ML20236J702
| ML20236J702 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/01/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236J684 | List: |
| References | |
| 50-348-98-03, 50-348-98-3, 50-364-98-03, 50-364-98-3, NUDOCS 9807080315 | |
| Download: ML20236J702 (49) | |
See also: IR 05000348/1998003
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l U.S. NUCLEAR REGULATORY COMMISSION (NRC)
i REGION II
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Docket Nos: 50-348 and 50-364
Report No: 50-348/98-03 and 50-364/98-03
Licensee: Southern Nuclear Operating Company (SNC)
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. Facility: Farley Nuclear Plant (FNP). Units 1 and 2
Location: 7388 North State Highway 95
Columbia. AL.36319
Dates: April 12 through May 30. 1998
Inspectors: T. Ross. Senior Resident Inspector
J. Bartley Resid?nt Inspector
R. Caldwell Resident Inspector
J. Zimmerman. NRR Project Manager
W. Kleinsorge. Senior Reactor Inspector
(Section M1.5)
G. Kuzo. Senior Radiation Specialist
(Sections'R1. R2. R3 R7 and R8)
L. Stratton. Physical Security Specialist
(Sections S2. S3. 54, and S8)
W. Sartor. Senior Radiation Specialist
(Sections P2. P3. P5. P6 P7. and P8)
J. Kreh. Radiation Specialist (Sections P2. P3.
P5. P6. P7. and P8)
Approved by: L. R. Plisco.. Director
Division of Reactor Projects
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Enclosure 2
9907090315 990701 .
PDR ADOCK 05000349 !
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I Notice of Violation 3
The NRC has concluded that information regarding the reason for Violation C.
! the corrective actions taken'and planned to correct the violation and prevent
l- recurrence and the date when full compliance was achieved is already
l adequately addressed on the docket in LER 97-10-01. However.-you are required
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to submit a written statement or explanation pursuant to-10 CFR 2.201 if the
description therein does not accurately reflect your. corrective actions or
, your position. In that- case, or.if you choose to respond. clearly-mark your i
response as a." Reply to a. Notice of Violation.'" and send it to the U.S.
Nuclear Regulatory Commission. ATTN: Document Control Desk. Washington. D.C.
-20555 with a copy to the-Regional Administrator. Region II and a copy to the
NRC. Resident Inspector at the Farley Nuclear Plant within 30 days of the date
of:the letter transmitting.this Notice of Violation (Notice).
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If you contest this enforcement action, you should also provide a copy of your 1
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response-to the Director. Office of Enforcement. United States Nuclear-
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Regulatory Commission. Washington DC 20555-0001.
Because your res)onse will be placed in the NRC Public Document Room-(PDR). to
the extent- possi ale, it should not include any' personal privacy. 3roprietary. 4
- or safeguards information so that it can be placed in the PDR wit 1out
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i. redaction. However. if you find it necessary to include such information, you
- should clearly indicate the specific information that you desire not to be
placed in the PDR. and provide the legal basis to support your request for
,
withholding the information from the public.
Dated at Atlanta, Georgia
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this 1st day of' July 1998
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Enclosure 1
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EXECUTIVE SUMMARY
l Farley Nuclear Power Plant. Units 1 and 2
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NRC Inspection Report 50-348/98-03. 50-364/98-03 l
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This integrated inspection included aspects of licensee operations.
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engineering. maintenance, and plant support. The report covers a 7-week
period of onsite resident inspector inspection and announced inspections by
regional inspectors.
Operations
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e Control Room professionalism and communications remained good. l
Operating crew demeanor, team work and conduct were professional and !
effective. Operator attentiveness to Main Control Board (MCB) i
annunciator alarms and response to changing plant conditions were ;
prompt. The operating crew consistently demonstrated a high level of '
awareness of existing plant conditions and ongoing plant activities.
(Section 01.1)
e The inspectors concluded that the licensee adequately prepared for and
then satisfactorily conducted Unit 2 midloop operations. (Section 01.3)
e The Unit 2 cycle 13 initial approach to criticality and restart were
well-briefed. deliberate, and conservative. The reactor core performed
within its design parameters. (Sections 01.5 and 01.6)
e The Unit 2 power ascension for Cycle 13. following the )ower uprate, was
conducted in a safe and controlled manner. The unit aclieved full power
without a significant personnel incident or equipment problem.
(Section 01.7)
Maintenance
o Maintenance and surveillance testing activities were generally conducted
in a thorough and competent manner by qualified individuals in
accordance with plant procedures and work instructions. Close
coordination was maintained with the main control room during
surveillance testing activities. (Section M1.1)
e The corrective actions for the March 28 and May 12 rod drop events were
not thorough, but the corrective actions follcwing the May 15. 1998
event appeared to be comprehensive and effective, pending completion of
the licensee's root cause determination. (Section M1.4) '
e A non-cited violation was identified for the licensee's failure to
report a manual reactor trip in a timely manner. (Section M1.4)
e Inservice Inspection (ISI) activities were conducted in accordance with
procedures and regulatory requirements. (Section M1.5)
Enclosure 2
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e The Inservice Inspection / Nondestructive Examination (ISI/NDE) program
lacked procedure qualification of high temperature liquid penetrant
examination. (Section M1.5)
Enaineerina l
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e The inspectors concluded that the licensee had established suitable
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programmatic guidance to ensure that the regulatory requirements of
10 CFR 50.59 would be met by the various onsite and offsite ;
organizations. However, the inspectors did identify several
programmatic deficiencies and inconsistencies. Training of safety
evaluation preparers and reviewers was adequate. (Section E1.1)
e Changes, tests and experiments were properly screened for 10 CFR 50.59 !
applicability, and adequately evaluated to ensure an unreviewed safety
question (US0) did not exist. Personnel preparing and reviewing safety
evaluatinos were qualified. However, the documentation that addressed
the US0 criteria in several safety evaluations lacked specificity and I
thoroughness. Furthermore, very few of the safety evaluation forms
provided any direct evidence of a cross-disciplinary review.
(Section El.1)
e A violation was issued to the licensee because the original safety
assessment for LER 97-10 was inadequate. In addition, the ability to l
safely shutdown and cooldown the plant from the HSDP was determined to l
have been in a degraded condition for about 12 years. This issue
remains under NRC review and was identified as an apparent violation.
(Section E8.1)
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e A violation of 10 CFR Part 50. Appendix B. Criterion XVI. Corrective ;
Action was identified. The licensee identified three conditions adverse '
to quality of Control Room Ventilation System Functional System Design ,
(FSD) Open Items, which were either inadvertently or inappropriately 1
closed and not corrected. (Section E8.5)
Plant Suonort
e A weakness in exposure controls and poor communications contributed to
the licensee exceeding its budgeted dose for the removal of Tri-Nuclear
equipment and filters from the U2 lower reactor cavity due. '
(Section R1.1)
e For U2 Refueling Outage 12 (U2RF12) activities, dose ex
exceeded original estimates due to expanded work scope,penditure
unexpected
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Residual Heat Removal (RHR) system maintenance problems, and elevated U2 -
Spent Fuel Pool dose rates. (Section R1.2) !
! Enclosure 2 j
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e Worker Shallow Dose Equivalent (SDE) exposures resulting from personnel;
contamination events and work activities during the U2RF12 activities
were evaluated properly and were.within 10 CFR 20.1201 linits.
'(Section R1.2)
e Controls for minimizing workers' internal exposure during U2RF12
activities were effective. (Section R1.3)
e Respiratory protection training, fit tests, medical qualifications, and
equipment status met 10 CFR 20.1703 requirements. (Section R1.4)
e Plant personnel observed working in the radiologically-controlled area
(RCA) generally demonstrated appropriate knowledge and application of
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radiological control practices. (Section R2.1)
e The evaluated Radiation Monitor System (RMS) equipment was installed
properly and the reviewed detector calibrations and functional tests
were conducted in accordance with and met procedural. 10 CFR Part 20.
and Offsite Dose Calculation Manual (0DCM) requirements. (Section R2.2)
-e For 1997. program activities to control. monitor and document liquid and
airborne radionuclides concentrations in effluents and in the offsite
. environment were implemented effectively. No significant environmental
impact was identified. Projected. offsite doses to the maximally exposed
individual were a small fraction of ODCM and'40 CFR 290 specified j
limits. (Section R3.1)
e Extensive delays in returning a community particulate air sampler to
service and lack of corrective actions to prevent recurrence was
~ identified as 'a program weakness. (Section R3.1)
'e The licensee Health Physics (HP) and Dosimetry (DOS) observation program
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continued to be im)lemented effectively and contributed to the reduced
personnel errors o) served for U2RF12 activities. (Section R7.1)
e. Emer :y Response Facilities (ERFs) were well-equipped and
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opf ct onally ready to support an emergency response. Emergency
respuose personnel were adequately trained and responded appropriately
to a scheduled drill. (Sections P2.1 and P5.1)
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.e . Changes to the Emergency Plan were made in accordance with
L 10 CFR 50.54(q). The emergency declaration on March 8. 1998. was made
'in accordance with the. Emergency Plan. (Section P3.1)
e~ The 1996 and 1997 Emergency. Preparedness'(EP) program audits met the
10 CFR 50.54(t) requirement for an annual independent audit of the EP
. prc; ram. (Section P7.1)
Enclosure 2
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e Security personn ? activities observed during the inspection period were
performed well. . te security systems and barriers were adequate to
ensure physical p;stection of the plant and complied with the Physical i
Security Plan. (Section S1.1) !
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e The failure to include a documented process in access control procedures l
for contractors to timely inform the Security Department of terminated
individuals contributed to a violation for failuce follow procedure to l
immediately terminate eight individuals' unescorted access.
(Section S2.1)
e The licensee had in place a sound strategy that was capable of l
protecting vital equipment from acts intended to cause a significant :
release of radioactivity. (Section S4.1) !
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Enclosure 2
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ReDort Details
Summary of Plant Status.
Unit 1 operated continuously at 100% rated thermal power (RTP) for the entire
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inspection period, with the exccption of several hours on May 21, when power
} was reduced to approximately 65%. due to a Steam Generator Feedwater Pump
L trip. The unit reached 360 days of continuous operation as of May 30,
surpassing the previous Unit 1 record of 357 continuous power' operation days.
Unit 2 was.in a refueling outage for most' of the inspection period. On
. May 15. operators attempted to restart Unit 2. However, the unit was manually
tripped due to a:dro ped rod. After making repairs- Unit 2 was returned to
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criticality on May 1 . The unit achieved 100% RTP on May 24 and continued to
operate at this level through the end of the. inspection period.
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I. Operations
- 101 Conduct'of Operations l
01'.1 Routine Observations of Control Room Ooerations (71707 and 40500)
Following the guidance provided in Inspection Procedures (IPs) 71707 and
40500, the inspectors conducted frequent inspections of routine plant
operations.
The inspectors observed that control room professionalism and
communications remained good. Operating crew demeanor, team work-and
conduct were professional'and effective. Operator attentiveness to Main-
Control Board (MCB) annunciator alarms and response to changing plant
. conditions were prompt. The operating crew consistently demonstrated a
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high level of awareness of. existing plant conditions and ongoing plant i
activities.
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The inspectors routinely reviewed the. Technical Specification (TS) y
Limiting Conditions for Operation (LCO) tracking sheets filled out by j
the Shift Foreman (SF). All tracking sheets for Units 1 and 2 reviewed {
by the inspectors were consistent with plant conditions and TS '
requirements. .
01.2._ Unit 2 Refuelina (60710) j
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The resident inspectors observed Unit 2 refueling activities from the j
Main Control Room (MCR)'and containment on April 27, 1998. .The !
,,' refueling was conducted in accordance with FP-APR-R12. " Refueling .)
~ Procedure J.M. Farley Unit 2. Cycle XII - XIII Refueling." Revision !
(Rev.) 0. Refueling activities observed were~ performed in a i
y well-controlled and methodical manner in accordance with procedures. '
Communications between the various stations were clear and concise.
Personnel.were familiar with the procedure ~and knowledgeable of the
process and systems. No significant incidents occurred during_ fuel _
o Enclosure 2 -l
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. handling and all observed fuel a'semblies s were landed in their
' appropriate locations. The inspector concluded that fuel handling was
accomplished in a professional and competent manner.
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01.3 Unit 2 Mid-loco Doerations
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'a, Insoection Scooe (71707)-
The inspectors observed licensee preparations _ for establishing midloop
conditions on Unit'2 in accordance with FNP-2-UDP-4.3. "Mid-loop
- Operations." Rev.'8. .The ins)ector reviewed FNP-2-UOP-4.3.
FNP-2-STP-18.4. " Containment iid-Loop and/or Refueling Integrity
Verification and Containment Closure." Rev.17. and. Generic Letter .
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- 88-17, " Loss of Decay Heat Removal." and verified selected portions of
.FNP-2-UOP-4.3. Section 2.0. Initial Conditions. The inspectors also
performed MCR observations during midloop conditions to verify
compliance with FNP-2-UOP-4.3.
b. Observations a Findinas
The inspectors interviewed Operations and Training supervisors and
determined that the operating crews.had been.provided basic midloop
training during the first training cycle.- In addition, the crew
scheduled to' initiate mid-loop conditions was to receive a detailed
pre-evolution briefi.ng prior.to going into mid-loop operations.
The inspectors reviewed several procedures that were needed for mid-loop
operation and found them to contain adecuate information and appropriate
detail to satisfy the concerns expressec in Generic Letter 88-17.
The inspectors observed MCR operations during mid-loop conditions that
were established and maintained from May 3 through May 5. All required
reactor vessel level indications were functioning properly and closely
monitored by the operators. No significant problems were identified by
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the inspectors.
c. Conclusions
The inspectors concluded that the licensee adequately prepared for and
then satisfactorily conducted Unit 2 mid-loop operations.
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01.4' Unit 2 Preparation'For Startuo and Mode Chanaes
The' inspectors periodically reviewed FNP-0-SOP-103 " Return to Service
' Checklist." Rev.10. and verified that mode-specific' lists were
up-to-date and complete. The inspectors verified that appropriate .
