ML20039D182
| ML20039D182 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/28/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-07-01.B, TASK-7-1.B, TASK-RR LSO5-81-12-080, LSO5-81-12-80, NUDOCS 8112310399 | |
| Download: ML20039D182 (6) | |
Text
-
f y
g
.g December 28, 1981 1
(
Docket No. 50-219 A
i LS05-81-12-080 RECElWO 9
[
DEC80 21*
Mr. I.-R. Finfrock, Jr.
Vice President - Jersey Central sammumusamme -
Power & Light Company mes uu888 Post Office Box 388
't Forked River, New Jersey 08731
+
Dear Mr. Finfrock:
i
SUBJECT:
SEP TOPIC VII-1.B TRIP UNCERTAINTY AND SETP0 INT ANALYSIS REVIEW OF OPERATING DATA BASE (0YSTER CREEK).
i Our letter. dated August 17, 1978 forwarded'a Safety Evaluation Report (SER) on this topic and several other topics.. A review of the operating _ history of Oyster Creek does not support the orig $nal SER. Accordingly, a revised SER is' enclosed. The staff now recommends changingisetpoints and/or detectors for some plant parameters.
~
s You are requested to examine the facts.upon which the staff has based its evaluation and respond either by confirmint that the-facts are correct, or by identifying any errors.
If in error, p' ease supply corrected information for the docket. We enc 4urage you to supply for the docket any.other mater--
ial related to this topic that might affs::t the staff's evaluatton. Your response within 30 days of the date you receive this letter is requested.
If no response is received within that-t,ime, weewill assume that you have no a
i comments or corrections.
i The need to actually implement these changes will be detemined during the integrated safety assessment. This safety evaluation may be revised in the -
2 j-future if you facility design is changed or ff NRC criteria relating to
~
this topic are modified before the integrated assessment is completed.
Sincerely.
i 1
Dennis M. CrutchfieH, Chief Q2gfoM$$,9 Operating Reactors Branch No. 5 PDR.
Division of Licensing p
II[/
N' N/A cc:
I' See next page cx L kuUSEEl(A)
J CHU IU3Ws s omer >
......As..s ta ted.S
}L SEPB:SilDL SEP gC,,,
0,RB,#,5 :
.,Gkj n,,a,,1,,,....
- sumur,
,Ri
.,],,b,],,, (f, M,g. WRussel.l........JLombar Crutchf.ield I
.w/.p.81..........mth.81...........w.wm...
.. u n n. m..,...uwm......m.W.m.
om>
Nac ronu sis tio-soj uncu o24o OFFICIAL RECORD COPY use mi-m.co
_._._..-_. _-__._.- ___.-__...___ __. _.~ _._,
(s@ G!%q k
UNITED STATES.
v
[
p, NUCLEAR REGULATORY COMMISSION E
WASHINGTON. D. C. 20555 g
/
December 28, 1981
^
%,*****/
Docket No. 50-219 LS05-81-12-080 Mr. I. R. Finfrock, Jr.
Vice President - Jersey Central Power & Light Company Post Office Box 388 Forked River, New Jersey 08731
Dear Mr. Finfrock:
SUBJECT:
SEP TOPIC VII-1.B TRIP UNCERTAINTY AND SETPOINT ANALYSIS REVIEW 0F OPERATING DATA BASE (0YSTER CREEK)
Our letter dated August 17, 1978 forwarded a Safety Evaluation Report (SER) on this topic and several other topics.
A review of the operating history of Oyster Creek does not support the original SER. Accordingly, a revised SER is enclosed. The staff now recommends changing setpoints and/or detectors for some plant parameters.
You are requested to examine the facts upon which the staff has based its evaluation and respond either by confirming that the facts are correct, or by identifying any errors.
If an error, please supply corrected information l
for the docket. We encourage you to supply for the docket any other mater-ial related to this topic that might affect the staff's eyaluation.
Your response within 30 days of.the date you receive this letter is requested.
If no response is received within that time, we will assume that you have no comments or corrections.
The need to actually implement these changes will be determined during the integrated safety assessment.
This safety evaluation may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the, integrated assessment is completed.
Sincerely, f
k hDennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing cc:
See next page i
Enclosure:
l As stated l
l
0YS1ER CREEK Docket N3. 50-219 Mr. I. R. Finfrock CC G. F. Trowbridge, Esquire Resident Inspector Shaw, Pittman, Potts and Trowbridge c/o U. S. NRC
- 1800 M Street, N. W.
Post Office Box 445.
Washington, D. C.
20036 Forked River, New Jersey 08731 J. B. Lieberman, Esquire Commissioner Berlack, Israels. & Lieberman New Jersey Department of Energy 26 Broadway 101 Commerce Street New York, New York 10004 Newark, New Jersey 07102 Natural Resources Defense Council 91715th Street, N. W.
Washington, D. C.
20006
.J. Knubel BWR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, New Jersey 07054 Deputy Attorney General State of New Jersey Department of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 Ms. Phyllis Haefner 101 Washington Street Toms River, New Jersey 08753 l
Mayor Lacey Township 818 Lacey Road Forked River, New Jersey 08731 U. S. Environmental Protection Agency Region II Office ATTN: Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Licensing Supervisor
,~
Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
0YSTER CREEK SEP TOPIC VII-1.B, SETPOINTS TODIC VII-1.B: TRIP UNCERTAINTY AND SETPOINT ANALYSIS REVIEW OF OPERATING DATA BASE I.
