ML18100A353
| ML18100A353 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 05/05/1993 |
| From: | Jason White NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18100A347 | List: |
| References | |
| 50-272-93-08, 50-272-93-8, 50-311-93-08, 50-311-93-8, 50-354-93-06, 50-354-93-6, NUDOCS 9305110084 | |
| Download: ML18100A353 (38) | |
See also: IR 05000272/1993008
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos. 50-272/93-08
50-311193-08
50-354/93-06
License Nos. DPR-70
- Licensee:
Public Service Electric and Gas Company
P.O. BOx 236
Hancocks Bridge, New Jersey* 08038
- * Facilities:
Salem Nuclear Generating Station.
Hope Creek Nuclear Generating Station
Dates:
March 14; 1993 - April 17, 1993
Inspectors:
Approved:
.J. R.
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Date
This* inspection report documents inspections of activities affecting public health and safety
during day and backshift hours, inchiding: . operations,. radiological controls, maintenance
and surveillance testing, emergency preparedness, security' engineering/technical support,
and safety assessment/quality verification .. The Executive Summary delineates the overall
inspection findings and conclusions.
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9305110084 930505*
ADOCK 05000272
Q
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EXECUTIVESUM:MARY
Salem Inspection Reports 50-272/93-08; 50-311/93-08
Hope Creek Inspection* Report 50-354/93-06
March 14, 1993 - April 17, 1993
OPERA TIO NS (Modules 60710, 71707, 71710, 93702)
Salem: The. licensee operated the Salem units safely. The inspector found the licensee's
actions taken in response to the March 16, 1993; Unit 2 reactor trip to be appropriate and
effective, as were their corrective actions and event follow-up. The Unit 2 seventh_ refueling
outage was Initiated following the March 16 reactor trip,* and the inspectors determined the
outage activities performed during *the inspection period to be well planned, coordinated, and '
. executed. Unit 1 operators were forced to reduce ullit power level several times during the
period due to marsh grass accumulation iil the circulating water system. The inspectors *
observed good operator perrormance during these events and noted that Opera~ons
management conservatively *managed unit power as a* result of the environmental conditions.
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Hope Creek:* The licensee operated the Hope Creek unit-safely. _The~e were no* significant
challenges to plant operation. **
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. Co:n:iinon: PSE&.G conducted a frre protectio~ drlll during the inspeetion that involved
assistance from offsite emergency response forces; The drill was a good exercise. of the site .
Fire Protection and Security Departments, both of which responded well~
RADIOLOGICAL CONT&OLS __ (Modules 71707, 93702)
Salem: * P~riodic inspector observation of station workers arid Radiation Protection personnel**
noted* g90d. implementation of ICldiological-controls and protection program requirements.
The inspectors noted good performance in this area especially with respect to the Unit 2
,.,-,,-~-------~----, -- -refueling outage* and* its--associated containment activities.
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Hope Creek: . Periodic inspector observation of station workers and Radiation Protection * .
personnel noted good implementation.of radiological controls and protection program
requirements. _An incident involving a non-licensed radioactive waste operator and apparent
violation of l~_censee procedures is unresolved;- .
MAINTENANCE/SURVEILLANCE' (Modules 61726, 62703)
Salem: . Inspection in this area found good perl~friiance in the routine maintenance and
surveillance activities performed at both Salem units. The licensee declared an Unusual
Ev_ent at p:nit 2 when a maintenance activity involving Service water system piping
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Executive Summary
replacement resulted in the inadvertent discharge of carbon dioxide gas in a vital area. The
inspector concluded that the licensee properly responded to the event, but an unresolved item
was opened pending the licensee's evaluation of the potential generic effects of the event at
both Salem units. The inspector closed an open item after determining that PSE&G has
acceptable programs to assure the Control *ofexpendable and consumable items.
Hope Creek: Two Technical Specification surveillance intervals were missed relating to the
high pressure coolant injection system isolation function, and the main steam isolation valve
. seali11g system valve stroke times. The latter issue is unresolved. The reactor recirculation
pump end-of-cycle and anticipated transient without scram trip breakers were found operable
and related surveillances were acceptable.
EMERGENCY PREPAREDNESS (ModuleS 71707, 93702)
- The*inspectors observed and participated in (1) portions of an unannounced off-hours
emergency preparedness drill at Hope Creek that theJicensee conducted to especially. test
- their automated callout system and. (2) in a routine monthly drill conducted at Salem. The .
inspectors* determined both drills to* be well conducted* and an effective exercise of the
licensee's Emergency Plan.*
SECURITY (Modules 71707, 93702)
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The inspectors determined that the licensee appropriately implemented security program
requirements. The inspector concluded. that the licens~ demonstrated a proactive approach
relative *to severe .winter storm planning and appropriately compensated for any degraded
conditions. The inspector also n.oted that the PSE&G Security Department performed well in
. the* April* 14, 1993, fire proteetion * diill that required the security force personnel to process
offsite emergency response forces into the Artificial Island protected area under simulated
emergency conditions.
ENGINEERING/TECHNICAL SUPPORT (Modules 37828, 71707)
- Salem: The inspeetors noted thatengineering personriel properly prioritized work activities.*
The licensee. engineering staff provided a good evaluation and safety-conscious resolution
whert the Salem units' control air system outboard containment isolation valves were
. determined to -be outside their *design basis.. *The licensee identified that the eniergericy diesel *
generator cooling water flow control valves had been installed with improper setpoints which
. constituted a condition outside* their design basis. The inspector noted good* engineering .
response to the disoovery by the licensee, although the item remains unresolved pending the
licensee's evaluation of the past effect the setpoint error could have had on the generators'
operability under design conditions .
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Executive Summary
Hope Creek: The inspectors noted that engineering personnel properly prioritized work
activities. A violation regarding the lack of testing of the standby start feature on Reactor
Building filtration, recirculation and ventilation system fans was closed.
SAFETY ASSESSMENT/QUALITY VERIFICATION (Modules 30702, 40500, 71707,
90712, 90713, 92700, 92701)
Salem: . The inspectors found generally positive acceptance by Salem operators of the re-
unification of the Salem operating crews and concluded the new shift schedule should have a
..... *positive effect on reactor operator and equipment operator morale.* The inspectors cloSed a *
- previously open item when they . concluded* that the containment isolation function for the
feedwater system continues to be in accordance with Technical Specification requirements: *
Hope Creek: Licensee follow-up to plant events was thorough and effective ..
. Common: . The NRR project managers (PMs). for .Salem and Hope Creek inspected the
licensee's 10 CFR 50.59 program'.. The PMs found several discrepancies in a relatively
small sample size, . which. indicated a weak:ness in the program,. and an apparent violation of
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TABLE OF CONTENTS
EXECUTIVE SUMMARY ...................................... ii
TABLE OF CONTENTS ............ ; . : . . . . . . . . . . . . . . . . . . . . . . . . . . v
1.
SUMMARY OF OPERATIONS ............................... 1
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Salem Units 1 and 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2
Hope Creek . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2..
OPERATIONS * ............ ; * ......... * ........ : .. ; . . .. . . . . . . l
2.1
- Inspection Activities ............. : .. *; ........... *; ... ; . . . . 1
2.2 * * Inspection Findings and Significant Plant Events . . . . ; . . * . . . . . . . . . . . . . * * 1
.2.2.1* Salem . * .. * ............. ; .* *. *.* . : . . . . . . . . . . . . . .. . . . . . . 1
2;2.2 Hope *creek . , ......*..... *; ...*... ; **. ; .. ; .. : .. ~ . 5
2.2.3 Common ........ * .................. * ..... *. . . . . . 5
3. .
RADIOLOGICAL CONTROLS ........................ _ . . . . . . . 6
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Inspection Activ~ties .......... : ............ ; .... , ... * . . 6 *
3 .2
Inspection Findings * : . * . .. . ; . * . . . . . . . * . . . , * . . . . . . . . . . . . . . . 6 *
3.2.1 Salem . *~ ........ ; .. ~ ......... : ...... * * ; ; * ~ * ** * *. *
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3.2.2 Hope Creek
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4. * . MAINTENANCE/SURVEiLLANCE IBsTING ............... *. . . . . . . 7
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4.1
Maintenance Inspection Activity : ; . . . . . . . . . . . . . . . . . . . . . . . : 7
4.2.
Surveillance Testing Inspection Activity .... ; ; ..... ~ .........
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4. 3
Inspection* Findings . . . ; . . . . . . . : . . . . . . . . * . . . * . . . *. . . . . . . 8
. 4.3.1 Salem '. .... * .. _ ..... * .. *. _ ....*... * ... * ... ~ ..... * .* : ... . ..
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4.3.2 Hope*t;reek ........... _:. -.. .- .............. .- ..... .
~MERGENCY PREP A&EDNESS . -. ~ * ......... ; .. ~ ....... : .... .
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Inspection Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
5.2 -* * ~Inspectien--Findings ;: ....*.* * ...... ; .................... .
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SECURITY ; . . . . . . . . . *. . * . . . . . . . . . . . .. . . ~ . . . . . . . . . . . . . . . . . 11
6.1 . Inspection Activity. . . . . . . . . . . ~- .. * . . . . . . . . . . . . . . . . . . . . .
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6.2
Inspection Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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7."" * ENGINEERING/TECHNICAL SUPPORT .. ; . -~ ... ; ~ .. ; . . . . . . . . . . . . 12
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Salem* .. .- . * .... ~ *; ." *. : .. * * ... ; ........ * ............. * .. * * 12
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Hope Creek .... ; .* .... : .. * .. _ .. *. _.'. *. * ................ * . . .. . .
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Table of Contents
Table of Contents (Continued)
8.
SAFETY ASSESSMENT/QUALITY VERIFICATION . . . . . . . . . . . . . . . .
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8.1
Salem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .
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8.2
Hope Creek ................ * .............. *: *.
