IR 05000317/1996010

Insp Repts 50-317/96-10 & 50-318/96-10 on 961130-970111. Violations Noted.Major Areas Inspected:Operations, Maintenance,Engineering & Plant Support

ML20135E554
Person / Time
Site:	Calvert Cliffs
Issue date:	02/27/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20135E550	List: ML20135E547 ML20135E551 ML20135E554
References
50-317-96-10, 50-318-96-10, NUDOCS 9703070136
Download: ML20135E554 (25)

Text

__ _ _ _ _ _ _ __. _

,. .. ,

.

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

License No DPR 53/DPR 69 Report No /96-10;50-318/96-10 Licensee: Baltimore Gas and Electric Company -

Post Office Box 1475 Baltimore, Maryland 21203 Facility: Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Location: Lusby, Maryland l

Dates: November 31,1996 through January 11,1997 '

inspectors: J. Scott Stewart, Senior Resident inspector H. Kirke Lathrop, Resident inspector Fred L. Bower lil, Resident inspector i Tim Kobetz, Senior Engineer, Spent Fuel Project Office, NRR i l

l

'

Approved by: Lawrence T. Doerflein, Chief i Projects Branch 1 Division of Reactor Projects

)

'

l

!

I

l

!

,

l

!

,

l l

l 9703070136 970227 l l

PDR ADOCK 05000317 i

'

! G PDR

,

c

.. .. . --. . . . - . -. .. ._ _-.

.

. .

EXECUTIVE SUMMARY 4 Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Inspection Report Nos. 50-317/96-10 and 50 318/96-10

This integrated inspection report includes aspects of BGE operations, maintenance, I l engineering, and plant support. The report covers a seven week period of resident ;

inspection and includes the results of an announced inspection by a spent fuel project i specialist Plant Operations e The inspectors identified that during fuel handling in the spent fuel pool, had a fuel .

handling event occurred, some of the radioactive material released may not have i i

passed through charcoal adsorbers prior to release to the environment. Also, the i inspectors found that BGE activities during the fuel handling were deficient in that pre-evolution briefings had not been conducted with control room personnel and a t controlled copy of the fuel handling procedure was not in the control roo t

'

l I * The inspectors found that operations personnel exhibited poor work practice and did l l not have a questioning attitude during venting of a chemical and volume control system filter. Control of the evolution using the safety tagging procedure was ineffective and contributed to a plant auxiliary operator mispositioning a vent valve l to an on-line purification ion exchanger resulting in a lowering of volume control tank level.

l l Maintenance e During fuel handling operations in the spent fuel pool, maintenance was conducted .

l

'

in the auxiliary building that was poorly planned in that an unmonitored path between the fuel handling area and the environment was intermittently created and compensatory actions to ensure that radioactive material would be contained were not prescribe ;

e The inspectors concluded that emergency diesel generator realignment maintenance l activities were very well planned and effectively implemented. Advance planning i for the maintenance was extensive and included walkdowns of the job with a l vendor technical representative, prefabrication of speciallifting and alignment tools, '

l dry runs on a spare EDG, a detailed risk assessment, and good coordinGion between maintenance and engineerin Enaineerina e The inspectors determined that BGE engineers conducted a thorough and rigorous examination of a piping defect and evaluated potential safety consequences of continued operation. The technical content of the engineering evaluation was excellent. The Plant Operating Safety Review Committee demonstrated a strong safety perspective and questioning attitude in their review of the potential nuclear safety consequences of the leak and the engineering assumptions used to justify operability.

i

.

Il

- - -. . - . . - . ._- - .- . .-

-

l

l t

• The inspectors deterrnined a BGE procedure did not provide adequate guidance to ;

ensure that a dry shielded canister would not be over pressurized during unloading i

'

operations. BGE engineering did not develop supporting documentation to !

determine the required re-flood rate, and select and test equipment required for canister re-flood operation .

• Due to recent industry events and as a voluntary initiative, BGE developed a project

. plan to perform a review of the Updated Final Safety Analysis Report to assure that the report accurately reflected the current plant design and operating practice * The inspectors found that BGE corrective actions to ensure that the as-built versus as-designed configurations for electrical separation barriers were inadequate. The specific weakness was the challenge to electrical separation resulting from damaged or missing marinite separation barriers. The inspector also found that some design documents did not reflect the as-built configuration Plant Suooort i

• During the conduct of operator rounds, the inspectors considered the actions of an auxiliary plant operator to be a very good demonstration of sound ALARA and radiation controls practices, j

l i

l l

l

i l

l l

!

i a

III l

l

._- . .. . -. .- - . . . - . .- - . - - - -

'

l

.

l TABLE OF CONTENTS

EX ECUTIV E SU M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii TAB LE O F CO NTE NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv

! Summ ary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l

+

'

l 1. O p e ratio n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ;

j 01 Conduct of Oper ations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 General Comments (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.2 Spent Fuel Handling Operations ........................ 1 01.3 Valve Mispositioning in the Chemical and Volume Control ,

System.......................................... 4 l

07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 l

07.1 (Closed) LER 5 0-318/9 6004 . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 J i

i 11. M a in t e n a nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

.