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'signoffs and reviews were completed for each mode change evolution.
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Enclosure 2
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01.5L Unit'2 Startuo and Initial Criticality
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I a. Insoection Scoce (71711r
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L The inspectors observed initial criticality of Unit 2 for Cycle 13 and
the commencement of zero power' physics testing. The inspectors also
observed thel pre-evolution briefing,
b. Observations and Findinos
On May'15. the inspectors attended the pre-evolution' briefing for Unit 2
- Cycle 12 initial criticality and . low power physics testing. Because
L this ~was an infrequently performed evolution, the Engineering Support .
! (ES) manager and test coordinator (i.e. nuclear engineer) conducted the
L . briefing per FNP-0-AP-92. " Infrequently Performed Tests and Evolutions."
Rev: 3. The briefing was attended by all affected parties and was
comprehensive.
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Unit 2 entered Mode 2 on May 15. when' operators began to wittidraw the
control rod shutdown banks. ~An inspector monitored the aaproach to
criticality during withdrawal of the control banks and su) sequent
. Reactor Coolant. System (RCS) boron dilution. .The inspectors verified
.that the-startup was performed in accordance with FNP-2-U0P-1.2.
"Startup of-Unit From Hot-Standby To. Minimum Load." Rev. 39, and
.FNP-2-STP-101. "Zero Power Reactor Physics Testing." Rev. O.
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The' approach to criticality was conducted in a slow, deliberate manner
in strict compliance with procedural instructions. Criticality was
!- achieved within expected bounds of the estimated critical concentration
.~(ECC) and predicted quantity of makeup water needed to dilute the RCS.
All reactivity alterations were precisely controlled and directly-
communicated .to the shift supervisor (SS) prior to implementing any
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^change. The inverse count rate ratio (ICRR) was plotted methodically
l during the entire evolution and reflected control over reactor
reactivity conditions by Operations and ES personnel. Overall. the
Cycle 12 approach to criticality was performed well.
Later that evening during the dynamic rod worth measurement on control
bank *D' with reactor power below the point of adding heat, control rod I
, F10 dropped into the core from approximately 216 ste)s. Operators
manually tripped the reactor and carried out applica)le emergency
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operating procedures. A blown fuse was found on the movable gripper for-
rod F10. The licensee promptly notified the NRC of this event pursuant
to 10 CFR 50.72_(b)(2)(ii). This was the third rod drop event for
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Unitf2 since March 28-(see Section M1.4).
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c. Conclusions
The Unit.2. Cycle 13 initial approach to criticality was all
pre-briefed. deliberate and conservative. The reactor core performed
.within its design parameters.
01.6 Unit 2' Restart and-Low Power Physics Testina
a. Insoection Scoce (71711)
On.May 17. the inspectors observed operators return the reactor to a
critical condition'by-pulling rods and continuing with low power physics.
testing.
b. ' Observations and Findinas
On May 17. after completing repairs to the rod control power cabinets
(see Section M1.4). o)erators recommenced Unit 2 restart. An inspector
monitored the approac1 to criticality during withdrawal of.the control
banks in accordance with FNP-2-UOP-1.2. "Startup of Unit From Hot-
Standby To Minimum Load." Rev. 39. and FNP-2-STP-101. "Zero Power
Reactor Physics. Testing." Rev. O.
The approach to criticality was conducted in a slow, deliberate manner
in strict compliance with procedural instructions. Criticality was
achieved within the expected bounds of the estimated critical position-
- (ECP). All reactivity alterations were precisely controlled and
directly communicated to the SS prior'to implementing any change. The
inverse count rate ratio (ICRR) was plotted methodically during the
entire evolution and reflected positive control .over reactor reactivity
conditions by Operations and ES personnel.
, c. Conclu~sions.
Overall. the Unit 2 Cycle 13 restart was well-controlled and the reactor
core responded ~within design expectations.
01.'7 Unit 2 Power Ascension
a.' Insoection Scooe (71707)
From May 17.through 22.~ the inspectors observed portions of the Unit-2
power ascension and operations, as conducted by associated operating
crews in accordance with FNP-2-UOP-3.1. " Power Operations." Rev 38.
g and FNP-2-ETP-4441. " Power Ascension Following Unit U) rate." Rev. O. In
H addition..the inspectors observed FNP-2-IMP-228,11. ~ils Power Range
Channel N44 Current Rescale." Rev. 18. and portions of FNP-2-ETP-4440.
" Steam Generator Water Level Control Test." Rev. O. at the 33%. power
plateau.
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b. Observations and Findinos
The main generator was synchronized to the grid on May 18 and achieved
full. power.on May 24. The Unit 2 power ascension and power operations
were well-controlled and consistent with FNP-2-U0P-3.1 and
FNP-2-ETP-4441 guidance. FNP-2-IMP-228.11 and FNP-2-ETP-4440 were
conducted in a controlled.. step-by-step manner and completed
1 ' satisfactorily.
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c. Conclusions
-The Unit 2 power ascension for Cycle 13. following the power uprate, was
conducted in'a safe and controlled manner. The unit achieved full
power without a significant personnel incident or equipment problem.
02 Operational Status of Facilities and Equipment
02.1 General Tours of Soecific-Safety-Related Areas-(71707)
related areas were performed by the inspectors
General
to observetours of safety condition of plant equipment and structures, and
the physica
to verify that' safety systems were properly maintained and aligned.
Overall material' conditions for Unit 1 and Unit 2 structures systems.
- and. components (E~3) were good, and safety-related system appeared to
be properly aligne1. Minor ecuipment and housekeeping problems
- identified by the -inspectors curing their routine tours were reported to
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the responsible 5S anFor maintenarice department for resolution.
02.2 Inspections of Safety ,vstems (71707)
InsSectors walked down the newly installed Unit 2 Trisodium Phosphate
Bascets and recently repaired emergency core cooling system (ECCS)-
containment sump screens to verify operability. Accessible portions of
these safety-related system components were verified to be adequately.
installed and in good cperating condition. The inspectors did not
identify any issues that adversely affected component operability.
02.3' Unit 2 Conta' inment Closecut Walkdown (71707)
.0n May 14. the inspectors toured the inside of Unit 2 containment
shortly after entry into Mode 3. 'The inspectors identified a slight
amount of debris which was removed prior-to unit startup. A few minor
leaks and housekeeping problems were reported to the Unit 2 SS for
action. Overall. the licensee did an adequate.' job in cleaning and
clearing out the containment.
- Enclosure 2
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02.4 Verification of Safety Taaaina (71707)
The inspectors verified that selected tagouts were implemented in
accordance with procedural requirements. The inspectors reviewed and
walked down selected devices tagged by the following tag orders (TOs):
e 98-0242-1.1C Coolant Charging Pump Auxiliary Lube Oil Cooling
Pump
e 98-1186-2. 2C Battery Charger
e 93-1843-2. Spent resin sluice pump
e 97-2742-2. Containment Spray Addition Tank
e 98-0467-2. Lower Equipment Room Air Handling Unit
e 97-2458-2. Turbine Driven Auxiliary Feedwater Pump
The inspectors verified that devices identified on the Tag Orders (TOs)
were properly tagged and that the administrative aspects of filling out
the tagging order forms were complete and correct.
The inspectors concluded that the reviewed safety tagging activities
were correct and met the procedural requirements for personnel safety
and equipment protection.
05 Operator Training and Qualification
05.1 Unit 2 Power Vorate License Condition Trainina
On April 29. the NRC issued Amendment No.137 to Facility Operating
License (FDL) No. NPF-2 and Amendment No. 129 to FOL No. NRF-8 for FNP.
Units 1 and 2. respectively. These license amendments authorized SNC to
operate both units of FNP at reactor power levels up to 2775
megawatts-thermal (MWt), which was a power increase from the original
license limit of 2652 Mwt. As part of the license amendments, the NRC
approved certain new license conditions, one of which was that SNC shall
complete classroom and simulator training regarding power uprate for
operations crews on both units prior to the Unit 2 restart (i.e. , before
entering Mode 2) from U2RF12. The Unit 1 reference simulator was also
to be temporarily modified to accommodate the training. On May 5. an
inspector interviewed responsible training instructors and management to
discuss the conduct of operator training pursuant to the new license
conditions.
The inspector reviewed training lesson plan OPS-56202A " Power Uprate."
which addressed power uprate changes to: 1) System and Control
Setpoints. 2) Technical Specifications. 3) Emergency Procedures, and 4)
Accident Analysis.' The inspector also reviewed applicable training
attendance sheets to verify operator attendance for classroom and l
simulator training. Classroom training was held for all licensed
reactor operators (RO).and senior reactor operators (SRO) during January
through April 1998. Simulator training was conducted for the operating
crews between April 22 and May 6,1998. Based upon the interviews and
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document reviews, the ins)ector concluded that the Appendix C license
conditions of Amendments 90s. 137 and 129. which required operator
training prior to Unit 2 restart, were satisfactorily fulfilled.
061 Operations Organization and Administration
I 06.1. Peer Review by World Association of Nuclear Ooerators (WANO) (71707 and i
L- -40500)
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The-inspectors reviewed the Final Report of the "WAN0 Peer Review of
c- Farley Nuclear Plant." conducted onsite during the month of July 1997.
F The inspectors' review of the Interim Report dated September 16. 1997,
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was documented =in inspection report.50-348, 364/97-14. After reviewing
l the. Final Report, the inspector concluded that there were no new safety
,. issues identified which would require NRC follow-up action or.
L reassessment of NRC perspectives regarding licensee performance.
08 Miscellaneous Operations Issues
-08.1Emolovee Concerns Proaram
a. Insoection Scooe (40500) I
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The inspectors performed a review of a sample of Employee Concerns
Program (ECP) files.
b. Observations and Findinas
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The licensee recently dedicated a full-time Jerson to serve as ECP
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' Coordinator to process employee concerns T11s individual was a Shift
Foreman and holds a Senior Reactor Operator (SRO) license. This i
individual has recently begun actively advertising the plant ECP and {
encouraging: people to submit concerns. l
The total concerns for'1995. 1996, and 1997 were 5' Concerns for 1998
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year to date totaled 30. A review of all. ECP packages for 1995. 1996..
and 1997-and 4 ECP packages for 1998 found two that had followup
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commitments. However there was no documentation indicating that these
commitments were. completed. !
~ 0re concern stated that there were no searches of individuals entering
or exiting the contractor Jarking area. Security had committed to l
review their. procedures. )ue to communications problems, security. '
reviewed the entrance / exit procedures for the wrong area. Consequently.
no actions were taken. ECP 3ersonnel acknowledged the error-when the
inspectors questioned them a)out the dis)osition of this concern. Upon
~
o subsequent' review.. the -licensee stated tlat no actions were recuired
l because searches were conducted prior to entering the protectec area. ,
J A1.so, the licensee had independently implemented random exit searches. '
Enclosure 2 .
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o
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8
i
The other concern was associated with inconsistent verification of
. controlled. leakage when swapping a charging pump. The licensee had
committed to change the Unit Operating Procedures (UOPs) and the System
~0perating Procedures (SOPS). This commitment was to be entered into an
informal tracking system.for procedure enhancements. The 50P changes
.were entered into the tracking system and incorporated, but the UDP
= changes were not entered. Consequently, these changes were not made.
However, the. licensee planned to incorporate the changes during the next )
revision cycle of the procedure.
c. Conclusions
There has been a significant increase in usage of the Employee Concerns
Program during 1998. For the ECP packages reviewed, one commitment was
not entered into the tracking system.
,3
II. Maintenance
M1 Conduct of Maintenance
M1.1 Maintenance and Surveillance Testina Activities (61726 and 62707)
Using the guidance provided in IP 61726 and IP 62707 the. inspectors
observed and-reviewed portions of selected licensee corrective and
preventive maintenance activities, and routine surveillance testing
including detailed reviews of the following:
"o FNP-2-STP-40.0, " Safety Injection with Loss of Off-Site Power
Test." Rev. 29
'e FNP-2-STP-40.3, " Phase A Isolation Test," Rev. 1
e WA485614. Replace 1A Condensate Pump bearing
e FNP-1-STP-109.1, " Power Range Neutron Flux Channel Calibration "
Rev. 10
e FNP-0-MP-7.3. " Turbine Driven Auxiliary Feedwater Pump Overspeed
Trip Setpoint Checks," Rev. 4
During the observation of FNP-0-MP-7.3, the inspectors noted several
. . unsuccessful attempts to perform the maintenance. Mechanics were unable !
to adjust the Turbine-Driven Auxiliary Feed Water. (TDAFW) overs)eed trip '
setpoint within the required tolerances. The source of the pro)lem was
p narrowed down to non-equivalent parts that were replaced on the
overspeed trip mechanism during the Unit 2 Refueling Outage 12 (U2RF12).
The licensee. in consultation with the vendor. concluded that a part of
l. the overspeed_ device was " custom fit" at the factory and that the off-
L :the-shelf component would not work. The licensee determined that the
entire overspeed device would be purchased and factory tested in the
future. After the original parts were put back into the overspeed trip
mechanism, a successful test was accomplished the following day. ;
Enclosure 2
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_mi_._____ __-.____i___--_______----- - - _ -
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7
9
'Other observed maintenance work activities and surveillance testing were
performed in accordance with work instructions, procedures, and
applicable clearance controls. Safety-related maintenance and
surveillance testing evolutions were properly planned and executed.
. Licensee personnel demonstrated familiarity with administrative and
radiological controls, Surveillance tests of safety-related equipment
were consistently performed in a deliberate manner in close
. communication with the Main Control Room (MCR). Overall, operators,
technicians and craftsman were observed to be knowledgeable,
experienced, and trained for the tasks performed.
