IR 05000528/1991053
| ML17306A470 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 02/04/1992 |
| From: | Louis Carson, Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17306A469 | List: |
| References | |
| 50-528-91-53, 50-529-91-53, 50-530-91-53, NUDOCS 9202240029 | |
| Download: ML17306A470 (17) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION REGION V
Report Nos.
50-528/91-53, 50-529/91-53 and 50-530/91-53 License Nos.
NPF-41, NPF-51, and NPF-74 Licensee:
Arizona Nuclear Power Project'(ANPP)
P.
0.
Box 52034 Phoenix, AZ 85072-3999 Facility name:
Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2,
3 Inspection at:
PVNGS Site at Mintersburg, AZ Inspection conducted:
December 30 - 31, 1991, through January 2 - 3, 1992 Inspection by:
r n
,
eactor a
1a )on pecla 1s a
e
>gne Approved by:
~Summal:
as, ie Reacto adiological Protection Branch a
e cygne Areas Ins ected:
Routine unannounced inspection of the licensee's process and effluent radiation monitoring systems (RMS) and followup on previous inspection findings.
Inspection procedures 84524, 84724, and 92700 were used.
Results:
The licensee continues to evaluate and improve their radiation monitoring systems.
Their use of the incident investigation process in connection with these efforts is considered a strength.
However, the decision to rely on the containment high range radiation monitor's americium "Keep Alive" source for calibration, as discussed in Section 3 e, will be considered an unresolved item pending additional technical review by NRC.
A perfo'rmance based weakness was observed at Unit 1 related to the failure of the control room staff to periodically check the operability of a radiation monitoring system multi-.point recorder and is described as a non-cited violation in Section 3 a.
9202240029 920204 PDR ADOCK 05000528
DETAILS
a.
Licensee
- J.'cott, General Manager, Site Chemistry J. Albers, Manager, Radiation Protection Operations
"J. Wilson, Project Manager, Project Management Department R: Sorensen, Chemistry/RMS Technical Services Manager
"P. Coffin, Compliance Engineer
- R. Rouse, Compliance Supervisor
- T. Murphy, RMS Supervisor
"R. Fountain, Quality Assurance (QA) 8 Monitoring Supervisor
"C. Gray, Unit-3 RMS Supervisor
"W. Blaxton, Unit-1 RMS Supervisor
"W. Wattson, RMS Plant System Engineer K. Kutner; RMS/Effluents Advisor D. Elkinton, QA Engineer
- J. Draper, Southern California Edison, Site Representative
"R. Henry, Salt River Project, Site Representative b.
NRC
"J. Sloan, Resident Inspector (*)Denotes those individuals present at the exit interview conducted on January 3,
1992.
I Additional discussions were held with other members of the licensee's staff.
2.
Onsite Followu of Licensee Event Re orts (LERs) and S ecial Re orts Item 50-529/90-05-01 (Closed):
This supplement to a 1990 SR informed the a
ra sa son e
uen monitors RU-143/144 were inoperable/out of service more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Both RU-143/144 were out of service to allow the scheduled calibration to be performed.
The calibrations actually took 312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br /> to perform for various reasons.
During that period, the lic'ensee used the preplanned alternate samplinq system to fulfill its safety obligation.
The licensee has since devised a plan for radiation monitor (RM) calibrations and surveillance (ST) to be performed without exceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Item 50-529/90-04-01 (Closed):
This supplement.to a 1990 SR informed the a
ra sa son e
uen monitors RU-145/146 were inoperable/out of service more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Both RU-145/146 were out of service to allow the scheduled calibration to be performed.
The'alibrations actually took 312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br /> to perform due to parts replacement and modifications.
During that period, the licensee used the preplanned alternate sampling system to fulfill l~ts safety obligation.
The licensee has since devised a plan that will allow RM calibrations and STs to be performed without exceeding 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Gaseous Waste S'tem:
Process and Effluent Monitors (84524 L 84724)
a.
Multi oint Chart Recorder 1J-S A-RR-0029, The inspector toured Unit-1 to observe remote and local, process and effluent RMS readings to determine operability, also, to determine if the lice'nsee replaced the multi-point chart recorder
, 1J-SgA-RR-0029 as referenced in an NRC Inspection Report 50-528/90-04.
During a Unit-1 control room tour on January 2, 1992, the inspector found that multipoint chart recorder 1J-S(A-RR-0029 was not printing any of its six RM data points on the chart paper.
This chart recorder provides a hard copy record of radiological data for RMs RU-29, RU-31, RU-33, RU-37, RU-148, and RU-150, as discussed in Section 11.5 of the licensee's Updated Final Safety Analysis Report (UFSAR).
The inspector requested that operations roll out the chart paper to determine the last time the chart recorder had printed its points.
When rolled out, the chart paper revealed that the recorder had not recorded the RMS data since December 30, 1991, when the chart paper was replaced.
The inspector asked the chemistry. staff and the Unit-1 control room staff who was responsible for the operation and maintenance of 1J-SgA-RR-0029.
The chemistry staff was responsible for checking the operation of 1J-S(A-RR-0029 on a weekly basis.
The chemistry
=
.supervisor gave the inspector a copy of the preventive maintenance (PM) work order (WO) that authorized cleaninq and inspecting the recorder on December 23, 1991.
508858, the inspector concluded that it was a detailed recorder PM package..
The inspector reviewed Unit-1 operations requirements for. assuring the recorder's operation.
Appendix A of Procedure 40DP-90P05
"Control Room Data Sheet Instructions" included the 1J-S(A-RR-0029 recorder check for day and night shift.
Procedure 40DP-90P05, Section 3. 11, states in part that:
"When checking control room back panels, the operators should check for normal configuration given the current plant conditions.