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8.3
10 CFR 50.59 Program inspection . . . . . . . . . . . . . . . . . . . . . . . .
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8.3.1 Apparent Violation . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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LICENSEE EVENJ' REPORTS (LER), PERIODIC AND SPECIAL
REPORTS, AND. OPEN ITEM FOLLOW-UP ......... ; . . . . . . . . . . .
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LERs and Reports . . . . . . . . . . . . . . . . . . . . . . . . . . ; . . . . * . .
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9.2 **. * Open Items .... ** .* ..... * .......... * ........... '. . . . . . * . .
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10.
- EXIT INTERVIEWS/MEETINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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Resident Exit Meeting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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10.~ Specialist Entrance and Exit Meetings . . . . . . . . . . . . . . . . . . . . . .
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1.
DETAILS
SUMMARY OF OPERATIONS
1.1
Salem Units 1and2
Unit 1 began the inspection period operating at 100% power. On March 15, 16 and 23,
1993, and through April until the end of the inspection period, the unit operated at mostly
reduced power levels due to the effects of seasonal marsh grass accumulation oii the
circulating water system and the main condensers (See Section 2.2.1.C).
Unit 2 also began the period at full power but tripped on March 16, 1993, due to a steam
- generator feed pump trip (See Section 2.2.1.A). PSE&G elected to maintain the unit
shutdown and enter the unit's seventh refueling outage slightly ahead ofschedule. The plant
- reached Mode 5 (Cold Shutdown) on March 15, Mode 6 (Refueling) on March 25, and the *
unit was de-fueled by April 5 arid remained so through the end of the.inspection period.
1.2 . Hope Creek
The Hope Creek unit operated at _power during the period.
2.
- OPERA TIO NS
2.1
Inspection Activities *
The inspectors verified that Public Service Electric and Gas (PSE&G) operated the facilities
safely and in conformance with regulatory requirements. The inspectors evaluated PSE&G's *
management control by direct observation of activities, tours of the facilities, interviews and
discussions with personnel, independent verification of safety system status and Technical
Specification .compliance, and review of facility records. . The inspectors _performed normal
- and back-shift inspections~* including deep back-:shift (39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />) inspections.
2.2
Inspection Filldiri.gs and Significant Plant Events
2.2.1 Salem
A.
Salem Unit 2 Reactor Trip -
- On March* 16, 1993, at 11:06 a.m., the Salem Unit 2 reactor automatiCany tripped from
100% power. At 11:04 a.m. the No; 22 steam generator feed pump (SGFP) had tripped on
fow suction pressure, -and the *operators initiated a turbine generator runback, in an attempt to
- reduce power to 60%. I:Iowever, before the runback was completed the reactor tripped on
No. 24 steam generator (SG) fow-low level. The licensee informed the resident inspector,
and the inspector arrived in the control room approximately five minutes after the reactor
trip. The licensee subsequently reported the event to the NRC Operations Center .
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Prior to the No. 22 SGFP trip, reactor and SG pressures and temperatures were staqle.
Control room operators received No. 22 SGFP "High-vibration" alarm and noticed that No.
22 SGFP had* tripped. Operators initiated a turbine load reduction. All SG levels trended
downward. An automatic reactor trip occurred when No. 24 SG reached its low-low level
trip setpoint.
The licensee entered the reactor trip procedures, Emergency Operating Procedure (BOP) -
Trip-I and 2, which required initiation of a manual steamline isolation because a high
auxiliary feedwater (AFW) flow rate resulted in lowering primary system average *
teinperature. Systems responded normally to the trip with. the followmg exceptions: (1) the
No. 24 SGfeed regulating valve (24BF19) failed open, and (2) the No; 23 AFW puinp .*
restarted even though rio valid start signal was present. The licensee cooled down the plant
and entered Mode 5 (Cold Shutdown).* Licensee management elected to *commence the Unit
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2 seventh refueling outage four days ahead of hs scheduled start date. The licensee formed a
Significant Event Response Team (SERT) to determine causes and corrective actions for the
reactor. trip.
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The licens~'s investigation *deterinined *the proximate cause of tile event to be the failure of a *
- condensate polishing (CP) system pressure control swit~h due to water intrusion from a
leaking valve. This caused the 24CP2 valve to open, diverting SGFP suction flow and
resulting in the actuation of the low suction pressure switch on the No;* 22 * SGFP. The
licensee.confirmed the cause during a condenSa.te system test oii March 18, 1993, and
. replaced the failed. pressure <<xmtrol switch. Tlie licensee determined the root cause to be a
- management/QA deficiency for failure to take timely actions to correct the leaking * *
.condensate polishing valve. The licensee determined that the 24BF19 valve was held in *the
open position by a piece of metal pipe, which was apparently from a broken chemical_ feed
line upstream of the. valve. this pipe failure was due to an original construction deficiency.
A -check of Unit 1 did not note the same deficiency. Licensee investigation into. the 23 AFW
. pump unexpected restart determined that a* start/stop valve *solenoid failure "allowed the steam
admission valve (MS132) to stay open. Thus, the pump restarted without an actual start
signal. The licensee replaced a faulty auto start relay contact. The licensee _submitted
Licensee. Event Report_ (LER) 93~05 for this event.
. The inspector* reviewed the operations logs and control room recorders, verified BOP .
implementation, interviewed onshift operators, and reviewed *and discussed the event with the
- SERT team and plant management. The inspector reviewed the-SERT report, LER 93-05.
and the AD-16 procedure (post reactor tiip review). The inspector found the licensee's
actions taken in response to the event appropriate and effective. 1be inspector'-s evaluation
of the licensee's corrective actions and. event* follow-up determined them to be appropriate.
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B.
Unit 2 Refueling Activities
On March 29, 1993, the licensee commenced core offload for the Unit 2 seventh refueling
outage. The inspector observed fuel handling activities from the fuel handling building,
containment refueling platform and the control room. The inspector noted good
coordination, communication and cooperation between the licensee and the Westinghouse fuel
handlers. The defueling process was performed with precision and professionalism. The
inspector noted that the fuel handling supervisors emphasized attention to detail, quality
control, and good* radiation work practices, even at the expense of expediency. The
. inspector interviewed the refueling senior reactor operator' the equipment operator' the
upender operators, and the Westinghouse shift engineer. All personnel were cogh~t of*
their duties and responsibilities and very knowledgeable of the procedure and process.
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The licensee. temporarily stopped refueling activities on two occasions. At 6:02 a.m. on
March 31, 1993, the spent fuel *pit (SFP) fuel transfer system upender operator inadvertently
lowered the upender while fuel assembly U-47 was being withdrawn.* The fuel assembly was
raised approximately 75% out of the upender frame when the upender _operator acd.dentally
bumped the '.'fraine down" button on the fuel transfer system control console; The upender
oper~tor immediately" pushed the "frat:ne stop" button. The upender frame moved
approximately five inches down. The spent fuel tool operator, stationed on the SFP bridge,
notified the refueling shift supervisor in containment. The shift supervision assessed the
- situation, then directed the frame upended,* arid the fuel assembly removed and placed in the *
spent fuel* pit. The spent fuel tool operator performed a visual inspection of the fuel *
assembly and upender basket. No damage was detected. The licensee resumed fuel handling
at 6: 14 a.m; on March 31. On April' 6, the U-47 fuel assembly *was the first .assembly to
undergo an ultrasonic test and visual *test. The licensee and Westinghouse fuel *
representatives performed. a thorough.examination of the assembly and concluded *that no
damage was done to the assembly. * .
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At 4:30 p.m. on March 31, 1993, fuel handling activities were halted when a weld broke on
the fuel transfer system conveyor car roller chain. The conveyor car supports the fuel
assembly as it is transported horizontally through a.tracked tubular.passageway connecting
the reactor side lower. cavity .and the fuel handling buildillg spent fuel pit. . The track- .
mounted conveyor is driven by .3: sprocket and chain drive mounted between the tracks. The
roller chain is welded to the bottom of the conveyor car and engages the floating drive
sprockets on the reactor side. The broken .weld allowed the. roller chain to disengage from
- the sprocket, preventing the transfer of fuel assemblies between the reactor C:avity and the
SFP. The licensee evaluated the situation and decided the. chain could best be repaired by
removing the conveyor car from the lower cavity and rewelding the roller chain on the
containment refueling deck. The licensee removed, repaired and replaced the trailing 20-foot
portion of the conveyor car. During this repair the licensee discovered five broken welds on
the leading 15-foot portion of the conveyor -car. These welds were also repaired. The
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licensee attributed the broken weld to roller chain wear and misalignment over time. The
core offload resumed at 6:44 a.m. on April 5, and was completed later that morning at 8:45
a.m.
The inspector discussed the two events, the fuel transfer system design and operation, and
the licensee's work plan with the cognizant reactor engineer. The inspector concluded that
the two events were unrelated, the work activities were well-planned and the licensee
exercised appropriate safety radiological precautions. The inspector discussed the inadvertent
upender operation. incident .investigation with the licensee. The inspector noted that the
resultant consequences were minimal and the licensee's immediate actions were acceptable.
Overall,. thein~pector found the Unit 2 offload appropriately planned, coordinated, executed,
and documented.
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C.
Unit l Power Reductious Due to.Circulating Water Concerns
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At various times. during and .through the majority of the second half of this. inspection period, .
PSE&G operated Salem Unit 1 at a reduced power leyel due to concerns related to the
. circulating water system .. * The circulating water. system uses the Delaware River as 3: source
of water to condense the steam exiting the main* turbine. For environmental reasons, the
State of New Jersey limits the allowed 24-hour average temperature rise for circulating water
to a maximum of 27 .5 °F .. At various times during the inspection period, because of silt and
grass collecting in condenser waterboxes, Salem Operations isolated individual* waterboxes to
clean them and lower the temperature rise across all of the waterboxes. In order_ tp avoid
... approaching the temperature limits, Salem *operators reduced unit power while waterboxes .
were isolated. During the second half of the inspection period, circulating water problems
were aggravated by a large amount of marsh grus collecting *in the river by the circulating
water intake structure. * This marsh grass phenomenon occurs this time of year at Artificial
Island .and was worsened this year by several rain storms. As a result of grass accumulation
on the circulating water traveling screens, circulating water flow was decreased and, at
times, circulating water pumps tripped off due to low suction pressure.
