ML20234E405
| ML20234E405 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 09/16/1987 |
| From: | Andognini G SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Partlow J Office of Nuclear Reactor Regulation |
| References | |
| GCA-87-492, NUDOCS 8709220378 | |
| Download: ML20234E405 (30) | |
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1 if Ih- )SMUD SACRAMENTO MUNICIPAL UTIUTY DISTRICT O P. O Box 1 30 Sacramento CA 95852 3,
91 52 3 p
p RN SEP 161987 GCA 87-492 James Partlow, Director S
i Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation Washington, DC 20555 Docket 50-312 Rancho Seco Nuclear Generating Station Unit #1 AUGMENTED SYSTEMS REVIEW & TEST PROGRAM INSPECTION (ASRTPI) RESPONSE
Dear Mr. Partlow:
As committed in our May 15, 1987 submittal, this letter transmits the District's response to the " Summary of Significant Findings" of the April 10,1987, NRC ASRTPI Report, 50-312/86-41.
This response describes corrective actions taken on programmatic findings.
Comments are response to specific findings within the programmatic responses. provided in The ASRTPI review enhanced confidence in the functionality of Rancho Seco systems and programs. The District's Expanded ASRTP Evaluation (described by June 28,1987 submittal, GCA 87-385) will provide further assurance that the plant is rea@ for restart.
The District is prepared for the followup inspection which will be conducted beginning September 28 to evaluate resolution of the items identified by the ASRTPI.
Sincerely, flybh.
rok
'G arl ndognini Chief Executive Officer, Nuclear Attachment cc:
A. D'Angelo, NRC, Rancho Seco G. Kalman, NRC, Bethesda (2)
J. Dyer, NRC, Bethesda J. B. Martin, NRC, Walnut Creek D. Grimes, NRC, Bethesda T. O. Martin, NRC, Bethesda gy 2
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RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald, CA 95638 9799;(209) 333 2935
t.j
. ~t EXECUTIVE
SUMMARY
The NRC's Augmented System Review & Test Program Inspection (ASRTPI) efforts
'.during the period. December 1,1986 - February 12, 1987, sought to assess the effectiveness of the System Review & Test Program (SRTP) process and evaluate the adequacy of the ongoing programs to ensure that systems will continue to function properly after restart.
This was accomplished by a detailed review of the SRTP as implemented on a sample of 2 of the 33 systems included in the SRTP with a partial review of 6 supporting systems.
The ASRTPI also provided an indepth review of existing programs for engineering and modification, maintenance, operations, surveillance and inservice testing, and quality assurance.
The ASRTPI Report (50-312/86-41), dated April 10, 1987, stated that "the team acknowledges that you [the District) had numerous improvement programs to remedy, in time, the majority of the concerns raised in the enclosed report."
The team also concluded that "the problem identification phase of the SRTP
[ Plant Performance and Management Improvement Plan (PP&MIP)] appeared to be generally effective." The team was concerned, however, that "several examples were noted where the resolution of the identified problems lacked sufficient engineering and operations involvement."
This response addresses the ASRTPI team's programmatic findings as recorded in Section 2 of the ASRTPI Report, " Summary of Significant Findings." This response demonstrates that ongoing operational programs and special corrective action programs are timely and effective.
The following broadly directed programs respond to the ASRTPI findings:
Expanded ASRTP Evaluation Engineering Action Plan Operator Readiness Program Technical Specification Compliance Verification Review a
Redirected Test Program w/ Deliberate Power Ascension Independent Reviews of Test Program and Engineering Calculations The specific findings of the ASRTPI have been processed in accordance with the District's Restart Scope List program. Corrective actions taken in response to the ASRTPI specific findings are reported in context within the programmatic responses.
The District believes its response to the ASRTPI findings accurately reflects the current status of management commitment to effective corrective actions for resolving identified deficiencies.
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.i NRC ASRTPI Report Item 2.1.1 Deoth of Investigation into Identified Problems l
"Although the SRTP [PP&MIP] problem identification process appeared generally I
effective, the inspection team identified instances where the licensee's investigation into the identified problems lacked sufficient engineering and operational depth.
The following are examples of technical concerns with the AFH [ Auxiliary Feedwater) system identified by the team that had remained l
undetected through the licensee's problem review process.
(1) Past testing of the AFH pumps has not demonstrated them to be capable of I
providing the flow required by Technical Specifications.
(2) The condensate storage tank (CST) pressure relief valves appeared to have been set above the design pressure of the tank and were not receiving the i
required inservice testing and the CST vacuum breakers appeared to be l
incorrectly sized.
J (3) The turbine overspeed trip setting for the dual drive AFH pump appeared to be set above the maximum speed rating for the electric motor connected to the common shaft.
(4) The SRTP evaluation of pump damage due to the runout condition experienced during the December 26, 1985 event did not consider potential pump l
l degradation. Additionally, the proposed AFH system design for restart, with the Emergency Feedwater Initiation and Control (EFIC) System modifications, was still susceptible to pump runout under certain situations. At the exit meeting, the licensee committed to install flow limiting devices in the AFH system to prevent pump runout."
District Response For items (1), (2), and (3) above, the District acknowledges that design documentation did not sufficiently justify the design. Additional analyses were performed to fully document the acceptability of the design for items (1) and (2). A modification in the design was required, combined with non-destructive examination to demonstrate acceptability of past operation for item (3).
The AFH technical concern related to AFH pump runout, item (4), was being adequately addressed by the SRTP.
(1)
Licensee Event Report (LER) 87-05, submitted February 12, 1987, reported the ASRTPI finding that the surveillance procedures had failed to conclusively demonstrate the required AFH pump capacities.
The LER stated the surveillance procedures did not provide sufficient administrativetive control of CST miscellaneous flow.
The LER also stated that calibration of the CST level instrument was not at a sufficient frequency or procedural rigor to justify its use for the surveillance test.
V Two additional issues relating'to AFH pump testing were raised by ASRTPI but not reported in LER 87-05. The first issue, documentation of the 60 gpm recirculation. flow rate, should be resolved by an original vendor letter stating the required flow of 60 gpm, and the associated original architect / engineer (Bechtel) orifice sizing calculation that establishes the flow as a plant design value.
Further validation of the 60 gpm i
recirculation flow rate will be achieved by planned testing and l
inspection.
The second issue, procedural requirements for connecting the plastic tubing sightglass to the CST during testing, has been addressed by additional procedural guidance clarifying the need to ensure the tubing is connected at both ends. The District acknowledges that previous procedures did not provide adequate administrative control on use of the plastic tubing sightglass.
Rancho Seco Technical Specification 4.8.1 requires that the AFH pumps be capable of delivering 780 gpm to a steam generator at 1050 psig. The District has submitted a proposed license amendment, with supporting technical analysis by Babcock and Hilcox, justifying a change from the 780 gpm requirement to a 475 gpm requirement.
The 475 gpm requirement is consistent with the AFH pump capacity requirements at other B&W units.
Based upon the actual AFH flow requirements to maintain adequate core cooling, pump performance has been maintained well above the level necessary to assure public health and safety.
The AFH pump surveillance procedures have been revised accordingly since the ASRTPI and are being used to demonstrate required pump capacities. A review of these procedures and recent test data will allow resolution of NRC followup item 86-41-28.
To assure a convenient, dedicated, and accurate measurement in the future, the District has accelerated resolution of AFH SSR Problem (3).
The full flow test line modification, ECN R-1188, will be completed prior to t
restart.
ECN R-1188 is available for review and closure of NRC open item 86-41-03.
(2) The CST pressure relief valve setpoint was procedurally controlled to 2.0 1 0.5 psig.