M1.2 Unit 2 Safety In.iection with Loss Of Offsite Power Test
The inspectors observed the preparations for and performance of
FNP-2-STP-40.0, ~ ~ Safety Injection with Loss of Off-Site Power 1st,"
.Rev. 29C, on April 25. The inspectors also observed portions of the
-follow-up testing on April 29. and reviewed the completed test package.
The test was satisfactorily completed with the exception of five valves,
inadvertently omitted due to personnel error, and several relatively
minor exceptions, which were noted and rescheduled.' While recording
' post-test. component']ositions, the licensee identified that five valves
.were not placed in tie required pre-test alignment because.the operator
performing the pretest, lineup ~ missed page 5 of Table 2. The licensee
' initiated.0ccurrence Report (OR) 2-98-162 to evaluate this occurrence.
These. valves, and the known test exceptions. were retested on April 29.
L The inspectors observed the retest and verified that all exceptions from
.the. initial test were included. The test and retests adequately
. verified SI operation with a loss of offsite power.
M1.3 Residual Heat Removal-(RHR) Heat Exchancer head Gasket Replacement
Theilicensee replaced the RHR heat exchanger (HX) head gaskets, under
.
Work Orders (W0s) M00203005 and M00168359. to eliminate small borated
l' water leaks. The job was complicated when seven of the studs which were
threaded through the tube sheet stuck and had to be cut out with
specialized equipment. This significantly delayed completion of work
and resulted in the dose for the job being almost double the budgeted
dose.
'
On May 1. a contractor noted that some of the. studs (one on each HX
l- endbell and multiple studs on the inlet and outlet flanges) did not have
complete thread engagement with the nuts. In most cases, only one or
l two threads were not: engaged. -The American Society of Mechanical
Engineers-(ASME) Code required full thread engagement. The condition
'
L
was missed by the craftsmen. the craftsmen's supervision, and inspection
personnel. The licensee issued OR 2-98-171. evaluated the condition'for
. current plant conditions initiated a formal root cause investigation,
and. installed longer studs where needed for full thread engagement.
' Based on initial calculations, the licensee determined that for the
Enclosure 2
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10
current plant conditions (i.e. Modes 5 and 6). the RHR HXs would meet
their intended function with the as-found thread engagement. Therefore,
based on having adequate bolt strength but not meeting the ASME Code,
the licensee considered this to be a degraded but operable condition.
On May 4. an inspector walked down other systems to determine if
'
inadequate thread engagement was a problem on other safety-related
systems. The inspector identified limited examples of potentially
inadequate thread engagement on the 1A Containment Spray pump supply and
discharge flanges.1C CCP discharge flange. and the 2B RHR pump casing.
After the walkdown, the inspector asked the Plant Modifications and
Maintenance Sup) ort (PMMS) manager if licensee 3ersonnel planned to
walkdown any otler safety-related systems for taread engagement
discrepancies. The manager stated that no further system walkdowns were
planned but that he would discuss the issue with senior plant
management. The inspector then presented his walkdown findings to the
licensee for resolution. On May S. the licensee commenced walkdowns to
identify thread engagement problems on other safety-related systems.
Based on the initial calculations for the RHR HX thread engagement and
the as-designed safety margins the inspector concluded that this was
not a significant safety concern. This issue is identified as
IFI 50-348. 364/98-03-01. Inadequate Thread Engagement, pending
inspector review of the licensee's walkdown results and evaluations.
M1.4 Drooned Rod Durina Rod Doerability Testina
a. Jnsoection Scooe (62707)
Inspectors observed control rod troubleshooting and reviewed the
applicable occurrence reports and Licensee Event Report (LER).
associated with the dropped control rods.
b. Observations and Findinas
On Nay 12. control rod K-2 (control bank A) was dropped during rod
operability testing. This was the same rod that dropped on March 28
during a scheduled Unit 2 refueling outage shutdown (refer to IR 98-02)
when a fuse associated with the stationary gripper coil blew. At that
time. the licensee's troubleshooting concluded that there was an
intermittent problem in the rod control power cable that crossed from
the reactor cavity to the reactor head. The suspected cables were
replaced during the outage. However, for the May 12th event, a fuse
associated with the movable gripper was found to be blown. In both
cases, operators manually tripped the subcritical reactor. The licensee
suspended further control rod operability testing and commenced
troubleshooting to identify the problem. Licensee personnel
subsequently concluded that the movable gripper coil fuse had
experienced a late failure due to the prior overcurrent condition that
blew the stationary gripper coil fuse. Maintenance replaced the movable
gripper coil fuse and exercised the bank several times.
Enclosure 2
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11
Following replacement of the fuse, the inspectors observed Instrument
and Control troubleshooting activities (e.g., electrical power current
traces) of each control rod while operators exercised applicable banks.
The. licensee did not report the May 12 event t'o the NRC until May 15.
Failure to re
by 10 CFR 50. port the manual
/2(b)(2)(ii) reactor
constituted trip withinThis
a violation. four non-repetitive.
hours as required
licensee-identified and corrected violation is being treated as a
Non-Cited Violation (NCV). consistent with Section VII.B.1 of the NRC
Enforcement Policy. This is identified as NCV 50-364/98-03-02. Failure
to Report-Manual Reactor Trip in a Timely Manner.
On May 15, another control rod dropped during zero power physics testing
(see Section 01.5). The licensee's NSSS supplier reviewed the event. A
potential concern was identified regarding suitability of fuse types for
the Control Rod Drive Motor Generator sets and the control rod system.
The licensee had on occasion installed a different type of fuse than-
recommended by the vendor, considered to be equivalent by the licensee.
All of the Unit 2 control rod drive fuses (approximately 160) were
.
replaced with the vendor's recommended fuses and the reactor was
restarted. The inspectors will continue to review the root cause
determinations and the associated LERs. The licensee's corrective
actions. appeared to have been successful and the repetitive dropped rod ;
. problem corrected..pending results of'the on-going Root Cause
Investigation.
c. Conclusions
The corrective actions for the March 28 and May 12 rod drop events were
L
'-
not thorough, but the; corrective actions following the May 15. 1998
event appeared to be comprehensive, pending completion of the licensee's
-root cause determination.
.
A non-cited violation was identified for the licensee's failure to
report a manual reactor trip in'a timely manner.
M1.5- ' Inservice Insoection Unit 2
a. Insoection Scooe (73753)
To evaluate the' licensee's inservice inspection (ISI) program and the 1
. program's implementation ~. the inspectors reviewed procedures observed
work in progress, and reviewed selected records. Observations were i
' compared with applicable procedures the Updated Final Safety Analysis i
Report (UFSAR) and ASME B&PV Code Sections V'and XI, 1989 Edition. No
. Addenda (89NA). l
,
Enclosure 2 'l
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Specific areas examined-included the following observations.: magnetic
particle (MT) examination of Item Nos. APR2-4350-20 and APR1-1300-S35:
manual ultrasonic (UT) examination of Item No. ARP2-4350-20.: visual
(VT-1) examination of Item No. APR1-4303-0V021(B): visual (VT-3)
. examination of: Item No. APR2-4613-SS-12297: and data acquisition
activities associated with eddy current -(ET) examinations of steam
'
'
generator-(S/G) tubing. The. inspectors reviewed selected completed
-
examination reports and_ procedures.
Procedures reviewed included: FNP-0-NDE-100.1.-'" Measuring and Recording
Techniques for NDE: Examinations." Rev. 2.: FNP-0-NDE-100.5 " Liquid
Penetrant Examination (Color Contrast and Fluorescent)." Rev. 4:
FNP-0-NDE-100.11, " Magnetic Particle Examination." Rev. 3:
~FNP-0-NDE-100.21. " Visual Examination VT-1." Rev. 1: FNP-0-NDE-100.22.
" Visual Examination VT-2." Rev. 2: FNP-0-NDE-100.23. " Visual Examination
VT-3." Rev. 2: FNP-0-NDE-100.32. " Qualification of Ultrasonic-
..
Instruments." Rev. 2: FNP-0-NDE-100.31. " Manual Ultrasonic Examination
l . of Full-Penetration Welds (0.200 to 6.0 Inches)." Rev 4. with .TCN 4A:
FNP-0-NDE-100.34._ " Manual Ultrasonic Examination of Welds in Vessels.".
Rev. 6: FNP-0-NDE-100.35. " Ultrasonic Thickness Examination Procedure."
Rev.~1: FNP-0-NDE-100.37. " Manual Ultrasonic Examination of. Reactor
Coolant Pump Flywheels." Rev. 2: FNP-0-NDE-100.38 " Manual Ultrasonic
Examination of Nozzle Inner Radius." Rev. 2: FNP-0-NDE-100.39. " Manual
Ultrasonic Examination of Bolts and Studs Greater than 2 inches in -
Diameter." Rev. 3: FNP-0-NDE-100.40. " Manual Ultrasonic Examination of
Centrifugal Charging Pump Case." Rev.1: and FNP-0-NDE-100.41 " Manual
Ultrasonic Examination of Cast Stainless Steel Pipe Welds," Rev.1. with
.TCN 1A.
The inspectors performed an independent evaluation of indications to
confirm the licensee's ISI examiners * evaluations. In addition, the
-
inspectors conducted an independent VT-3 inspection of the following
su) ports previously examined by the licensee to confirm their.results:
AP11-4301-2HR-R155 and APR2-4619-SS-12459.
1
The inspectors reviewed records for the nondestructive examination (NDE)
personnel and ecuipment utilized to perform ISI examinations. The
records includec : NDE equipment calibration and materials certification
and NDE examiner qualification, certification, and visual acuity.
=The inspectors observed activities associated with insertion and
y expansion of S/G tube sleeves.
b. ' Observations and Findinas
Unit 2 S/G tubing was subjected to ET examination. This examination was
planned to include: -bobbin - 100% full length: + Point rotating pancake
-(RPC) 100%- top of tubesheet (TTS) 3-inches hot leg. 20% TTS 3-inches
cold leg. row 1 U-bends S/G 2A. row 2 U-bends S/G 2C. and all bobbin
indications. For the alternate repair criteria program, the ET.
Enclosure 2
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13
i
examinations were to include all tube support plate (TSP) bobbin
indications over 2 volts. all dents over 5 volts, and large mixed
residuals (33 in each S/G). Due to finding 2 tubes with indications in
the S/G 2C cold leg TTS + Point. the + Point program was expanded to 100%
of the inservice tubes. Due to finding an outside diameter stress
corrosion cracking (ODSCC) indication at one TSP intersection in S/G 2B.
that had a large mixed residual signal the + Point program was expanded
to look at the next 66 largest residuals. No indications were
identified in the expanded sample of 66. !
Procedure FNP-0-NDE-100.5. Rev. 4. Appendix A-2. provided step by step
instructions for the performance of PT examinations in the temperature
range 145 F to 325 F using the following Sherwin Doubl-Chek * visible
solvent removable family: KO-17 Penetrant. D-350 Developer. and KO-19
Cleaner. ASME B&PV Code Section V, paragraph T-647 requires procedure
qualification for PT examinations that are to be conducted outside if
the range of 60 F to 125 F. The licensee indicated that at present. (
it did not have any of the high temperature penetrant consumable
materials. The inspectors determined that no examinations had been
conducted in accordance with FNP-0-NDE-100.5. Appendix A-2. The
licensees *s approval and issuance of a PT procedure for examinations
outside of the range of 60 F to 125 F. without first performing a
qualification in accordance with T-647. was considered an inadvertent
omission of the licensee's ISI/NCE programs.
The inspectors observed several personal safety concerns regarding
improperly secured ladders and Jersonnel working more than six feet
above the floor without pro)er land rails or safety harnesses. The
, inspectors reported these o)servations to the licensee who took
immediate corrective actions to address these issues.
ISI examinations observed / reviewed were conducted in accordance with
approved procedures, by qualified and certified examiners ising
certified / calibrated equipment and materials.
c. Conclusion
i
The Inservice Inspection / Nondestructive Examination (ISI/NDE) program !
lacked procedure qualification of high temperature liquid penetrant j
examination.
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Enclosure 2
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.14
.III. Enaineerina
E1- . Conduct of Engineering
' E1.1' .10 CFR'50.59 Safety Evaluation Proaram And Implementation
a. 'Insoection SCoDe (37001) 'I
I
LBy letter dated December 2, 1997. the licensee submitted Revision
, (Rev.) 14 to'the Updated Final Safety Analysis Report (UFSAR) for the
t , time. period of November 5.1995, to June 4.1997. This letter also
,
included a Summary Report of all changes. tests, and. experiments (CTEs)
L that were completed under the provisions of 10 CFR 50.59 over the same
period. The licensee's December 2 summary included approximately 135 =
changes. ;
m
k '
The ' inspectors' conducted.a review of the licensee's program for meeting
L , the regulatory requirements of 10.CFR 50.59 and. examined its .
3 implementation. ' The inspectors. reviewed applicable administrative and
controlling procedures, training materials and numerous safety
evaluations.. including associated UFSAR Rev. 14 changes. In addition,
the inspectors attended a meeting by the Plant Operations Review
Committee-(PORC) that included review and approval of several safety _
L evaluations. The inspectors also reviewed recent audits conducted in
the area of 10 CFR'50.59 safety evaluations.
~
?
b. ' Observations and Findinas
Proaram Review
Thetins)ectors reviewed onsite Administrative. Procedure (AP)
'..
FNP-0-A)-88; " Nuclear Safety Evaluations." Revs. 2 and 3. and corporate
Farley. Nuclear Project procedure G0-NG-42. ~50.59 Evaluations." Rev. 4
In addition. -the inspectors reviewed Nuclear Engineering Procedure
(NEP)'8-102, " Preparation Of Safety. Evaluations." Revs. 6 and 7. and
Nuclear Engineering Procedure Instruction (NEPI) 4-0, " Design Change
Packages." Rev. 2. The principal offsite design organizations.
'
Southern Company Services =(SCS) and Bechtel, used NEP 8-102 and NEPI 4-0
, . for conducting safety evaluations of plant design changes. At the time
l- : of.- the inspection, the licensee was transitioning from using NSAC-125.