- The effects of the ciiculating water system events presented a challenge to the Salem . _ - . .
operating .crews as they were forced to adjust plant power to match the available circulating
water* system configuration and to adhere to the water temperature-rise limits. The operators
- re"ducedpower to approximately 90% on March 15, 16 and 23, 1993, to accommodate *
- *circulating water system restrictions arid, during the first week of April, power .level waS
varied several times between. 60 % and full power. From that time through. the. end of the
report period, Operations management made the decision to maintain power level between
- *10% and 80% until the environmental conditions cleared .
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The resident inspector observed the challenge these conditions presented to the Salem
operators, especially on the night of April 14, when up to four of the six circulating water
pumps were lost from service. The inspector noted good performance on the part of the
operating crews, and the assistance they received from Salem maintenance and site services
personnel, *in operating the unit and managing to keep it on line.. The inspector also noted
Salem Operations management's decisfon to operate at a lower power level during these
conditions to be conservative and prudent.
2.2.2 * -Hope Creek
- The Hope Creek unit remained at or near *fun powe~ during th~ period.* The inspectors
monitored steady~state unit operations and performed routine inspection activities. The
inspectors concluded that the licensee safely operated and maintained th*e *unit during this
inspection period.
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2.2.3 Common
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A.
Fire Protection.and Security Offsite Assistallce Drill __ _
On April 14, 1993, PSE&G conducted an after-hours drill involving the site Fire protection
and Security Departments -and which was designed to require assistance* from off site
- *emergency response forces. The drill scenario consisted of a simulated fire and personnel
- injuries at the new *warehouse -facility inside the -protected area at Artificial Island. The -
- primary focus of 1he drill was to exercise the PSE&G Nuclear Fire Protection Department
and its ability to combat the fire, loCa.te and assist the injured personnel, and to solicit the
help of_and integrate *the participation of offsite forces. The drill also required the
partiCipation of the PSE&G Site Protection Department in that site security forces were -
required to process the offsite_ response personnel futo the protected area via emergency
- procedures.
On the day of the drill, the PSE&G Nuclear Security Manager and the Senior Nuclear Fire
Protection Supervisor briefed the NRC resident inspectors on the drill scenario and _
expectations for *licensee performance. -. Followmg the performance of the drill, the inspector
discussed the drill with the Senior Fire ProteCtion Supervisor, reviewed the licensee's drill
critique and viewed a video tape of various aspects and highlights of the drill. In addition* to
_ *reviewing the drill results, the inspector toured the site Fire Protection Department's facilities
anci equipment and diseussed the DePartffient's capabilities with-the Supervisor. The
inspector concluded that the drill had been a. worthwhile exercise of the PSE&G Fire
Protection and Security Departinerits and that the Fire Protection Department -maintains an
ability to* respond to site emergencies very well. -
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3.
RADIOLOGICAL CONTROLS
3.1
Inspection Activities*
The inspector verified on a periodic basis PSE&G's conformance with the radiological.
protection* program.
3 .2
Inspection Findings
3~2.1 Salem
- A~
Containment Tours
The fuspecto~ periodiCany toured the Salem Unit 2 containment during the *current refueling
outage i)eriOd. Items checked included access controls, use of anti-contamination clothing,
worker radiation practices, dosimetry and exposure controls, decontamination procedures,
tool control, and work in progress. The inspector found the radiation protection personnel
very kiiowledgeable, extremely visible in and out of containinent, actively involved in
. radiological .. controls, and* keenly interested in minimizing exposure to. workers .. The * * * *
inspector observed the licensee's use of appropriate radiological precautions and controls
. during the core offload and weld repair of the refueling conveyor car; The inspector noted
. good radiation work practi~s and. a good ALARA. consciousness. * * * *
3.2.2 Hope Creek
. A.
. Improper Personnel Entry Into Radiological Controlled Area *(RCA)
- an Monday, April 12; 1993, during an informal audit of radioactive. waste operators' time*
. sheets and RCA access records, a licensee supervisor noted an .apparent discrepancy between
one operator's stated work start tiine and RCA access times. When questioned about the
discrepancy, the operator stated that he had reached his assigned work area, the radwaste
control room,. after transiting portions of tlie turbine building, including the maintenance "hot
control point to obtain his. dosimetry. The licensee determined that this method of entry had
occurred twice, once on Saturday, April 10, and again on April 12.
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The licensee concluded that these acts were violations of administrative p~ocedures
NC.NA.AP-ZZ-024, "Radiation Protection. Program," and HC.SA.AP-ZZ-0046, "Radiological
Access Program." The individual's dosimetry was immediately pulled and access to the *
RCA barred. * Thiough a review of applicable aceess records and time sheets of a number of
other radwaste workers, the licensee determined that this violation of RCA requirements was
apparently an isolated case, as no other such dis~repancies were found.
Longer term actions
.. were being developea when. the inspection period ended.
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The inspector discussed this incident with health physics and plant management. The
inspector was concerned that these violations of station procedures had apparently been.
committed by an experienced radiation worker on at least two occasions. This issue is
unresolved pending completion of the licensee's corrective actions and NRC review (URI
5().;354(93-06-01).
4.
MAINTENANCE/SURVEILLANCE TESTING
- 4.1
Maintena1,1ce Inspection Activity
- The inspectors observed selected maint~nance activities on safety.,related equipment to.
ascertain that the licensee conducted these activities in accordance with approved procedures,
Technical Specifications; and appropriate industrial codes and standards.
The inspector observed portions of the following activities:
WorkOrder(WO) or Design* *
. Unit.
. Change Package (PCP)
Description
Salem 1
Various
Salem 2 *
Hope Creek
..
Circulating water and main condenser*
repair and cleaning
Repair of fuel transfer systein conveyor
car
Replace transmission seal on Technical
Support Center (fSC) chiller BK403
The maintenance activities inspected were effective with respeet to meeting the safety .
objectives of the maintenance program.
4.2
Surveillance. Testing Inspection Activity
The inspectors performed detailed technical procedure reviews, witnessed in-progress
surveillance testing, and reviewed completed surveillance packages. The inspectors verified
that the surveillance tests.were performed in accordance with Technical Specifications,
- approved procedures, and NRC regulations.
.
.
.
The inspector reviewed the following suI'Veillailce tests with portions witnessed by the*
- *. inspector: * *
.*
Salem
Hope Creek
Hope Cr~k
Procedure No.
S2.0P-ST.DG-0002(Q)
OP-ST.KJ-001
OP-iS.BD-001
8
Unit 2 Emergency Diesel Generator
2C Operability Test
11A
Monthly Surveillance Test
Reactor Core Isolation Cooling
(RCIC) Jockey Pump *92-Day
Inservice Test
The surveillance testing activities inspected were effective. with respect to meeting the safety
objectives of the surveillance testing program.
4.3
Inspeetion Fb:i.clings
4.3.1 Salem *
.
. ..
A.
Unusual Event at Unit 2 Due to Toxic Gas DiScharge
On April 3, 1993, with Unit 2 in the Refueling Mode; Salem Maintenance workers were
performing service water piping replacement work in the 78-foot elevation mechaniCal
penetration area. Due to silt clogging of the pipe, and unknown by the workers, the piping
. to be replaced CQUld not be properly. drained. When the workers cut the line, water sprayed
into the overhead of the mechanical penetration. area. Approximately five hours later, the
carbon dioXide fire. protectiori system discharged into the. adjacent 78-foot elevation electrical
penetration area ... The licensee immediately *evacuated the effected area and declared *an
Unusual Event due.to the discharge of a toxic gas which was a threat to personnel in a vital
area. The Salem operating crew properly notified the NRC Operations Center and resident
inspector of the event and the emergency *declaration. The licensee terminated the Unusual
Event approximately one hour after the event; upon restoration of the electrical penetration
area to normal fa1bitability conditions. and determilling *no personnel injuries had resulted*
from the event.
.
-
...
Following a subsequent investigation, the licensee determined the cause of the inadvertent
discharge of the carbon dioxide system to be. water intrusion mto the. carbon dioxide system
control panel located in the mech.anical penetration area. The licensee's. investigation also
revealed that the orily potentially significant plant equipment effected by the event was the
.
11A
11 reactor vessel level indication channel, which was not required to be in service at the
- * time of the event but was adversely affected by the cold temperature resulting from the
discharge. *The resident inspector verified that PSE&G had properly implemented their
Emergency Plan and that the proper personnel protection actions had .been taken subsequent
to the event. The inspector also discussed the event with the Salem Quality Assurance
9
inspector who led the licensee's investigation, and concluded that PSE&G had properly
responded to the event and determined its root cause in a proper fashion. At the end of the
report period, however, the licensee had yet to determine if the ability of the water to intrude
and collect in the affected control panel was a design flaw or a result of improper
construction or installation. Until this question has been resolved and corrective measures
identified, this item will remain open (URI 50-272 and 311/93-08-01).
B.
Open Item Follow-up
(Closed) Unresolved Item 50-272&311/91-16-01: Review of existing* programs to assure
acceptable control of Safety-related expendable arid consumable items. The inspector
reviewed the licensee's procedure that certifies reactor and secondary plant bulk chemicals.
The procedure, No. SC.CH-CA.ZZ-:0401(Q), provides sampling and analysis i:equirements
for selected bulk chemicils that may be a source of impurities for plant systems. The
inspector also determined that expendable and consumable items (chemicals) are classified,
labelled and controlled per the requirements specified in Nuclear Administrative Procedure
(NAP) No.* 38, "Chemical Control Program." The inspector conciuded that the licensee has
acceptable.programs to assure*control of expendable and consumable items, and therefore
. closed this unresolved item .
. 4.3.2 Hope Creek
.
.
- .
.
A.