The working pressure and the nominal design overpressure for the tank, designed to API 620, were both documented at 2.0 psig.
Recent analysis of the CST demonstrates a design allowable overpressure of 5.0 psig.
Documentation of the original CST hydrotest, performed in i
March 1972 to Hydrotest Procedure CHP 101, demonstrates the tank capable of withstanding 6.0 psig overpressure. A recent calculation demonstrates that the CST loop seal blows out at approximately 4.9 psig and is thus j
the tank's ultimate overpressure protection.
The CST pressure relief valves maintain the tank's working pressure. The procedural control on their setpoint has been revised to 2.0 + 0.0/-0.5 psig.
The structural integrity of the CST has not been challenged by its previous operation.
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w The District reexamined the design basis documentation of the CST's vacuum protection.
The maximum negative pressure and outflow conditions evaluated were for maximum gravity drain makeup to the hotwell (outflow from the CST) with AFH pumps operating and an assumed loss of nitrogen makeup to the CST.
The CST vacuum breakers are single failure proof under these conditions.
Each CST vacuum breaker can provide the necessary protection to prevent the CST from exceeding its low pressure design limit.
The possible case of alignment of the CST to hotwell vacuum is precluded by operating procedure, which requires breaking of main condenser vacuum at low level in the CST. The documentation that the CST vacuum protection satisfies 10 CFR 50, Appendix A, Criterion 34 single failure requirements is now in place.
LER 87-14, submitted March 5,1987, reported three deficiencies in inservice inspection (ISI) procedures discovered in February 1987.
Two deficiencies were uncovered during the District's systematic review and rewrite program for surveillance procedures.
The third deficiency, failure to perform a required five-year inspection on one of the CST l
relief valves, was identified by the ASRTPI_ team.
LER 87-14 reported this required inspection was missed due to lack of coordination between the computerized valve test history list and the surveillance test scheduling procedure.
It also reported the relief valve would have fulfilled its function if called upon, as was shown by testing subsequent to identification of the issue.
5,oad scope corrective actions to prevent recurrence were already under way and are presented under Item 2.3.1,
" Surveillance and Inservice Testing Program."
The design basis documentation of the CST overpressure and vacuum protection has been formalized, and ISI procedural weaknesses regarding CST relief valves have been reported and corrected.
This information is available for review and resolution of NRC followup item 86-41-15.
(3) The ASRTPI team identified the turbine overspeed trip for the dual drive AFH pump could be procedurally set above the documented maximum speed rating of the electric motor connected to the common shaft.
The overspeed trip point for the turbine driver was administratively controlled at 4450 rpm i 200 rpm.
The attached Hitachi motor has a documented nominal rated overspeed of 4320 rpm.
The District has been working with Hitachi to assess the real motor overspeed capability.
Hitachi documented that all the motor's components except the rotor upper end ring and the fan vane rivet can withstand a 4650 rpm overspeed and remain within Hitachi's recommended allowable stress.
These two components exceed Hitachi's recommended allowable stress values by less than 17..
Hitachi's recommended allowable stresses are at least three times less than the associated material yield stresses.
The motor was removed, disassembled and a detailed visual examination was performed.
The Hitachi recommended dye pennetrant test was performed on suspected components (fan vane rivets and rotor upper end rings) with no abnormalities noted.
Numerous other tests and measurements were made which showed no unusual indications.
The rotor was rebalanced and motor reassembled and tes ed.
All tests indicate that the motor is capable of l
performing its design functions. --
The District also has been working eith the turbine's manufacturer, Terry Dresser Rand, to re-establish the turbine's overspeed trip band.
The administrative control on the turbine trip is being changed to 4450 rpm +
50 rpm /-200 rpm. A maximum overspeed trip setpoint at 4500 rpm will assure that all Hitachi motor components can withstand future
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hypothetical turbine overspeeds without exceeding Hitachi's recommended I
allowable stress.
The District has no procedural record of either past turbine overspeed trip testing or changes to the trip setpoint. AFH SSR Problem (31) identified that the AFH P-318 turbine trip setpoint was not i
adequately tested and may be incorrect.
The actual trip setpoint has been determined by testing.
The initial trip setpoint per Special Test Procedure STP.1025 was 4481 rpm.
Related issues concerning AFH pump testing are also being addressed.
Testing to determine the minimum required shutdown time before restarting the turbine after a trip was performed with the turbine and motor coupled to the pump. A caution against restarting the turbine prior to the tested minimum shutdown time will be placed in revised operating procedures prior to restart.
Engineering analyses to determine whether the maximum overspeed setpoint of the turbine could overpressurize the AFH system are almost complete.
The results of these analyses and testing will allow closure of NRC open item 86-41-01.
(4) The SRTP evaluation of pump damage due to the runout condition experienced during the December 26, 1985, event considered potential pump degradation.
AFH SSR Problem (43) identified that problem and provided the validated and approved resolution to contact the pump vendor and take appropriate actions based on the vendor's evaluation prior to restart.
The AFH pump vendor. Hayward Tyler Incorporated, responded in May 1987 to the District's inquiry about potential pump runout damage.
Hayward Tyler's evaluation of the transient data concluded that the runout conditions experienced would not impose any damage to the pumps that would affect operability or performance.
Hayward Tyler recommended increased monitoring of pump vibration and temperature levels at both inboard and outboard bearings be conducted for a short time period, about two months for P-318 and about four months for P-319.
For P-319, Hayward Tyler also recommended coupling alignment verification between pump and driver.
The District is implementing Hayward Tyler recommendations.
These actions will allow closure of NRC open item 86-41-04.
The District is not installing flow limiting devices in the AFH system to prevent pump runout.
The NRC's February 15, 1985, Safety Evaluation Report (SER) on the District's response to IE Bulletin 80-04 concluded that "(1) you [the District] have demonstrated that the AFH pumps will survive the runout flow condition for the 30 minutes required for operator action for both the current and proposed upgraded Emergency feedwater System (EFIC System) and (2) your analysis demonstrates that the resulting containment pressure and the reactivity consequences with the EFIC in operation are bounded by the FSAR analysis." The District has reviewed the potential for AFH pump runout and determined its previous position is acceptable.._ _ _
The District is installing flow limiting venturis to assure that the maximum allowable AFH flow into a single steam generator does not exceed the B&W limit of 1800 gpm.
The District identified the discrepancy between the maximum possible AFH flow and the B&W limit in the PP&MIP investigation phase, as documented by AFH SSR Problem (54).
Resolution of this issue has been assigned a restart priority to provide documented compliante with Once Through Steam Generator (OTSG) operability requirements in Technical Specifications.
The potential for exceeding the 1800 gpm limit was being greatly reduced with the installation of EFIC and other AFH-related plant modifications.
The limit is set below the onset of significant flow induced harmonic vibration, a phenomenon which cannot be calculated precisely.
OTSGs of similar design have experienced flows higher than the stated limit and have returned to service with no damage. At the District's request, B&W reaffirmed the current limit of 1800 gpm and confirmed the possibility of raising that limit in excess of 2600 gpm.
B&H felt that additional testing and/or investigation would be required to fully substantiate such an increase.
In parallel, the District investigated the possibility of installing flow limiting venturis to provide positive passive assurance that maximum AFH flows would comply with present limits.
Since flow limiting devices would reduce the flow reaching the generators in the AFW design basis scenario, justification for a reduced minimum AFH flow was necessary to assure core cooling in all cases.
B&W provided the technical justification for a 475 i
minimum AFH flow requirement, which the District proposed in a recent i
License Amendment as previously stated.