H -
" Guidelines For 10 CFR 50.59 Safety Evaluations,~ to NEI 96-07. -
" Guidelines For 10 CFR 50.59 Safety Evaluations," dated September 1997.
. Revision 2 of FNP-0-AP-88 Rev. 6 of NEP 8-102, and Rev. 2 of NEPI 4-0
were, based on NSAC-125: the other procedures were recent revisions to
4 (endorse NEI 96-07. The: licensee has committed to fully implement
NEI 96-07 by June 30, 1998.
Enclosure 2
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15
After reviewing the above procedures, the inspectors concluded that the
licensee had established suitable programmatic guidance to ensure that
the regulatory requirements of 10 CFR 50.59 would be met by the various
onsite and offsite organizations. However the inspectors did identify
several program inconsistencies, as described below:
1) The onsite, corporate and design organization transition
from NSAC-125 to NEI 96-07 was poorly coordinated. At the
time of the inspection, onsite personnel were using
FNP-0-AP-88. Rev. 2. based on NSAC-125 (Rev. 3 was approved
but not issued): whereas. corporate project personnel were
using GO-NG-42 based on NEI 96-07 since January 15. 1998.
Also, offsite design organizations were required to use two
guidelines at the same time with one based on NSAC-125
(i.e.. NEPI 4-0) and another based on NEI 96-07 (NEP 8-102.
Rev. 7).
2) Although the offsite design organizations conducted the
majority of all actual safety evaluations (i.e.. addressing
the unreviewed safety question (US0) criteria), the level of
detail of their procedural guidance was minimal compared to
the onsite and corporate project organizations. NEPI 4-0
lacked much of the general and specific guidelines contained
in FNP-0-AP-88 and GO-NG-42. [However, this was previously
recognized by the licensee who was preparing to implement a
new, more detailed instruction PDI 5.8-102. " Preparation of
Safety Evaluations (10 CFR 50.59)." based on NEI 96-07.]
3) Lack of written guidance for addressing the necessity of
reverifying safety evaluations for changes that are not
implemented after many months or years.
4) Conduct of cross-disciplinary 3 reparation / review of safety
evaluations was not addressed )y 10 CFR 50.59 program
procedures. Only FNP-0-AP-1. Rev. 36. " Development. Review,
and Approval of Plant Procedures.' makes any direct
reference to cross-disciplinary reviews of safety
evaluations and even that applies only to the 10 CFR 50.59
screening. Onsite and offsite design change control
procedures were vague and unclear regarding
cross-disciplinary reviews of 10 CFR 50.59 safety
evaluations.
5) Definition and explanation of safety margin in FNP-0-AP-88.
Rev. 2 was inconsistent with NRC Inspection Manual
Part 9900: 10 CFR Guidance issued April 1996 as
~10 CFR 50.59 Interim Guidance on the Requirements Related
to Changes to Facilities. Procedures and Tests (or
Experiments)." and revised in October 1997. [ Note.
Enclosure 2
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16
inspectors did not review NEI 96-07. or licensee NEI 96-07
related procedures, against the NRC interim guidance.]
6) Responsibilities of the Manager, described in FNP-0-AP-88.
Revs. 2 and 3. for approving safety evaluations that do not
pass the 10 CFR 50.59 screening, are not addressed.
10 CFR 50.59 Screenino Process
l
The inspectors reviewed 20 completed safety evaluation forms for CTEs l
that the licensee determined did not satisfy the requirements for '
performing an evaluation of the USQ criteria of 10 CFR 50.59. For these
safety evaluation forms, only Section B. "10 CFR 50.59 Applicability." ,
was filled out, which determined that a 10 CFR 50.59 safety evaluation i
was not required. The 20 CTEs selected were screened for 10 CFR 50.59 !
a) placability during the time period from August 1997 to April 1998. ;
T1e inspectors did not identify any CTE that was improperly screened for j
10 CFR 50.59 applicability.
1
Safety Evaluations - US0 Criteria
The inspectors selected about two dozen safety evaluations of
safety-significant CTEs that were determined to be 10 CFR 50.59
applicable to verify that each of the individual safety evaluation
preparers / reviewers / approvers were qualified to conduct these
evaluations. The inspectors also selected 17 safety evaluations of
safety-significant CTEs for a detailed review on the completeness and
adequacy of the answers to the US0 criteria of 10 CFR 50.59. All of the
CTEs selected included a variety of systems and different engineering
disciplines. However, they were almost exclusively related to plant
design changes requests for engineering assistance (REAs), and as-built !
notifications (ABNs). Very few procedure changes ever met the i
10 CFR 50.59 applicability determination. and the licensee rarely I
performed tests and experiments not described in the UFSAR. Of the '
safety evaluations reviewed. the majority were performed by offsite
design organizations. i
Unlike the FNP site and corporate project, the inspectors found that the
offsite design organizations did not maintain any lists of qualified
preparers / reviewers / approvers, but rather relied on the individual l
engineering supervisors and managers to keep track of who was cualified )
in their. areas of responsibility. This practice made it very cifficult ;
for the inspectors to independently verify the qualifications of :
personnel from offsite organizations. Consequently, the inspectors had
to rely on licensee assurance that offsite design personnel were
qualified. based on their review of individual personnel files. The
. j
inspectors did verify that FNP site and corporate project personnel were -
qualified to conduct and review the selected safety evaluations. During
this effort. the inspectors also observed that very few of the safety
. evaluation forms provided any evidence of a cross-disciplinary review
Enclosure 2
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _________a
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which was consistent with the lack of pra rammatic guidance in this
area. In response to this observation, the licensee maintained that
cross-disciplinary reviews of safety evaluations would normally. be
covered during the design input review, and verification. However, the
t
inspectors determined that neither NEP 4-1. " Establishing Design Input
Requirements." Rev. 4. for preparing design input records or NEP 4-9.
" Design Verification." Rev. 4. for conducting design verifications
provided any explicit guidance for multi-discipline reviews of safety
evaluations. Consequently, even after reviewing numerous design change
packages (DCPs) associated with the selected safety evaluations and
reflecting on NEP 4-1 and 4-9. the inspectors were unable to conclude
that safety evaluations were s)ecifically receiving multi-discipline
reviews. Failure to perform taese reviews may be a contributing factor
to the lack of necessary detail in safety evaluations as discussed
below.
Of the 17 safety evaluations reviewed in detail for adecuacy and
completeness of their answers. the inspectors did not icentify a CTE
that involved a USQ. However. many of the safety evaluations provided a
minimal or insufficient level of detail in answering the questions to
address the 10 CFR 50.59 US0 criteria. In general, the information
contained in the " Background and Description" portions of Section A.
" Activity Summary." of the safety evaluation form tended to be quite
detailed. Also, the responses to the questions of Section B.
~10 CFR 50.59 Applicability." were suitable. But the answers to the
questions in Section C.~ ~US0 Criteria.~ were typically very summarized
and lacked specificity. For several of the safety evaluations, the
Section C answers were so brief and generalized that. by themselves.
they would have been inadequate. However, in almost all of these cases,
the reader was able to obtain sufficient information from the
description in Section A to satisfy the appropriate question of
Section C. The major problems with this approach were that it made
reviewing the safety evaluation more difficult suggested that the
preparer did not understand the scope of each question, and was
inconsistent with the NSAC-125 and NEI 96-07 guidance for providing
complete and thorough answers to the seven questions addressed by the
descriptive information.
Some particular examples of safety evaluations that provided inadequate
detail in Section C to address the USQ criteria of 10 CFR 50.59. but
where the information could basically be found in Section A. were as
follows:
-
10 CFR 50.59 Evaluation. Rev. 5. for DCP 96-0-9012-2-006:
-
10 CFR 50.59 Evaluation. Rev. 3. for DCP 97-0-9182-0-004:
[
-
10 CFR 50.59 Evaluation for DCP B-97-1-9192-0-003:
-
10 CFR 50.59 Evaluation for ABN 95-0-0589: and.
-
10 CFR 50.59 Evaluation. Rev. 1. for DCP 95-2-8932-1-004.
Enclosure 2
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18
In addition to these safety evaluations. there was an examale where
neither Section A nor the answers to Section C of the 10 C R 50.59
l safety evaluation form provided enough detail to determine that a USQ
did not exist. However, in this instance, the licensee was able to
demonstrate to the inspectors that, even though the safety evaluation
forms lacked the information needed to answer the questions for
i determining a' US0. there was sufficient documented basis in the
associated DCP and/or references listed in Section A to determine that a
USQ did not exist. This example was the 10 CFR 50.59 Evaluation for
DCP 95-2-8932-1-004.
10 CFR 50.59 does not specify the manner in which a safety evaluation j
should be documented. As such, failure to provide sufficiently detailed
answers to the "seven questions" in Section C does not specifically
constitute a noncompliance with 10 CFR 50.59. However, the guidance in
NEP 8-102 clearly states that " the safety evaluation should be written
as a stand-alone [ emphasis added] document with sufficient detail." It
also states that "a thorough description is required because other
personnel reviewing the documentation may not be familiar with the
physical plant." There was nothing in NEP 8-102 (nor FNP-0-AP-88 or
GO-NG-42) to suggest that the answers of Section C could rely upon
information in Section A, the DCP package, or references in order to
address all elements of the subject change that could reasonably affect
a US0 determination. Quite the contrary, program guidance recommended
completeness and specificity. Adequate documentation to address the US0
criteria is considered a weakness in the implementation of the
licensee's 10 CFR 50.59 program.
10 CFR 50.59 Summary Reoort Descriptions
The inspectors compared 13 summary descriptions of CTEs reported to the
NRC pursuant 10 CFR 50.59 in the December 2, 1997. letter to the
description of changes contained in the actual 10 CFR 50.59 safety
evaluations. Inspection report (IR) 96-07 had identified examples in {
the licensee's previous 10 CFR 50.59 report to the NRC that were either {
incomplete, did not clearly identify the nature of the change, or used I
plant-specific acronyms that were not readily recognizable. During this I
review, the inspectors did not identify any of these exam)les and '
concluded that the descriptions of changes contained in t7e most recent
10 CFR 50.59 summary report were complete and adequately described the
change.
UFSAR Chances Resultina From CTEs
The inspectors reviewed ten design changes identified in the
10 CFR 50.59 summary report and compared them to the actual changes
contained in Rev. 14 to the UFSAR. No discrepancies were identified. i
j
Enclosure 2
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_ _ _
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4
19
Trainina Associated with 10 CFR 50.59 Proaram Activities
The inspectors reviewed the licensee's training program contained in
the Farley Technical Staff and Management document TSM-510. " Nuclear
Safety Evaluations." July 1996, and associated training material used by
the training department. The training material included: (1) personnel
requirements for aerforming 10 CFR 50.59 evaluations: (2) an FSAR
overview: (3) an :SAR search program use: (4) examples of 10 CFR 50.59
evaluations and screening material for 10 CFR 50.59 applicability; (5)
10 CFR 50.59 evaluation guidelines: (6) NSAC-125 and guidan a related to
its use: and (7) administrative procedure FNP-0-AP-88. Rev. 2.
The inspectors also reviewed FNP-98-0067-TRN. " Designation of Qualified
Reviewers." which contains a matrix of personnel that are qualified to
prepare and/or review 50.59 safety evaluations. Individuals were
selected for 10 CFR 50.59 training based on need and department
managers' recommendations. The training department maintained a list of
trained individuals and the next due date for refresher training. The
inspectors verified that those individuals listed in FNP-98-0067-TRN
have maintained their training current. However, the inspectors noted
that the FNP-0-AP-88 requirement that " Personnel who prepare review. )
or approve 10 CFR 50.59 evaluations will be trained every two calendar
years." was inconsistent with the guidance of Technical Staff and
Managers Curriculum Guide for TSM-510. The licensee's actual practice
of retraining conformed most closely with the curriculum guide rather
than FNP-0-AP-88. Although there is no specific 10 CFR 50.59 I
requirement for refresher training, the licensee was informed of the
conflict between TSM-510 and FNP-0-AP-88.
Prior to November 1997. qualifications to perform 10 CFR 50.59 I
evaluations were based on maintaining training current and successful
completion of the one-day course on TSM-510. However, since that time,
individuals that attend the TSM-510 course were given a written multiple
choice exam at the end of the course. Approximately 150 of 300 people
listed in FNP-98-0067-TRN. have taken the written exam. The licensee
has indicated that the remaining individuals will be given a written l
exam when refresher training is taken. '
During the transition to NEI 96-07. Rev. O. " Guidelines for 10 CFR 50.59
'
Safety Evaluations." and the process of revising FNP-0-AP-88, the
training department was also updating associated 10 CFR 50.59 training
material as applicable. In anticipation of the NEI 96-07 transition and
to provide onsite and offsite safety evaluation preparers / reviewers with
more in-depth and comprehensive training. the licensee contracted for a
special one-day training course, primarily duririg the Summer and Fall of
1997. The SNC 10 CFR 50.59 Evaluation Training Program of " Meeting the
10 CFR 50.59 Evaluation Challenge - A Program to Achieve Excellence."
was given to all qualified preparers / reviewers. The inspectors reviewed
f the training manual used and found it to be comprehensive and thorough.
Enclosure 2 ,
!
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L Audits of'10 CFR 50.59 Proaram Implementation
Very few audits of the 10 CFR 50.59 program and its implementation have
i been performed. Historically, any audits of licensee safety evaluation
activities were typically included as part of broader audits of other
programs'(e.g... design change control). However, an audit of SNC-Farley-
Support-Nuclear Engineering and Licensing was conducted during July 1997
that. included implementation of G0-NG-42. Rev. 3. Although no audit
finding reports were identified, one comment was made regarding lapsed
training-for certain individuals. Also, during the team inspection, a
L
'
site-specific spot audit was in progress "to obtain the status of, as
well as determine.~ the degree of consistency in implementation of
'
NEI 96-07." The inspectors . reviewed the audit report and audit notes
,
associated with these audits. Both audits were of very limited scope
L and detail. The overall paucity of auditing-in this area provided the
inspectors with ' insufficient information to conclude that the licensee's
audit program was effectively assessing conformance with 10 CFR 50 59. .
