.
2.1 Reactor Vessel Steam Dome Pressure-Hioh The Licensee letter of May 15, 1987,1 Attachment 1 proposed changes to Technical Specification pages 3/4 3-1, 7, 9, 11, and 29.
These changes were for a one-time extension of about 31 days to both the Channel Calibration and Logic System Functional Test (LSFT) for the Reactor Vessel Steam Dome Pressure-High instrument of the Reactor Protection System instrumentation and for the Reactor Vessel (RHR Cut-In Permissive)
Pressure-High instrument, and Isolation System Response Time of the RHR I
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System Isolation Actuation instrumentation.
The purpose of this change is to allow the above-listed surveillance to be delayed from their scheduled due date of August 15, 1987, until the first refueling outage scheduled to begin September 15, 1987.
The requested extension of the surveillance already has accounted for the maximum extension of 25% allowed by Technical Specification 4.0.2 to the existing required 18-month surveillance interval.
Evaluation:
The Licensee provided the following technical justification for requesting these surveillance extensions.
1.
For the Channel Calibration extension, the present allowance of 15 psig for channel drift between the Technical Specification "allowetle valves" of $1079.7 psig and $150 psig and the respective
" trip setpoints" of $1064.7 psig and $135 psig for the reactor vessel pressure-high ;f the Reactor Protection System and RHR Isolation System, respectively, more than compensates for any additional drift that may occur during the requested extension period.
Licensee calculations indicate that the calculated drift for the requested extended surveillance period was 12.7 psig.
Thus, drift is acequately specified anc neither the Tecnnical Specification " trip setpoints" ncr the " allowable valves" require any change from their present values for the requested surveillance period extension. Channel Checks and Channel Fur.-tional Tests are specified and executed for the Reactor Vessel Steam Dome Pressure-High and Reactor Vessel (RHR Cut-In Permissive) Pressure-High instruments in Technical Specification Tables 4.3.1.1-1 and 4.3.2.1-1, respectively.
2.
For the LSFT, Licensee results of a search of the industry Nuclear Plant Reliability Data System (NPRDS) indicated a very low probability for equipment failures with regard to manual switches and auxiliary relays for the subject surveillance period and requested extension.
Channel Checks and Channel Functional Tests are specified and executed per Technical Specification Tables 4.3.1.1-1 and 4.3.2.1-1.
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addition to the automatic isolation feature, administrative controls do not allow placing the RHR system in the cooling mode until reactor pressure has been reduced to less than 135 psig.
3.
Maintenance or calibration of instruments on the reactor pressure vessel sensing legs has resulted in spurious Engineered Safety Features actuations.
Based on tne above justification provided by the Licensee and the relatively short extension requested (about 31 days) compared to the required 18 month + 25% extension and the fact that the subject instrumentation does not exhibit sudden catastrophic failure or out-of-specification performance in these time frames, it is judged that the requested extensions for the Channel Calibration and LSFT are acceptable. Also, per Footnote (g) of Technical Specification Table 4.3.1.1-1 and Footnote (b) of Table 4.3.2.1-1, trip unit setpoints are calibrated at least once per 31 days.
However, it is noted that although the Licensee has also requested the same time extension for the RHR System Isolation, Reactor Vessel (RHR Cut-In Permissive) Pressure-High, Isolation System Response Time, no justification for this change was presented.
Table 3.3.2-3, Isolation System Instrumentation Response Time of the Technical Specifications indicates that the response time for the Reactor Vessel (RHR Cut-In Permissive) Pressure-High instrument of the RHR System Isolatian is "NA".
It is recommended that the words " Isolation System Response Time and" be deleted from the asterisk footnote at the bottom of proposed Technical Specification page 3/4 3-11.
2.2 Main Steat Line Flow-High The Licensee letter of May 15, 1987,1 Attachment 2 proposed changes to Technical Specification pages 3/4 3-11, 26, and 29. These changes were for a one-time extension from August 22, 1987, to no later than September 15, 1987, to the Channel Calibration, Logic System Function Test I
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C (LSFT), and Isolation System Response Time surveillance of the Main Steam Line Flow-High instrument of the Main Steam Line Isolation actuation instrumentation.
The purpose of this change is to allow the above-listed surveillance to be delayed from their scheduled due date of August 22, 1987, until the first refueling outage scheduled to begin September 15, 1987.
Evaluation:
The Licensee previoed the following technical justification for requesting tne surveillance extensions.
1.
For the Channel Calibration extension, the present allowance of 5 psid for channel drift between the Technical Specification " allowable value" of 178 psid and the " trip setpoint" of 173 psid more than compensates for any additional drift that may occur during the extension period.
Licensee calculations using standard setpoint methodology (GE Topical Report NEDC-31336) indicate a channel drift of 1.41 psid for a 30-month surveillance interval. The 173 psid trip setpoint converts to a 140% of rated mass flow setpoint [using flow element calibration data to convert from differential pressure units (psid) to main steam line flow).
The 140 % analytical limit is t,o protect a safety limit of I'0 *. of rated mass flow in the event of a postulated guillotine break of a' main steam line.