M1 Conduct of Maintenance .................................. 6 ,

l M1.1 Routine Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . 6 i

M1.2 2A Emergency Diesel Generator Realignment . . . . . . . . . . . . . . . 6 M1.3 Routine Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . 7 l M1.4 Reactor Trip Circuit Breakers .. ....................... 7 l 111. Engineering ................................................... 8

El Conduct of Engineering . ................................. 8 E1.1 General Comments ................................. 8 E1.2 Degradation of the Safety injection Pump Recirculation Piping . . . 8 E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . 9

, E3.1 Deficient Procedure for Unloading a NUHOMS Cask . . . . . . . . . . 9 l E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . 11 l E Update Final Safety Analysis Report (UFSAR) Review Project .. 11 l

E7.2 (Closed) Unresolved item 50-317&318/96-08-01 Cable i Separation Issues ................................. 12 l E8 Miscellaneous Engineering issues (92902) . . . . . . . . . . . . . . . . . . . . . 15 l E (Closed) Unresolved item 50-317&318/93-25-01 . . . . . . . . . . . 15 l E8.2 (Closed) Unresolved item 50-317 and 318/96-04-01 Inoperable l

LPSI pump circuit breaker due to bent trip-paddle problem. . . . . 16 I V. Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 R4 Staff Knowledge and Performance in Radiation Protection and C h e mi st ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

.

.

V. M an a geme nt M e eting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 l

,

X1 Exit Me eting Sum m ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 iv

!

_ _ . _ _ _ .. ._. __ .___. . _ . . _ . . . _ _ _ _ . . -._ .._._- _.-_..-_ _ _ __ _ _ _.__._ _. _ .. ;

i Table of Contents (cont'd) ,

t

'

ATTACHMENT Attachment 1 - Partial List of Persons Contacted

-

Inspection Procedures Used  !

-

Items Opened, Closed, and Discussed

'

-

List of Acronyms Used

I

,

.

$

)

f I

'

e I

.

-

t

!  !

!

i i

[

l i

!

!

,

e i i  !

l V l

,

! \

l l

.

l

_ , ._,,., . . _- _ _ _ . - . ._ _ _ _

,

i Report Details Summarv of Plant Status

Unit 1 remained at full power during the inspection perio l Unit 2 operated between 95 and 100 percent power during the inspection period. (See I 01.1 General Comments)

1. Operations 01 Conduct of Operations ' l 01.1 General Comments (71707)

Overall plant operations were conducted with a proper focus on continued nuclear safety. A deficiency in the cooling system for the Unit 2 main transformer required that reactor power be reduced 20 megawatts electric at 50 degrees fahrenheit (*F)

ambient temperature and 20 additional megawatts for each additional 10 degrees of outdoor temperature. When temperature dropped, power could be restored. These power swings were frequently conducted during the inspection period without complication. BGE planned to repair the transformer during the upcoming Unit 2 refueling outage. As a result, Unit 2 operated between 95 and 100 percent power throughout the inspection perio .2 Spent Fuel Handlina Ooerations Insoection Scope During a plant walkdown, the inspectors observed that air from the fuel handling area was flowing into the auxiliary building while spent fuel was being moved in the spent fuel pool. The circumstances of the observation were reviewed by the inspector Observations and Findinas

)

On January 10, during a routine plant walkdown, the inspectors observed that  !

spent reactor fuel was being moved in the spent fuel pool to prepare for the upcoming Unit 2 refueling outage. Simultaneously, the inspectors found that a door i from the spent fuel pool area into the auxiliary building stairwell was ajar, with !

indications that air was flowing out of the spent fuel handling area into the auxiliary l building through the doorwa The inspector noted that the basis for Technical Specification 3/4.9.12, " Spent Fuel j Pool Ventilation System," stated that "The limitations on the spent fuel pool

' Topical headings such as 01, M1, etc., are used in accordance with the NRC standardized reactor inspection report outline found in MC 0610. Individual reports are not expected to address all outline topic ,

i

_ - -

.. . . - . . - .. - -- - ~ _ - -.-- -- . -..- -. .-

.

.. j i

J,

{

-

'

'2' .

'

ventilation system ensure that all radioactive material from an irradiated fuel l assembly will be filtered through the HEPA filters and charcoal adsorber prior to f

,

discharge to the atmosphere." The auxiliary building ventilation system did not j

. include charcoal adsorbers. The ventilation limitations were repeated in the Calvert }

( Cliffs UFSAR Section 9.8, " Spent Fuel Pool Ventilation" and Section 14, " Accident l Analysis". The inspector also noted that the UFSAR, Section 9.8, stated that the i

, - spent fuel pool ventilation system was capable of maintaining a negative pressure j with respect to ambient and surrounding areas of the auxiliary buildin l

.

'

The inspectors questioned operations department supervisors as to the adequacy of l the fuel handling area ventilation during the spent fuel movement. Plant operators j secured the fuel movements until the open door was repaired and shut. On j

> additional questioning from the inspector, BGE initiated an investigation and *

identified that work was being conducted oa the auxiliary building ventilation l '

system and that two o.* the three supply fans for the system were out of servic l Then, a BGE engineering review was conducted which idontified that because o' i.he  !

maintenance, the auxiliary building ventilation system was out of balance and j l confirmed that air flow was being directed from the fuel handling area into the ,

) auxiliary building. Since fuel handling was in progress when the fuel handling l

,

ventilation system was out of balance, on January 10, BGE made a report to the  :

,

.