After considering the options, the District decided to modify the AFH system by installing flow limiting venturis at a location between the AFH cross connect and the control valves, upstream of the OTSGs.
The flow limiting devices limit maximum AFH flow to a steam generator to about i
1000 gpm. This facility change (Engineering Change Notice ECN R-1672)
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will allow closure of NRC open item 86-41-02.
AFH pump runout can still occur with the AFH system design planned for plant restart.
The flow limiting venturis will provide pump runout protection for the cases of one pump feeding into one generator and two pumps feeding into two generators.
In both cases, the AFH lines are assumed intact.
For the remaining cases of either one pump feeding two i
generators, or one pump (or two pumps) feeding through breached AFH lines, the plant operators should be able to correctly diagnose and terminate the runout condition within 30 minutes using available information.
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the event of one pump feeding at runout conditions ir.to two generators, l
the over 1300 gpm supplied to the generators greatly exceeds the 475 gpm maximum required flow to maintain core cooling and would produce an obvious primary system overcooling. Nonetheless, additional operator procedural guidance and training are being planned.
These actions will j
permit closure of NRC open item 86-41-34.
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NRC ASRTPI Report Item 2.1.2 Problem Resolution Prioritization l
"At the time of the inspection, the SRTP priority system and restart plan did not identify all problems that were to be corrected before restart.
The team j
identified several problems that affected safe plant operation and were not J
currently scheduled for completion before restart. At the exit meeting, the licensee committed to correct the identified problems affecting safety and provide the NRC with a list of all problems that would be corrected before restart."
District Response The District's priority system has been modified to establish a concrete and l
manageable restart scope. This process facilitates the formation of an l
objective, tangible Restart Scope List (RSL).
It also provides the basis for l
a phased approach to continued plant performance improvement.
The priority system was discussed with the NRC on several occasions in 1986.
The District's RSL was submitted to the NRC on September 3, 1987.
l The District's priority system assures that actions will be completed prior to restart which:
"a.
assure the plant remains within the post-trip window b.
assure compliance with Technical Specifications c.
minimize the need for Operator Action outside the control room within the first ten minutes of an event."
The District's methodology assures that actions necessary to enhance reactor safety are identified and incorporated into corrective action programs.
Safety is enhanced by this second level of actions that reduce challenges to safety systems and enhance the ability to remain within the post-trip window.
These actions will be initiated as promptly as practicable, but are planned and scheduled to avoid conflict with restart priorities.
The ASRTPI team reported on eight Main Feedwater (MFW) problems considered to "significantly affect plant reliability" which were not previously classified as restart priority actions. Current restart plans have work proceeding on all eight of the identified problems with six of the eight MFH problems to be complete before restart, as committed in the District's May 15, 1987 ASRTPI submittal.
Many of the concerns raised by the ASRTPI team in the four electrical SSRs reviewed (125 Vde,120 Vac vital, 480 Vac, and 4160 Vac) can be classified as:
- 1) operating procedure enhancements, and 2) improvements of indication /
alarming on loss of electrical equipment.
The District committed to resolve these problems in an orderly manner within the confines of the District's RSL process.1 Representative examples of these issues follow. -
The 120 Vac SSR Problem (6) suggests a procedural enhancement to include a load schedule in Operating Procedure.A.62, "120 Vac Vital System." While a load schedule in this procedure is not required to assure safe plant operation, its inclusion would be a human factors improvement.
The plant can be operated by
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referencing controlled drawings which include load schedules.
An Electrical Alarm and Annunciation System Review has been initiated to address 4160 Vac SSR Problem (25), 480 Vac SSR Problems (26), (10), and (19),
and 120 Vac SSR Problems (12), and (2). Two of these pertain to indication of dc control power to switchgear breakers.
Loss of control power would result in the loss of switchgear breaker (s) in a single electrical distribution train.
This would constitute a single random failure in the electrical system, an
. occurrence for which the plant's design assures safe operation. Credit is taken for shift walkdowns already performed by operations.
Enhanced procedural guidance will be developed prior to restart as a result of the ongoing review.
Additional recommendations associated with this issue are also under review.
The 120 Vac SSR Problem (7) identified the need to update the response to IE Bulletin 79-27 as a priority 1 issue.
The re-review will evaluate the impact of bus failures on the operators' ability to use plant shutdown procedures to reach safe shutdown.
Procedural enhancements or plant r.: modifications identified by this effort are being considered in accordance with the District's restart priority system.
i NRC ASRTPI Report Item 2.1.3 SSR Control I;
'" Selected System Status Reports (SSRs) did not appear to be properly controlled considering their importance as a basis for-the NRC development of the restart Safety Evaluation' Report (SER)."
l District Resoonse Rancho Seco Administrative Procedure AP.93, " System Status and Investigation-
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. Reports," was issued March 13, 1987. AP.93 was based on commitments delineated j
in the Rancho Seco Action Plan, and requirements in AP.92, "SRTP Description and Organization" and the District's priority system.
It formalized the
. policies being implemented. AP.93, Section 6.4.7, "SSR/ SIR Updates" provided for control of technical changes to the problem priorities, problem resolutions, and test program.
These changes are required to undergo the same review and approval cycle outlined for the SSR/ SIR major revisions:
System 1
Engineer, Nuclear Design Engineer, Test Review Group, Performance Analysis Group (PAG), and CEO, Nuclear.
AP.93, Section 7.2, specifies that " controlled copies of each approved revision and interim change shall be distributed in the following manner:
one copy to the. PAG chairman; two copies to the NRC [ Resident Inspector's Office), and one copy maintained by an SRTP designee."
AP.93 implements the SSR control process and is available for review and closure of NRC open item 86-41-43.
As an input to the NRC, the Rancho Seco Restart Report, Rev. O, submitted'on December 1,1986, was structured in a form compatible to support preparation of the NRC's SER for plant restart. This report, as revised in July 1987, focuses the results of the programs being implemented by the Action Plan. The revised Restart Report is the base document from which the Restart Action Plan Report (RAPR) is being developed.
The RAPR will be an integration of the Restart Report, the Action Plan, and the RSL.
It will be the single source 1
documenting the plans the District has implemented to improve performance and restart Rancho Seco. - - - - - - - - - -. - - - - - -.. - - - - - - - - - - - _ _ _ _ _ _ _
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NRC ASRTPI Reoort Item 2.2.1 Ouality of Modifications Designed Durina This Outaae "The following deficiencies were identified with modifications being accomplished during this outage and not reviewed by the SRTP:
(1) After installation of the larger BA and BB batteries, certain circuit breakers on 125 Vdc buses SOA and SOB will apparently be too small for interrupting short circuit current.
I (2)
Inadequate implementation of design requirements resulted in the Interim Data Acquisition and Display System (IDADS) computer inputs being incorrect for-the 125 Vdc bus failure and the AFH pump runout alarms.
-(3) Modifications to the instrument air system appeared to provide incomplete analyses for environmental qualification, specify' incorrect components to accomplish the intended design function, and incorrectly display installation of components on the fabrication drawings."
District Reso_onig At the time of the ASRTPI, significant phases of implementation and testing for modifications designed during the restart outage had not been completed.
Programmatic control of installation, testing, and release of facility changes is achieved by Administrative Procedure AP.44, " Plant Modifications - ECN Implementation." The modifications identified above would not have significantly affected plant safety if implemented as designed.
Field Problem Reports issued during the current outage are being reviewed and trended to ensure continued improvement in the pre-installation design review process.
(1) The de short circuit calculation to verify the compatibility of the new batteries with existing de panel boards did not consider temperature effects.