Missed SurveilhinceS on Main Steam Isolation Valve Sealing System (MSIVSS)
Valves and High Pressure Coolant lnjection*(HPCI) Systein
. On March 23, 1993, during a review of several outage work orders for which the original
retest activities were not available, the licensee -discovered that several Valves in the MSIVSS
had not been surveilled as required by Technical Specification (TS) 4.0.2. Both valves (HV-
2512B.and HV-5829B) were refurbished during the fourth refueling outage (September-
November 1992). At that time, the post-maintenance tests apparently had not required
timing the valve stroke. The licensee could find no documentation that the surveillance
procedure (OP-IS.KP-103)_had been performed until March 1993. The licensee concluded
that both valves were operable as, at the time of discovery, both were within their current
surveillance* frequency.
LER. 93-01 discussed the circumstances surrounding a missed* Technical Specification
- required surveillance on the high pressure eoolant injection system (HPCI) isolation delta
temperature instrumentation. Due to a procedure inadequacy* covering the use of primary
and backup instruments, technicians failed to perform* a channel calibration on the A2 logic .
channel after . the primary instrument had been repaired* during the fourth refueling outage.
The inspectors noted that the safety significance of this event was minimal as other tests
performed before, during and after the surveillance was missed indicated that the A2 channel
. was capable of performing its isolation function. Related HPCI isolation logic channels were
also functional. The licensee's corrective actions included performing the appropriate logic
10
channel surveillance and procedural revisions to identify requirements for spare channel
usage and restoration to normal configuration. The inspector noted that this LER was.
generally well-written. However, in noting that there had been one similar occurrence, the
licensee incorrectly referenced LER 86-09 (The correct LER number was 89-06.). The
inspector brought this minor discrepancy to the licensee's attention, who indicated that a
corrected LER would be submitted. The LER remains open.
The inspector reviewed the event and concluded that there was minimal safety . significance to
the* missed surveillances; However, the. inspeetor* noted that this* was a ~orid instance of a
missed TS required surveillance during this reporting period. The licenSee's review of these
everits is ongoing. These events are unresolved pending completion of the licensee's review
and implementation of corrective actions (URI 5~354/93-06-02). *
B.
Recircui~tion Pump Trip Logic* Surveillance ..
.
.
On March 4, 1993, the licensee at the Washington Nuclear Power Station, Unit 2 (Hanford)
reported (reference EN No. 25i90) to the NRC that the end-of-cycle recirculation pump trip
breakers were inoperable due to pever having been. surveilled. * The licensee diseov~red this
fact during a design review. ** Because of the similarities petween Hope Creek and Hanford
- regarding the reactor recirculation systems, the inspectOr reviewed PSE&G's surveillance.
procedures and discussed this event with licensee operations and maintenance supervision.
The* inspector determined that both*the anticipated transient without scram (ATWS) breakers .
and the end-of-cycle pump trip breakers (two breakers for each function) were tested to .
demonstrate operability. Each breaker is tested individually with the control logic circuits
- tested from each process input to the breaker trip coil. . That the breaker trips open when its
- - ** * --- * associated trip coil energiZes is also demonstrated. Based on this review and discussion, the
inspector ooilcluded that Hope.Creek's recirculation pump trip breakers were operable and
that the surveillance procedures were adequate _to demonstrate operability.
5. ** . ***EMERGENCY PREPAREDNEsS
. 5.1
Inspection Activity ..
The inspector reviewed .PSE&G's conformance with 10CFR50.47 regarding implementation
of the emergency plan and procedures. In addition, the inspector reviewed licensee event
. notifications and* repc)rting requirements per 10CFR50.72 and 73.
5~2
Inspection Findings-
A.
Off-Hours Drill at Hope Creek.
In order to evaluate the effeetiveness of their emergency response organization's automated
callout system* and to demonstrate the ability to activate their. em.ergeilcy response facilities
_(ERF.s) . .within an hour of notification of emergency responders, PSE&G oonducted an
11
unannounced off-hours emergency preparedness (EP) drill at Hope Creek early on the
morning of March 25, 1993. As part of the drill, the licensee contacted the appropriate
personnel via their pagers, and those employees reported to their assigned positions at an
ERF (either the Hope Creek Operations Support Center, the Hope Creek Technical Support
Center, the Emergency Offsite Facility or the Emergency News Center).
The NRC resident inspector was appropriately notified by the Hope Creek control room
communicator during the drill and, subsequent to the drill, discussed the drill conduct and
results with the PSE&G BP Manager and his staff. The inspector observed portions of the
licensee's automated callout system, reviewed the accountability logs kept by PSE&G during
the drill, and determined the drill had been appropriately conducted and that PSE&G had
accomplished the requirements of their Emergency Plan for ERF manning.
- B. *
Routine Emergency Preparedness Drill Conducted at Salem
On March 31, 1993, the PSE&G Emergency Preparedness (BP) organization conducted a
routine monthly drill at the Salem facility. The drill involved the simulated sabotage of the
Salem service water intake structure and a subsequent loss-of-coolant accident, which
required the licensee to man the Salem Operations Support Center, the Salem Technical
Support Center (TSC) and the Emergency Off site Facility, and to declare a General
Emergency per their Event Classification Guide.
The NRC resident staff participated irt the drill at the TSC and the Salem simulator control
room and determined that PSE&G personnel followed the appropriate procedures and
performed well during the drill, and that the drill provided a good exercise of the licensee's
Emergency Plan and organization.
6.
SECURITY
6.1
Inspection Activity
PSE&G's verified regularly the conformance with the security program, including the
. adequacy of staffing, entry control, alarm stations, and* physical boundaries.
6.2
Inspection Findings
A.
Security Operations Thi.ring the March 12-15, 1993 Winter Storm
The inspector reviewed the licensee's security operations during the severe winter storm that
occurred March 12-15, 1993. (The inspector previously reviewed storm preparations and
associated conduct of plant operations in NRC Inspection 50-272, 311, 354/93-02.)
12
The licensee initiated plans for staffing and reviewed security plan contingencies prior to the
storm arrival on March 12, 1993. During the storm, security staffing was maintained by
augmenting the onshift personnel, including holding personnel beyond their shift change in
order to sleep. During the storm, the licensee compensated for degradations that occurred to
portions of the security hardware caused by high winds and rain/snow.
The inspector discussed these security operations with security management personnel and
with selected guard force members. The inspector also reviewed a licensee security report
regarding this storm. event. The inspector concluded that the licensee demonstrated *a
proactive approach to storm planning and appropriately compensated for the degraded *
conditions. *
B.
. PSE&G Offsite Assistance Drill
.
.
On April 14, 1993, PSE&G conducted an after-hours drill of the site Fire Protection and
. Security Departments which required the solicitation and integration of off site emergency .
response forces (see Section 2.2.3.A of this report) .. The inspector concluded that the site
security personnel had performed well in expediting the in-processing of the required *offsite
personnel per emergency prOcedures as part of the dnll scenario ..
7.
ENGINEERING/TECHNICAL SUPPORT
7.1 .. Salem
A. * * Unit 1and2 Containment Isolation Valves Determined to be Outside.Their
Design Basis
During the review of a design change to repla:ce the solenoids for the control air system air-
. operated outboard containment isolation valves on both Salem units, the licensee determined,
on March 4, 1993, that the valves were not as described in the Salem Updated Final Safety
Analysis Report (UFSAR). The UFSAR states that automatic isolation valve closures are fail
safe, i.e. closure is initiated, upon loss of voltage and/or control air. The identified valves,
- which are normally .open, fail "as;.:is" on loss of l25 VDC power to their solenoid actuators;
the solenoids inust be energized tci ()pen the valves and to clOse them. The valves. are
designed to perform their isolation* function upon receipt of a Phase A isolation signal, but *
. the valves would be unable to close if 125 VDC power was not available.
. .
Upon discovery of this ci>ndition, which is an as-built condition for both u~ts, the licensee.
- initiated an engineering evaluation to determine if this as-built configuration is appropriate
with no changes. Factors considered in the licensee's evaluation were: the reliability of the
125 VDC electrical system; the presence of the inboard control air containment isolation
valves and the fact that they are mechanical check valves; the availability of the two
.. *Emergency Control Air. Compressors to maintain *pressure in the control air header and the
fact that minimum header pressure would be 65 psig, greater tha:n containment design
- - - - - - - - - - - - - - - - - - - - - - - - - -
13
pressure of 47 psig; and that the outboard isolation valves still go closed upon loss of air
pressure. The licensee's evaluation concluded that the valves' fail as-is configuration
provides a level of safety consistent with lOCFRSO, Appendix A, General Design Criterion 56, which requires, in part, that
11 *** upon loss of actuating power, automatic isolation valves
shall be designed to take the.position that provides greater safety.
11 Based on that conclusion,
the licensee prepared a Justification for Continued Operation (JCO), performed a
.
10CFR50.59 Safety Evaluation, reviewed and approved both at a Station Operations Review
Committee (SORC) meeting, and submitted a request for licensee amendment to the NRC, in
accordance with 10CFR50.90, in order to change the description of these valves in the
UFSAR. *
When the licensee originally identified the above.identified discrepancy, theyproperly *
notified the NRC Operations Center in accordance with 10CFR50. 72 and. the r¢sident. .
inspector. The resident inspector diseussed the condition with PSE&G engineering staff, *
examined the.licensee's JCO and 10CFR50.59 -evaluations, .attended the related SORC
meeting and reviewed PSE&G's Licensee Event Report (See Section 9.1) and 10CFR50.90
submittals. The inspector concluded that while P$E&G's lack of awareness of the Salem
plants' as-built configuration. was a weakness, but .also that the licensee performed well in
evaluating and- resolving this issue and that the as-built configuration of the control air .. *
outboard containment isolation *valves did not adversely affect the Safe operation of the Salem .
plants.
B.