Power, chart paper, door position, etc.
should be checked.
If all is normal, the operator should so indicate by placing a check mar k (/) in the block. If all conditions are not normal, a note should be made at the bottom of the page in the remarks section explaining the condition.
This would be considered an abnormal reading and dealt with as such.
The inspector examined control room data sheets with the 1J-S(A-RR-0029 recorder check from December 30, 1991, to January 2, 1992.
The inspector found that all the recorder checks for six sh>fts were marked indicating all conditions were normal.
Procedure 40DP-90P05, Section 3. 17, states in part that:
"Recorder charts shall be appropriately marked at the start of each day with the date and time by the individual assigned to work that area.
Each shift should check each chart for synchronization with the Control Room clock, proper inking, and initial near the beginning of the'hift."
The 'inspector found that the chart recorder 1J-S(A-RR-0029 had not been verified in accordance -with the above procedure section for at least six shifts, since December 30, 1991.
This was a violation of the licensee's Technical Specification (TS) 6.8. 1 which requires that written procedures shall be established, implemented, and maintained covering the activities recommended in Appendix A 'of Regulatory Gui'de (RG) 1.33, Revision 2, February 1978.
RG l. 33 Appendix A. 1. h, requires administrative procedures for controlling "Log En'tries, Record Retention, and Procedure Review."
This violation is not being cited because the criteria specified in Section V.A. of the Enforcement Policy were satisfied (50-528/91-53-01).
The licensee took prompt corrective action in restoring the chart recorder to pr oper operation.
On January 9,
1992, the licensee issued a Unit-1 Night Order to the Unit-1 operations crew,'hich detailed this potential violation.
The Order stressed management's expectations for correctly checking the control room data sheets, and the chart recorder.
The Order was reviewed and signed by the Unit-1 operators, and the Order was issued to the Units 2 and 3 operations staff.
The Order stated that a Condition Report Disposition Request (CRDR) No. 1-2-009 was written'to investigate this chart recorder problem.
Additionally, the inspector observed a Unit-1 reactor operator adequately perform the multipoint recorder check on January 3, 1992.
The inspector had no further concerns in this matter.
Incident Investi ation Re ort (IIR) Radiation Monitorin S stems The inspector examined the results of IIR 3-1-90-65, completed December 12, 1991, in which the licensee assessed the extent of RMS licensing document discrepancies.
During a 10 CFR 50.59 safety evaluation on revising the RMS alarm setpoint procedure, the licensee found that several RMS design basis documents were incorrect (i.e.
TS, UFSAR, Design Criteria Manual
RMS Description). 'ince this concern had safety implications, the licensee's Plant Review Group (PRG) tasked nuclear instrumentation
controls engineering (NICE) to lead an investigation.
This NICE IIR consolidated severaT different licensee reports on RMS problems such as:
Problem Resolution Sheet (PRS),
equality Deficiency Report ((DR),
CRDR, Engineerinq Evaluation Report (EER) and a vendor report.
This IIR 3-1-90-65 raised the following questions about the RMS:
Mhat was the calculational basis of the setpoints for the RMS?
Why do the UFSAR, RMS description, design criteria and other relevant plant documents not correctly ref'lect the present field equipment configuration?
What administrative procedural faults permitted the RNS documents to become inaccurate?
The licensee's IIR found:
The RNS design basis calculations 13C-S(001 were superseded.
Calculations for most RNS setpoints were not readily available.
RMS setpoints were different from one design document to another.
Design changes were implemented without updating RNS documents.
RMS temporary modifications (T-mod) were in place for too long.
Adherence to document control guidelines were not mandated.
Situations existed that allowed RMS changes without updating design record.
T-mods were installed without NICE concurrence.
RNS changes were not evaluated for operation prior to installation.
RMS field configuration and design documents were deficient and inaccurate.
The licensee's IIR conclusions and corrective actions were as follows:
Thei e was no adequate source or justification for design basis calculations 13C-Sg001 and RMS setpoints.
The superseded setpoint design criteria and basis for RMS will be re-established during the Setpoint and Design Basis Reconstitution program, which will be complete in June 1992.
Stricter administrative controls guidelines wi 11 be established to ensure that RMS design changes are updated promptly and accurately incorporated into the appropriate licensing documents.
RMS field installed configurations will be compared to the TSs, UFSAR and other licensing design documents.
The inspector concluded that the IIR process provided an integrated approach to understanding and resolving RMS problems.
The inspector
reviewed the operability status of the RMS, and found it to be as required by the TS and UFSAR.
The inspector. reviewed the current RMS setpoints as described in Section 3(c) of this report.
The RMS adequately performed its designed safety objectives.
The, inspector had no further concerns in. this matter.
RMS Set pints Basis vs Re ulator Guide 1. 105 The. inspector examined the current status of the RMS setpoint program with respect to operational safety.
In Section (b) of this report, NICE committed to re-establishing new RMS design basis setpoints by June 1992.
The licensee's setpoint program was committed to Regulatory Guide (RG) 1. 105, "Instrument Setpoints,"
November 1976, Revision 1, by the UFSAR Chapter 1.8.
The licensee's Engineering Evaluation Request (EER) No. 90-Sg-100, dated November 28, 1990, re-evaluated the applicability of RG 1. 105 to safety related RMS setpoints in regards to calculating instrument loop uncertainty errors and setpoint errors.
The EER also examined the basis of the current operating RMS setpoints.
The inspector examined the EER's findings and discussed them with the RMS engineer who dispositioned EER 90-S(-100,
,The inspector had discussions with NICE who will be performing.the setpoint reconstitution and writing TS interpretations on setpoints/limits.
The EER 90-Sg-100 found that the original RMS setpoint calculatiori basis were superseded, but it did not mean that the current RMS setpoints used for operations had no documented basis.