NRC in accordance with 10 CFR 50.72.b.1.ii(b), for a condition outside the design  !

basis of the plan l

'

l i The inspectors were later informed that maintenance was in progress to replace the I 11 and 12 auxiliary building supply fan discharge gravity dampers. The supply fans

, nad been stopped when work started on January 6. The work was completed on .

j i January 10. Fuel handling operations were conducted on January 8,9, and 1 During the maintenance,' supply fan dampers were sequentially removed from the system and replaced. The work order specified that any plant condition was adequate to support the maintenance and the system engineer stated that fuel handling operations were not considered when planning the work The work conducted on the ventilation supply resulted in a breach of the ducting when the supply dampers were removed. The breach provided a path to the outside environment that was not monitored for radioactive release. The inspectors were informed that during the fuel movements, fuel handling and auxiliary building exhaust ventilation ran continually and air flow remained from the outside into the building through the breach and back to the environment through mo.nitored flow path The inspectors also found that the fuel movements were conducted by two contractor personnel using fuel handling procedure FH-340. Both contractors had completed fuel handling qualifications administered by BGE. The fuel handling area ventilation system exhaust was aligned for the fuel movement with a charcoal adsorber in service. A prepared set of fuel movements were specified and completion of the moves was documented on the appropriate form. The fuel movements involved two procedures; FH-340, " Component Movement in the i

i

!

.- _ , . . _ , -

_ . . _ _ , _ _ _ _ _ _ __ _ . - _ _ ,- - ,i

- . - - - . -- - - - -.- --

t 4 P l

.

Auxiliary Building," and Operating instruction OI-25A, " Spent Fuel Handling i Machine." '

Two procedure compliance discrepancies were identified by the inspectors. FH-340 ,

step 2.1.B, stated that the controlled copy of the procedure shall be maintained by '

the control room when core components were being moved in the spent fuel poo *

The inspectors found that no copy of the procedure was in the control room and the controlled copy was maintained by the fuel management group in the engineering departmen FH-340, Attachment FH-340-1, " Spent Fuel Pool Component Moves," stated that the operators and the control room supervisor will be briefed as part of the pre-evolution brief. The briefing instruction specified that communications during fuel moves, ventilation lineups, and actions to be taken if a radiation monitor alarms or if a fuel handling incident occurred would be included in the briefing. The inspectors

'

found that the briefing did not include any licensed or control room personnel, but !

instead was conducted by the fuel management group with the contractors. The control room operators were aware that fuel handling was planned and had placed charcoal adsorbers in service per Operating Instruction 22D and were informed by ;

the contractors when fuel handling had started and stopped. However, control room personnel had not reviewed the fuel handling precautions and procedures and had not briefed or prepared for actions in event of a fuel handling problem. No evacuation routes for personnel in the auxiliary building had been planned or discussed with applicable work groups. The work group assigned the auxiliary i building ventilation work were working in an area only accessible through the spent J fuel handling area and had not been informed that fuel moves were in progress or informed of their responsibilities in event of a fuel handling inciden CFR 50, Appendix B, Criterion V stated, " Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, and drawings". The failure of BGE to estab!ish conditions that assure that had a fuel handling event occurred, all of the released gases would be filtered through charcoal adsorbers and the failure to follow the briefing and procedure control guidance in procedure FH-340, were in the aggregate, a violation of NRC requirements. (VIO 50-317&318/96010-01) l l

When informed of the inspector findings, BGE management initiated a review of spent fuel pool operations, including fuel handling evolutions and ventilation adequacy in different operating modes. Also, a review of ventilation adequacy in

'

other areas of the auxiliary building was initiated. BGE management also informed i the inspectors that the management expectation was for procedure compliance in plant operations and that this expectation was not met in the fuel handling

operations observed by the inspector ;

i i

-

l l

I

'

i l Conclusions '

The inspectors found that BGE activities during fuel handling in the spent fuel pool were deficient in that pre-evolution briefings had not been conducted with control room personnel and a controlled copy of the fuel handling procudure was not in the ,

, control room. Also, due to a procedure inadequacy, had a fuel handling event !

I occurred, some of the radioactive material released may not have passed through charcoal adsorbers prior to release to the environmen .

During fuel handling in the spent fuel pool, maintenance was conducted in the l auxiliary building that was poorly planned in that an unmonitored path between the fuel handling area and the environment was intermittently created and compensatory actions to ensure that radioactive material would be contained were not prescribe BGE management responded promptly to the inspector findings by declaring the fuel i pool ventilation system out of service until the various modes of operation could be l evaluated. BGE management also initiated a review of fuel nandling operations and !