Channel Checks and Channel Functional Tests are specified and executed for the Main Steam Line Flow-High instrument per Technical Specification surveillance Table 4.3.2.1-1.
Also, per footnote (b) to Technical Specification Table 4.3.2.1-1, calibration of the Main Steam Line Flow-High trip unit setpoint is required at least once per 31 days.
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using a fault tree model of the Main Steam Isolation Valve (MSIV) isolation logic with basic event f ailure rates based on conservative WASH-1400 data,2 indicates MSIV isolation logic failure at I
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~3 1.06 x 10 / demand and 1.41 x 10 / demand for 18.and 24-month surveillance intervals, respectively.
(The Licensee's letter had these failures per demand interchanged. A correction will be issued.)
3.
GSU has determined, based on the group isolation logic of the isolation actuation instrumentation, that the subject surveillance should be performed while in cold shutdown due to the high risk of placing the plant in a scram condition.
Based on the above justification provided by the Licensee and the relatively short. extension. requested (about 25 days) compared to the l
required 18 tt.onth + 25% extension and' the f act that the subject q
instrumentation does not exhibit sudden catastrophic fr.ilure or out of specification performance in these time frames, it is judged that the requested extensions for the Channel Calibration and LSFT are acceptable.
However, it is noted that although the Licensee has also requested the same time extension for the Main Steam Line Flow-High instrument for the Isolation System Response Time in Technical Specification 4.3.2.3, no justification for this change was presented.
The RBS FSAR, Section 15.6.4, Steam System Piping Break Outside Containment, takes credit for isolation valve closure time and a value is specified for the subject instrument in Technical Specification Table 3.3.2-3, Isolation System Instrumentation
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Response Time.
Although the short time extension requested compared to the required 18 month + 25% extension surveillance interval is such that significant changes to the Isolation System Response Time would not be anticipated; without any Licensee justification, it is recommended that the words "Isolat19n System Response Time and" be deleted from the asterisk footnote at the bottom of proposed Technical Specification page 3/4 3-11.
2.3 Low Reactor Vessel Water Level i
l The Licensee letter of May 15, 1987, Attachment 3 proposed changes to Technical Specification pages 3/4 3-11, 26, 27, and 29.
These changes were 6
for a one-time extension of about 30 days to the Channel' Calibration, Logic System Functional Tegt (LSFT), and Isolation System Response Time surveillance of the fcilowiny 4.nstruments:
Reactor Vpssel Water Level-Low Low Level 2 (Primary Containment Isolation)
Reactor /essel Water Level-Low Low Low Level 1 (Main Steam Line Isolation)'
Reactor Vessel Water Level-Low Low Level 2 (Secondary Containment Isolation)
Reactor Vessel Water Level-Lew Low Level 2 (Reactor Water Cleanup System Isolation).
The purpose of this change is to allow the above stated surveillance to be delayed from their scheduled due date of August 16, 1987, until the first refueling outage scheduled to begin September 15, 1987.
Evaluation:
The Licensee provided the following technical justification for requesting the surveillance extensions.
I 1.
For Channel Calibration extension, the present allowance of 4 inches of water for channel drift between the Technical Specification
" allowable values," either -47 or -147 inches, and the respective
" trip setpoints" of either -43 or -143 inches, more than compensates for any additional drift that may occur during the extension period.
Licensee calculations, concurred in by the vendor, indicate that the calculated drift for 24 months between surveillance is 3.76 inches.
k The existing surveillance interval of 18 months plus the 25% extension allowed by Technical Specification 4.0.2, plus the additional 30 days requested by the Licensee, accumulates to 23.5 months.
Thus, drift is t.
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-adequately.specified and the Technical Specification " trip setpoints" and " allowable values" for the subject instrumentation do not require any change.from their present values for the requested surveillance period extension.
1 2.
For the LSFT, Licensee calculations performed using a f ault tree model L
of the Main Steam Isolation Valve (MSIV) isolation logic with basic event failure rates based on conservative WASH-1400 data,2 indicates
-3 MSIV isolation logic failure at 1.06 x 10 / demand and
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1.41 x 10 / demand for 18-and 24-month surveillance intervals, respectively.
(The Licensee's letter had these failures per demand interchanged.
A correction will be issued.)
A review of the Nuclear Plant Reliability Data System (NPRDS) wks performed by the Licensee to gather data on manual switches and auxiliary relays. This review found no reports of manual switch failures and a mean time between failure (MTBF) of 239,- 278 hours0.00322 days <br />0.0772 hours <br />4.596561e-4 weeks <br />1.05779e-4 months <br /> (about 27 years) for auxiliary relays.
These data show that the basic design of the logic is highly reliable and is unlikely to fail due to l
the delay in testing.
The logic for Reactor Vessel Water Level-Low Low Level 2 operated successfully including ali Level 2 isolations during a loss of feedwater event on March 1,1986 (LER 86-021).
3.
GSU has determined, based on the group isolation logic of the
' isolation actuation instrumentation, that the subject surveillance should be performed while in cold shutdown du'e to the high risk of placing the plant in a scram condition.