The short circuit current supplied to 125 Vdc buses SOA and SOB from the batteries increases approximately 157. from the nominal operating temperature of 77'F to the maximum hypothetical operating temperature of 1
104*F.
Failure to account for this temperature effect in the intended design could have resulted in exceeding the nominal circuit breaker rating for one of the two redundant buses, SOB, in the event of a bolted bus fault with an elevated battery temperature.
This is not a common mode failure and SOA is not susceptible; thus, SOA, independent and redundant of SOB, would provide vital 125 Vdc power.
The short circuit analysis was conservative in that it considered a bolted (no contact resistance) fault and did not consider either additional resistances between the batteries and buses, inductance effects, or temperature effects on cable conductivity.
In the unlikely event of a short circuit at elevated battery temperatures, it is not likely that the SOB circuit breakers would be unable to interrupt realistic values of the short circuit current.
The maximum recorded battery room temperature is i
less than 95'F.
Class 2 recirculation fans are available.
The District identified the need to examine battery room cooling in 125 Vdc SSR Problem (9).
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Facility change ECN R-0608, Rev.1, installed a smaller cable between battery BB and bus SOB.
The increased rasistance will ensure that the short circuit current does not exceed the nominal rating of the bus circuit breaker.
DC short circuit Calculation Z-DCS-E0612 has been revised and is available to provide resolution of NRC followup item 86-41-16.
(2) Design errors were made in the implementation of design requirements of IDADS computer inputs for the 125 Vdc bus failure and the AFH pump runout alarms.
The IDADS alarm for 125 Vdc bus failure was intended by its associated Design Basis Report to signal inadequate bus voltage at 105 volts.
Due to inadequate coordination between the Electrical and Instrument and Controls (I & C) design disciplines, the IDADS alarm input was prescribed to be digital and the alarm was set to trigger at approximately 60 1 5 volts.
The operator's first notice of slowly degrading bus voltage would thus be loss of equipment operation rather than an appropriate IDADS alarm.
The 125 Vdc buses are independent and redundant and the battery backups are designea with ample capacity.
The IDADS alarm design was developed circa 1983, before the implementation of the Inter-Disciplinary Review Notice (IDRN) process, which has significantly improved the cooperative multidisciplinary design process.
ECN R-1551 is being implemented to provide an analog signal to the IDADS and alarms at 105 V (degraded bus voltage),125 V (bus off battery charger and on batteries),
and 141 V (bus overvoltage condition).
The IDADS alarm for AFH pump runout was to include a short delay to provide time for the pump to come up to speed so that an unintended alarm would not be received on startup; however, due to an error in translating the existing design from one set of drawirigs to a new set of drawings, the time delay was omitted.
Thus, if actually implemented, the AFH runout alarm would erroneously signal for a short period each time the AFH pumps were started. Other indications would show normal, adequate pump performance and the alarm would clear quickly.
The function of post-modification testing is to detect problems of this nature.
This I
l alarm delay is being installed via the IDADS computer software.
Implementation of the corrected designs will resolve NRC followup item 86-41-21.
(3) The District agreed to install, in conjunction with EFIC, a two hour back-up supply of air to assure that the AFH control valves, MFH control valves, startup feedwater control valves, and Atmospheric Dump Valves (ADVs) will function properly even in the event of loss of offsite power or other loss of normal air supply.
The back-up system is Seismic Category I, and except for low pressure alarm indication, functions totally independent of electrical power supplies.
Proposed operability and surveillance requirements for back-up air power were submitted in Proposed Amendment 152, Revision 2. _ _ _ _ _ - - _ _ _ _ _ _ -
Extensive testing of this Seismic Category I plant enhancement was planned to assure design operability and reliability.
The ASRTPI team recorded
.several concerns with the reliability and documentation of the back-up air system:
(a) ASRTPI Report Section 3.1.2(1) noted that overpressure protection for the main and startup feedwater control valve actuators appeared inadequate.
The hypothetical case of a back-up air system pressure control valve' failure so system air pressure from the nominal 2400 psig bottle supply would be regulated to a pressure above the valve i
actuator rating of 150 psig, but below the system rupture disk j
setting of 225 psig, had not been considered credible.
In response to this hypothetical event, airsets (pressure regulators) rated at 250 psig will be installed upstream of the valve operators and the tie-in to the normal instrument air supply to provide overpressure protection. To further improve system reliability, the relief valve setting is being reduced to 120 psig and the rupture disk design rupture pressure is being reduced to 175 psig.
(b) ASRTPI Report Section 3.1.2(2) noted that seismic qualification of certain back-up air system components had not been documented.
The necessary seismic documentation has now been completed.
The added control of ECN closure checklists implemented with the Engineering Action Plan should help ensure all design package documentation is completed. A related issue, environmental qualification of EFIC OTSG sensing line excess flow check valves, identified by NRC open item 86-41-08, has also been completed.
(c)- ASRTPI Report Section 3.1.2(3) expressed concern with the application of excess flow check valves in the back-up air system design.
The District considers the use of these valves, which were in stock and available, to be acceptable. Unlike many other check valves, the i
excess flow check valves provided the capability for valve position indication.. If their use had adversely affected system reliability, it would likely have been discovered during preoperational testing.
To reduce vulnerability to potential delays resulting from system reliability testing, the excess flow checks are being replaced with soft-seated simple check valves which provide positive shutoff for reverse flow conditions.
Further, periodic test procedures are being established to assure system operability.
l (d) ASRTPI Report Section 3.1.2(4) questioned whether the pressure regulating valves were appropriately designed for their specified application.
The ASRTPI team stated that the currently specified pressure regulating valves, Circle Seal Pressure Regulators SR-830, were not designed for tight shutoff under zero demand conditions.
The valve manufacturer, Circle Seal Controls of Anaheim, CA, was contacted. They informed the District that under their internal QA program each model of pressure regulator is tested to the design pressure (5000 psig for the SR-830 series) for one minute.
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l The acceptance criteria for this test is no bubbles for one minute.
Appropriate action will be taken if preoperational testing or normal operating
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experience-indicate the valves are not leak tight.
Proposed Technical Specifications on the back-up air supply require it to be maintained operable or the plant will enter the associated limiting condition of operation (LCO).
(e) ASRTPI Report Section 3.1.2(5) noted that fabrication drawings for the pressure regulating valves and excess flow check valves were 1
'found to show the valves with improper. orientation. As noted, for i
the case of the excess flow check valves the in/out ports were properly labeled despite the valve outline being shown backwards.
In accordance with plant procedures, fabrication of the panel boards l
with the excess flow check valves was proceeding based on the preliminary drawings in parallel with the issuance of final drawings.
l The valves were correctly installed during fabrication. Verification of the final drawings to the fabrication prior to installation into the plant had.not been performed. The drawing depiction error was i
of no consequence because it was fabricated correctly, and if it had l
not been fabricated correctly, the design would not have functioned and the error would most likely have been discovered during preoperational testing.
Similarly, the incorrectly drawn pressure regulating valves, if installed as drawn, would most likely have been discovered when they failed to perform as designed in their preoperational test.
The Field Problem Report mechanism is employed to provide feedback to the design group from the construction and testing groups.
The root cause of these discrepancies identified in thE back-up air system design was a lack of continuity in the design responsibility for the system. Hith recent management commitment to the system engineer /
l system design engineer concept, this should not occur in the future.
Information will be available to provide resolution to the specific back-up air system concerns associated with NRC followup item 86-41-06.