Emergency DieSel Generator Cooling Water Flow Outside Design Basis
During sen1ice water piping upgrade work on the Unit 2 Emergency Diesel Generators
- -** * --* * (EDGs) on April 7, 1993,--iiie*licensee*found an error in the setpoint of the differential
---* -- -pressure controllers for the-val:veswhich modulate service water flow to the ED"Gjacket
water coofors and lube oil coolers at both Salem units. The field setpoint matched the *
PSE&G. system description, i.e. a 6 psig . drop across the coolers. .. PSE&G believed this
value to*be the value specified by the manufacturer's design, however, the manufacturer had
-~~-,c" _________________ s~ified this value. fo_i:_~c;Q .cooler, not for the total pressure drop across both coolers in
. series as was found. The licensee determined the result of this error to be an approXimate
16 % *reduction in the 700 g3.llon per* niinute design *flow .rate of service water through the.
coolers, and this resulted in a conservative determination of the EDGs only being operable if
service water temperatures remain befow 60°F. Once this condition had been identified,
PSE&G made the proper notifications to the NRC, generated design change requests to
properly set the controller setpoints, and initiated an engirieering evaluation to determine the .
historical design basis significance ()f the situation.
The NRC resident inspector discussed the discrepancy with PSE&G engineering and *
determined that the discrepancy did not immediately impact EDG operability, in that river
water temperature did not rise above 60°F prior to the implementation of the flow controller
design change; * The inspector also verified through .licensee data. that the EDGs had not .
- eXp.erienced any heat ioad problems during their lifetime and that the design changes were
adequately implemented in a timely 111anner. By the end of th~ inspection period, the
14
licensee had not completed the evaluation of the past effect the setpoint error could have had
on EDG operability under design conditions. Until that evaluation has been reviewed by the
NRC, this item will remain open (URI 50-272 and 311/93-08-02).
7 .2
Hope Creek
A.
Open Item Follow-up
(Closed) ViolatiOii (50-354/92-03-04); Inadequate Filtration, Recirculation and Ventilation
system (FRVS) Testing. Ori JUiy 17, 1992, the licensee responded *to a Notice of Violation .
(NOV) involving FRVS surveillance testing in which the automatic start function of the
standby ventilation. unit was not periodically tested. The licensee committed to *a number of
correetive actions, as detailed in their response to the NOV (Letter NLR-N92097, dated July
17' 1992). The inspector reviewed the licensee's response and ensuing corrective actions and
determined that:
. The licensee modified surveillance proCedure HC~OP-ST.SM-002, "Primary
Containment Isolation System/Reactor *Building and Refuel Floor Corit:ainlilent *
.. Isolation Functional *Test-18 Months, ... to include testing of the eontrol logic for both
- the auto-lead and standby functions of the FRVS fans, including the proper operation
of the two-minute time delay component. The procedure also referenced the *
appropriate acceptance criteria.
The licensee performed an evaluation of plant systems to identify any system transfer
. function whose failure to transfer could result in a loss of the system's. safety
function. The report was comprehensive and thorough. . The :review identified one .
- similar instance. The licensee properly documented their finding in Incident Rep0rt 92-184; Their corrective actions were appropriate. The.licensee had ctl.so *
implemented a number of other recommendations effecting non-safety related systems.
The licensee performed a test of the auto-lead and standby functions of the FRVS fans
during the eighteen month . surveillance tests on the A . and B. emergency. diesel
generators in October 1992. The response time for both standby farts was 120.5
seconds (US-125 seconds was required).
Based on the foregoing, the inspector concluded that the licensee .had acceptably addressed
the issues cited in the violation and therefore closed the Violation.
-.
15
8. *
SAFETY ASSESSMENT/QUALITY VERIFICATION
8.1
Salem
A.
New Shift Schedule for Salem Operations Nuclear Control Operators and
Equipment* Operators
On April 4, 1993, the Salem Operations Department re-aligned the shift schedule of the
reactor operators (ROs) and equipment operators (EOs) such that the ROs and EOs would be
.* working a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift schedule to match the rotation schedule of the .shift senior reactor
operators (SROs). The SROs had been pfaced on the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift rotation on.November 15,
1992, but the*ROs/EOs maintailled an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rotation due to their union's objection to the 12
hour schedule (see NRC Inspection 50-272 and 311/93-01). Since that time, enough Salem
ROs/EOs were attracted to the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> rotation that ihe union dropped its objection, and
Salem _management obliged the ROs/EOs and re-united the ROs/EOs and* the. SROs on
. common shift sch~ules.
- .
.
.
The NRC resident staff noted thaf the split-:-shift configuration of the Salem operating *crews*
had not adversely effeeted lieensed operator individual or t~m performance, and the shift*
overlap had, in fact, helped to foster Operations Department unity. In discussions with
licensee operations following the April 4 change, the inspector found generally* positive
acceptance of the shift re~unification and concluded the new scheclule should have a positive
effect on RO/BO morale. The inspector will continue. to monitor the transition to the new *
- shift schedule and its effect on operator performance.
B.
Open Item Follow-up
(Closed) Unresolved Item 50-272&3~1/92-01-05:. Concenis associated with. environmental
qualification (EQ) for the feeclwater system *stop check isolation valves (BF-22s) .. The
inspector found that BF-22 valves do not currently meet all EQ requirements to function as
containment isolation valves (CIVs). For the interim;the licensee will continue to rely on
the mai~ and bypass feedwater regulating. valves (FRVs) to function as the CIVs (per
- Teehnical Specifications). The licensee plans. to fully qualify the BF-22sand then revise the
Technical Specifications to replace the FRVs with the BF-22s for CIV purposes .. The
inspector concluded that the contfilnment isolation function for the feedwater system *
.continues to be in. compliance with Technical Specifications~ This item is closed.
8~2
Hope Creek .
Licensee follow-up to plant events was thorough and effective.
.,
16
8.3
10 CFR 50.59 Program inspection
A.
Overview and Objective
The NRR*project managers (PMs) for both Salem and Hope Creek.inspected th.e licensee's
Safety Evaluation Program (10 CFR 50.59 program) from February 23 through 26, 1993,
and March 30 and 31, 1993. The PMs performed the inspection in accordance with
Inspection Procedure 37001, "10 CFR 50.59 Safety Evaluation Program," issued December
79, 1992. .
.
.
.
.
.
.*
.
.
.
The objective .of the inspection was to verify *that the licensee implemented a safety
evaluation program in conformance with 10 CFR50.59, "Changes, Tests and Experiments
11
(CTEs). The objective was accomplished by (1) reviewing thelicensee's* procedures to
verify that they conform to the 10 CFR 50.59 rule; (2) reviewing the licensee's training
program; and (3) reviewing a sample of the licensee's 10 CFR 50.59 reviews and safety
evaluations (SEs). The PMs noted that the licensee performs applicability reviews to.
determine whether 10 CFR 50.59. applies to a propose<l CTE, and performs a safety
evaluation (when it is determiried that 10 CFR 50.59 applies) to cietermine whether the .
- .proposed .CTE involves an unteviewed safety question .. Accordingly,* the PMs reviewed a*
sample of CTEs that receive an applicability review and subsequently required a Safety
. Evaluation; and a sample of CTEs .that were reviewed for 10 CFR 50.59 applicability but.did
- *not require a. lQ CFR 50 .. 59 Safety Evaluation, a~ determined by the licensee.. A list of the
CTEs that were reviewed by the PMs is* containea fu Attachment 1.
B.
- * *Procedur~ Review, Salem/Hope* Creek
Admmistrative Procedure NC.NA*AP.ZZ-0059(Q), Revision 0, "10 CFR 50.59 Reviews and *
Safety Evaluations" (NAP-59), is the licensee's governing document for fO CFR 50.59
reviews and safety evaluations. The procedure was previously reviewed as documented. in
inspection reports 50-272/91~26; 50-311/91-26 and 50-354/91-19.
Durfug this inspecti.on. the PMs .reviewed NAP-59 in accordance. with the gl!idance provided
in Inspection Procedure 37001. The PMs detenitlned that NAP-59 is well written,
. comprehensive, and adequately addressed implementation of the 10 CFR 50.59 rule.
- C.
Training Review, Salem/Hope Creek
PMs determined that the training program is excellent, overall .. The. program provided an in-
. depth and extensive discussion of the lO*CFR 50.59 rule, NAP-59, and recent NRC and
industry 10 CFR 50.59 guidance. However, the PMs detected an eleinent where the
guidance given by the training department is incorrect and may cause inadvertent violation of
the 10 CFR 50.59 rule.*
.*
17
Training module 0905-002.14B-5059ZZ-OO, "10 CFR 50.59 Training," contains an "open-
reference" exam that cites an incorrect answer to a posed question. Specifically, question
number 1.19 requires students to give examples of when they must clearly answer "YES" to
the question, "Does the proposal change the facility as described in the SAR [i.e., safety
. analysis report]?" The provided answer states,
"There are NO cases where, categorically, we must answer this question, 'yes.'
Management stresses the fact that we must think smart, and look at everything we
review from.the perspective _of 'can_ this change impact the safe operations of the
plant?"'
The PMs determined that this answer is incorrect and misleading since it chariges the scope
. of the question from the intent of 10 CFR 50.59. The PMsnoted that the 10 CFR 50.59 **
rule is intended to be applied .in two steps. First, determination of whether 10 CFR 50.59
applies [The 10 CFR 50.59 rule applies if the licensee is changing a structure, system, or
component (SSC) or a procedure described in the licensee's final SAR (FSAR) and if the *
FSAR description of the SSC (or procedure) being changed *would be affected by the
change].* The Safety significance of the change is considered following the first
determination, i.e., does the change involve an unreviewed safety question. The licensee's
NA:i>-59 procedure correctly identifies this two step 10 CFR 50.59 process. However, the
response to the question incorrectly suggests that the questions "Does the proposal change the .
facility as* described in the SAR?", and "Can this change impact the safe operation of the
plant?"* are the same, and conrequently may lead to improper determination of 10 CFR 5059 *
- applicability. *
D. *
Implementation ;Review
Hope* Creek*
.
.
.
.