auxiliary building ventilatio l 01.3 Valvo Micoositionina in the Chemical and Volume Control System ' Insoection Scope The inspectors rev'ewed a valve mispositioning occurrence at Calvert Cliffs Unit l l Obs3rvations and Findinas On January 15, a radiation protection technician requested control room operators i open a vent valve for the 22 purification filter in the chemical and volume control system. Prior to the request, the filter had clogged, a work order had been prepared, and a tagout had been issued to isolate the filter. Control room operators told the inspector that the technician made the request so that a radiation protection l survey could be complete An auxiliary building operator was instructed by control room personnel to open the purification filter vent valve, 2-CVC-122. Instead, the operator opened 22-purification ion exchanger vent valve,2-CVC-140. The operat.or did not read the valve label resulting in the wrong valve being operate Since the ion exchanger was in service at the time, partially depressurized reactor

,

coolant passed from the ion exchanger through the vent valve ari into the waste l gas and miscellaneous waste processing system. In response, the volume control

'

tank level dropped approximately one inch (60 gallons) when a control operator suspected a problem and requested the auxiliary building operator to shut the vent valve. On returning to the valve, the auxiliary building operator observed that the wrong valve had been manipulated and informed control room personne . . _ .- _ ..

,

i t

in response, control room personnel requested radiation controls personnel to conduct surveys of the waste gas header to determine if radiation level changes had l occurred in the auxiliary building and to verify the extent of contamination of the i header by reactor coolant. No unusual radiation levels were detected. The l operating crew reported that no unusual radiation alarms had occurred during the event and a review was conducted which verified that there had been no measurable release of radioactive material through the plant vent. Subsequently, a drain valve on the waste gas header was opened to drain the coolant from the i normally dry waste gas header, but no liquid passed through the drain valv Further BGE investigation identified that the 60 gallons of coolant had entered the 11 miscellaneous waste receiver tank through an open vent line on the tank. The l operations department determined the occurrence to be a significant event near I l miss and an issue report was written. As followup action, BGE management j ( reviewed safe plant operations and the need to complete self-verification prior to l l operating plant equipment with all operating personnel, including the operator who l had mispositioned the ion exchanger vent valve.

l l The inspectors became aware of the mispositioning in discussions with control room personnel during a control room walkdown. The inspectors reviewed the event and found that a safety tagout and work package to support replacing the filter had l been issued on January 15. Although not tagged, a note on the tagout stated that l the vent valve would be operated by the work group after a concrete shield block l was removed by plant mechanics to access the valve. Otherwise the valve was I inaccessibl When the control room was contacted to open the vent valve, a safety tagging technician was contacted to authorize opening the valve, and permission was given.

l

'

Control of the shield plug was apparently not raised in this discussion and direction was given to the auxiliary operator without mentioning that shield plug removal was necessary. Control room personnel did not review either the safety tagout nor the work package prior to giving direction to open the valve and no written procedure l directing the venting was available in the control room. On questioning, BGE ,

'

management stated that the evolution was controlled by the Calvert Cliffs safety 1 l tagging procedure.

l The inspectors were informed that the involved auxiliary building operator had not l previously positioned valves in either the purification or the purification filter

! systems and was unsure of the proper valve location. The inspectors found that control room personnel did not have a questioning attitude when the decision to vent the filter was made because the removal of the shield plug to access the valve i was not considered and no written procedure step was sought directing the actio T,ie mechanical work group vented the filter on January 15 and completed the work

. on January 16.

!

, Conclusions

'

The inspectors found that operations personnel exhibited poor work practice and did not have a questioning attitude during a venting evolution. Control of the evolution

!

i l

.

using the safety tagging procedure was ineffective and contributed to a plant operator mispositioning a vent valve to an on-line purification ion exchange Quality Assurance in Operations 07.1 (Closed) LER 50-318/96004: Missed Surveillance Due to Less than Adequate Technical Review of Surveillance Test Procedure. BGE identified that a Unit 2 surveillance test procedure for verification of containment closure did not verify a closed containment as specified by Calvert Cliffs Technical Specification 3.9. Specifically, a steam generator sample drain valve was not verified shut when the steam generator was open to the containment atmosphere during core alteration The discrepancy was identified by BGE during a review of the test procedure. The inspectors reviewed the LER and verified completion of the long term corrective actions including a technical review of containment closure for fuel movement and i that the applicable procedure had been appropriately revised. The issue was !

considered a Non-Cited Violation, consistent with Section Vll.B.1 of NUREG 1600, 1 NRC Enforcement Policy. The LER is closed, ll. Maintenance M1 Conduct of Maintenance M 1.1 Routine Maintenance Observations Using Inspection Procedure 62707, the inspectors observed the conduct of maintenance and surveillance testing on systems and components important to safety. The inspectors also reviewed selected maintenance activities to assure that the work was performed safely and in accordance with proper procedures. The inspectors noted that an appropriate level of supervisory attention was given to the work depending on its priority and difficulty. Maintenance activities reviewed included:

MO2199601746 21 Charging Pump Suction & Discharge Valve Replacement MO2199604378 23 Saltwater Pump Volute Cleaning Due to Low Flow MO2199304705 Replace 21 AFW Pump Turbine Stop Valve Position Switch M01199603843 EQ Replacement of Solenoid Valve on 12 Component Cooling Heat Exchanger M1.2 2A Emeraency Diesel Generator Realianment The inspectors reviewed and observed selected portions of scheduled corrective maintenance conducted to relieve crankshaft strain on the 2A emergency diesel generator (EDG) during plant operation. NRC inspection report 50-317 & 318/96-06 documented the inspectors' previous review of maintenance act;vities that identified crankshaft strain of -0.00275 inches. At that time, BGE was attempting to meet an acceptance criteria of +0.001 to -0.001 inches. In November 1996, the EDG vendor provided BGE information that the crankshaft strain acceptance criterion had been revised to 0.000 to + 0.001 inches.