Additionally, the EER clarified that RMs RU-30, RU-31, RU-37, RU-38, and RU-145 were safety related.
The EER stated that RG 1. 105 was not applicable, and RMS setpoints were in 'accordance with the Offsite Dose Calculation Manual (ODCM),
The EER further justified the view that each TS setpoint contained a
high degree of conservatism.
This view was consistent with the licensee's TS Interpretation (TSI)¹ 201 dated June 12, 1987, which stated in part:
"Absolute values listed in the TS are assumed to be limits which are not exceeded in the safety analysis.
To,maintain validity of the safety analysis, these values must not be exceeded."
However, the inspector noted that TSI¹ 201 was superseded by TSI 13-07-00 effective September 10, 1991, which states in part:
"Values in the TS were derived using the criteria of RG 1. 105 and such conservatism has already been applied and therefore further "inaccuracies" or "tolerance"'annot be applied since the margins used in the safety analysis would be=compromised."
The inspector discussed with NICE and the RMS engineer the contradictory TSI positions with regard to RG 1. 105.
The RMS engineer issued a
TS interpretation change request on January 3,
1992, to clear up the TSI problem.
The RMS engineer stated that RMS setpoints listed in the TS contain a sufficient safety margin, and the TSI needs to reflect that items specifically called out as
"setpoints" and be treated as such.
NICE held a meeting on
January 3, 1991, and gave a memorandum to the inspector on the RMS setpoint program, its applicability to RG 1.105, EER 90-S(-100, and the TSIs.
NICE committed that reevaluation of the RMS setpoints will be completed by June 1992.
NICE decided that RG 1.105 did not strictly apply to RMS setpoints.
However, NICE concluded that under the setpoint reconstitution program PVNGS was required to comply with RG 1. 105, to be consistent with, industry standard ISA-S67. 04-1988,
"Setpoints for Nuclear Safety-Related Instrumen-tation."
I The inspector determined that the issues were:
whether or not the original RMS setpoints, 13-S(001, during the design basis, had an adequate safety margin calculated into the setpoints to assure that the safety limit parameter would not be exceeded due to instrument. loop and RMS inaccuracies; when the original 13-S(001 design basis calculations were superseded, did subsequent setpoint determinations re-establish a safety margin.
The inspector concluded that based on the licensee's efforts on the Setpoint Reconstitution program and the EER:
The licensee originally complied with the intent of RG 1. 105 to have adequate safety margins in the RMS instrument loops.
Although the licensee superseded the original RMS design basis calculations (13-S(001),
the current RMS setpoints are still consistent with 13-SQOOl, and the ODCM, The RMS Setpoint Reconstitution will reassure that the intent of RG 1. 105 Revision 1 is met.
The inspector examined the licensee's current
"RMS Effluents Monitor Setpoint Calculations for 1991," dated January 30, 1991.
These setpoint.calculations were based on 1X failed fuel mix as specified by procedure 74RM-9EF42.
The licensee determined that the 1991 setpoints would be unchanged from the 1990 setpoints.
The inspector verified that alarm/trip setpoints for the RMS were maintained in accordance with ODCM and applicable procedures.
The inspector had no further concerns with this matter.
d.
RMS Li ht Emittin Diodes (LEDs) for Source Checks The inspector examined the process and effluent RMS source check program to determine if it was in compliance with TS Table 4.3-8.
On April 30, 1991, the licensee submitted a proposed change to TS Table 4.3-8, "Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements."
On September 26, 1991, the NRC's Office Nuclear Reactor Regulation (NRR) approved the proposed TS change as Amendments Nos.
56, 43,and 29 for Units 1, 2, and. 3 respectively.
The TS Amendments specifically allowed the licensee to use LEDs as source checks for noble gas activity monitors RMs RU-12, RU-141, RU-143, and RU-145.
The licensee's TS 1.33 defined a Source Check as "a qualitative assessment of channel response when the sensor is exposed to a source of increased radioactivity."
TS Tabl,e 4.8.3-8,
"Table Notations,"
number (7) was added,.and it states that,
"LED may be utilized as the check source in lieu of a source of increased radioactivity."
A major consideration in the NRC's decision to allow this TS change was the.
licensee's EER 90-Sg-094 which stated that the RMs had no credible failure that could be detected by a radioactive source that a
LED could not detect also.
The inspector examined the licensee's procedures 74ST9S(04 and 74ST9S(06 for performing source checks on the RNs dur'ing the interim period prior to approval of the TS Amendment.
Licehsee Event Report (LER) No. 90-012-00, dated December 26, 1990, addressed the licensee's use of LEDs contrary to the TS 1.33 definition.
The LED issue in LER 90-012-00
,also, addressed in NRC Region V Inspection Report 50-528/91-13, stated that the licensee would submit a TS change proposal to the NRC by April 30, 1991.
The. LER explained that procedural controls were 'in place for. the RNs to be source checked using a radioactive source.
The inspector confirmed that the procedures had provisions to use radioactive sources as alternate check sources.
The inspector had no further concern in this matter.
RMS Calibrations and Detector/Laborator Com arisons The inspector verified that surveillance requirements for the RMS were being maintained and implemented by the methods allowed by TS Table 3.3-6, TS Table 4.3-8, NUREG 0737, Table II.F. 1-3, and UFSAR Chapter 11. 5. 2.
Process and Effluents RMS The inspector discussed RMS calibration and detector/laboratory comparison programs with the RMS engineer, the RMS/effluents supervisor, and a
RNS technical advisor.
The licensee routinely performed. cross checks between effluent lab samples and RMS readings as additional verification of RMS accuracy.
If there was a 30K variance between an effluent lab sample and a
RMS reading, the RMS supervisor was notified for advice.