However, the _ inspectors did note .that routine auditing of 10 CFR 50.59
1 activities were-not specifically required by the audit program as
defined by TS', UFSAR, and Operations Quality Assurance Policy. Manual
(00APM).
c. Conclusion-
The inspectors concluded that the licensee had established: sufficient
programmatic guidance' to ensure .that the regulatory requirements of
10 CFR 50.59 would be met.by the various onsite and offsite
organizations. However, the inspectors did identify several
programmatic deficiencies and inconsistencies. Training of safety
evaluation preparers and reviewers was adequate.
Changes.. tests and experiments were properly screened for 10 CFR 50.59
applicability, and adequately ~ evaluated to ensure an unreviewed safety
. question-.did not exist. Personnel preparing.and reviewing. safety
-'
evaluations were qualified. . However, the documentation that addressed
lthe.US0 criteria in several safety evaluations lacked specificity and
- thoroughness. Furthermore, very few of the safety evaluation forms
provided any direct evidence of a cross-disciplinary review.
Licensee audits of the 10 CFR 50.59 program were few in number and very
limited-in scope.and detail.
,
<
Enclosure 2
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E8 Miscellaneous Engineering Issues (IP 92903)
E8.1 (Closed) Unresolved item (URI) 50-348.364/98-01-04. Inadeouate Safety
Assessment For Mis-wired Hot Shutdown Panel MOVs
'
(Closed) Licensee Event Report (LER) 50-348.364/97-10-01. Motor Operated
Valve Local - Remote Control Circuit Wirina Discrepancies
a. Inspection Scope
.
The inspectors reviewed revised LER 97-10-01. applicable Abnormal
Operating Procedures (AOP). licensee's response to a request for
engineering assistance (REA), and conducted walkdowns of the equipment
in question.
b. Observations and Findinas
In May 1997 the licensee discovered that the control power circuitry
for five motor-operated valves (MOVs) were mis-wired. The mis-wired
,
condition defeated the electrical isolation fuse design that ensured
1 that these five MOV's would be operable from the Hot Shutdown Panel
3
(HSDP) during .a fire in the MCR or cable spreading room. The licensee
promptly repaired the improper wiring and submitted LER 97-10 for
operating in a condition outside of the design basis.
As referenced in Section E8.1 of IR 98-02. 10 CFR 50.73(b)(3) requires
the licensee to assess the safety consequences and implications of
reportable events, including the availability of other systems or
components that could have performed the same function as that which had
failed. After reviewing the safety assessment of the original
LER 97-10. the inspectors concluded that the licensee failed to perform
an adequate assessment of the safety consequences and implications of
the mis-wired MOV's during a fire in the MCR or cable spreading rooi.1
that would necessitate implementing the unit-specific A0P-28.1. " Fire or
Inadvertent Fire Protection System Actuation in the Cable Spreading
Room.~ or AOP-28.2 " Fire in the Control Room.~ Furthermore, the
licensee failed to describe availability of other systems, components,
or manual actions to compensate for the loss of MOV functions. Failure
to perform an adequate safety assessment constitutes a violation of
10 CFR 50.73(b)(3) and is identified as VIO 50-348. 364/98-03-04
Inadequate Safety Assessment for Mis-wired Hot Shutdown Panel MOVs.
Based on this violation. URI 50-348.364/98-01-04 is closed. The
licensee revised its original LER, especially the safety assessment, to
provide a better understanding of the risk and safety significance of
the reportable event. Additionally, no new corrective actions for the
event were identified. The inspectors reviewed the revised safety
assessment and concluded it adequately addressed the safety consequences
and implications of the event, as well as, described the availability of
other systems. components. or manual actions to compensate for the loss
of the MOVs.
Enclosure 2
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22
The licensee's revised safety assessment determined the possible adverse
impacts of postulated plant fires in the cable spreading rooms (CSR) and
the main control room (MCR) upon the ability to shutdown and cooldown
the plant considering the mis-wired MOV's. For these limiting fires,
the consequences of inadequate electrical isolation of these five
mis-wired MOVs on the HSDP and MCB are discussed below.
Power Operated Relief Valve (PORV) Block Valves
i
Postulated fires in the MCR and CSR could have resulted in both d
mis-wired PORV block valves becoming inoperable from the HSDP. However.
the licensee concluded that a failed open PORV could be manually
de-energized and thereby closed to re-establish the RCS pressure
boundary. by using procedural guidance previously approved by the NRC in
a SER for an Appendix R exemption. The inspectors verified that this
procedural guidance was available at the Hot Shut Down Panel.
Component Coolina Water (CCW) to Secondary Heat Exchanaer Isolation
Valve MOV 3047 and Refuelina Water Storaae Tank (RWST) to Charaina Pumo
Suction Isolation Level Control Valves (LCVs) 1158 and 115D
The more significant concern identified by the licensee's safety
assessment was the potential that a MCR or CSR fire coincident with a
'
LOSP could have resulted in a LOCA due to loss of cooling to the RCP
seals. [ Loss of cooling to the RCP seals would require a loss of seal
injection and CCW flow to the thermal barrier heat exchangers.]
Under certain fire-induced failure conditions, the centrifugal charging
pump (CCP) suction could be' lost, resulting in possibie vapor binding
and damage to the operating CCP. In addition, a fire-induced s)urious
valve closure could isolate CCW supply flow to the RCP thermal >arrier.
Restoration of CCW flow or charging flow would then recuire manual
actions because of the loss of control of MOV 3047 anc LCV-115B and
LCV-115D. from the HSDP due to blown MCR fuses. According to the
licensee's assessment, the redundant CCPs would have been available to
start-from the HSDP. but only after manually opening the RWST to CCP
suction MOVs (and venting the CCPs, as necessary) to reestablish seal
cooling through seal injection. Similarly. CCW flow could only be
restored by manual operation of MOV 3047. However, these manual
actions, without specific operator training or procedures. would have
significantly delayed restoration of seal flow.
'The inspectors walked down the MCR and CSR wiring for the components of
concern to determine the probability of a fire affecting both the CVCS
(letdown. CCPs. VCT discharge valves) and CCW MOV 3047. The inspectors
found that MOV 3047 and the CVCS were on the same section of the MCB.
,
approximately 12 feet apart. The respective cables dropped vertically
f from the switches through floor penetrations directly to the CSR. The
majority of the cables in that section of the MCB had a braided
stainless steel jacket. There were no vertical separators in the MCB to
Enclosure 2
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23
provide 3hysical isolation. Once in the CSR, the cabling split north
and sout1 to the respective "A" and "B" train cable trays. The
inspectors concluded based on the 3hysical layout, that it would take a
major MCR or CSR fire to affect botl charging flow and MOV 3047 to cause
a loss of RCP seal cooling and subsequent seal LOCA.
During the 12-year period that the affected MOVs were mis-wired, no
fires or other significant' plant events occurred that necessitated
taking control of the MOVs at the HSDP.
The licensee's assessment concluded that it would be extremely unlikely
that a fire of sufficient magnitude to adversely impact components as
described above to occurr. The licensee considered it even more
unlikely that a fire of this magnitude could occur coincident with an
LOSP. A significant' fire in the main control' room was not considered by
the licensee to be a credible event because it is continuously occupied.
)ower. circuits are minimized, and combustibles are limited. The CSRs
lave automatic fire su
combustible. materials.ppression The CSRssystems, fire alarms,
are also equipped withand limited
manually
actuated low pressure carbon dioxide fire suppression systems. The
licensee also considered that the advance of a postulated fire would.not
have been instantaneous- and' components adversely affected by the fire
would not have.been affected simultaneously. 0)erators in the MCR would
have been alerted to component malfunctions eitler through indications,
alarms.-or procedural steps.
The licensee concluded that'while the MOVs were mis-wired, the ability
-to achieve safe shutdown for certain )ostulated plant fires coincident
with a LOSP was degraded. This possi)le loss of capability to shutdown
and cooldown the plant from outside the MCR; as required by 10 CFR 50.
Appendix F is~ identified as an apparent violation. EEI 50-348,
364/98-03 3 HSDP Loss of Function.
In addition to the corrective actions of LER 97-10-00, the licensee
- reported in LER 97-10-01 that the Abnormal Operating Procedures for
responding to fires in the CSR or MCR were revised. The revised
)rocedures now require operators (if time allows) to open LCV-115B and
.CV-115D prior to evacuating the control room in order to minimize the
potential for. losing suction to the operating CCP. This action was
considered an enhancement to the procedure by the licensee. These
)rocedure changes were verified by the inspectors. The revised
_ER 97-10-01 is considered closed.
c. Conclusion
- A violation was identified because the original safety assessment for
LER 97-10 did not completely address the safety consequences and
implications of the possible failure of five mis-wirec motor-operated
valves at the Hot. Shutdown Panel during a control room or cable
spreading room. fire. The subsequent supplemental LER did provide
Enclosure 2
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.
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24
sufficient information for an adequate safety assessment of the event.
In addition, an apparent violation was identified due to the I
determination that the licensee's ability to safely shutdown and '
cooldown the plant from the HSDP was in a degraded condition for about
12 years.
E8.2- (Closed) VIO 50-348. 364/97-11-04 Failure to Imolement a Test Proaram j
for Service Testina of the TDAFW Batterv
The licensee responded to this VIO in correspondence dated December 17,
1997 and initiated Corrective Action Report (CAR) 2321. The inspectors l
reviewed the licensee's written response and completed CAR. verified j
implementation of corrective actions, and reviewed the test data for the l
Unit TDAFW battery service test. This VIO is closed. !
E8.3 (Discussed) VIO 50-348. 364/97-11-03. TDAFW Batterv Installation and
Check Valve Test Deficiencies
The licensee responded to this VIO in correspondence dated December 17,
1997, and initiated CAR 2322. This CAR was not complete at the end of
this inspection period. The inspector verified that the TDAFW Battery
racks were rebuilt per the applicable drawings. This VIO will remain ;
open pending review of the completed CAR and the check valve test i
deficiencies corrective action.
E8.4 Modification of Penetrations for GL 96-06
The inspectors verified that relief valves were installed in i
Penetration 30. Pressure Relief Tank (PRT) Makeup, and Penetration 31.
Reactor Coolant Drain Tank (RCDT) Drain, per the licensee's letter dated
May 23. 1997, in response to GL 96-06 " Assurance of Equipment
Operability and Containment Integrity During Design-Basis Accident
Conditions."
-EB.5 (Closed) URI 50-348. 364/98-01-05. Failure to Track and Correct
Conditions Adverse to Quality :
(Closed) IFI 50-348. 364/98-01-06. Control Room Ventilation Testina I
a. Inspection ScoDe (37551)
The inspectors reviewed a variety of tracking list data and closure l'
documentation, interviewed personnel, and walked down the systems.
l
b. Observations and Findinas
The subject of 1.icensee Event Report (LER) 97-13. " Operating Outside the
Design Basis Due to Control Room Exhaust Isolation Dam)ers Not Closed."
originated from Open Item CRV-007, identified during t1e Control Room
Ventilation (CRV) Functional System Description (FSD). Historically.
Enclosure 2
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the plant had operated with the CRV exhaust dampers open. But during a
self-assessment of the CRVS. questions arose regarding the design basis j
requirements for positioning these exhaust dampers and FSD CRV-007 was '
opened. In April 1995, a design evaluation was performed offsite to
address CRV-007. This evaluation concluded that the CRVS emergency
) pressurization system was designed to operate with these dampers shut. ,
lowever, the results of the evaluation were misinterpreted at the site '
and the dampers were procedurally left open. In August 1997. during the
FSAR verification arocess, the open damper position was again questioned
by the licensee. _icensee personnel determined that the plant had
o)erated outside its design basis while the exhaust dampers were open. 1
T1ese dampers were then shut and LER 97-13 was issued (see IR 98-01. l
Section E8.2) ;
As part of the corrective actions for.LER 50-348. 364/97-13. the
licensee conducted a review of all the closed out "Open Items"
previously identified during the CRV FSD and Safety System i
Self-Assessment (SSSA) and ascertained that 5 of 19 open items had been
'
closed out without any evidence that the recommended corrective actions
were implemented. Responsibility for two of these items (0 pen Items
CRV-010 and CRV-019) had been assigned to the site.
Open Item _ CRV-010 originally identified that some areas surrounding the
main control room (MCR) could be pressurized to greater than 0.125 l
inches water gauge (w.g.), thus allowing unfiltered in-leakage, greater !
than assumed. into the MCR. Bechtel letter AP-21274, dated June 7. !
1995, completed the evaluation and identified two s) aces where single l
failures could cause a room adjacent to the MCR to 3e pressurized J
greater than 0.125 inches W.G. This letter provided recommendations to
resolve the concern of over-pressurizing areas next to the Control Room
and thereby not allow greater than assumed unfiltered in-leakage into
.
the MCR. These recommendations were provided to the site from corporate
engineering via a letter (NEL 95-0189). dated July 6. 1995. The Bechtel ;
letter also recommended closing item CRV-FSD-010. " Control Room
'
Pressurization from Adjacent Areas." and it was removed from the
corporate tracking list. However during this time the o)en item was
also inadvertently removed from the site tracking list. _ater in
November 1997, during the review of completed CRVS FSD, and SSSA Open
Items as a corrective action for LER 97-13. the licensee determined that
no evidence (e.g. revised procedures. etc;) could be found to ascertain
that the recommendations for CRV-FSD-010 had ever been implemented or
dispositioned at the site. This item was subsequently reopened, and the
proposed recommendations were still being evaluated at the end of this
inspection period.
j
'
0)en Item CRV-019 was originally concerned with weaknesses in testing
t1e pressurization system to support the allowable o)en penetrations in
the control room boundary. During the SSSA of the CRV system.