!

l l

. - . . - - . - .- -. -. - - - - .. . . - . _ ~ .

.

i

.

7  ;

Relief of the strain was accomplished by shimming the generator to obtain an improved alignment to the engine. The final crankshaft strain of -0.00025 inches was determined to be acceptable by BGE and the manufacturer. However, BGE  ;

would monitor for changes in the strain during periodic surveillances. Also, an issue report has been entered into the corrective action system to further evaluate the discrepancy between the as-left crankshaft strain and the November vendor lette The inspectors concluded that the 2A EDG realignment maintenance activities were . I very well planned and effectively implemented. Advance planning for the maintenance was extensive and included walkdowns of the job with a vendor '

technical representative, prefabrication of special lifting and alignment tools, dry = ,

runs on a spare EDG; a detailed risk assessment, and good coordination between '

maintenance and engineering personnel. Although the issue of the conflict between the as-left strain and the November 1996 vendor letter required an engineering ,

review prior to returning the EDG to service, the issue was resolved and the engine j was returned to service without challenging the allowed outage time. The *

maintenance work order was effectively implemented with strong support provided l by vendor technical representatives and system engineering personne .!

M1.3 Routine Surveillance Observations -

The inspectors witnessed / reviewed selected surveillance tests to determine whether approved procedures were in use, details were adequate, test instrumentation was ,

properly calibrated and used, technical specifications were satisfied, testing was

.

i performed by qualified personnel, and test results satisfied acceptance criteria or were properly dispositione ;

The surveillance testing was performed safely and in accordance with proper l procedures. The inspectors noted that an appropriate level of supervisory attention was given to the testing depending on its sensitivity and difficulty. Surveillance testing activities that were reviewed are listed below:

!

STP M-200-2 Reactor Trip Circuit Breaker Functional Test i S.7 v-70-2 Monthly Test of "A" Train Containment Cooling Units, lodine Removal Units, and Penetration Room Exhaust Filter l STP O-65B-2 21 Service Water Subsystem Valve Quarterly Operability Test  :

I M1.4 Reactor Trio Circuit Breakers 2  ;

! The inspectors informed the licensee of a condition identified at other nuclear ,

l facilities that involved the testing of the undervoltage and shunt trip electricallogic paths for the Reactor Trip Circuit Breakers (TCBs). Either condition would cause the l TCBs to trip, however, the test that was historically performed did not test each of the trip devices independently. The inspectors and BGE personnel reviewed j Technical Specification 4.3.1.1.1, and applicable diagrams and surveillance test

'

procedures and concluded that the same conditions did not exist at Calvert Cliffs.

The Calvert Cliffs technical specification did not have a specific surveillance

. requirement (similar to the other nuclear facility) to verify the independent i

h

.- -. . . . - _. _-- . . .. _ -- - . .-

,

,

.

B ,

operability of the undervoltage and shunt trips. However, the testing performed by BGE was complete. The reactor protective system matrix functional test performed !

quarterly verified the logic matrices and the matrix relays. This procedure included .

verification of the change of state of installed indicating lights provided in both the ?

shunt and undervoltage trip paths. The reactor TCB functional test was performed monthly to independently verify the response time of both the undervoltage and shunt trip devices and verify operat!~ of the TCBs. The inspectors concluded that :

the TCB testing performed by BGE was appropriat ;

lit. Enaineerina -

l E1 Conduct of Engineering (37550) .

E General Comments  :

On December 27,1996, during the process of reviewing design specifications in preparation for purchasing new Dry Shielded Canisters (DSCs) for the Independent Spent Fuel Storage Installation (ISFSI), BGE identified conflicting information concerning the weight of fuel assemblies at Calvert Cliffs. A 1992 fuels vendor letter identified the bounding maximum weight as 1300 pounds versus a 1995 letter 1 that identified the bounding weight as 1327 pounds. ISFSI Technical Specification l 3/4.1.7 specified that the maximum assembly mass including control components i shall not exceed 1300 pounds. BGE entered this discrepancy into their corrective action process. Additionally, BGE cancelled two scheduled DSC loadings and postponed all future loadings until the issue was resolved. BGE personnel informed

'

the inspectors that the safety significance.of this issue was low due to the design of the canisters. The safety significance was based on a review of the system; however, BGE had not quantified the available margin for their dry fuel storage activitie E1.2 Dearadation of the Safety Inlection Pumo Recirculation Pioina Insoection Scoce (93702)

The inspectors reviewed the issues involving the discovery of a through-wall leak in the ASME Class 2 combined recirculation pipe for the Unit 1 emergency core ,

cooling system (ECCS) pump ) Observations and Findinos On December 11, a maintenance worker noted moisture on grouting near a 4-inch pipe in the Unit 1 component cooling water pump room. The stainless steel schedule 10 pipe (4"-HC-23-1005) provided a recirculation flow path from the safety injection and containment spray pumps back to the refueling water tank (RWT) during testing and other times when the pumps were in operation but not )

'

injecting into the reactor coolant system. The inspectors responded to the site after l l

i

!