Fourteen completed 10 CFR 50.59 reviews and safety evaluations were reviewed by the
Hope Creek PM. * This represents about 5 % of all the 10 CFR 50.59 reviews and SEs that
were completed between A11gust 1991 and December _1992, as documented by the licensee in
their monthly operating reports .. Additionally~ thirty-two oompleted 10 CFR S0.59
..
. .
applicability reviews were ins~ted, i.e., items for which the licensee determined that 10
CFR 50.59 did not apply and required no SE. The sample was drawn from a list of items
provided py the licensee and generally covered calendar .year 1992.
There were* no safety significant problems that were identified during the inspection, for
reviews and safety evaluations that were completed under 10* CFR 50.59 requirements.
However, there were instances when the licensee did not follow its own procedure. For
example: *
,.
18
NAP-59 paragraph 5.1.1 states in part "The description [of the CTE] shall be specific
and unambiguous. It shall also include a discussion of the applicable design,
operation and regulatory requirements that relate to the proposal." Contrary to this
requirement, there were three 10 CPR 50.59 reviews and SEs that did not contain an
adequate description of the change. For example, design change package (DCP)
4Ec.:.3111 Package 4, does not include a discussion of the design, operation or
regulatory requirements that relate to the DCP. Other reviews and SEs that did not
have an adequate description of the change were: DCPs 4HX-0331 and 4EC-3002,
package 1,_ ...
- * . . NAP'.'"59 paragraph 4. 7 states "The 10 CPR 50.59 Review and Safety EvaluaP.on shall
. address all phases of the change; test or experiment,* including the installation, * *
removal and *testing phases .... " Five of the 10 CFR 50.59 reviews and. safety
evaluations *for DCPs and Temporary Modifications (T-mods) that were reviewed by
the inspector did not contain this required discussion. For example, TMR 91-046
does .not contain any discussion of the installation: or testing phases for this T-mod;
and. 4EC~3226 does not contain a discussion of the testing phase for this DCP. Other.
DCPs and T-mods that did not have an adequ_~te_discussion of all phases of the CTE
were: DCPs 4HX-3342 and 4EC-3182, package 9; .and T-mod TMR 92-020.
Similar in nature to the 10 CPR 50.59. reviews and SEs discussed above, there were*
instances when the licensee did not follow .its own procedure relative to items that were
reviewed for 10 CFR 50.59 applicability, but did not require an SE (i.e., the licensee .
determined that 10 CFR 50~59 was not applicable). For example:.
- The entire description for Revision 4 to procedure HC.IC-DC.ZZ-070 states "The
proposed procedure revision rewrites the prOcedure to bring* it in accordance with the *
. vendor recommended method of testing *and Calibration." This description is not
specific and it does not.contain any discussion of the applicable design, operation and
regulatory requirements. Additionally, Revision 5 to procedure HC.IC-TR.AB-
OOl(Q) does not contain a discussion of the applicable design, operation and
regulatory requirements_.. .
NAP-59 paragraph 5.2.2 states in part "The 10 CFR 50.59 Review shall set forth the .
. SAR sections reviewed, and the basis used in making the determination. A simple
statement of conclusion is not sufficient, nor is* merely restating the question in the
- form of an answer.* *The level*of detail.must be sufficient to allow an independent
reviewer to verify the conclusion, and to permit review by external. organizations
(i.e., the NRC)." Revision 3 to.procedure HC.OP-AP.ZZ-Olll(Q) does not
"UFSAR [i.e., updated FSAR] Section 10.4.4 was reviewed and it is determined that
the SAR is not affected by these modifications." However, the inspector's review
determined that UFSAR Section 10.4.4 does not apply to the system involved in the
DCP. The UFSAR Section referenced should have been 10.4.5.
...
19
The procedure compliance discrepancies noted above are not safety significant by themselves.
However, they indicate that the licensee is being less .critical in this area than is required.
The licensee was previously informed of similar NAP-59 procedure compliance discrepancies
in inspection report 50-272/91-26, 50-311/91-26 and 50-354/91-19. Since NAP-59 was
written to implement the requirements of 10 CFR 50.59, deviations from. the guidance in
NAP-59 could lead to a violation of 10 CFR 50.59..
For example, NAP-59 paragraph 6.2 defines changes in the facility as described in the SAR
as ", .. modifications that affect the design, function or method of performing the function of a
structure, system* or component described in the SAR.* These changes are not limited to
- structures, systems or components specifically described in the SAR, *since changes to
com.ponents not sJ>eCifically described in the SAR can affect the design or operation* of
systems or components that are described in. the SAR. ... Paragraph 6.2.l further states that
changes include " ... Operation with known setpoint drift or degradation of equipment due to
creep; fatigue,* corrosion, or erosion." Notwithstanding these specifications, the following is
an example of a case that was improperly screened from the need to perform* a SE: .
. .
.
- .
.
.
Relati:ve. to DR .HTE 92~230, th~ licensee supported a "u~-as-is" disposition for .
unqualified. gauges in the gland seal portion of the high pressure coolant injection
(HPCI) system.** In the 10 CFR 50.59 review, the licensee states, "The pressure
gauges are not described [in the UFSAR]." The *PM determined that this statement is
incorrect. These gauges* are described in UFSAR Figure 6.3-2 as being within the
- "Q" boundary. Furthermore, *in order to resolve this DR, the licensee changed.the
normal position of the isolation valves for these gauges from open to closed.**
However, UFSAR Figure 6.3~2 clearly shows the isolation valves for. these gauges as
being normally open. Since the licensee changed the facility as described in the SAR,
a 10 CFR 50.59 SE should have been performed.
.
.
.
.
This example constitutes a violation of 10. CFR 50.59(b)(l), which states, in part, that
records of changes to the facility as described in the UFSAR "must include a written safety
evaluation which provides the basis for the determination that the change, test, or experiment
does. not involve an.unreviewed. safety question." In addition, the Technical Specifications
(TS), Section 6.5. 1.6.e. and Section 6.5 .2.4.2.a. requires* the Station Operatlo11s Review
.Committee (SORC) and the Offsite Safety Review Group (OSR), respectively, to review all
safety evaluations completed under the provisions of 10 CFR 50.59. Because the licensee**
. determined that 10 CFR 50.59 did not apply to this change, a safety evaluation was not ..
prepared. Therefore, a: SORC and.QSR review was not performed as required by the TS.
(Section 8.3.1 pertains to the apparent violation.) *
Salem
Twenty-two completed 10 CFR 50.59 reviews and safety evaluations (SEs) were reviewed by
the project manager (PM). This represents.about 5%*of all the. to*CFR 50.59 reviews and
.
.
.
SEs that were completed between July 1991 and December 1992, as documented by the
licensee in their monthly.operating rei)orts .. Additionally, thirty-six completed 10 CFR 50.59
- .
.
. .
.
.
.
20
reviews were inspected ,i.e., items for which the licensee determined that 10 CFR 50.59 did
not apply and required no SE. The sample was drawn from a list of items provided by the
licensee and generally covered calendar year 1992 .
. For reviews and safety evaluations that were completed under 10 CFR 50.59 requirements,
one 10 CFR 50.59 procedure review relative to NC.NA-AP;ZZ-0036(Q), "Control of
Information System and Computer Resources," did not meet the licensee's procedural
requirement that sufficient detail be included to allow a reviewer to independently arrive at
.. the same conclusion.
.
.
.
-
.
.
. .
. The following comments ooncern items that were reviewed under 10 CFR 50.59, but did not
required a 10 CFR 50.59 SE (l.e~; the licensee determined that*10 CFR 50.59 was not
applicable to. these items):
Deficiency Rep0rt 92-024 addressed coating of the 22 RHR Pump Room* Cooler
tubesheet, but was not complete in that it was not a stand-alone document.. The
method of repair' which represented the change being made was not discussed and the
engineering evaluation that addressed the issue was not referenced in the 10 CFR * **
- 50.59 .Review.*
Deficiency Report 92-644 addressed the deficiencies found during testing of valve
1SJ135, but containe<:l errors in the evaluation in that it indicated that the maximum
calculated torque exceeded the continuous. duty torque by 13 % * *It actually exceeded
the torque by 30. 8 %. The incorrect maximum calculated thrust value was used
.
throughout the evaluation. Subsequent to this finding, J>SE&G identified the
following to the PM:
a.
The use of the maximum.CaJ.9ulated thrust in the evaluation is being
reconsidered. .Either the measured or calculated value of thrust '\\Vill be used in
a revision of the deficiency report;* depending on which value is most
conservative.
b.
should have been 30.8%. This will be corrected in the revised deficiency
report.
- . In addition, there were two incorrect 10 CFR *50.59 applicability detefminations; These were
both temporary modifications (T-mods). The T-mods were TMR 92-031 that provided
instructions for disconnecting the normal .power (vital bus lB) and reconnecting a temporary
power supply (vital bus lC) to the. No. 12 Auxiliary BuildiIJ.g supply fan; and TMR 92-043
that installed a blank flange in the serviee water* system. The following pertains:
21
TMR 92-031 was evaluated by the licensee as not changing the Updated Final Safety
Analysis Report (UFSAR). The licensee's 10 CFR 50.59 review stated that the
reasons that this T-mod did not change the UFSAR were: The fan would be
inoperable during the "lB" bus outage and this T-mod would provide temporary
power to make the fan operable; The function of the fan remains unchanged,
therefore, this T-mod is not a change to the SAR. However, the PM found that the
Auxiliary Building Ventilation System and the vital bus connection of the supply fans
are included in Tables 8.3-2 and 8.3-3, and Figure 8.3-4A in the UFSAR. In
addition, TMR 92-031 referenced TMR-006 for a discussion of the separations
requirements which did require a 10 CFR 50.59 Safety Evaluation .
. TMR 92-043 was evaluated as* not changing the UFSAR because the blank flange
served to isolate portions of the service water header that were in service. There is a
manual valve (22SW414) installed in.the system and the blank flange was installed
upstream of that valve to ensure positive (leak tight) isolation. However, this changed
the system as shown on Figure 9.2-lB in the. UFSAR.