Assessment Observation CRV-MECH-02, dated November 17. 1995 (later
designated as Open Item CRV-019) identified some potential weaknesses in
Enclosure 2
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26
the testing to support MCR allowable open per.etrations calculation. The
response to CRV-MECH-02 (NEL 96-0069. dated February 27, 1996) addressed
several specific issues, but did not look at' the larger issue of MCR
boundary degradation. During corrective actions for LER 97-13. former
"Open Items" from the FSD/SSSA were reviewed to verify adequate closure.
The licensee identified that the response to close out this issue did
not address the boundary degradation issue (NEL 97-0526. dated
December 19. 1997) and recommended that periodic testing be accomplished
to assure that the MCR penetration opening allowance was conservative.
On February 2. 1998, the licensee established an administrative Limiting
Condition of Operation (LCO) that prevented additional breaches in the
control room boundary. However, on March 5. 1998, corporate engineering
sent an evaluation (NEL 98-0088) to the site which recommended against
testing to validate the pressurization performance, but did state that
the plant may want to consider adding a test to determine boundary
degradation. In addition, the letter re-issued a Bechtel letter which
confirmed that the 21.21 scuare inch opening was conservative. At that
time the restriction on adcitional boundary breaches was lifted.
However, no additional boundary breaches were required. After
discussions with the licensee. Engineering Support conducted a test on
March 25. 1998. for the CRV system that included air flow data. The
licensee's calculation determined that only 10 square inches of opening
could be allowed and still maintain the required MCR over-pressure. The
licensee re-instituted an administrative LCO to prevent CR boundary
breaches greater than the new calculated area. On April 21. 1998, the
licensee added a temporary change to FNP-0-AP-16. " Conduct of Operations
-Operations Group." Rev. 27, which removed the 21.21 square inch
administrative limit and referenced a data sheet in the Plant Curve
Book. which is updated quarterly to reflect the boundary degradation of
- the control room.
Until this time, the testing being done did not verify that the control
room minimum pressure could be maintained with a 21.21 square inch
opening, the licensee's administrative limit.
An inspector's review of the most recent surveillences. FNP-0-STP-26.2.
" Control Room Pressurization / Filtration Operability Test" for 'A' and
- B' Trains." Rev. 12. indicated that the system can maintain the minimum
control room pressure in its current condition and that the licensee is
cognizant of the need to maintain the control room boundary integrity.
- Open Items CRV-007. CRV-010. and CRV-019 were conditions adverse to ;
quality that were not adequately corrected. In each of the three i
previously described cases, the licensee had originally identified a l
deficiency and then either inadvertently or inappropriately closed it v
- out. The licensee then re-identified the items and either has taken or i
is completing corrective actions on the individual issues. Failure to !
adequately correct conditions adverse to quality is identified as i
VIO 50-348. 364/98-03-06. Inadequate Corrective Actions for MCR 1
l
Enclosure 2
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~ Ventilation System. ' Also, the inspectors expressed concern to the
licensee that FSD open items for systems in addition to the CRVS may
have been closed with inadequate corrective actions. By the end of this 1
inspection period, the licensee had not reviewed or reverified the
adequacy of the corrective actions closed out for other FSD systems.
l
l
c. Conclusiq01
P
- A violation of 10 CFR Part 50. Appendix B Criterion XVI, Corrective
Action was identified. The licensee identified three conditions adverse
to quality of Control Room Ventilation System Functional System Design
-(FSD) Open Items, which were either inadvertently or inappropriately
. . closed and not corrected.
IV. Plant Sucoort
R1 Radiological Protection and Chemistry (RP&C) Controls (83750, 84750,
86750)
RI.1 Radiological Controls (83750)
m
E i a. Insoection Scooe
L Radiological controls associated with ongoing Unit 1 ('U1) routine
l_ ' operations and with Unit 2 Refueling Outage 12 (U2RF12) activities were
< reviewed and evaluated by the inspectors. Reviewed program areas
'
Li included area-cleanliness and housekeeping, area postings. radioactive
o >
material and waste'(radwaste)-container labels. high and locked-high
radiation area controls, and procedural and radiation work permit (RWP)
implementation. The inspectors made frequent tours of the
U
Radiologically Controlled Areas (RCAs) and directly observed worker and
Health Physics Technician (HPT) performance during selected tasks. l
, Established Radiation Protection (RP) program guidance and
!
'
implementation were compared against commitments detailed in the Updated
Final Safety Analysis Report (UFSAR). and in procedural. Technical
l Specification (TS), and 10 CFR Part 20 requirements.
L
"
b '. 0 observations and Findinas-
High.and locked high radiation area controls were' established and
L. maintained in accordance with TS requirements. Area postings and
4 container labels were proper for the radiological conditions and met
procedural. TS, or 10 CFR 20 Subpart J requirements. Improvements were
"
noted in labels provided for containers of radioactive materials.-
Contamination and radiation surveys were conducted in accordance with
procedural requirements. Radiation and contamination survey results met
established regulatory and procedural limits.
1
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On April 29, 1998, the inspectors observed work activities and HP
practices associated with removal of Tri-Nuclear equipment and filters
from the lower reactor cavity conducted in accordance with Specific RWP
No. 298-2491. Revision (Rev.) 1. During completion of the task., a
remote handling-tool was damaged causing the dose expenditure to
escalate to approximately 985 millirem (mrem), exceeded the budgeted
dose-of 900 mrem. The following poor radiological practices, job
. planning weaknesses, and communication issues were identified and
discussed with licensee representatives:
. Mock-u) training was not provided for removing the spent filters
from t1e vacuum system and for transferring the material to the
Unit 2 drumming room. Design differences between the previous and
current vacuum system model, which now required a specific
alignment of the filters within their housi.ngs for proper removal,
was not identified. Improper alignment during the initial . attempt-
'
to remove a filter from the vacuum system resulted in excess force -
being applied and the remote handling device being damaged.
- Methods and controls to limit personnel exposure were minimally
effective. ' Plans for use of an extension tool while transferring
the filters in containment was abandoned after problems were
' identified during removal of the initial filter from the vacuum
system: extensive exposure time was required to manually tie and
untie ropes to bags used to hoist the eight vacuum system filters
from the lower to-the upper cavity area: and on several occasions.
workers entered designated exclusion areas during transfer of the
filters. Also. when a supplemental teledose monitor.'provided to
a Health Physics Technician (HPT) handling a spent filter alarmed
as a result of an improper dose rate alarm setting. the individual
returned to transfer an~ additional filter prior to change-out of
the alarming unit.
.. Planning and communication weaknesses were identified during a
post-job briefing and from followup discussions with participating
operations, maintenance HPT and "As Low As Reasonably Achievable"
(ALARA) staff. For example, maintenance workers were
knowledgeable of general area dose rates associated with the
filters, but were not fully aware of the significant hazard from
the filter contact dose rates and the importance of using remote
tools in handling the filters. Furthermore, previous maintenance
staff-safety concerns regarding use of the remote tool to remove
the filters from the vacuum equipment and potential contamination
concerns from ropes used to suspend the vacuum equipment in the
cavity were not' incorporated into planning for the current task..
p
1
Enclosure 2
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w__=_-___ _ _ - _
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. - -
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29-
c. ' Conclusions
Radiological controls were established and maintained in accordance with-
. procedural.-TS and 10 CFR 20 Appendix J requirements.
A weakness in exposure controls and poor communications contributed to
-the licensee exceeding its budgeted dose.for the removal of Tri-Nuclear
equipment and filters from the U2 lower reactor cavity due.
R1.2 External Exoosure (83750)
a. Insoection Scooe
The inspectors discussed and reviewed deep dose equivalent (DDE) and
shallow dase equivalent (SDE) exposures to workers involved in U2RF12
activities. Personnel contaminations, documented as personnel
contamination events (PCEs), i.e.. dispersed contamination greater than
or equal to.(a) 5000 disintegrations 2
per minute per 100 square
centimeters (dpm/100cm ) and specks = 100000 dpm/ probe. areas, were
reviewed and discussed.
Dose assessment methods and assumptions, where applicable, were reviewed
.for technical adequacy. : Dose results were, compared against 10 CFR
Part 20. limits,
b. Observations and Findinas
Estimated dose data, as measured by Digital Alarming Dosimeter (DAD) for
U2RF12 activities, were reviewed and discussed with responsible staff.
-As of April- 30, 1998, dose expenditure.for outage activities.
approximately 169.064 person-rem, exceeded the original projected dose
expenditure of 155.956 person-rem. The licensee identified problems
with Residual Heat Removal (RHR) pump maintenance activities, expanded
scope of mid-loop valve maintenance, and unexpected elevated dose rates.
in the U2 spent fuel pool (SFP), contributing to the elevated person-rem
expenditures. From review of selected Occurrence Reports and
l
discussions with licensee staff, the inspectors verified that RHR
maintenance and-SFP dose expenditure issues were being reviewed and
evaluated.
As of April 29. 1998, approximately 29 personnel contamination event
!
'
reports were documented with only one event requiring a skin SDE
determination. For the affected individual. a hot particle located on
the u
-(SDE)pper.right forearm
of ap3roximately resulted
7.76 rem. in an assigned
Licensee assumptions shallowand dose details equivalent
regarding p1ysical: location.' length of. exposure and isotopic
characteristics of particle were appropriate. The inspectors noted that
all assigned doses were within 10 CFR 20.1201 limits.
I
Enclosure 2 I
? l
.-
L___-_-______-__ -
-
.
-.
. .
'
- . ..
' - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
- __ . _ _ _ - _ _ - _ --. . _ _ . _ _ .
. .
,
l.
30
c. Conclusions
For U2RF12 activities, dose expenditure exceeded original estimates due
to expanded work scope, unexpected maintenance problems, and elevated U2
spent fuel ' pool dose rates.
Worker SDE_ exposures resulting from personnel contamination events and
work activities during the U2RF12 activities were evaluated properly and
were within 10 CFR 20.1201 limits.
R1.3 Internal Exoosure (83750)
a. Insoection Sccoe
Results of selected investigative whole-body count (WBC) analyses
conducted during the U2RF12 outage 'were reviewed in detail.
b. Observations and Findinas
From review of WBC analysis records of workers' positive radionuclides
intakes, the inspectors identified one individual whose initial WBC
analyses data resulted in an assigned committed effective dose
equivalent (CEDE) exceeding 10 mrem. The inspectors noted that as of
April 29,1998, approximately 12 investigative WBC analyses were
conducted as a result of specific events, usually documented in
-Radiation Worker Performance Observations, which could cause or indicate
potential radionuclides intakes resulting in internal exposure. The
estimated maximum iiltake was approximately 158 nanocuries (nC1). >
resulting in an assigned CEDE of 12 mrem. The ins)ectors verified that
the 12 mrem CEDE was added to the DDE to provide t1e total effective
dose equivalent (TEDE) documented in the individual's official exposure
records. No other evaluated worker intakes exceeded 10 mrem. i.e. 0.2
percent of the annual limit of intake (ALI) required to be documented by
licensee procedures.
c. Conclusions
Controls for minimizing workers' internal exposure during U2RF12
activities were effective.
R1.4 Respiratory Protection (83750)
a. .Insoection Scone
Respiratory protection program implementation for U2RF12 activities was
reviewed and evaluated. The review verified training, fit testing, and
medical qualifications for selected licensee and contractor personnel
who were supplied and used respiratory protection equipment.
Enclosure 2
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i
.
!
L
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31
Licensee activities were reviewed and evaluated against procedural and
10 CFR 20.1703 requirements.
L b. Observations and Findinas
Workers using respiratory protective equipment during U2RF12, were fit
tested. medically qualified, and trained in accordance with procedural l
requirements. l
l
c. Conclusions ,
,
Respiratory protection program implementation for U2RF12 activities met -
established procedural and 10 CFR 20.1703 requirements.
R2 Status of Radiological Protection and Chemistry Controls Facilities and
Equipment
R2.1 Radiologically Controlled Area (RCA). Units 1 and 2 (71750) i
Overall cleanliness of the RCA remained good. Plant personnel observed
working in the RCA generally demonstrated appropriate knowledge and
application of radiological control practices. Health physics
technicians generally provided positive control and support of work
activities in the RCA.
R2.2 Radiation Monitorino Systems I
a. Scope
Design and calibration issues were reviewed and discussed for selected
Radiation Monitoring System (RMS) sampling equipment and detectors. ;
Design issues for the RE-29B. Plant Vent Post-Accident Vent monitor were '
reviewed. Calibration activities for the U2 Containment High Range
Monitor (CHRM) RE-27A were observed and resultant calibration data were
reviewed and discussed. In addition, design issues associated with
effluent stream flow pathways for the RE-29B particulate sampler were
reviewed and verified.
Installed equipment was evaluated against recommendations specified in
American National Standards Institute (ANSI) N13.1-1969. American
National Standard Guide to Sampling Airborne Radioactive Materials in '
Nuclear Facilities. Calibration activities were evaluated against
applicable sections of the Updated Final Safety Analysis Re) ort (UFSAR).
Technical Specification (TS), and Offsite Dose Calculation ianual (ODCM)
requirements. In addition, calibration activities to meet a March 14
1983 Order to implement and maintain licensing commitments associated
with Three Mile Island (TMI) Action Item II.F.1 for the CHRMs special
calibrations were reviewed.
t
Enclosure 2
,
L_____
- _ _ . _ _ _ _ .
,1 ,
<
.
32
b. Observations and Findinas
.The installed U1 RE-298 RMS sample line flow path was found to be
acceptable. The U2 CHRMs electronic calibrations and functional tests
and isotopic calibration checks were conducted in accordance with
Surveillance Test Procedure FNP-2-STP-227.18. "In-Containment High Range
Radiation Area Monitor R-27A." Rev. 9. and Radiation Control and
Protection (RC&P) Procedure FNP-2-RCP-272. " Isotopic Calibration Check ,
of the Unit 2 Containment High Range Area Monitors." Rev. 3. The
calibrations were conducted by electronic signal substitution for all
range decades above 10 Roentgens per hour (R/hr). No regulatory issues
were identified for the test and calibration data reviewed.
c. Conclusions
The evaluated RMS equipment was installed properly and the reviewed
detector calibrations and functional tests were conducted in accordance
with and met procedural. 10 CFR Part 20. and ODCM requirements.