,

.

l

.  !

l 9 l being informed of the finding and observed BGE efforts to characterize and resolve the issu BGE engineers determined that the moisture was coming from a section of the pipe which could not be isolated from the RWT. Using several non-destructive examination (NDE) techniques, BGE determined that the moisture resulted from a small pinhole leak at a welded support joint, and that this condition had probably existed since original construction. The leakage could not be readily quantified, but the surface of the weld would appear moist about 15 minutes after being drie ;

Additionally, a Code repair, as required by the plant's Technical Specifications (TS), j could not be conducted when the ECCS pumps were required to be operable. The <

reactor would have to be shut down and cooled to below 200 I l

BGE engineers evaluated the leak and conducted a risk assessment. The results of l the evaluation and potential corrective actions were presented to the plant i operational safety review committee (POSRC) on December 14. The POSRC ,

concluded that risk to safe plant operation was minimal and no POSRC member had '

a safety concern. The conclusion was based, in part, on the engineering evaluation j which stated that the indication was unlikely to propagate because of the very low )

stresses (compared to design allowable) on the pipe. A compensatory measure to

]

evaluate the flaw for growth was specified to validate the engineering assumptions '

until the defect could be repaired. In a 5-4 vote, the POSRC recommended to the plant general manager that an ASME Boiler and Pressure Vessel Code exemption be requested from the NRC until p!.snt conditions were conducive for a Code repai The minority of POSRC voteis believed that continued plant operation with the i existing defect was contrary to TS requirements. The plant general manager accepted the majority POSRC recommendation and BGE submitted their exemption request on December 19. The request was under review by the NRC when the inspection period ended, Conclusions )

i The inspectors determined that BGE engineers conducted a thorough and rigorous ,

examination of the piping integrity defect and evaluated potential safety consequences of continued operation. The technical content of the engineering I evaluation was excellent. The POSRC demonstrated a strong safety perspective and questioning attitude in their review of the potential nuclear safety consequences of the leak and engineering assumptions used to justify operabilit E3 Engineering Procedures and Documentation E Deficient Procedure for Unloadino a NUHOMS Cask Insoection Scone (60855)

A follow-up inspection was conducted to assess BGE corrective actions for two NRC identified weaknesses in independent spent fuel storage activities.

l l

l l

- .-. - .-- . - - . - . _ .

k i

,.  !

b. Observations and Findinas  ;

in a previous inspection, the NRC identified two weaknesses regarding independent '

spent fuel storage activities. These issues were described in NRC Inspection Report 50-317&318/96-07. One weakness was that the procedure used to unload fuel from a spent fuel cask did not control re-flooding the cask such that over- ,

pressurization of the cask would not occur. The second issue was that the procedure did not contain a method to sample the cask for damaged fuel prior to removing the dry shielded canister (DSC) shield plug. The inspector interviewed l

BGE staff and reviewed the following documents:  !

• Calvert Cliffs Technical Procedure ISFSI-02, Revisions 1,2 and 3, ,

" Independent Spent Fuel Storage Installation (ISFSI) Unloading" .

!

• Calvert Cliffs Independent Spent Fuel Storage installation, Materials License ,

No. SNM-2505, as amended July 21,1995, i The inspector found that BGE made changes to ISFSI-02 to include a step to sample the DSC atmosphere following removal of the outer cover plate. The change was l appropriate and the inspectors had no further concerns in this are !

I BGE also added a step, prior to re-flooding the cask to ensure the Nuclear I Engineering Unit had completed a calculation to determine maximum flow rate of j

'

water into the canister such that canister pressure remained below 10 psig. The

'

l inspectors had two concerns with the procedure change. ISFSI-02 was updated to i l add a step to " Ensure Nuclear Engineering Unit has completed calculations to determine maximum flow rate of SFP water into DSC such that DSC pressure remains below 10 PSIG". However, BGE did not have e procedure in place to

support the nuclear engineering unit in performing the calculation and selecting ,

l appropriate equipment to re-flood the DS Also, BGE had not performed a preoperational test of the proposed re-flood system to ensure it would adequately control the re-flood rate of fuel pool water into the DSC which is a safety related component. In addition, BGE had not performed a bounding analysis to determine what flow rates would be required to ensure that the cask would not be over-pressurized. Although the procedure was updated in an attempt to address NRC concerns documented in inspection Report 50-317&318/96-07, BGE did not develop supporting documentation to determine the required re-flood rate and select and test equipment required for DSC re-flood operations. This failure to develop supporting documentation was a violation of 10 CFR 50, Appendix B, Criterion V, which required that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances (VIO 50-317&318/97010-02). The inspector followup item related

!

to previously identified NRC concerns was closed. (Closed

.

IFl 50-317&318/96-07-01).

a

The inspector also found that the pressure gauge used to monitor DSC pressure during fill of the DSC was located down stream of the DSC and provided a

.