Engineering Action Plan The Engineering Action Plan, Rev. O, submitted April 17, 1987, describes l
actions initiated or planned to enhance the engineering design process at Rancho Seco. Areas of concern were identified in independent reviews of
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engineering activities by the NRC ASRTPI, INPO, and the SMUD management team.
j The Engineering Action Plan identifies numerous review and programmatic d
development activities to support plant restart and to establish the Engineering foundation necessary for continued operation.
l The plan includes actions to enhance:
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l Design change and modification process j
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j Design Review of work performed during the current outage 1
Engineering baseline of Rancho Seco design
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Management programs to identify and correct problems.
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In the area of design change process improvements, construction specifications are being updated, design data elements of the components in the Master Equipment List (MEL) are being validated, and a configuration status accounting l
process is being established.
The Design Change Package program is being revised to ensure adequate processes for review of proposed design changes, i
comprehensive definition of technical requirements, implementation of changes into. plant configuration and associated technical documents, and pre-review, j
issuance, and' closure of completed design change packages.
A Design Assurance Engineering function has been established to review and.
concur with package closure of restart ECNs (short term) and assist with the detection, evaluation, and resolution of programmatic and procedural weaknesses (long term). Until.the Design Change Package program is fully implemented, an ECN closure checklist is being used to verify completion of ECN documentation, and to record details on the as-built ECN review.
In the area of Engineering baseline development,' System Design Basis (SDB) documents have been written to recapture key design requirements and criteria.
In the process, an in-depth review of system design documents, their validity, and interface requirements is being completed.
The long-term objectives of the SDB documents is contained in Nuclear Engineering Administrative Procedure (NEAP) 4121 " System Design Bases." SDBs have been prepared for all B&W-designed systems. These documents are currently in SMUD Engineering review, which will be' completed before restart. Ultimately, SDBs will be in place for all plant systems important to safe and. reliable operation.
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i NRC ASRTP_LRecort Item 2.2.2 Ouality of Enaineerina Calculations
" Examples of deficiencies were noted in the design calculations reviewed by
. the team including the use of incorrect methods, assumptions, design inputs i
and -acceptance criteria.- Additionally, in some instances calculations did not exist to support the design analyses."
Di stricMpsoonse q,
oT(aEngineeringActionPlan,describedunder-Item 2.2.1,"Qualityof N0 deifications Designed During This.0utage," includes several activities geared l
to improving the quality of engineering calculations. A " good practice" guideli_ne is being prepared to define acceptable technical and quality requirements.for Nuclear Engineering calculations.
The long term 3DB reconstruction effort will identify and correct calculation deficiencies.
This is being accomplished by B&W and Bechtel in a program under way for the review of existing calculations for correctness of methods, assumptions, inputs, references, and acceptance criteria.
The quality of baseline f
engineering calculations and of new calculations will be assured.
In addition, j
a special independent verification of all restart ECNs has been completed by l
Bechtel.
I The independent calculation review performed by Bechtel focused on the technical adequacy and completeness of existing calculations.
The scope involved a detailed review of 260 restart ECNs.
This r'eview resulted in 173
.obsf rvations issued to the Engineering Department for disposition.
Many of these observations resulted from insufficient attention to detail in documenting calculations. A SMUD Engineering review has been completed.
Efforts are under way to resolve each observation by correcting errors, and providing missing information.
It does not appear that any of the observations will result'in the need to modify plant procedures or hardware. Outstanding corrective action and observations closeout activities are ongoing.
Calculation deficiencies identified in the ASRTPI report are being resolved.
The revised calculations will be available for review and resolution of NRC followup items 86-41-25, 86-41-17, 86-41-18, 86-41-19, and 86-41-05.
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l' JRC~ ASRTPI Report Item 2.2.3 Ouality of Drawing Control "Significant deficiencies were noted in the control of system drawings used (M
for plant operations and design engineering projects."
District Response At the time of the ASRTPI, procedures to control' the distribution of drawings and Drawing Change Notices (DCNs) were under preparation.
Nuclear Administration has since released two procedures for the Document Control Section. RSAP-0503, " Design Change Document Control." and RSAP-0505, "SDC Distribution Control," were approved and issued on June 1,1987.
These procedures establish standard methods by which Site Document Control (SDC)
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receives, processes, and distributes drawings and documents that affect the Site Drawing and Configuration Control. QA is scheduled to perform an audit of SDC activities prior to restart.
The ASRTPI' team found controlled drawing files were not being maintained up-to-date. Document Control personnel completed a 1001, audit of the 35 SDC/NEDC (Nuclear Engineering Document Control) maintained yellow / white stick
. files on February 6,1987. ' All deficiencies were categorized during the audit and immediately corrected.
The immediate actions taken to improve controls were:
- 1) updating the computerized DCN file, 2) performing a 1001. audit of stick files, and 3) establishing a formal internal document control audit schedule.
Short term steps to improve controls include:
- 1) expediting the document I
control procedure preparation, 2) developing " NORMS" Information Handling System to assure availability of on-line information and status, and 3) establishing Configuration Management Program interfaces to schedule and control workload impacts.
Long term measures which will be pursued to assure continued quality of document control are:
- 1) evaluating personnel position profiles and attempting to establish promotional opportunities to reduce the present high turnover rate, 2) implementing a satellite document control program to provide for reduction of document distribution while allowing improved access to current information, and 3) maintaining a formal audit program on an ongoing basis consistent with the established schedule.
The document control procedures and plant documents will be available for review and evaluation of progress toward resolution of NRC followup item 86-41-24.
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RBC ASRTPI Report Itam 2.3.1 Surveillance and Inservice Testina Programmatic Concerns "The surveillance and inservice testing program was found to have deficient f
procedures, improper procedure implementation, and inadequate evaluation o?'
r test results."
District Resgonig Surveillance Proaram - At the time pf the ASRTPI, the surveillance program h'd y
a not been upgraded to support plant restart.
The existing surveillance program
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was being maintained while major upp'ade. iMtiatives were proceeding.
The three primary areas of concentration for programmatic improvements are:
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Procedure rewrite effort I
Computerized scheduling development a Day to day administration enhancements I
i In response to NRC Inspection 85-23 and the NRC SALP report issued September 9, 1985, the District committed to:
- 1) revise all surveillance f
procedures to require documentation of calibration data f# all instrumentation ' j l
other than Control Room instrumentation, and 2) perform a complete evaluation of surveillance procedures against a writer's guide and rewrite as necessary.
The completion of this task of rewriting / reformatting all surveillance
,f procedures was originally targeted for cct91etion in early 1987.
1 The surveillance procedure rewrite affyt is proceeding'. As of August 1987, 123 revised procedures have beer apprc>ed for use.
The revised procedures include the required documentation of histrument calibration data in accordance
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with the Writer's Guide for Surveillance Procedures, AP.2.2 4 The present focus of the rewrite effort will ensure that those survei M ance procedures required for plant startup and to maintain operable stattis during power operation will receive an engineering review and be revised to assure technical adequacy prior to restart.
The remaining procedures will be revised during their biannual review to assure technical adequacy.
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, Several restart-related activities affect.the technical aspects of the surveillance procedures.
Equipment modifications and Technical Specification The completion of the upgrade of the 7, amendments must be incorporated. surveillance procedures is therefore scheduled in s 9
of a number of other restart activities.
t Surveillance procedure rewrites are bging accomplished by procedure writers idej'icated to this task.
Prepared procedures are reviewed for adherence to the
'a-Procedure Writer's Guide and for technical adequacy prior to release for intppendent reviews.
Independmf riviews for technical adequacy are performed
'by'a combination of the foll:4 % g: <SKTP system engineer, Test Review Group (TRG),andeithertheTestWordngGroup(THG)orotherappropriatemulti-discipline reviews consistenf Mth AP.2.00, " Rancho Seco Procedure Manual Instruction."