I
R3 Radiation Protection and Chemistry Documentation (84750) !
R3.1 Radiological Effluent and Environmental Monitorina Reports l
a. Insoection Scone l
l
Data and conclusions documented in the 1997 Annual Radiological !
Environmental Operating Report and the 1997 Annual Radioactive Effluent !
Release Report were reviewed and discussed. The contents and l
conclusions of the reports were evaluated against the applicable '
sections (SS) of TSs 6.8 and 6.9.1. and S 7 of the Offsite Dose i
Calculation Manual (0DCM).
b. Observations and Findinas
The inspectors verified that the 1997 Radiological Environmental
Operating Report was prepared and submitted in accordance with TS and
ODCM requirements. Based on trend data for radionuclides concentrations !
in offsite environmental matrices at control and indicator stations, no '
discernible offsite effects or trends were demonstrated from plant
effluent discharges to the environment. The licensee properly
determined the controlling receptor to evaluate the maximum dose to a
member of the public beyond the site boundary based on releases and
current land-use census data. From review of the 1997 environmental
monitoring program sam) ling deviations required by ODCM Section 7.1.2.4. l
the inspectors noted t1at community airborne particulate monitoring
,
station number (No.) 1108 was inoperable from approximately November 18.
1997, through January 27. 1998, due to construction at the electric
substation which supplied power to the equipment.. Farley Nuclear Plant
(FNP) Occurrence Report No. 973135. generated in response to finding the
! power off on November 25, 1997. initially documented that power would be
4
l Enclosure 2
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_ - _____ - -_-____ _ _ -
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33
interrupted for approximately two weeks. but an attached note indicated
that as of January 13. 1998. power had not been restored to the sampling
equipment. Corrective actions to 3revent recurrence, such as the use of
portable samplers or securing anotler source of electrical power within
the immediate vicinity was not addressed. The inspectors identified the
lack of detailed corrective actions to prevent recurrence of the
l extensive out-of-service condition due to a known power supply
l interruption. extending past the original two-week estimate, as an
environmental monitoring program weakness.
The 1997 Annual Radioactive Effluent Release Reports was submitted in
accordance with TS and ODCM requirements. For the raport period,
calculated offsite doses from liquid and gaseous effluent releases were
a small fraction of the ODCM limits.
c. _ Conclusions
for 1997, program activities to control, monitor and document liquid and
airborne radionuclides concentrations in effluents and in the offsite
environment were implemented effectively. No significant environmental
impact was identified. Projected offsite doses to the maximally exposed
individual were a small fraction of ODCM and 40 CFR 190 specified
limits.
Extensive delays in returning a community particulate air sampler to
service and lack of corrective actions to prevent recurrence was
identified as a program weakness.
R7 Quality Assurance in RP&C Activities
R7.1 Licensee Self-Assessment Activities (83750. 84750. 86750)
a Insoection Scooe
The ins)ectors reviewed implementation and status of the licensee's
Health 3hysics Observation program. Program implementation and results
were evaluated against commitments initially documented in an
October 25, 1996 licensee response to a Notice of Violation regarding
improper dosimetry use.
b. Observations and Findinas
The inspectors noted that observations of both Health Physics (HP) and
Dosimetry (DOS) practices continued. The observed data were assessed
for the most current 1000 observations made and sorted into 26 separate
ty)es of poor practices or issues. Each identified item was
subsequently assigned to one of twenty-three separate work groups.
Results routinely were presented to upper management and workers. For
identified error rates exceeding five and fifteen percent of the HP and
DOS practices observed. Occurrence Reports were initiated and additional
Enclosure 2
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__
_ _ - _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ . _ _ - _ - _ _ _
.
.
...
,
34
l
L co'rrective' actions taken. From review of the licensee weekly
!
- observation trend data for the U2RF12 outage, the inspectors noted that
the percent errors for both HP and DOS issues were reduced by
approximately 50 percent- relative to weekly data collected during the
I 3revious. Unit.1 Refueling Outage 14 (U1RF14) activities. For the
!
J2RF12 the majority of identified ~ issues were associated with incorrect
- radioactive material . handling, violation of HP boundaries, spread of
contamination..and incorrect dress out;
- - c. Conclusions-
The licensee HP and' DOS observation program continued to be implemented
effectively and contributed to the reduced personnel errors observed for
-U2RF12 activities.
R8. Miscellaneous RP&C Issues (83750, 84750)
- R8.1 (C1osed) IFT 50-348.364/98-01-07: Review Licensee Actions ~ to Imoro've
Radioactive Material Container Label Effectiveness.
'In response to. inconsistencies and poor practices noticed for .
<
radioactive material container label types and information required by
10 CFR 20.1501; the licensee had assigned a senior HPT responsibility to
'
review and provide oversight. of the subject program area. Based on
improvements-in the radioactive material. container labeling program
activities-noted during the current inspection. period, this item is
closed'.
~ P2 Status of Emergency Preparedness (EP) Facilities, Equipment, and
Resources
- P2.1 Facilitv Insoection
a2 Insoection'Scoce (82701)
The inspectors examined the licensee's emergency response facilities
'(ERFs) and equipment to assess their adequacy and to determine whether
they were maintained in a state of operational readiness as specified in
the Farley Emergency Plan.
..
- b. Observations and Findinas
The' inspectors toured the Control Room. Technical Support Center (TSC).
Operational Support Center (OSC). Emergency Operations Facility (EOF),
and the alternate E0F. Selected equipment, supplies, and communications
L systems within these facilities were inspected. All tested equipment
r and systems were found to be in operable condition. The facilities were
L well-maintained.
b '
L Enclosure 2
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35
c. Conclusions
.ERFs were well-equipped and operationally ready to support _an emergency..
- P3- Emergency ~ Preparedness (EP) Procedures and Documentation
P3.1 Emeraency Plan
a. Insoection Scoce=(82701)
'The' inspectors reviewed recent revisions to the Emergency Plan to
' determine whether' changes were made in accordance with 10-CFR 50.54(q).
and )lan implementation. In addition..the implementation'of the plan of
' Marc 1 8. 1998. was reviewed.
,
-b. Observations ~and Findinas
The current revision of-the Emergency Plan was administrative in nature.
.The Emergency Plan was implemented'on March 8.1998, with a Notification-
-
of-Unusual Event (NOUE) due to high river water level. There was a
partial TSC activation and review of documentation revealed that the
requi. red notifications were completed in a timely manner,
c. -Conclusions
Changes to the Emergency Plan were made in-accordance with
. -The NOUE on March'8. 1998..was made in accordance with
~10 CFR 50.54(q)1an.
the Emergency P
P5 Staff Training and Qualification in EP.
P5.1 Trainina of Emeroency Resoonse Personnel
a. Insoection Scooe (82701)
The inspectors evaluated the training program for the Emergency Response
. Organization (ERO): through review of program documentation and
observation of licensee training' functions.
.
'
b. Observations and Findinas-
-The licensee conducted a program of periodic integrated response drills
~
'(typically six per- year)' to enhance the training for ERO personnel. In
an effort to gauge:the effectiveness of the emergency res)onse training
pr gram, the inspectors observed a previously scheduled ERO training
dr ll.on May_21. ERO personnel- activated the ERFs in a timely manner
and responded capably ~to the simulated emergency, which included event
classifications e f Alert. Site Area Emergency. and General Emergency.
Minor problems wdh the ER0's . response efforts. were identified by
licensee drill monitors for corrective action. The inspectors. also
Enclosure 2
L _ . -_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
D
p
...
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36
L
lL ' observed an E0F tabletop drill on May 19 involving real-time setup of-
o
- the- facility.and a round; table discussion of staff functions and -
interfaces,
c. Conclusions
The conduct of regular integrated drills enhanced the quality of ER0-
.
training. Drill monitors effectively identified response problems for
L. corrective actions. 'ERO personnel were adequately trained and responded
l- appropriately to a simulated event.
_P6- EP Organization and Administration
P6.1 EP Proaram Organization
i
'
a. Insoection Scooe (82701)
The inspectors reviewed this area to determine if changes in personnel
had occurred which could adversely affect the management and
implementation of the EP program.
[ b. ' Observations and Findinas
The organization of the EP program was reviewed and discussed with
L. -- -licensee management representatives. Two changes to the EP' organization
were noted. .The position of Emergency Planning Technician was.
reassigned in' September 1997-to an individual who had 3reviously been a
member of the radiation protection group at Farley. T11s individual's
- 3 professional development included a one-week training course in EP in
Jecember 1997.
A new Emergency Management Director for Houston. County was recently
aapointed. According to licensee management representatives, this
clange had not had a negative impact upon the working relationship
between the licensee and Houston County. The inspectors were informed
~
that no other significant changes in management )ersonnel for offsite
interface / support agencies had occurred during tie past two years.
c. Conclusions:
. No degradation had occurred in the EP program since the previous
inspection.
1
.. ,
Enclosure 2
L.
1
_ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ - _ _ . - - _ _ - _ _ _ _ _ - . - _ _ _ _ _ _
__
,
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37
P7 Quality Assurance in EP Activities
P7.1 10 CFR 50.54(t) Audit of Emeraency Preparedness Proaram
a. Insnection Scone (82701)
The inspectors reviewed this area to assess the quality of the required
annual audit of the emergency preparedness program, and to verify that
the audit met the requirements of 10 CFR 50.54(t).
b. Observations and Findinas
The inspectors reviewed documentation associated with the following EP
program audits conducted by the licensee's Quality Assurance (0A) group:
e Safety Audit and Engineering Review (SAER) Audit of Farley Nuclear
Plant-Emergency Planning Report No. 97-EP/16.
e SAER Audit of Farley Nuclear Plant-Emergency Planning
Report No. 96-EP/16-1.
The audits were thorough and independent, and the nature of the j
identified issues indicated inclusive understanding of the EP area by
the auditors. The audits provided evidence of the licensee's ability to J
l
self-identify emergency preparedness program issues.
c. Conclusions
The 1996 and 1997 EP program audits met the 10 CFR 50.54(t) requirement 3
for an annual independent audit of the EP program.
P7.2 Effectiveness of Licensee's Corrective Action Proaram for EP Issues
a. Insoection Scope (82701) <
The inspectors reviewed this area to evaluate the licensee's program for
identifying, tracking, and resolving problems in emergency preparedness.
b. Observations and Findinas
The licensee formally identified and tracked EP issues by means of the
" Emergency Planning Punchlist." The licensee's list of open EP items is
used to track all substantive findings, including many improvement items
derived from drill critiques and carried at the lowest priority.
Although the punchlist was maintained by the Emergency Planning
Coordinator and was not integrated with any plant-wide tracking system,
it was periodically distributed for updating by the assigned group for
each item. This method was effective for resolving identified EP
I
deficiencies and issues.
Enclosure 2
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38
c. [pnclusions
The licensee's program for identifying, tracking, and.resolv'ng problems
in EP was effective.
P8 Miscellaneous EP Issues
P8.1 (Closed) Insoector Follow-uo Item (IFI) 50-348. 364/96-14-01: Exercise
Weakness--Significant Emeraency Information Was Not Communicated to the
Acorooriate Emeraency Manaaer in a Timely Manner.
This exercise weakness from the 1996 full-participation exercise was
identified because significant emergency information was not
communicated to the appropriate emergency manager in a timely manner.
.The training drill observed by the inspectors on May 21, 1998. provided
the opportunity for the inspectors to focus on the transfer of emergency
information from the interim Emergency Director (ED) in the simulator to
the ED in the TSC. and later from the ED to the Recovery Manager in the
Enrgency Operations Facility. In all cases, the transfer of
inform:.ition was done clearly with repeat-backs to assure understanding,
and it was done timely. This item is closed.
51 Conduct of Security and Safeguards Activities
S1.1 Routine Observations of Plant Security Measures (71750)
The inspectors verified that selected portions of site security program
plans were being adequately implemented. Disabled vital area doors were
properly maned and controlled. Security personnel activities observed
during the inspection period were performed well. Site security systems
were adequately maintained and functional to ensure the physical
protection of the plant. However, the inspectors did identify two minor
instances in which Security personnel were not attentive to equipment
3roblems that adversely im) acted effectiveness of physical security
Jarriers: 1) Inoperative iCR door card reader green light (contrary to
alant policy egress was allowed without verifying green light) and 2)
1roken door latch on bullet hardened door outside PAP (door was blocked
open, rather than disabling latch and leaving door shut. Although not
specifically addressed by the Physical Security Plan (PSP), these
barriers were in a degraded condition without compensatory measures in
place. Once notified Security promptly resolved each instance.
Enclosure 2
l
l . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
_ --_- -_-----
,
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A
39
S2 Status of Security Facilities and Equipment
S2.1 .P_rotected and Vital Area Access Control
a. Insoection Scope (81700)
i The inspectors reviewed the PSP to determine if the licensee's access
i
control program for personnel and packages met the commitments specified
therein.
b. Observations and Findinos
On April 21. the inspectors observed a reliability test conducted at the
Primary Access Portal (PAP). A disabled weapon was placed inside a
lunch cooler that was carried by a licensee individual with unescorted
access. The search officer immediately discovered the weapon and
detained the individual. The officer located in the final access
control booth appropriately locked the protected area turnstiles and
summoned assistance.
The inspectors reviewed records for 17 favorably terminated individuals
with respect to the licensee s action to remove unescorted access.
Procedure FNP-0-SP-11. ~Badgire Procedures." Rev.13. required that
changes in personnel access require'aents caused by termination of
employees will be reported immediately to the Security Site Manager.
The necessary action will be taken to remove the individuals' name from
access. In addition, procedure FNP-0-AP-42. " Access Control." Rev. 26.