.

nonconservative indication of the actual pressure of the cask. In addition, the pressure gauge was specified by BGE to read 0 - 100 psi, minimum. Based on industry practice, the inspector considered this scale to be too large to accurately monitor a pressure of less than 10 psig as required by the BGE procedure. Industry practices such as those specified in ASME Boiler and Pressure Vessel Code, Section Ill, required that gages used in testing shall be graduated over a range not less than 1 1/2 times nor more than 4 times the test pressur The inspectors were also concerned that BGE had not adequately demonstrated re-flooding of the cask during the original dry run of DSC activities which were performed to address Condition 15 of License No. SNM-2505. Although the condition did not specifically state that a demonstration of re-flooding the DSC should be performed, the condition stated that the activities should not be limited to only those listed. The inspectors determined that since re-flood of the cask was required prior to retrieval of the fuelit would have been appropriate to perform a dry run of the re-flood during the demonstration of Condition 15.g, " Removing the cask lid and cutting open the DSC (length may be truncated) assuming fuel cladding failure."

c. Conclusions The inspectors found a significant weakness in the methodology used by BGE to determine that independent spent fuel storage installation unloading could be conducted safely. Specifically, BGE had not demonstrated the ability to re-flood the DSC such that overpressurization would be prevented. Also, BGE had selected a pressure gage that was nonconservative because the indicating scale did not

,

conform to industry practic E7 Quality Assurance in Engineering Activities E Update Final Safety Analvsis Report (UFSAR) Review Proiect Due to recent industry events and as a voluntary initiative, BGE developed a project plan with an overall goal to perform a review and ensure that the UFSAR accurately reflects the current plant design and operating practices. This plan was intended to be accomplished in at least two phase The first phase included review of approximately 50 selected UFSAR sections, using the guidance in NRC Regulatory Guide, RG-1.70. " Contents of Final Safety Analysis Reports (FSAR)". The review was to ensure that the UFSAR accurately and adequately described the design and operation of the plant. BGE system engineering, design engineering, and operations personnel performed these review BGE initially identified approximately 40 issues that were entered into the BGE corrective action program for resolution. BGE also planned to complete a root cause analysis to assess whether there were generic problems with the UFSAR change processes or their implementation. Examples of the types of discrepancies included:

-- . -- . - - .- - --. .

.

s

.

.  !

!

'

• Conflicting statements within the UFSAR concerning the ECCS pump [

'

minimum-flow isolation valves. One UFSAR section identified that these valves automatically close on a recirculation actuation signal (RAS), whereas !

another portion identified that this automatic feature is normally locked ou * Conflicting statements within the UFSAR concerning the 36 minute minimum time to switch over to the recirculation mode. One UFSAR section identified i that this minimum time was based on operating two HPSI pumps, in the i injection phase, whereas another section indicated that 36 minutes was ,

based on the operation of three HPSI pump ,

• The UFSAR stated that each of the twelve containment pressure transmitters has an individual sensing point whereas, in the plant, there were three transmitters common to each sensing poin :

l Also, in review of the spent fuel handling issue (See 01 F .he inspectors found that UFSAR Section 9.8.2.3 " Spent Fuel Pool Ventilation," .ated that an air supply j system consisting of two 50 percent capacity air handling units provides ventilation :

for the spent fuel pool area. However, the inspectors were informed by BGE that l the supply fans had been removed from service and not operated for more than five '

years. Instead, air was supplied only from leakage into the area from adjoining areas. The inspectors considered that the discrepancy could have been identified j during the ongoing BGE UFSAR initiativ i l

BGE planned to develop a second phase that will expand the scope and depth of the UFSAR reviews based on the findings of the initial reviews. BGE informed the inspectors that they plan to make a submittal to NRC concerning scope and schedule for completion of the project. No operability issues had been identified during the reviews. Enforcement action regarding design issues identified during the BGE review was Unresolved (URI 50-317&318/96010-03), pending completion of the BGE initiative and NRC inspection of the completed review. The unresolved item is consistent with the General Statement of Policy and Procedures for NRC Enforcement Actions, NUREG 1600, as published in the Federal Register, Volume 61, Number 203, Page 5446 l E7.2 (Closed) Unresolved item 50-317&318/96-08-01 Cable Seoaration issues i I inspection Scope

,

The inspectors reviewed safety related cable separation issues associated with URI 50-317&318/96008-01, including the status of electrical cable separation barriers and related modifications performed in 1990; BGE's corrective actions for a prior NRC violation; BGE actions to update the configuration control drawings for separation barrier installation changes; and whether appropriate administrative controls have been applied to the storage and retention of the applicable project record .

.

b. Findinas and Observations NRC violation 50-317/89-27-05 involved discrepancies in the BGE control of electrical cable separation barriers. NRC Inspection Report (IR) 96-08 identified an apparent lack of ownership for the cable separation barriers and the associated modifications performed to restore these barriers to licensing basis condition Discussions with BGE personnel indicated that BGE had assigned engineers responsibility for cable separation. System engineers were responsible for cable separation for the individual systems including the need to conduct periodic walk downs on portions of the barriers tmed on accessibility, maintenance history, and potential for damage. A project ensmeer was assigned responsibility for closeout of the 1990 barrier modification package (FCR 90-10).