The licensee did not write a procedure for this comparison process, nor was it specifically committed to in the UFSAR.
However, the effluents comparison process routinely assured that permit release rates and RNS setpoints were not exceeded.
The licensee's UFSAR Chapter 11.5.2 does not require isotopic calibrations of the RNS, only a single point calibration to confirm detector sensitivity.
Full isotopic calibrations were performed at the factory, and the factory provided field calibration sources and reference decay curves.
The UFSAR states that the RMS detector geometries cannot be altered, therefore, subsequent calibrations were based on known correlations between the detector response and
field calibration standards.
The inspector pointed out that TS Table 4.3-8 Notation (3) requires the initial channel calibration. be performed using one or more certified National Institute for Standards'and Technology (NIST) radiation sources or factory
- obtained standards that were traceable to NIST.
The licensee repiesentative stated that the initial calibration on the RMS was per'formed at the factory, and that only the last sentence in TS Notation (3) applied.
That last sentence requires subsequent channel calibrations to use sources that were related to the'nitial calibration.
The licensee gave a copy of EER 89-S(-157, completed June 14, 1991, to the inspector to examine.
, EER 89-Sg-157 was an energy response test of eight RMs using three different beta radiation sources (Tc-99, Cl-36 8 Sr-90) that were similar to what was used at the factory initial calibration.
The test objective was
,to determine if primary "In Situ" calibrations were needed on RMS '
effluent and process monitors as suggested by EER 86-Sg-030.
The results of the test suggested no "In Situ " calibration was necessary, because none of the RMs tested had a response which differed by more than 15% from the factory calibrations.
The inspector concluded that the process and effluent RMS response was acceptable.
Containment Hi h Ran e Area Monitors The inspector examined the area radiation monitoring instrumentation program.
The calibration and test requir ements for this part of the RMS were in TS Table 3; 3-6 and UFSAR Chapter 11.5.
The inspector examined the calibration methods for Containment High Range Monitors (CHRMs) RU-148 and RU-149.
The licensee's UFSAR Chapters 11.5 and 18. II.F. 1.3 committed them to additional requirements for the Containment High Range Monitors found in NUREG-0737, Table II.F.1-3.
The licensee's TS required that CHRMs RU-148 and RU-149 receive channel calibrations every 18 months'
channel calibration is defined in TS 1.4 as;
"The adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which. the channel monitors.
The channel calibration shall enc'ompass the entire channel including the sensor and alarm and/or trip functions..."
NUREG-0737, Table II. F. 1-3 requires that CHRMs have in situ calibrations for at least one decade below 10 Roentgen/hour (R/hr)
by a calibrated radiation source.
Additionally, the NUREG states the original laboratory calibration is not an acceptable position due to the possible differences after installation.
The licensee's representatives gave the inspector the following documents for review:
Procedure 36ST-9S(08,
"Radiation Monitoring Calibration Test for New Scope Area Monitors," and Procedure Change Notice (PCN)
No. 3, dated October 21, 199 EER 90-S(-107, dated December 20, 1990, on CHRMs RU-148 and RU-149 evaluated the licensee position on complying with NUREG-0737.
The EER suggested that in situ tests be performed, since the original installation at Palo Verde did not include the test.
Instruction Change Request (ICR) 18057, was initiated December
1990, and completed October 23, 1991.
The ICR allowed the CHPMs to use internally mounted Am-241 sources for meeting the intent of NUREG 0737 Table II.F. 1-3, with regard to the in situ calibrations and a radiation source at least 1 decade below 10
'/hr.
The inspector's review centered around two questions related specifically to the licensee's CHRM design:
Was the Am-241 internal radiation source equivalent to the NUREG-0737 calibrated radiation source at least one decade below 10 R/hr, and not an electronic calibration?
Was the NUREG-0737 in situ calibration requirement satisfied by the CHRM internal test Am-241 radiation source?
The inspector reviewed vendor documents on the CHRM type ionization chambers, factory calibrations, and the internal Am-241 source.
Additionally, the inspector reviewed licensee s
CHRM calibration data, calibration procedure, and the test results of EER-90-Sg-107.
According to the vendor data, the operating characteristics of the Am-241 internal test source was equivalent to a 1 to 5 R/hr source.
Each Am-241 source generates a continuous current of about lE-ll to 5E-11 amps at the tsme of primary calibration, which was equivalent to gamma response sensitivity in units of amps/R/hr.
The licensee s
EER-90-Sg-107 tested the CHRMs (RU-148/149) for Units-l, 2,
3 by comparing the internal radiation source results to the primary calibrat)on.
The internal source test results for all CHRMs were within 20% of the primary calibration data.
The licensee's ICR No.
18057, allowed calibration procedure 36ST-9S(08 for the CHRM to have an acceptance criteria of +30%.
The licensee used this test data to justify why in situ calibrations, as suggested by the EER were unnecessary.
The inspector noted that vendor documents referred to the internal radiation source test method as an indication of adequate electronic calibration, and that reference may have lead to confusion on whether PVNGS met the NUREG-0737 requirements of "In situ" calibration by electronic signal substitution for all ranges above 10 R/hr."
The licensee's procedure 36ST-9S(08, clearly used electronic signals for all ranges above 10 R/hr, during CHRM calibrations, and the internal radiation source for the range one decade below 10 R/h ~
4.
As previously stated in this report, the licensee's UFSAR requires only a single point calibration with the factory provided source and reference curve for verifying RMS sensitivity.
On the question of the licensee performing in =situ calibrations on the, CHRMs, again, the UFSAR stated that the RMS detector geometries cannot be altered, therefore, subsequent calibrations were based on known correlations between the detector response and field calibrations standards.