Section 7.5.3. required that individuals' names be removed from the
appropriate access list immediately upon termination of need.
Of the 17 records reviewed the inspector determined that 8 of the
individuals did not have their unesc'orted access removed from the
security computer ranging from 1 to 11 days after the individual had
been terminated. All eight individuals had access to protected and
vital areas: however, no individual accessed those areas after
termination. Security removed access upon notification.
The inspector determined that although procedures did exist.
clarification was needed as to contractors * responsibilities. Neither
procedure had a process in place to ensure that contractor personnel who
no longer needed unescorted access were immediately removed from the
security com) uter. The inspector reviewed a Change Order (CO) for a
contractor w1ich was currently providing work at FNP and found that the
C0 simply stated to follow access procedures.
The failure to immediately report terminations of 8 employees to the
Security Department is identified as Violation 50-348, 50-364/98-03-07.
Failure To Promptly Terminate Security Access.
Enclosure 2
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. _ _ _ _ _ -- _ - - - _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _
, *
.
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'
40
l c. Conclusions-
The licensee was a)propriately searching individuals and packages prior
i to entrance into t1e protected area. The failure to include a
,
'
documented process in access control procedures for contractors to
timely inform Security of terminated individuals contributed to a
violation for failure to immediately terminate 8 individuals' unescorted
access.
! S2.2 Protected Area Detection
a. Insoection Scone (81700). a
l
The inspectors evaluated the licensee's protected area detection
capability to determine if provisions of Section 5.3 of the PSP were
met.
b. Observations and Findings
On April 22. the inspectors observed the licensee test two perimeter
zones. Both zones alarmed appropriately. The inspector additionally i
performed a walkdown of the aerimeter and determined that the design. l
placement, and coverage of tie' intrusion detection system met the
'
requirements specified in the PSP.
c. Conclusions :
A test of two perimeter zones identified that they alarmed
appropriately. A walkdown of the perimeter intrusion detection system
identified that design. placement, and coverage met the requirements of 4
the PSP. l
S2.3 . Protected and Vital Area Barriers
a. Insoection Scooe (81700) !
Section 3 of the PSP outlined protected and vital area barriers that are
in place at FNP. The inspectors evaluated those barriers to ensure that
the criteria were being met,
b. Observations and Findinas
The inspector performed a walkdown of 3rotected and vital area barriers.
Fences and gates were intact and met tie overall height requirement.
Manholes were appropriately secured and isolation zones were free and
clear to assure a distinct field of vision. Protected area barriers
were separated from vital area barriers.
Enclosure 2
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_ - - _ _ _ _ _ _ _ _ - - - _ _ - _
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I
41 ,
l
Vital area barriers were appropriately in place and contained no )
openings greater than 96 square inches. Vital c:rea doors were locked I
and alarmed. Access was controlled by a security focce member, card
reader, or the Central / Secondary Alarm Stations. The inspector
accompanied a security officer who performed vital area door checks as
part of his post. All vital area doors were secured and alarmed in the
Central and Secondary Alarm Stations when opened. Vital area
penetration points were secured by locks, alarms, or welded in place.
FNP-0-SP-30. " Declassification of Vital Area / Systems / Equipment." Rev. O.
dated March 11, 1994, was reviewed by the inspector. The licensee
devitalized four valve boxes during Mode 5 of the outage. This
devitalization of equipment met procedural requirements.
c. Conclusions
Protected and vital area barriers were appropriately placed, maintained,
and secured as specified in Section 3 of the PSP. The licensee followed
procedure to devitalize equipment during Mode 5 of the outage. j
S3 Security ana Sifeguards Procedures and Documentation
l
S3.1 Security Procram Plans
a. Insoection Scoce (81700)
To determine if requirements were met, the inspectors reviewed Rev. 8 of
the Training and Qualification Plan, which was subnitted under
b, Observations and Findinos
Revisions to the Training and Qualification Plan met the requirements of
10 CFR 50.54(p). Administrative changes and clarification statements
were also noted.
c. Conclusions
A revision to the Training and Qualification Plan did not decrease the
effectiveness of the plan and met the requirements of 10 CFR 50.54(p).
l
l
!
I
f Enclosure 2
l
_,
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.
42
l S4 Security and Safeguards Staff Knowledge and Performance
S4.1 Resoonse Capabilities
a. Inspection Scooe (81700)
The inspectors reviewed and evaluated the licensee's response force
strategy to determine if the licensee was capable of engaging an
adversary force to preclude penetration of vital area barriers and any
act intended to cause a significant release of radioactivity.
b. Observations and Findinas
The inspector reviewed the Security Response Plan. Rev 5 and drill
critiques for the last two quarters. Three of the four res)onse teams
participated in the drills and the fourth team performed taaletop
exercises. In addition, the inspector discussed with licensee
representatives the current strategy. Target sets established by the
licensee for the 1995 Operational Safeguards Response Evaluation (OSRE)
remained current.
c. Conclusion
The licensee had in place a sound strategy that was capable of
protecting vital equipment from acts intended to cause a significant
release of radioactivity.
S8 Miscellaneous Security and Safeguards Issues
S8.1 Actions on Previous Insoection Findinas (92904)
(Ocen) IFI 50-348. 364/97-02-01: Failure to Provide Locks of
Substantial Strenath to Prevent Tamnerina
The licensee had changed FNP-0-SP-10. " Patrol Procedures." Rev. 16. to
require the motor patrol to physically check the locks every four hours.
once per shi ft. In addition. the locks selected by the licensee were
susceptible to damage by hand tools, creating a possible vulnerability.
The licensee had purchased more substantial locks. However, the
evaluation of the lock covers was still underway. This IFI remains open
pending the completion of the licensee's evaluation.
Enclosure 2
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43.
- (Closed)'IFI 50-348; 364/97-13-01
- Questionable Planned Biometrics
l ; Implementation.
'The licensee was installing and testing the use of biometrics for access
. control in the protected areas. The-licensee had determined that the
l
Service Water- Intake Structure (SWIS) a separate protected area, would
ibe controlled by portable biometric units. Procedures were in ) lace to=
process employees from one protected area to the other.. This III is
. closed.
~ S8.-2 Protection of Safeauards Information
a. =Insoection Scooe (81810)
An evaluation of the licensee's program to protect Safeguards
Information-(SGI) under the provision of 10 CFR 73.21 was conducted.
! b. ' Observations and Findinas
.The inspector reviewed and evaluated FNP-0-AP-72. " Protection of
Safeguards Information." and determined that all components of
10 CFR 73.21 were incorporated. The licensee currently-is storing
, Safeguards Information at various locations. The inspector toured all
areas and randomly checked Ge1eral Services Administration-(GSA)
approved safes to ensure that they were locked. In addition, in
Document-Control, the inspector selected non-Safeguards a) proved
. containers and selected files to ensure that'SGI was not )eing stored at
these locations' All.SGI was appropriately stored.
.
Through discussion with licensee representatives, the inspector
determined that SGI was logged transported, and given to only those
individuals with fingerprints.on file and with a need to know.
- c. ' Conclusions
Safeguards -Information was appropriately handled and stored as specified d
-in'10 CFR-73.21.
F2 .-Status of Fire Protection Facilities and Equipment
F2.1'-(Oce'n)'URI 50-348; 364/98-01-10: Pre-Action Sorinkler System Failures
-(71750)
On May 19. 1998, a. conference call was held between the resident staff,
~','
.NRR. Plant Farley personnel, and Farley Project personnel in Birmingham
to; discuss the status'of this issue. The licensee reported that an ,
Enclosure 2
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..
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44
equipment problem root cause team had been assembled, including
representation from the nuclear industry. The team had met at the
l- . vendor's site and reviewed the problem. No conclusion concerning the
failure cause had been identified. However, the team continued its
evaluation.
V. Manaaement Meetinas and Other Areas {
X1 Review of Updated Final Safety Analysis Report (UFSAR) Commitments
While performing the inspections discussed in this report. the
inspectors reviewed the applicable portions of the UFSAR that related to 1
the areas inspected. The inspectors verified that the UFSAR wording was I
consistent with the observed plant practices, procedures and/or
parameters. The inspectors identified that FSAR Section 5.2.2.3 stated,
" Pressurizer pressure is sensed by fast response pressure transmitters l
with a time response of better than 0.2 seconds." This is faster than {
the acceptance criteria of 0.23 seconds used by the licensee for testing i
the pressurizer pressure transmitters. This is not safety-significant
because the pressurizer pressure instruments currently installed have
response times faster than 0.2 seconds. This was provided to the
licensee for resolution.
X2 Exit Meeting Summary
i
The inspectors presented the inspection results to members of licensee l
management on June 4 and June 25, 1998, after the end of the inspection I
. period. The licensee acknowledged the findings presented. l
The inspectors asked the licensee whether any materials examined during I
the inspection should be considered proprietary. No proprietary
information was identified.
1
PARTIAL LIST OF PERSONS CONTACTED
Licensee
R. Badham. Supervisor, Safety Audit and Engineering Review (SAER)
P. Crone. Engineering Support (ES) Performance Supervisor
K. Dyer. Security Manager, Farley Nuclear Plant
T. Esteve. Planning & Control Supervisor
R. Fucich ES Manager
S. Fulmer. Plant Training and Emergency Preparations Manager
S. Gates. Administration Manager
D. Grissette. Operations Manager
R. Hill. General Manager
D. Jones. Configuration Management Manager
W. Lee, Emergency Preparedness Coordinator (corporate office)
Enclosure 2
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45
i
l T. Livingston. Chemistry Superintendent
R. Martin. Maintenance Team Leader
M. Mitchell.'HP Superintendent
C. Nesbitt. Assistant General Manager. Plant Support
W. Oldfield Nuclear Operations Training Supervisor
L. Revels. Assistant Security Manager. FNP
M. Stinson. Assistant General Manager. Operations
R. Vanderbye. Emergency Planning Coordinator
G. Waymire. Technical Support Manager
G. Wilson. SNC Corporate Senior Engineer
R. Winkler. Engineering Group Supervisor. PMMS
B. Yance. Plant Modifications and Maintenance Support Manager ,
1
'
NRC
J. Zimmerman NRR Project Manager
l
l
INSPECTION PP9CEDURES USED
IP 37001: 10 CFR 50.59 Safety Evaluation Program ,
IP 37551: Onsite Engineering !
IP 40500: Effectiveness of Licensee Controls In Identifying. Resolving and
Preventing Problems
IP 60710: Refueling Activities ',
IP 61726: Surveillance Observations ;
IP 62707: Maintenance Observations
IP 71707: Plant Operations !
IP 71711: Plant Startup from Refueling !
IP 71750: Plant Support Activities
IP 73753: Inservice Inspection .
IP 81700: Physical Security Program for Power Reactors
IP 81810: Control of Safeguards Information l
IP 82701: Operational Status of the Emergency Preparedness Program
IP 83750: Occupational Radiation Exposure
IP 84750: Radioactive Waste Treatment, and Effluent and Environmental
Monitoring
IP 86750: Solid Radioactive Waste Management and Transportation of
Radioactive Materials >
IP 92700: Onsite Follow-up of Written Reports of Nonroutine Events at Power
Reactor Facilities
IP 92903: Followup - Engineering
IP 92904: Followup - Plant Support
l
l
l Enclosure 2
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46
ITEMS OPENED, CLOSED, AND DISCUSSED
Tyng Item Number Description and Reference
Ooen
IFI 50-348, 364/98-03-01 Inadequate Thread Engagement
(Section M1.3).
IFI 50-364/98-03-03 Rod Control Fuse Failures (Section M1.4).
VIO . 50-348, 364/98-03-04 Inadequate Safety Assessment for Mis-wired
Hot Shutdown Panel MOVs (Section E8.1).
EEI 50-348, 364/98-03-05 HSDP Loss of Function Inadequate Safety
Evaluation (Section E8.1).
VIO 50-348, 364/98-03-06 Inadequate Corrective Actions for MCR
Ventilation System (Section E8.5).
VIO 50-348, 364/98-03-07 Failure to Promptly Terminate Security
Access (Section S2.1).
Closed
i
NCV 50-364/98-03-02 Failure to Report Manual Reactor Trip in a
Timely Manner (Section M1.4).
LER 50-348, 364/97-10-01 Motor Operated Valve Local - Remote
Control Circuit Wiring Discrepancies 1
(Section E8.1).
URI 50-348, 364/98-01-04 Inadequate Safety Assessment for Mis-wired
Hot Shutdown Panel MOVs (Section E8.1).
VIO 50-348, 364/97-11-04 Failure to Implement a Test Program for
Service Testing of the TDAFW Battery
(Section E8.2).
URI 50-348, 364/98-01-05 Failure to Track and Correct Conditions
Adverse to Quality (Section E8.5).
IFI 50-348, 364/98-01-06 Control Room Ventilation Testing
(Section E8.5).
IFI 50-348. 364/98-01-07 Review Licensee Actions to Improve
Radioactive Material Container Label
Effectiveness (Section R8.1).
Enclosure 2
,
$
.____..__....n _ _ . _ _ . . _ _ _ _ -. __ -
_ -
,-- _ - _ - - - . _ - - - . - - - _ - . - - _ - - - _ . - - _ - - ------___ ---_- - - _-- --__---- - _ -_-- _ - _ -- _ -_ _ _
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47
- - IFl 50-348, 364/96-14-01 Exercise Weakness--Significant Emergency
Information Was Not Communicated to the
Appropriate Emergency Manager in a Timely i
Manner (Section P8).
'
IFI 50-348, 364/97-13-01- Questionable Planned Biometrics
Implementation (Section S8.3).
Discussed
VIO 50-348, 364/97-11-03 TDAFW Battery Installation and Check Valve
Test Deficiencies (Section E8.3).
URI 50-348, 364/98-01-10 Pre-Action Sprinkler System Failures
(Section F2.1),
.
'
IFI ~50-348.-364/97-02-01 Failure to Provide Locks of Substantial
Strength to Prevent Tampering
(Section S8.1).
I
Enclosure 2