.
.
.
.. -
-
.
..
.
. The failure of the licensee to identify the two T-mod~, d~scribed above, as changes to the
UFSAR constitutes a violation*of lO*CFR 50.59(b)(l), which states in part that records of
changes t6 *the facility as described in the UFSAR "must include a written safety evaluation *
which provides the basis for the deterntlnation that the change, test, or experiment d0es not
involve an unreviewed safety question." -In addition, the technical. Specifications {TS),
Section 6.5 .1. 6~e. and* Section 6.5 .2.4.2.a. requires the Station Operations Review
Committee (SORC) and the Offsite Safety Review Group (OSR), respectively, to review all
... safety evaluations completed under the provisions of 10 CFR 50.59. Because the licensee *
determined that* 10 CFR 50.59 did not apply to these changes, a safety evaluation was not
prepared. Therefore, SORC and OSR reviews were riot performed as required by the TS.
(Seetion 8.3.1 pertains to the apparent violation)
8.3.1 Apparent Violation*
The safety significance of the i11dieated apparent violations (as detailed in Section 8.3 D.
relative to Hope Creek and Salem) is low.* However, sinee there .. were several discrepancies *
found in a _relatively small sample size, in aggregate, these findings indicate a weakness in
the licensee's implementation of the 10 CFR 50.59 program, and are considered as examples
of an apparent violation of the requirements of 10 CPR 50.59 (VIO 50-354/93-06-03; VIO .
- 50-272 and 50-21/93-08-03}..
.J
22
9.
LICENSEE EVENT REPORTS (LER), PERIODIC AND SPECIAL REPORTS,
AND OPEN ITEM FOLWW-UP
9 .1
LERs and Reports
PSE&G submitted and reviewed for accuracy and evaluation adequacy the following special
and periodic reports.
.
.
Salem and Hope Creek Monthly Operating Reports for March 1993.
Salem and Hope Creek Annual Personnel Exposure and Monitoring Report for 1992.
Salem Unit 2 Special Report 93-1 regarding the inoperability of radiation monitors
2R45B and 2R45C.
Hope Creek 1992 Annual Environmental Operating Report.
The inspector concluded that the licensee appropriately issued the above reports~ *
Salem LERs
Unit 1
LER 92-26-02 is a supplemental LER which addressed three additional events
(radiation monitoring system ESF actuations) which had the same root cause
(increased containment activity) as the first event. The inspector monitored the
licensee's efforts in this area, and closed this LER.
.
.
LER 93-04 discussed an automatic reactor trip from 100% power due to an equipment
failure (overtemperature differential temperature gain selector switch). The inspector
reviewed this event in NRC Inspection 50-272/93-02, and closed this LER.
. *
LER 93-05 concerned a reactor protection system actuation (reactor/turbine trip
signal) while in Mode 3 (Hot Standby) due to personnel error. The inspector
reviewed this event as described in NRC Inspection 50-272/93-02, and closed this
LER ..
.
.
.
.
.
. . LER 93-06 described a Te:chnical Specification required shutdown due to the loss of
- * one offsite transmission network. The inspector reviewed this event in NRC
Inspection 50-272/93-02, and closed this LER.
LER 93-07 concerned two Technical Specification (TS) 3.0.3 entries more than one
analog rod position indicator (ARPI) per bank became inoperable. Actual. control .rod
positions were subsequently verified for the associated ARPis. For each occasion, TS
. 3.0.3 was exited within one hour. The inspector noted that the licensee's
investigation and corrective actions were appropriate, and closed this LER.
)
..) ,.
- 1
-~~
23
LER 93-08 discussed a design concern associated w~th control air containment
isolation valves. See Section 7.1.A of this report for details. The inspector closed
this LER.
LER 93-09 described a Technical Specification 3.0.3 entry due to a failed boric acid
storage tank level indication. The inspector reviewed this event in NRC Inspection
50-272/93-02, and closed this LER.
Unit 2 .
- .1:-ER 93-05 (See Seetion 2: l.B). This LER is dosed.
Hope Creek.
- *
LER 93-01 (See Section 4.3.2.A). This LER remains open.
9.2
Open Items* ..
. The inspector.reviewed the following previous. inspection items during this inspection.* These***
items are tabulated. below for cross reference purposes .
. Site**
Salem
Report. Section .
. 272&311/91.-16-01 * 4.3.1.B
272&311/92.:01-05 .. 8.1.B
. Hope Creek
354/92-03-04
7.2.A
10.
EXIT INTERVIEWS/MEETINGS **
10.1. Resident Exit Meeting
Closed
CloSed**
Closed
The inspectors met with Mr. C .. Vondra and Mr. R. Hovey and other i>SE&G personnel
periodically.and at the end of the inspection report period to summarize the scope and
fmdings of their iilspection. activities.*
Based on NRC Region I review and discussions with PSE&G, it was determined that this
. report does not contain informa~on subject to 10 CFR 2 restrictions.
.,
..*
J
.*
24
10.2
Specialist Entrance and Exit Meetings
Date(s)
3/29-4/2/93
4;5;;.9193
4/5-7/93
__ , ____ *.
,,,cc<:***
---**-*-*--------
Subject
Inservice
Radiological .
Controls .*
Security*
Inspection
Report No.*
Reporting
Inspector
50-272&311/93-09 * McBrearty
Inspection
50-272&311193-10
Nimitz
50-272&3 ll/93:. ll; . Albert
... 50"'.354/93-07.
- .
.,
j,..
ATTACHMENT 1
50.59 EVALUATIONS AND SCREENED OUT PACKAGES REVIEWED
HOPE CREEK
A...
DESIGN CHANGE PACKAGES.
1.
2.
4EC-3342 .
6..
. 4EC-3182/09
B.
PROCEDURES
. Modified the logic of the E and F Filtration Recirculation
Ventilation System recirculation fans. (Fr()m Mar 92
Hope Creek Monthly Operating Report (MOR))
Added time delay into the closing circuit of the alternate
- infeed breaker in the slow and dead bus transfer schemes
to prevent the alternate infeed from closing* too soon and
to enable the sequencer to reset after a bus transfer.
(From* Apr 92 MOR)
... Repl~ced mechanical snubbers with hydraulic snubbers.
(From Aug 92 MOR)
.
This DCP replaced 2" schedule 80 pipe with schedule 40 .
- * pipe. (From Dec 92 MOR)
This DCP diverted Service Water from the Cooling
Tower.Basin and Cooling Tower Bypass.Line to a*
- manhole* in the yard. * (From Dec 92 MOR) *
This DGP changed the power supply* short circuit .
protection of field wires ori lE instrument loops by
replacing fuses with resistors. . (From Dec 92 MOR)
1..
NCNA-AP.ZZ-007l(Q) * Revision 0 - Describes a .zero defect fuel performanee
2. .
HC.IC-LC.AE-OOOS(Q)
program that will prevent or mitigate the impact of failed
-fuel on plant operations. The procedure was d~veloped
to satisfy the recommendations of INPO SOER 90-02,
"Nuclear Fuel Defects." (From Aug 91 MOR) *
Revision 0- This procedure installs jumpers to bypass
the 20% total feedwater flow interlock to the
recirculation pump speed limiter to preclude an actual
runback from occurring during the transmitter
calibration. (From Mar 92 MOR)
~\\
Attachment 1
2
3.
HC.SA-AP.ZZ-0052(Q)
Revision 7 - Provides guidance for the station
departments involved in ensuring that water chemistry
parameters are maintained in accordance with the
appropriate vendor and industry guidelines. (From Aug
92 MOR)
C. TEMPORARY MODIFICATIONS CT-mods)
- L * ** 91-046
2.92-020
Modified the circuit* for the measurement of river water
temperature. Three of*four temperature detectors are
currently providing unreliable.readings. (From Sep 91
MOR).
- Installed Control Air tubing between a pressure control
valve in the Gaseous Rad waste system* and its associated
instrumentation. (From Aug 92 MOR)
.
.
D.
. UFSAR CHANGES AND DEFICIENCY REPORTS
1.
6.2.4.4.3
2.
.HMD 92-009
3.
RTE 92-010 . *
E.
SCREENED OUT ITEMS
DCPs *
. 1.
4HC-339, pkg 1 *
2.
3.
Assigns total observed leakage through the outboard
MSIV only . when leak rate .testing is performed between
the MSIV and the MSSV. (From Jan/Feb 92 MOR).
.
.
.
.
Addresses a through:..wal1 leak: on a Station Service Water
. . instrument line~ *(From Jail/Feb 92 MOR)*
- Addresses the installation of schedule 40 pipe instead of
schedule 80 pipe at several SSW 1" and l.5" root valve
lines. (From Oct 92 MOR)
. Replaces existing NaOCl storage tanks with Durakane
411.iined tanks for the Circulating Water
Hypochlorination System.
Refurbishes two SOOKV Type SFA gas circuit breakers.
Modifies the Reactor Core Isolation Cooling (RCIC)
system flow *controller Setpoint raise aiid lower circuit.
_)
Attachment 1
3
4.
4HE-0002,pkg 1
5.
Custom fits a new hinge and disc to the seat inside the
inboard feedwater containment isolation valve lAEV-
003.
Installs carbon steel angles along the top and bottom of
the FRVS straightening vanes.
Procedures
1.
HC.CH-GP.ZZ-0006(Q)
Revision 0- Provides guidance to the Chemistry
Department in the event of a SCRAM.
.
..
2.
HC.CH-EO.SH-0004(Q)
Revision 6 - Deletes steps that have been integrated into
HC.CH-EO.Sff*0005(Q).
3.
HC.IC-CC.AB-04l(Q)
Revision 15 - Incorporates new setpoint values for
channels A, C, Band J.
4. *
. HC.IC-DC.zz.:..070 *
Revision 4 - Rewrites the procedure to bringitin .*
- accordance with the vendor recommended method of
testing and calibration.
5.