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The surveillance procedure rewrite effor'fwith associated reviews for technical adequacy is expected to identify and rcsolve testing method problems similar to those identified by the-ASRTPI.
TM, specifics of corrective actions on the AFH pump flow surveillance testing, NRC followup item 86-41-28, was addressed under Item 2.1.l(1).
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NRC open item 86-41-29 noted that, without justification, the acceptance criteria for the stroke tide (.of' identical ait operated AFH flow control valves were different. A Special< Test M ocedure (STP) has been prepared and will be performedtoreesQblishthebaselinetestdataforanumberofAFHnon-manual valves. Similar re-establis! pent of l'aseline test data is also proceeding on other systen.sy
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NRC followup item 86-41'-30 identitled several deficiencies with the AFH 4
surveillance procedures. A procedure' d ror that resulted in failure to check the backseat of an AFH pump dhcharge chttk valve was corrected.
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of ASME Section XI rsquired fixed locations for bearing temperature measurements of AFH pumps is being incorporated into procedures.
Fixed thermocouple are being installed by ECN R-1948.
Specification of ASME Section XI required ALERT and ACTION RANGCS for pump differential pressure and pump flow for the AFH pumps is being; incorporated into procedures.
i NRC followup item 86-41-31 identifidi,severalconcernsregardingbattery surveillance and testing.
The upgrading of the battery maintenance and surveillance procedures to be consistent with~IEEE Standard 450, " Recommended Practice for Maintenance Testing and Rep?acement of Large Lead Acid Storage Batteries 'for Generating Stations aad Substations," will be completed by September 1987.
The importance of ?ccurate and complete surveillance data entry has been re-emphasized.
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The Olstrict is proceeding with the previously Lcommitted development of a computeriz H surveillance procedure schedu1 W hy' December 30, 1987 Its i completion Ehould help prevent occurrences of mhsed surveillance such as the 3" a d missed CST relief valve testing noted under NRC followup item 86-41-30 and previously addressed in this response under Item 2.1.1(2).
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The adequacy'of existing control procedures for day to day administration of f'
the surveillance program is-being reviewed.
New administrative procedures l'
will be written for surveillance performance, surveillance data review, and~
scheduling.
The formal procedure prepared for the trending of pump and valve
-test data will provide resolution of NRC followup item 86-41-27.
The necessity of strict adherence to surveillance schedules is being emphasized.
Effective coordination between Licensing and Surveillance is being established to assure licensing actions such as Technical Specification changes are appropriately reviewed and incorporated.
The routine duties and responsibilities of all-personnel within the Surveillance Program are being clearly defined. The
. implementation of the system engineer concept enhances the surveillance program by providing a directed overview for assuring system operability.
l The District continues to progress toward readiness of the Surveillance Program to sup? rt power operation.
Technical Specification Comoliance/ Verification Proaram Supporting the Surveillance Program upgrade, Nuclear Licensing instituted a Technical Specification Compliance / Verification Program to:
- 1) assure that procedures are in place.to implement all requirements of the Technical Specifications, and 2) establish a basis for verifying continued compliance after restart.
This program reviewed the Technical Specifications for observations in the following general categories:
Technical Specification concerns (requirements in Bases, Surveillance o
requirements in Section 3, LCOs in Section 4, deficient requirements)
Implementation concerns (missing, unavailable, or deficient procedures for l
implementing requirements)
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1 Process standards and USAR concerns (inconsistent with Technical Specification requirements)
The following methods were used:
Line-by-line review of Technical Specification requirements.
- Administrative review of procedures to evaluate adequacy in fulfilling Technical Specification requirements.
Line-by-line evaluation of NRC Region V's Evaluation of Rancho Seco's Technical Specifications, transmitted April 30, 1987.
Response is being prepared.
Formation of a computerized intaba which cross references procedures and requirements, and can be upda ad arJ maintained.
Inservice Testina (IST) Proaram - As reported by the ASRTPI, formal NRC approval of the District's IST program expired on June 17, 1986.
The program for the first ten year interval was approved September 25, 1984.
The expiration of the IST program was reported in a July 17, 1986, submittal from J. E. Hard to F. J. Miraglia. The submittal provided the following commitment: )
"The 10 year ISI program, which will be based on the issue and addenda of the ASME Code incorporated by reference.in 10 CFR 50.55a(b) twelve months prior to June 17 1986, will be submitted to NRC for review no later than June of-1987."
In LER 86-21, submitted October 24,.1986, the District stated its understanding of compliance until a new approved program could be developed, "It should be noted that the Code Reliefs documented in the i
September 25,.1984 NRC letter from J. F. Stoltz to R. J. Rodriguez will I
continue until a new relief document has been issued by the NRC for the next 10-Year IST interval.
By letter dated July 21, 1986 from J. E. Hard to F. J. Miraglia, the' District committed to provide the new 10-Year ISI/IST Program by June 1987. This submittal will involve a complete review of the ISI/IST Program."
.In a March 9,1987, submittal from J. E. Hard to F. J. Miraglia, the District clarified its commitment to submit only the revised IST program by June 1987.
The' applicable ASME Code Section IHA 2400 (a) permits the extension of the 10 year ISI interval because the plant had been shut down for a period exceeding one year..Since the extension does not apply to the IST program for pumps and valves.the applicable Rancho Seco ISI Code editions and addenda will be different than the IST Code editions and addenda.
The IST program code edition and addenda will be those incorporated by reference in 10 CFR 50.55a twelve months prior to June 17, 1986.
The ISI Code editions and addenda will be those incorporated by reference in 10 CFR 50.55a twelve months prior to the end of the extended ISI interval, which is approximately six months after restart.
By June 25, 1987, submittal from G. C. Andognini to F. J. Miraglia. the District forwarded the updated IST Program Plan consistent with the program requirements delineated in 10 CFR 50.55a(g)(s) and the ASME Boiler and Pressure Vessel Code, Paragraph IHA-1400(c).
The District requested a meeting be convened to determine the IST testing necessary to support plant restart.
The
. District also requested interim program approval to " ensure the rapid implementation of the more recent ASME Code requirements as well as assist the District by requiring the revision of plant procedures only once, prior to restart." Representatives from the NRC and Rancho Seco are working together to establish IST program compliance.
The proposed IST program plan submitted June 25, 1987, is available for review and resolution of NRC followup item 86-41-26. __
4 NRC.ASRTPI Report Item 2.3.2 Control of Plant Systems and Eauioment Status Trackina
" Deficiencies were identified with the implementation of administrative procedures for the control of plant systems and equipment status tracking."
District Response The Nuclear Operations Department is implementing a new administrative procedure to enhance control of plant systems and equipment status tracking.
The stated purposes of this Administrative Procedure, AP.90, " Work And Test Authorization (HATA) Program," are to:
Document the authorization of work and testing that affect the operational 1
status of components and systems in the Power Block Track the operational status of components and systems in the Power Block Prevent the removal from service of a Technical Specification required e
system / component while the redundant train is inoperable A HATA form is processed fo 11 testing and work to be performed which affects Power Block equipment.
Testing includes Routine Tests (RTs), Surveillance Procedures (SPs), and STPs. Work performed in accordance to the work request control system includes plant modifications, corrective maintenance activities, and preventive maintenance activities. HATA forms require a description of the test / work to be performed, a determination of related Technical Specification requirements, a determination of any system equipment / trains rendered inoperable, and the plant condition required for performance of the test / work.
The approved HATA forms are maintained in the Control Room while work / testing are in process.