INTRODUCTION Paragraph (c)(1)(ii)(A) of 150.36, " Technical Specifications," of 10'CFR Part 50 requires that, where a limiting safety system setting -
is specified for a variable on which a safety limit has been placed, the setting be so chosen that automatic protective action will cor-rect the most severe abnormal situation anticipated before a safety limit is exceeded.
Operating experience has shown that there is need for guidance in the selection of required instrument accuracy and the settings that are used to initiate automatic protective actions and alarms.
Abnormal Occurrence Reports submitted by operating utilities between January 1972 and June 1973 record the most frequent abnormal occur-rence as the drift of the protective instrument setpoint outside the limits specified in the technical specifications.
The single most prevalent reason for the drift of a measured para-meter out of compliance with a technical specification is the selec-tion of a setpoint that does not allow a sufficient margin between the setpoint and the technical specification limit to account for inherent instrument inaccuracy, expected vibration, and minor cali-bration variations.
In some cases, the setpoint selected was numer.
ically equal to the technical specification limit and stated as an absolute value, thus' leaving no apparent margin for error.
In other cases, the setpoint was so close to the upper or lower limit of the instrument's range that the instrument drift placed the setpoint be-yond the instrument's range, thus nullifying the trip function.
Other causes for drift of a parameter out of conformity with a technical specification have been instrumentation design inadequacies and questionable calibration procedures.
II.
REVIEW CRITERIA The current licensing criteria which govern the safety issue are identified in Table 7-1 of the Standard Review Plan.
The most signif-icant of these criteria are:
1.
Regulatory Guide 1.105, and 3.
i
, III.
RELATED SAFETY TOPICS AND INTERFACES The scope of review for this topic was limited to avoid duplication of effort since some aspects of the review were performed under -
related topics.
The related topics and the subject. matter are iden-ance criteria and review guidance for its subject matter.(the accept-tified below.
Each of the related topic reports contain Thermal-Overload Protection for Motor Operated Valves III-10. A Component Integrity
~
.III-11 Environmental Qualification III-12 IV-1.A
- Operation with less Than All Reactor Coolant Loops in Service RHR Reliability V-10.B High. Pressure / Low Pressure Interface V-ll ECCS Actuation System VI-7.A.3 The following topics are dependent on the present topic information for completion:
ECCS Single Failure Criterion VI-7.C.2 VI-10. A Testing of ESF Systems REVIEW GUIDELINES IV.
Protective instruments and alarms in nuclear power plants are provided with adjustable setpoints where specific actions are either automati-cally initiated, prohibited, or alarmed.
For example, pressure sensors typically are installed on main steam lines to measure steam pressure.
These sensors initiate corrective action if the steam pressure decreases to the predetermined and preset value that would result, for example, from a steam line break. Setpoints (e.g., pressure, differential pres-sure, flow, level, temperature, power, radiation level, time delay) correspond to certain provisions of technical specifications.that have been incorporated.into the operating license by the Commission.
Violation of these Limiting Safety System Setpoints (LSSS) implies an intrusion into the margin of safety between the limits and possible core damage.. Accordingly the actual setpoints used should be offset from the LSSS to account for instrument errors unless the instrument errors were used in selecting the LSSS.
Furthermore, even when instrument error is a part of the LSSS, a set-point which is close to either extrene' ~of an' instru' ment' range is un-
~~~l
~
~
acceptable. This type of design is unaccep. table because repeatabili_ty considerations may ca~u'se the trip point to end up outside the range of.
}
~
the instrument.
No amount of testing can assure that such random events will not occur.
O n
.1
a.
V.
EVALUATION As a part of the application for conversion to a full ters operating
' license, the licensee has submitted setpoint drift data for the period of January 1970 to December 1974. This data shows numerous violations of the Technical Specifications on several sensors and that some sensors are not suitable for the setpoint used.
Sensors RE23 A, B, C & D (closes MSIV on low steam pressure) have shown unacceptably high drift rates because of the large span (20 to 1400 psig.) compared to the LSSS and actual; setpoint o'f ~825.psig.
Sensors RE18 A, B, C & D (Auto depressurize on low low level), and RE02 A, B, C & D (Core Spray and Isolation on low low reactor water level) have setpoints at the extreme low end of their ranges.
The sctpoints for most operating licenses. issued prior to 1977 were evaluated in a generalized manner.
In this approach, the discrete com-
~
ponents of each of the margins to safety in trip setpoint values are not evaluated on a individual basis but are included in an overall safety margin.
Each setpoint value is based upon the most limiting transient -
or postulated accident condition associated with the bases for that set-point.
The magnitude of this safety margin and the resulting setpoints are established to ensure that there is a low probability of the margin being removed by an adverse combination of instrument-calibration error, instrument error and instrument drift.
The staff believes that this method is acceptable and the high drift rates for RE23 present a report-ing and alignment problem and that safety margins are adequate based upon experience to date.
However, the staff can not accept a LSSS at the end of an instrument scale because reliability considerations may cause the trip point to end up out-side the range of the instrument.
No amount of testing can assure that such random events will not occur.
l VI.
CONCLUSION The' staff has concluded that the LSSS for the actual setpoint for RE23 should be changed to eliminate Licensee Event Reports due to the " drift" of this instrument.
The staff has also concluded that the setpoints for RE02 and RE18 should be increased to a point where the margin to extreme range is at least equal to the instrument accuracy, or more suitable sensors (with differ-ent ranges) should be provided.
.