HC.IC-LC.FC-OOl(Q)
Revision 2 - Incorporates two technical changes.
6.
HC.IC-SC.BH-002.(Q) *
Revision 0 - Created to test and calibrate the standby
liquid control system storage tank level transmitters ..
7.
9.
HC.IC-TR.AB-OOl(Q)
Revision_5 - Challges the total time response acceptance
cri~ria._
HC.MD-AP.ZZ-,0014(Q)
Revision _7 - Changed format to comply with guidelines
of NC.NA-AP.ZZ-0032(Q);
HC.MD-GP.ZZ-002l(Q).
Revision 3 - Various changes were accomplished by this
revision.
10.
HC.MD-PM.KJ-005(Q)
lL
-HC.MD-ST.KE-OOl(Q)
Revision 5 - Revised procedure for biennial review. *
Revision 9 - Incorporated changes that were required by
the implementation of DCP 4EC-1043.
12.
.HC. OP-AP .ZZ-01 ll(Q)
Revision 3 - Revised procedure for biennial review.
- ,
,, .
..
./ *
Attachment 1
4
T-mods
1.92-005
Provided power feed to UPS load disconneet switch.
2.92-015
Installs pressure and .flow transmitters* to components
1AEPDT-N002A/N002B and 1APT-3686A/3686B in
support of a Unit Heat Rate Evaluation.
3.
Allows use of polar crane auxiliary hoist while the main
hoist is de-energized for maintenance .
4.
92-033*
. . Installs a temporary transformer in panel lB~C-156.
5.
- 92-034
Abandons 11 LPRM cables and subsequent temporary
routing of additional cables.
6;92-035.
- Addresses the use of a silver bronze. pressure sensing
tube in lieu of stainless steel.
DRs
L
HTE-92-122
Dispositions the condition of the backwash outlet flanges
on the C Ser\\rice Water strainer lC-F-509.
2.
HTE-92-124
Addresses the damaged concrete lining. on Service Water
discharge line EA-24"-STJ'."002 .
3.
HTE-92-148
. Addresses the presence of material anomalies on the
Reactor Feed .pump anti...:vortex darri.
4.
H".rE-92-230
Supports use-as-is disposition for pressure gauges in the
gland seal portion of the High Pressure Coolant Injection
. (HPCn system.*.
5.
HTE-92-232
Supports use-as-is disposition for Control Rod Drive
.(CRD) 3015 not meeting the acceptance criteria for
friction testing identified*in procedure RC.OP-FT.BF-
0004(Q), Revision 2*.
6.
HMD-92-159
Addresses the repair of the seating surface of the disc to
testable swing check valve lBCV-033.
7.
HMD-92-176
Restored HPCl turbine shaft gland seal area to acceptable
surface finish .
I
.,.!
Attachment 1
5
8. *
HMD-92-250
9.
HIC-92-202
SALEM
Repaired crack in FRVS flow straightening vane by
drilling a hole at the end of the crack.
Repaired LPRM detector cable outer jacket tear.
A. DESIGN CHANGE PACKAGES
. . 1. . . lEC-3205
-2 ..
. lEC-3195, Pkg 1
3.
.. lEC-3186, Pkg. 1
4 ..
lSC-2267 Pkg. 2 ..
5.
. lEC-3162 Pkg 1
6.
2EC-3110 Pkg l
7.
2EC-3087 Pkg 1
8.
2SC~2267 Pkg 1 .
B. PROCEDURES.
1.
NC.NA-AP .ZZ-0036 (Q)
RVLIS Refueling.* (From Unit 1 Dec .. 92 MOR)
Lube Oil Storage Facility Revitalization Project FC-0001
Units l and 2. (I<rom Unit 1 Oct. 92 MOR)
Steam Generator Feed Pump High Discharge Pressure
Trip. (Froin Unit 1 July 92 MOR) .
SEC ~ontainment Spray Actuation: ( From *unit * 1 *May.**
- 92*MOR)
.
Installation of Turbine Auto Stop Oil Systein Filters .
. (From Unit t April 92 MOR) .
Allowable Value and Setpoint for Containment Hi-Hi
Pressure. (From Unit 2 March. 92 MOR)
RHR Monitoring During Mid-Loop Operations. (From
Unit 2 Feb. 92 MOR)
.
. .
Safeguards Equipment Cabinet Control Electronics Unit
Replacement, Revision 1. (From Unit 2 Jan. 92 MOR)
Control of Iri.formation System aild Computer Resources.
.. (From Unit 1 Nov. 92 MOR)
2;
NC.NC-AP .ZZ-0013 (Q)
Control of Temporary Mods, Revision 1. (From Unit l
April 92 MOR)
.
3.
Sl.OP.AB.ROD-0004 (Q) * Rod position Indicator Failure. (From Unit l*fan. 92
MOR)
Attachment 1
6
4.
Sl.OP-SO.RC-0005 (Q)
Draining the RCS, Revision 2. (From Unit 1 July 92
MOR)
5.
TSI.OP-SO.AF-0001 (Q)
Aux Feed Operation. (From Unit 1 June 92 MOR)
C.
1.92-057
2..
TMR 92-015
3. .
TMR 92-037
Installation of Temporary Air Dryer. (From Unit 1 Sept.
92 MOR)
Removing/Returning 2A 125VDC Bus FroiniTo Service. * *
(From Unit 2 Feb. 92 MOR)
Monitoring Temperatures Inside Pressurizer Enclosure.
From Unit 1 June 92 MOR)
- .. .
.
-
.
D .. SAFETY EVALUATIONS. DEFICIBNCIES. SAR CHANGES. AND TECH SPEC *
INTERPRETATIONS
1.
S-O-AF-MSE*0812
- 2.
s/E wo 920411111 *
3.
SMD-92-735 .
4.
DR SMD 92-262
5.
SCN# 92-42
6.
TSI#3.7.1.1
E. SCREENED OUT ITEMS
1.
Potential Cavitation of the Auxiliary Feedwater Pumps.
(Safety Evaluation) (From Unit 1 Sept 92 MOR)
Breaching a Penetration Seal. (Safety Evaluation) (From
1,Jnit 2 May 92 MOR)
.
.
.
12 Service Water* Return From 12 CCHX Wall Thinning ..
(DR) . (From Unit 1 Sept 92 MOR)
.
Primary Water Storage Tank. (DR) (From Unit 1 May
92 MOR)
Updating. SAR - Control Room Habitability. (SAR
change) *(From July 92 MOR).
Operation of. Salem Units 1 and 2 with Reduced Main
Steam Safety Valve Flows* *(Tech Spec Interpretation)
(From Unit* 1 July *92 MOR)
a.
92.:011,. Removal ~f Reverse. Power Relay (Salem 2)~ .. **
Attachment 1
7
b.92-031, Jumpers and Lifted Leads to Supply Temporary Power
during lB
Bus Outage.
c.92-017, Clamp on Orifice and Seal of 13MS200.
d.92-026, Jumpers and Lifted Leads to Supply Temporary Power to #12 Spent
Fuel Pool pump during Bus lB Outage
.Fuel Pool pump.during Bus lC Outage
f.
- . 92-,029,* Removal of_MaJ1ipulator Crane West Trolley Limit of.Travel Bumper
g.92-043, Isolation of 22 Service Water Chiller Header
2.
- Deficiency Reports
.
. .
. . -
. .
.
.
a.
. * SMD 93-012, CFCU Inlet/Outlet Flange Repair.
b.
SMD 92~ 708, Unit 2 Reactor Trip Breaker Roller Assembly Out-of- .
Specification.
c.
SMD 92-532, * ISI - 13 Steam Generator Object Removal.
d.
SMD 92-664, Evaluation of l 1SJ40 Closing Thrust .
.
.
. .
e.
SMD 92-644, Measured thrust for 1SJ135 Higher than Maximum .
f.
SMD 92-615, Thrust for 1CV116 Lower than Required
g.
SMD 92-182, Airlock Door Hinge Pin Indications
h.
- sMD 92,.024, 22RHR Pump Room Cooler Tubesheet CorrosiOn
i.
SMD 92-546, Indication of Pipe to Valve Weld (12MS167)
j.
. SMD 92-362, No Limiter Plate oil 1CS16" Torque Switch.
k.
SMD 92-261; Spring Can Hanger, Unable to Adjust
3.
Procedures
a.
SC;DE-AP.ZZ-0055 (Q), Detailed Procedure for EiC Monitoring Program.
Attachment 1
8
b.
SC.RC-TI.ZZ-0190 (Q), Software Control
c.
S2.0P-SO.PZR-0003 (Q), Pressurizer Relief Tank Operation
d.
S2.RE-RA.ZZ-0008 (Q), Post Refueling Initial Criticality
e.
2-11-8.3.4, Draining the Reactor Refueling Cavity
f.
.* lIC,.14.3.002, Response Time Testing
g.
. si.op:.so.WG-0008, Discharge of No~ 11 Waste Decay Tank to the Plant
Vent
h.
- 11-153.2, Containment Entry
i.*
. SC.MD-GP.ZZ-0022, Torquing of Fasteners
j .. *
Sl.IC-CC.RM-0064 (Q), Plant Vent Radiation Monitor Channel Calibration
Procedures * *
k.
2IC-4.5.060, Calibration of Radiation Monitors
1.
S2.RE~RA.22-0002 (Q), Inverse Count Rate Ratio Puring Control Rod
. Withdrawal
4.
Design Change Packages
a.
- 2EC,.3154, Change to the AMSAC Diagnostic .Software
b.
2EC-3150/1, Change the Circuit Breakers for the Vacuum Pumps
c.
2EC-3137, Change the Circulating Water Intake Screen Wash Strainer Motor
Circuit Breaker
- d.
lEC-3200, Install a Hot Water Heater in the Turbine Building
- .-
.-.
.
.
.
.
- .
. '
e.
lSC-2269', Install Cable.and Raceways to Support Salem Eleetrical
Distribution Project
f.
2EC-3085/1, Changes Protective Relays for Main Generator Flashover
Protection