Additionally, the licensee's EER 90-Sg-107 anQ ICR-18057 validated their position with regard, to in situ calibrations on CHRMs and NUREG-0737.
The technical merits of the licensee's method for meeting NUREG-0737 requirements using the internal Am-241 source was discussed with NRR and will be considered an unresolved item pending further review (50-528/91-53-01).
An unresolved item is a matter about which more information is required to ascertain whether it is an acceptable item, a deviation, or a violation.
The licensee's RMS'rograms appeared to meet the safety objectives of the TS, UFSAR Chapter 11.5, and the ODCM.
One non-cited violation, and one unresolved item were identified; no deviations were identified.
Exit Interview The inspector met with the individuals denoted in Section 1 at the conclusion of the inspection on January 3,
1992, The scope and findings of the inspection were summarized.
The licensee was informed of the non-cited violation discussed in Section 3(a).
The licensee acknowledged the inspector's observatio ~)
n
~ 0 C
o 0 +oeo+
~ It~lint&Is U.S. NUCL R REGULATORY COIVlMISSIO Revision 2'Nml dry 1986 RE ULATORY U IDE OFRCE OF NUCLEAR REGULATORY RESEARCH REGULATORY (Task IC INSTRUMENT SETPOINTS FOR A. INTRODUCTION Criterion 13.
"Instrumentation and Control,"
of Appendix A, -General Design Criteria for Nuclear Power Plants," to
CFR Part 50,,"Domestic I.icensing of Production and'tilization Facilities," requires, among other things, that instrumentation be provided to moni-tor variables and systems and that controls be provided to maintain these variables and systems within prescribed operating ranges.
Criterion 20,
"Protection System Functions,"
of Appendix A to 10 CFR Part 50 requires, among other things, that the protection system be designed to initiate operation of appropriate systems to ensure that specified acceptable fuel design limits are not exceeded.
Paragraph (c)(1)(ii)(A)of'll 50.36, "Tcchnical Specifi-cations," of l0 CFR Part
r'equires that, where a
~
~
limiting safety system setting is specified for a variable on which a safety limit has Itcen placed, the setting be so chosen that automatic protective action will correct the most severe abnormal situation anticipated without exceeding a safety limit. It also requires the licensee to notify the NRC of any automatic safety system mal-functions, to review. the mat ter, and to record the results of the review. Setpoints that exceed technical specification limits are considered a malfunction of an automatic safety system.
This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations for ensuring. that instrument setpoints are initially within and remain within the technical specification limits.
The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.
'The substantial numbcc of changes ltt this cevtsion has made It impractical Io indicate fhe changes wilh lines In the macgln.
USNRC REGULATORY GUIDES Ragulaloty Guides are lssuad fp dasctlba and make avsllsbl ~ to the public mafhoas acceptable to Iha NRc slaff of Implamanlln9 specific parts of lha Commission's tagulsf lens, to dallnasla tech-nluuas usaa by the staff In evaluating spaclfle problams or posfu-Isfaa accidents, or to provide guidance to applicants.
Ragulsloty Guides sta noi substllulas fat tayulsf lans, snd compliance wllh I
sham ls nol tauultea. Mefhoas snd solullens different from those sat put In the guides will be szcepfsbla If they ptovlaa a basis for the
'findings tattulslta Io Inc Issuance or continuance of a patmll or license by Zha Commission.
GUIDE 1.105 0104)
SAFETY-RELATED SYSTEMS Any information collection activities mentioned in this regulatory guide are contained as requirements in 10 CFR Part 50, which provides the regulatory basis for this guide.
The information collection requirements in
CFR Part
have been cleared under OMB Clear-ance No. 3150I1.
B. DISCUSSIOM Revision I to Regulatory Guide 1.105, "Instrument Setpoints," was published in November 1976 in response to,the large number of reported instances in which instrument setpoints in safetycelated systems drifted outside the limits specified in the technical specifications.
Using the method described in RcvLdon I to ReguLitory Guide 1.105 and additional criteria on establishing and maintaining sctpoints, Subcommittee SP67.04, Setpoints for Safety-Related Instruments in Nuclear Power Plants, under: thc Nuclear Power Plant Standards Committee of the Instrument Society of America (ISA) has developed a standard containing minimum requirements to bc used for establishing and maintaining setpoints of individual instrument channels in safetycehted systems.
This stan-dard is ISA-S67.04-1982,
"Setpoints for Nuclear Safety-Rehted instrumentation Used in Nudear Power Plants."'e Some key terms used, throughout ISA-S67.04-1982 are not defined or have unclear applications. For czsn-venience, thc following information is provided: (I)
the definition of the term "safety hmit" Is contained in 0 50.36 of 10 CFR Part 50, (2) the term
"allowable value" as used in the standard is consistent with the usage in the bases sections of the Standard Technical Specification (STS),aae (3) the term
"upper setpoint aaCoples ata available from the Instrument Society of Amctka, P.O. Box 12277, Raseatch Tciangla Pack. North CatoBna 27709.
aaaNUREGAI03, Revision 4, "Slandatd Technical Spaclfice-fiens for Babcock and Wilcox Pcessutized Walec Reaciuts":NUREG-0123, Revision 3, "Slandacd Technical SpccNaafions fot Gcnccal Elecftlc Boiling Wales Reactors (BWRIS)": NUREG4212, Revision 2, "Slandstd Tcchnical Specifications for Combustion Ettgfnacrttzg Ptassucized Water Reacfots"; and NVREG4I452, Revision 4 "Stan-datd Tcchnical Speci licafions for Wasilnghouse Pcassutizcd Wales Reaafots." Capias of NUREG<cties documents may ba putchasad from the Supetfnfcttdenf of Documents, US. Gpwtnmepf Ptfttt~
Ing Oflice, Post Office Box 37051, Washiagfotl, DC 20013-7052.
wtltfan comments msy ba submitted to the Rules and ptps<<utes Stanch, DRR ADM, U.S.