An implementation transition period is allowing organization of current work documents already in the field and the Control Room HATA files relating to these work documents. Operations will evaluate using the HATA program to track Abnormal Tags, Safe Clearance Tags, Nonconformance Reports (NCRs), and other plant control mechanisms for which centralized operations status files are useful.
I Additionally, a comprehensive Operator Readiness Program is being conducted to l
assure adequate administrative processes are in place to control operations activities. The Operator Readiness Program will assure that the Operations Department is ready for power operations through detailed review of the following six Operations Department Performance Requirements:
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Operations Organization and Administration - Operations organization and administration ensures effective implementation and control of operations activities.
II.
Conduct of Ooerations - Operational activities should be conducted in a l
manner that achieves safe and reliable plant operations. --
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III. Elan! Status Controls - Operations personnel are cognizant of the status of piant systems and equipment under their control and ensure that systems and equipment are controlled in a manner that. supports safe and reliable plant operations.
IV. Ooerator Knowledae and Performance - Operator knowledge and performance supports safe and reliable plant operations.
V.
Ooerations Procedures and Documentation - Operations p ocedures and documents provide appropriate direction and are effectively used to support safe operation of the plant.
VI. Operations Facilities and Eauioment - Facilities and equipment effectively
. support plant operations.
These Performance Requirements are based on the INP0 Performance Objectives and Criteria for Operating Plants.
For each Performance Requirement, criteria have been established for meeting the requirement.
The basis for meeting each criterion will be documented, reviewed, and approved by the Manager of Nuclear Operations and Director of Operations and Maintenance.
The AGM, Nuclear Power Production and CEO, Nuclear will review and approve each of the six requirements to ensure the intent is met before restart. An independent reviewer will also review the program and results.
The following management activities will be used to assure operator readiness, which includes adequate plant control and equipment status tracking:
Operations management continue to stress the need for procedures and l
verbatim compliance thereto
- Operations internal audits to verify compliance to plant status and control procedures Operator training emphasizes communications, teamwork, transient diagnostics, and compliance to prc:0dures on both off-normal and routine plant evolutions Operations department support staff expansion and corresponding reduction of administrative burden on shift supervisors Optimization of operating crew composition to distribute and balance nuclear experience, technical skills, and supervisory skills Attitude change to operator " ownership" of the plant, and, with that responsibility, recognition of the accountability to ensure things are done the right way.
Operations implementation of administrative procedures for control of plant systems and equipment status tracking will be ready for review and resolution of NRC followup item 86-41-32.
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NRC ASRTPI ReDort Item 2.3.3 Ouality Assurance Programmatic Concerns l
"The Rancho Seco Quality Assurance (QA) program had previously been identified as a major problem area.
Improvements had been initiated in the QA program, l
but the team identified significant deficiencies in this area because the improvements were not implemented at the time of the inspection.
These improvements were delayed as a result of QA involvement with the SRTP process and, consequently, the QA program was not ready to support an operating plant."
1 District Response The responsibility for tracking the Restart Scope List activities has been assigned to Implementation and Scheduling. QA continues to review the implementation of this process for conformance to 10 CFR 50, Appendix B requirements, although their in-line involvement in the SRTP/PP&MIP process has been reduced.
0A Audit Proaram - As noted in the ASRTPI Report, an independent audit of the QA Audit Program was performed in November 1986 by a contractor to the District l
to identify areas for audit process improvement. Audit Report No. 86-02 was i
issued on January 15, 1987.
Eight open items required responses. The QA Department submitted its audit responses to the Management Safety Review Committee (MSRC) on August 5, 1987.
The recently established Quality Subcommittee of the MSRC is reviewing these responses.
The eight audit open items identify a total of 25 findings. As of August 1987, corrective actions had been taken and completed on all but five of these i
findings. Actions are underway to correct these five findings.
Based upon the actions completed and underway, the District's Nuclear Audit Program is now considered to be in compliance with license requirements.
The program-matic actions taken as a result of this District-initiated program audit will satisfy the identified ASRTPI concerns with the QA Audit Program and progress toward closure to NRC open item 86-41-38.
Audits performed during 1986 were not always performed in conformance with the monthly schedule because of manpower or schedule requirements dictated by i
startup tasks.
Several auditors were not performing audits, but were on i
temporary assignment to special projects.
The Quality procedures have been revised to require quarterly audit schedules.
This maintains commitments to the MSRC regarding the performance of audits and gives Quality the scheduling flexibility to effectively perform the high priority tasks.
In addition, the following steps have been taken to strengthen the audit group:
Audit personnel are now located at the Rancho Seco Site to increase the i
group's effectiveness.
Audit personnel, including the Audit Supervisor, have returned to the audit organization.
Five more permanent auditors have been authorized.
The audit group has been augmented with contract personnel.
These steps will increase the timeliness of audits performed during 1987.
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To increase the effectiveness of corrective actions, the audit response format has been extensively revised. Audit response forms state that responses should describe the cause and extent of the adverse condition, define the actions to be taken to correct the existing condition, identify responsible personnel, and state completion dates for all actions. Audit responses are approved by the responsible Department Manager.
The QA Corrective Action Tracking System has also been completely revised since the ASRTPI.
The present tracking system lists incomplete corrective actions separately and tracks them until they are completed.
Corrective Action Lists are distributed monthly to all responsible Department Managers. All audit corrective actions are tracked as separate "Open Items" until they are completed by the responsible department and verified by QA.
The old QA Followup List has been reviewed. A number of old items were closed out and removed from the Followup List during this review.
To facilitate more effective MSRC reviews of audits, the audit report format has been extensively revised to present audit results in a more understandable manner and to stress conclusions regarding program adequacy.
All audit reports are distributed to each MSRC member.
The list of outstanding audit open and corrective action items are distributed to the responsible Department Managers.
In addition, the MSRC Quality Subcommittee reviews, among other things, audit reports and corrective actions.
OA Surveillance Program - The ASRTPI Report correctly identified that checklists were not used to perform previous QA surveillance activities.
QAIP-2, " Quality Assurance Surveillance Procedure," reissued June 2, 1987, has been revised. QAIP-2, Item 5.3 now states, "A checklist may be prepared if deemed necessary."
The revised QAIP-2 now includes a dedicated form for documentation and tracking of surveillance findings.
This will strengthen the corrective action mechanism used to resolve QA surveillance findings.
The surveillance program, as outlined in QAIP-6 and reviewed by the ASRTPI, was considered experimental. A Formal Corrective Action mechanism was under review, pending assessment of the program's overall effectiveness. Prior to having " Findings" on QA Surveillance, Quality would leave the surveillance open if the deficiency was considered significant enough. Copies of these surveillance were distributed to members of the responsible organization. Quality monitored these significant deficiencies through followup surveillance.
In the future, surveillance findings will be trended together with other relevant items in accordance with newly developed QAIP-16, " Trend Analysis," issued June 2,1987.
The ASRTPI team also had a concern that there were no working files of completed surveillance reports in the QA office for management review.
Revised QAIP-2, Item 6.0 now states, "A working file of QA Surveillance Reports and associated documents and correspondence shall be maintained in the offices of the Quality Department for reference purposes for a minimum of one year.
Working file material should be discarded after one year."
The QA Surveillance Program is progressing toward closure of NRC open item 86-41-39.
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06_lertical Audit Proaram - The first QA vertical audit of selected station L
modifications was conducted from April to June 1987.
The following modifications formed the basis of this audit:
EFIC, TDI Diesels Electrical l
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Tie-In, Auxiliary Turbines System and Hot Leg Level Measuring /0TSG Level Measuring Systems. The purpose of-the audit was to assess the operational readiness of the selected modified system by determining whether:
The systems are capable of performing the required functions l
Engineering, design, and construction are in accordance with approved proredures and satisfy applicable codes and standards l
System test requirements and acceptance criteria demonstrate required system functions Changes to licensing documents, procedures, and training programs resulting l
from modifications have been appropriate.