Based en the above justification provided by the Licensee and the relatively short extension requested (about 30 days) compared to the required 18 month + 25% extension and the fact that the subject instrumentation does not exhibit sudden catastrophic failure or out-of-specification performance in these time freres, it is judged that c
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l the requested extensions for the Channel Calibrations and LSFTs are acceptable.
Further, as noted in Section 2.2 of this report, Channel Check 3 and Channel Functional Tests are specified and executed for Low Reactor Vessel Water Level for the subject isolation actuation l
Instrumentation per Technical Specification surveillance Table 4.3.2.1-1.
Also, per fcetnote (b) of Technical Specification Table 4.3.2.1-1,
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calibration of the Low Reactor Vessel Water Level involved trip unit setpoints are required at least once per 31 days.
however, it is noted snat although the Licensee has also requested the same time extension for the involved Low Reactor Vessel Water Level, Isolation System Response Times in Technical Specification 4.3.2.3, no justification for this change was presented.
Response time values are specified per the subject instruments in Technical Specification Table 3.3.2-3, Isolation System Instrumentation Response Time. Although the short time extension requested compared to the required 18 month + 25%
extension surveillance interval is such that significant changes to the Isciation System Response Time would not be anticipated; without any t.icensee justification, it is recommended that the words " Isolation System Response Tin'e.and" be deleted from the asterisk footnote at the bottom of proposed Technical Specification page 3/4 3-11.
2.4 Reactor Protection Instrumentation System The Licensee letter of May 15, 1987,1 Attachment 4 proposed changes to Technical Specification pages 3/4 3-30, 41, 42, 43, 54, 58 and 3/4 5-5.
These changes were for a one-time extension of surveillance from their due da te of August 25, 1987, until the first refueling outage scheduled to begin September 15, 1987.
The surveillance were on several Emergency Core Cooling System Actuation instruments and Reactor Core Isolation Cooling System Actuation Instrumentation.
Subsequently, the Licensee performed these surveillance prior to their due date of August 25, 1987, during an unexpected short outage. The Licensee, by letter dated June 30, 1987, withdrew the request for the changes in Attachment 4.
Therefore, of the Licensee's subject letter,1 is not further addressed in this report.
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CONCLUSIONS Based on the evaluations presented, GSU's requested one time extension of surveiliances was found acceptable for the Channel Calibration and Logic System Functional Test, but not for any Isolation System Response Time for the following instruments.
Reactor Vessel Steam Dome Pressure-High (Reactor Protection System)
Reactor Vessel (RHR Cut-In Permissive) Pressure-High (RHR System Isolation)
Main Steam Line Flow-High (Main Steam Line Isolation)
Reactor Vessel Water Level-Low t.ow Level 2 (Primary Containment Isolation)
Reactor Vessel Water Level-Low Low Low Level 1 (Main Steam Line Isolation)
Reactor Vessel Water Level-Low Low Level 2 (Secondary Containment Isolation)
Reactor Vessel Water Level-Low Low Level 2 (Reactor Water Cleanup System Isolation).
The requested Extension of the required 18-month surveillance interval accounted for the maximum 25'; extens' ion allowed by Technical i
Specification 4.0.2.
The purpose of the change is to allow the listed l
instrument surveillance to be delayed from their scheduled due dates (beginning August 15,1987), until the first refueling cutage scheduled to begin September 15, 1987. As the Licensee did not provide any justification for the involved Isolation System Response Times, the recuest for extension of their surveillance is not acceptable and it is recommended that this part of the request be deleted from Technical Specification page 3/4 3-11 in the asterisk in 4.3.2.2 and in the footnote.
It is also noted that although the proposed asterisk footnote on Technical Specification page 3/4 3-11 refers to a Note d on Table 4.3.2.1-1 (in Attachments 1, 2, and 3 of the Licensee's letter ), no Note d appears in that table. However, the Licensee's letter of May 11, 1987, 10
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,,f Attaciiment3,doesprovide.aNote(d)tothesubjectT,ble. As both Licensee letters '# provide changes to the same Table 4.3.2.1-1 and as 1
the indivic'cui changes were to be accumulated, reference to Note (d) is acceptable.
Finally, Attachment L cf the Licensee's submittal was withdrawn as the involved surseillences were completed during an unexpected outage.
This report, therefore, does not provide an evaluation of Attachment 4.
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REFERENCES 1.
J. C. Deddens letter to NRC, " River Bend Station Unit-1 Docket No. 50-458," RBG-25955, May 15, 1987, Gulf States Utilities Company.
2.
NUREG 75/014 (WASH-1400), Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, U.S. Nuclear Regulatory Commission, October 1975.
3.
J. C. Deddens letter i.o NRC, " River Bend Station-Unit 1 Docket l
No. 50-458," RBG-26186, Gulf States Utilities Company, June 30, 1987.
4.
J. C. Ceddens letter to NRC, " River Send Station-Unit 1, Docket No. 50-458," REG-25917, Gulf States Utilities Company, May 11, 1987.
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