Nuclear Rayulafpty Comm~
Washing ion, D6 20555.
The yuldas ata Issued In the fallowing fan broad dlvlslanst 1. Power Reactors 6, Ptoducts 2. Rasastch and Test Reactors 7. Ttsnspotlstlpn 3. Fuels snd Mslatlsls Fsclllilas b. occupational Health 4. Envltenmanlsl ana Slllny 9. Anultusz and Financial Review 5. Mslatlsls snd Plant Ptofacf lan lb. General s <<laa was ls<<aa at let eonsldatallen of comments cacalvad from fha public-Comments ana suyyasllans for Improvements In these yulaeS are anaeutayea St all umaS, ana yuldaS Will be taVISad, aS Jpptoprlsle. Io accommodate commanfs sna Io reflect naw Infotma.
zion ot experience.
Capias of Issued guides msy ba putchssad at Zha curtant Govatrlmant Ptlnllny Office ptlca. Infotmsfion on cuttant Gpo ptlcas msy ba obfslnad bv cantscllng the Suparlnlanaant of Documents, UW Government ptlnllng Office post'ffice Box 37052 Wsshlnyfon, DC 20013-7052, telephone f902)275-2060 or I202I275-217 limit" as used,in Figure I of the standard is the same as "trip setpoints" as used in the aforementioned STSs in that drift above the "upper setpoint limit" (standard)
or "trip setpoint" (STSs) requires readjustment.
Paragraph 4.3 of the standard specifies thc methods for combining uncertainties in determining a trip set-point and its allowable values. Typically, the NRC staff has accepted 95% as a probabQity hmit for <<rrors. That is, of the observed distribution of values for a particular error component in the empirical data base, 95% of the data points will be bounded by the value selected. If the data base follows a normal distribution, this corres-ponds to an error distribution approximately equal to a "two sigma" value.
C. REGULATORY POSITION ISA-S67.04-1982, Sctpoints for Nuclear Safety-Reiated Instrumentation Used in Nuclear Power Plants,"
establishes requirements eaceptablc to the NRC staff for ensuring that instnuuaa setpoints in safety-related systems are inithlly vsidun and remain within the technical specification Ihaits. The last section of ISA-S67.04-1982 lists additisrsai standards that are referenced in other sections of ahe.standard..
Those rei'erenced standards not endorsed hy.a regulatory guide (or incor-porated into the
~siations)
also contain valuable information and, if umph, should be used in a manner consistent with current aqpdations.
~
I Section
requires that
"software qualification" bc documented.
Although there is ao generauy accepted definition in thc nuclear industry for software qualifica-tion; the industry has used ANSI/IEEE.ANS-743.2-1982,
"Application Criteria for Programmable Digital Computer Systems in Safety Systems of Ãudcar Power Generating Stations,"
for verification and va5dation of computer software used in safety-related systems.
Reguhtory Guide 1.152,
"Criteria for Programmable Digital Com-puter System Software in Safety-Related Systems of Nuclear Power Plants," endorses this standard.
Some of the considerations in documenting sctpoint drift are (I) the degree of redundancy of the channels for which the allowable limits have been excecdcd, (2)
the type of instrument, inchding the instrument's designed accuracy, function, and plant identification number, (3) the allowable value in the technical specifi-cations, (4) the "as left" setpoint from prior surveillance, (5) the measured setpoint, (6) tbe amount of adjustment in the reported occurrence and the current
"as left" setpoint, and (7) the history of previous testing and the amount of any drift and adjustment in previous testing.
, D., IMKEIIENTATIOM The purpose of this mction is to provide information to applicants.and Iiccnaes.regarding the NRC staff's plans for using this reguhtosy guide.
Except in those ca+>> in which the applicant or li-censee proposes an amcpcable alternative method for complying with specie portions of the Commission's regulations, the methoi'a described in. this guide-will bc used by the NRC aaK in the evaluation of instru-ment setpoints for safe~lated systems with respect to the technical specitcataon limits for thc following nuclear power plants:
1. Plants for which shc oonstruction permit is issued after February 1986.
2. Phnts for which th= opaating.
license applica-tion is docketed 6 mo~.or. morc.aftcrgebruary.1986.
3. Phnts for which the applicant "or licensee 'vol-untarily commits-to the pzcnisions of this guide.
1.105-2
[I
~
VALUE/IMPACTSTATEMENT
~ ~
1. 1IACKGROUND guidance on establishing and maintaining setpo~ in response to the needs that werc apparent from (I) a.
continuing large number of reportlble occurrences and (2) the licensing review of methodology for speciflriag aOowable values and trip sctpoints.
The most common cause of a ac!point in a safety-rclatcd system being out of compliance with plant technical specifications has been the failure to auow for a sufficient margin to account for instrument inaccura-cies, expected environmental drift, and minor calibration variations. For example, in some cases, the trip sctpoint selected was numericauy equal to the allowable value and stated as an
"absolute value,"
thus leaving no apparent margin for drift. In other cases, thc trip setpoint was so close to the upper or lower limit of the range of the instrument that instrument drift placed the setpoint beyond the range of thc hstrument, thus nullifying the, trip function. Other-general causes for a aetpoint being out of conformity with the technical specifications have been 'instrument design Inadequacies and questionable calibration procedures.
2. YALUE/IMPACTASSESSlfENT 2.1 General ISA-S67.04-19S2 is considered stateef-theW acth-
'dology for specifying and reviewing technical specifica-tions on allowable values and trip setpoints, and mem-bers of the industry have incorporated this staldard into their internal procedures.