System modification evaluation considered the following:
Conformance to codes and standards Consistency with licensing and design requirements Availability of analyses to support design Appropriateness of calculational assumptions, methods and inputs Application of analyses results to other design documents Completeness, accuracy, and consistency of design drawings Adequacy of procurement, construction, and inspection requirements Availability of seismic and environmental qualification analyses Adequacy of post-modification testing requirements Continuity between design changes and operator training, operating procedures, surveillance test procedures and maintenance procedures.
The review portion of the QA vertical audit has been completed.
Corrective l
actions to resolve the observations resulting from the review are ongoing.
Samples of the audit observations were submitted in the District's July 28, 1987, submittal from G. C. Andognini to D. M. Crutchfield, which provided the Expanded ASRTP Evaluation methodology. Copies of the QA vertical audit report have been provided to representatives of Region V, NRR, and the ASRTPI team. Due to the Expanded ASRTP Evaluation, no further QA vertical i
audits will be performed prior to restart.
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N1C ASRTPI Report Item 2.3.4 Corrective Action Programs
" Licensee corrective action programs had not been managed effectively in the past and at the time of this inspection adequate management attention was stjil not being applied to this area."
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/bistrictResoonse Since the ASRTPI, the District has made the transition from restart organization to a permanent nuclear organizational structure.
The Deputy General Manager has been replaced by the CEO, Nuclear, G. Carl Andognini.
SMUD employees have been chosen for positions of Director of SRTP and Director of Nuclear Quality.
New SMUD employees have been recruited to perform as Assistant General Manager of Nuclear Power Production, Director of Nuclear Operations and Maintenance, Manager of Nuclear Operations, and Manager of Nuclear Licensing.
Each manager has defined a charter for his department which will be supplemented with a list of clearly defined accountabilities.
Management controls are being established to ensure that NRC findings, INP0 recommendations, and QA observations will be resolved in a timely manner.
No contractors are now employed in top management level positions (CEO, AGMs I
and Directors).
The Operations Advisor Program has been implemented to provide j
experienced contractor personnel as shift advisors.
Contractor personnel l
hired as shift advisors under this program held senior reactor operator licenses on B&W commercial nuclear power plants.
The Rancho Seco management believe the nuclear organization is fully capable of supporting the scheduled restart date and NRC open item 86-41-44 should be closed.
Onacing Corrective Action Proarams - As reported in the ASRTPI Report, in the past Rancho Seco did not have a single, uniform mechanism to ensure that "significant conditions adverse to quality" were reported to the appropriate I
levels of management, as required by 10 CFR 50, Appendix B, Criterion XVI. A draft procedure to address this problem was reviewed by the ASRTPI team.
Quality Assurance Procedure 27, " Corrective Action," has been issued.
The ASRTPI team expressed concern that Rancho Seco's trending program did not relate deficiencies to a common cause. A new trending procedure, "QAIP-16,
" Trend Analysis," has been issued.
The objectives of this trend program report, which will replace the "NUMARC Trend Report," are to:
Identify those trends which indicate a significant, recurrent problem or a decreasing level of efficiency with a particular component, system, procedure or program.
Determine the relative extent and severity of a specific problem area once it has been identified.
Provide management with the information necessary to identify root cause(s) for adverse trends in order to assist in formulating effective corrective and preventive action. l
Identify problems and causes which are common to otherwise unrelated areas.
Report the status of those corrective or preventive actions taken for previously identified trends.
The first trend report to QAIP-16 should be available in November 1987.
The ASRTPI team also expressed concern about the District's response to previous NRC inspection report findings.
The establishment of Licensing's Computerized Commitment Tracking System (CCTS) provides management with an effective tool to address ongoing compliance.
The District presented a number of planned programmatic improvements in Section 4B.11, " Commitment Management,"
of the Rancho Seco Action Plan. Many of these committed actions toward development of a long term commitment management process have been completed.
As of August 13, 1987, the NRC, working in conjunction with Licensing's CCTS, has reviewed and closed over 340 open items.
The District considers 182 remaining items as open.
The " Corrective Action" procedure, " Trend Analysis" procedure, and CCTS process are available for review and closure of NRC open item 86-41-40.
LRS Manaaement ADoraisal Report - The open recommendations of the LRS Management Appraisal Report, issued in November 1984, are being tracked by Licensing's CCTS.
The District has committed to re-review the LRS Report items to determine which should be fully resolved prior to restart. The results of this management review of the LRS items will close NRC open item 86-41-41.
Expanded ASRTP Evaluation - The ASRTPI Report states, "At the exit meeting on February 12, 1987, the licensea was informed that, as a result of the apparent weaknesses relating to operating and engineering depth and detail, additional measures to revalidate the adequacy of those reviews would be appropriate.
The licensee shared the same concern and agreed to assess the need for further actions."
The District provided the methodology for its Expanded ASRTP Evaluation in a July 28, 1987, submittal from G. C. Andognini to D. M. Crutchfield.
That submittal explained that the Expanded ASRTP Evaluation will assess system functionality for certain select safety systems using methodology similar to the early 1987 NRC ASRTPI.
The Expanded ASRTP Evaluation will determine the adequacy of activities in support of restart.
It will evaluate the effectiveness of programs established to ensure safety during plant operations.
The Expanded ASRTP Evaluation is a comprehensive assurance effort which complements but does not replace or invalidate other ongoing improvement and corrective action programs. As previously described in the Rancho Seco Action Plan, the SRTP will determine operability for select safety systems and document operability findings in the Rev. 2 SSRs.
These Rev. 2 SSRs will reflect the results of integrated system testing. _ _ _ _ _
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Some of the methods employed by the Expanded ASRTP Evaluation are:
Proven ASRTP-type inspection methods - team approach, detailed-technical review, freedom from checklists with flexibility to follow leads, creative questioning, open communications.
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Six system teams, each'with four to six members from Rancho Seco Operations,
. Maintenance, System Engineering and Nuclear Engineering Departments, and others independent from the Rancho Seco Restart Organization.
Team members chosen for their technical expertise.
Sampling will include assessment of the adequacy of the following activities important to system operability:
- SSR problem resolutions
- ECNs implementing SSR resolutions Functional testing described in SSR
- Conformance of design process to codes and regulation
- Technical analyses supporting modifications
- Technical Specifications reflecting design
- Test procedure implementing SSR testing Operations procedures complete and consistent with design
- Maintenance procedures addressing vendor regs., EQ req., and industry standards
- Maintenance and operator. training The Quality Department will periodically review the program to monitor conformance to evaluation plan, and review completeness of documented results.
The District has submitted the first twelve Expanded Augmented Systems Review and Test Program (EASRTP) Inspection Reports.
These reports provide the findings of the EASRTP Inspection team on the following systems:
Emergency Diesels
- Main Steam Integrated Control Decay Heat Removal
- Nuclear Raw Hater Control Room and Technical Support Center HVAC Systems 120V AC Vital
- Main Feedwater Non-Nuclear Instrumentation
l Seal Injection and Makeup I
Nuclear Cooling Hater e
Radiation Monitoring i
e All potential corrective actions resulting from the EASRTP evaluation /
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validation process are being processed and prioritized in accordance with established Rancho Seco procedures.
The District believes the Expanded ASRTP Evaluation effort will provide an i
important assurance that Rancho Seco's systems and programs are ready for safe operation and will close NRC open item 86-41-42.
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