Further, Iiaragiaphs 50.73(a)
and (b) of 10 CFR Part 50 define whca aa LER is required and what is to be included in an LER respectively.
RcviYion 1 to Regulatory Guide 1.105, "instrument Setpoints,"
was issued in November 1976 in response to the large number of instailces reported in Licensee Even!
Reports (LERs) o!
setpoints drifting outside thc hmits specified in thc technical specifications.
Revision I provided general guidance for (I) specifying setpoints (by considering instrument drift, accuracy, and range)
and (2)
having a
securing device for thc set-point adjustment mechanism.
- I The method described in Revision I to Regulatory Guide 1.105 has been incorpoiatcd into an Instrument Society of America Standard, ISA@67.04-1982,
"Set-points for Nuclear Safety. Related Instrumentation Used in Nuclear Power Plants."
Revision 2 to Regulatory Guide 1.105 was developed to usc the guidance of ISA-S67.04-1982.
This revision provides morc specific 2B Yalue The value to NRC operations and industry is that there would be (I) a systematic method for specifying and reviewing technical specifications on allowable vahics and trip setpoints, (2)
more sophisticated mcdiods for specifying technical specifications, (3) a rcductiaa m sctpoint readjustments, (4) less chance for unwarranted reactor shutdown, and (5) fewer LERs and other report-able occurrences, from the allowable limits of setpoints being exceeded.
Impact Thc impact would be minimal as ISA-S6?.04-1982 represents current industry practice that has been codified in a national consensus standard.
1.105-3-
O.
LJ.S. NUCLEAR RE TORY COMMISSION REGULATORY OFFICE OF STANDARDS DEVELOPMENT REGULATORY INSTRUMENT A. INTRODUCTION
. Criterion 13, "Instrumentation and Control," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Production and Utilization Facilities,"
requires, among other things, that instrumentation bc provided to monitor variables and systems and that
.,
controls be provided to maintain these variables and systems. within prescribe operating ranges.
Paragraph (c)(l)(iiXA) of )50.36,
"Tcchnical Specifications," of 10 CFR Part 50 rcquircs.that, where a limitingsafety system setting is speciTied for a variable on which a safety limit has been placed, the setting bc so chosen that automatic protective action will correct thc most scvcrc abnormal situation an-
=
ticipated before a safety limit is excccdcd.
This.'guide describes a method acceptable to the NRC staff for complying with the Commission's regulatIons with regard to ensuring that the instru-anent setpoints in systems important to safety initially are within and remain within the specilied limits. Thc Advisory Committee on Reactor Safeguards has bccn consulted concerning this guide and has concurred in the regulatory position.
B. DISCUSSION Operating expcricncc 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 occurrence as thc drift of the protective instrument setpoint out-side the limits spectTied in the tcchnical specifications.
n Lines indicate substantive chsntes from previous issue.
. ~
GUIDE Revision
November 1876 GOIDE 1.105 V SETPOlrcTS The single most prcvalcnt reason for the driftof a measured parameter out of compliance with a technical specification is the selection of a setpoint that does not allow a sufIicient margin betwccn the setpoint and thc tcchnical speciTication limit to ac-count for inherent instrument inaccuracy, expected vibration, and minor calibration variations. In some cases, the setpoint selected was numerically equal to thc tcchnical spcciTication limit and stated as an ab-solute value, thus leaving no apparent margin for er-ror. In other cases, thc setpoint was so close to thc upper or lower limit of thc instrument's range that the instrument drift placed the setpoint beyond the instrument's range, thus nullifyingthc trip function.
Other causes for drift of a parameter out of confor-mity with a technical specification have been in-strumentation design inadequacies and questionable calibration procedures.
Thc following terms arc listed with the definitions used in this guide:
1. Instrument accuracy the degree to which an indicated value conforms to an accepted standard value or a true value..
Protective instruments and alarms in nuciear power plants are provided with adjustable setpoints where specilic actions are either automatically in-itiated, prohibited, or alarmed. For example, pres-sure sensors typically, arc installed on main stcam lines to measure stcam prcssure.
These sensors in-itiate corrective action ifthe stcam prcssure decratscs to the. prcdctcrmincd and preset value that would result, for cxamplc, from a steam linc break. Sct-points (e.grs prcssure, differential prcssure, fiow,
~ level, temperature, power, radiation level, time dctay)
correspond to certain provisions of tcchnical speciTtcatioris that have been incorporated into thc operating license by the Commission.
USNRC REGULATORYGVIDES
~Stffvtatory Cuties we issued so describe and mote avai4blo to tho pubfrc
~vwthods ecctetab4 la cle acne staN ol implsrnsntinp specihc parts ol the CammiaaiOn'S rtsulatrana ta deSVWSte CtChniquea uaed by the Stan in eValu aors sptcifrc problems or peatutwtd acciAnls, or to provrde evidence to appli cats Repvtatory Cvides are not substnutes for resv4tions. snd compliance mech them i~ noc rcqvired asvthods an4 sotnwns 4iNerent Nom chose set ovt in mw svides vvirlbe ac ca ptabte rlthey provide ~ basis for the findinst requisite to
<<w issuance or continvance of a permit or tcenae by tlwCommission.
Comment ~ an4 suppsstions for improvements in these Suidts aro encourassd aa aa times. and Svides wra be revrae4. as approprwt ~. to accommodat ~ corn.
~nenu and to rsliecl new mlonnacionar eaptrience. Thrs Suide was rtviaed os ~
~ eavn of substantive comawnta received from tlw pvblic and addktional staff 1. Power lleactors a
ReaearChandTeat RaaatOra 0 fuel ~ and Materials Sacithits Snvironmental and Sitins s Materi ~l~ ond p4nt protection S, Products T. Transponation S. OccvpatwnslHeahh 0 Anthrust Rtv4w 10. Oentraf Copies ol published SuiAs msy be obtaine4 by wrhten request indcatinp ttw divisions desired to the U 5, ffuc4ar Reav4tory Commission. Wsshinaton. 0 C.
aoMS. AltentiOn; Oireetar.ofrrst Of Standard a Deyttapmant, comments shovld lw sent to Ihe secretary ol tho commission. U s ssvc4w Resulatory commrss4n. wssbinacon, oc aosss. Attention, Doctetrrvp ad Serwce Sectbn.
The Svides sre issvedm the fotcowinu ten broad divrsions
~
s g
2. Drifta change in the inputwutput relationship of an inrtrument over a period of time.
3. Nerginthe differcncc between a limitingcori-dition and an operating condition.
4. Range the region within which a quantity is measured, received, or transmitted.
5. ~ety limit a limit on an important process variabk that is necessary to reasonably protect the in-tegrity of physical barriers that guard against un-controlkd release of radioactivity.
6. Setpolnt a predetermined level'at which a bistabk device changes state to indicate that the quantity under surveillance has reached the selected value.
7. Span the algebraic difference between the up-per and lower limits of the range.
8. Technical apecif Ication llmllthe limit prescribed as a license condition on an important process variable for safe operation.
9. Systems important to aafetythose systems that are necessary to cnsurc (1) the integrity of the reactor coolant pressure boundary, (2) the capability so shut down the reactor and maintain it in a safe condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of IO,CFR Part 100, "Reactor Site Criteria."
C. REGULATORY POSITION The following are applicable to instruments in systems important to safety:
I. Tbc setpoints should bc established with suf-Gcient'argin between thc technical specification limits for the process variable and the nominal trip setpoints to allow for (a) the inaccuracy of the instru-ment, (b) uncertainties in thc calibration, and (c) the snstrument drift that could occur during thc interval bet~ ecn calibrations.
? All setpoints should be established in that por-tion ofthc instrument span which ensures that the ac-curacy, as required by regulatory position 4 below, is maintained. Instruments should be calibrated so as to ensure the required accuracy at the setpoint.
3. The range selected for the, instrumentation should encompass the cape>>ted operating range of the process variable bang monitored to the extent that saturation docs not ncgatc the required actioa of the instrument.
4. Thc accuracy ofall sctpoints should be equal to or better than the accuracy assumed in the safety analysis, which considers the ambient temperature changes; vibration, and other enVironmental coadi-tions. The instruments should not anneaL stress relieves or work harden under design conditions to the extent that they willnot maintain the required ac-curacy.
Design vcriTicatioa of these instruments should bc demonstrated as part of the instrument qualiTication program recommended in Regulatory Guide 1.89, "Qualification of Class IE Equipmcnt for Nuclear Power Plants."
5. Instruments shouM have a securing dcvic>> on thc setpoint adjustment mechanism unless it can be demonstrated by analysis or test that such devices will not aid in maintainmg the required setpoint ac-curacy and minimizing setpoint changes. The secur-ing device should be dcsigncd
.so that it can be secured or released without altering the setpoint and should be under administrative control.
6. The assumptions used in selecting thc sctpoint values in regulatory position 1 and the minimum margin with respect tn the limiting safety system set-tings, setpoint rate of deviation (drift rate), and the relationship of drift rate to testing interval (ifany)
should be documented.
D. IMPLEMENTATION The purpose of this section is to provide informa-tion to applicants and licensees regarding the stafFs plans for utilizing this regulatory guide.
Except in those cases in which the applicant proposes an acceptable alternative method for com-plying with specified portions of the Commission's regulations, the method described herein willbe used in the evaluation of submittals in connection with construction permit applications docketed after December 15, 1976, l
Ifan applicant wishes to usc this regulatory guide in dcvcloping submittals forapplications dockacd on or before December I5, 1976, thc pertinent portions I
of the application will be evaluated on thc basis of this guide.
s r'+h.w'...
] 105 2 i,
~
~
roe brut-UHanaT)ON ONLY l1(lib.(((fj~>y h
e a
A
~,
TABLE II.F.1-3 CONTAIN>1EHT NGH-RANGE RADIATION.GENITOR
~ I REgUIREHENT RESPONSE REDUNDANT DESIt'N AND QUA'FIGATION SPECIAL
'
CALIBRATION SPECIAL
'NV IRONHENTAL gUALIFIGATIONS The capabi)ity to detect and measure the radiation level within the reactor containment during and following an accident.
t 1 rad/hr to 1C rads/hr (beta and ganesa)
or alternatively 1 R/hr to 10~ R/hr (gamma only).
60 keV to 3 HeV photons, with linea~ energy response
+ 20Ã)'or photons of.0.1 HeV,to,3 HeV.
Instruments aust Ee accurate enough to provide usable information.
E A minimum of tom physically separated aonitors (i.e.
aonitoring vkoely separated spaces within containme'nt).
~ Category 1 instruments as described in Appendix A, except as listed be)car.
In situ calibration by electronic signal substitution is acceptable for all range decades above 10 R/hr.
In situ calibration for at )east one decade below 10 R/hr sbMa ) be by means of ca)ibrated radiation'ou'rce.'he origina)
laboratory calibration is not an acceptable position due to the possible differences after in situ installation.
For high"range calibration, no adequate sources. exist,,
so an alternate ms provided.
'alibrate and type-test representative specimens of detectors at sufficient points to demonstrate linearity through all scales up to LO R/hr.
Prior to initial use, certify cali-bration of each detector for at least one point per decade of range between 1: R/hr and 10s R/hr.
3-106 II. F:1-13 tie