The inspectors found that some of the original project records related to the walkdown, engineering evaluation, and repair of separation barriers for the project plan appeared to be quality records that were not stored in the records vault and were not available for general use by engineering personnel. The inspectors questioned whether the appropriate administrative controls had been applied to the storage and retention of the applicable project records. Discussions with BGE personnel identified that BGE would be reviewing these records as part of the closecut of the FCR 90-10. BGE personnel indicated that copies of the inspection records and engineering evaluations were filed as quality records with the ;

nonconformance reports (NCRs) and maintenance orders (MOs) generated during i FCR 90-10. Therefore, filing the originals would duplicate existing record However, the inspectors questioned whether the BGE inspection reports for areas found acceptable had been filed and maintained as quality records. This issue was ;

still under review by BGE and BGE personnel indicated that the index of the NCRs I and MOs were not currently retained as quality records, but would probably be added to the records syste The inspectors also found that the 1990 modification (FCR 90-10) to address the NRC violation for inadequate cable separation had not been closed since the effort 1 was completed in 1994. During further reviews with BGE personnel, the inspectors I found that FCR 90-10 was stillin working status. This modification was among a large number of modifications where the work had apparently been completed, but j the modification had not been closed out. BGE personnel indicated that their quality I l assurance department had identified that this was an issue requiring BGE l

'

( management attention. One effort to address the backlog and manage the

! closecuts included the development of a relevant performance indicator. Review of

,

the recently developed indicator revealed that there are approximately 375 i modifications that have had the work completed in the field, but the final closeout, including updating the configuration control documentation, was not complete. On questioning by the inspectors, BGE personnel could not identify the oldest outstanding modification or the average age of the backlog of modifications awaiting closecu BGE personnel identified that the drawings for Unit 2 had not been updated to reflect the as-built conditions for changes made to the separation barriers by FCR

.

.

90-10. BGE generated an issue report to adaress this issue. The inspectors identified two instances where the as-built conditions and the drawings for the Unit 1,45 foot switchgear room, did not agree. Discussions with BGE personnel indicated that their reviews concluded that the Unit 1 drawings were updated during the FCR 90-10 process. Additionally, marked-up drawings were developed for Unit 2 during the FCR 90-10 process; however, the controlled drawings were not updated and the marked-up drawings could not be located. As discussed below, the inspectors concluded that BGE did not complete the corrective actions identified i by their response to NRC violation 50-317/89-27-05. BGE told the inspectors of plans to walkdown the Unit 2 drawings that were known to have not been update These Unit 2 drawings included some Unit 1 and common areas. BGE planned to determine the extent of the condition and specify corrective actions based on the results of the walk downs and identified deficiencies. The inspector concluded that the configuration control related to cable separation had been inadequate to ensure that design documents reflected the as-built configuration !

!

The inspectors initially identified three examples where barriers did not meet the electrical separation criteria in FSAR chapter 8.5, " Separation Criteria," and design I document E-406, " Design and Construction Standards for Cable and Raceway".

During this inspection period, BGE and the inspectors identified eleven additional cable separation related issues during system walkdowns. The deficiencies included missing or cracked barrier materia The inspectors reviewed BGE's response to NRC violation 50-317/89-27-05 documented in a March 9,1990 letter to the NRC, and reviewed the effectiveness of the corrective actions. The inspectors concluded that two of these corrective actions were not effectively implemented or were inadequate to preclude recurrence. These issues were (1) failure to ensure that the design documents reflected the as-built configurations; and (2) failure to ensure that the as-built versus as-designed configurations continued to meet the criteria on a long-term basi The inspector also questioned whether BGE had established a clear understanding of l the licensing basis and criteria for electrical separation. Specifically, the inspector questioned the licensing basis and criteria for electrical separation of cables passing through free air. For example, cables passing between cable trays, cables passing from conduit to cable trays, or cables passing from trays and conduits to penetrations. Currently, BGE design document E-406 showed that a separation j barrier must be sealed at the penetration caused by a cable exiting a cable tray to ,

enter a conduit; however, E-406 did not require the protection and separation from l redundant channels for this cable passing through free air. Discussions with BGE personnel indicated that this issue had been identified and documented as an issue in their correctiva action system. The inspectors found that the lic9nsing basis l criteria for electrical separation relative to redundant cables passMg 1.hrough free air was not clea As discussed above, the corrective action to ensure that the design documents reflected the as-built configurations was found incomplete. BGE's corrective actions to ensure that the as-built versus as-designed configurations continued to

'. l

!

l

-

15 I meet the criteria on a long-term basis were inadequate to preclude the challenge to electrical separation resulting from damaged or missing marinite separation barrier The incomplete corrective action and the inadequacy of the corrective actions to preclude recurrence were violations of 10 CFR 50, Appendix B, Criterion 16,

" Corrective Actions" (VIO 50-317&318/96010-04). Unresolved item (URI 50-317&318/96-08-01) is close c. Conclusions ,

i The inspectors concluded that the BGE corrective actions to ensure that the as-built )

versus as-designed configurations for electrical separation barriers were inadequat i The specific weakness was the challenge to electrical separation resulting from l damaged or missing marinite separation barriers. The inspector also found that the j design documents did not reflect the as-built configuration The inspectors concluded that following issues present potential challenges to maintaining the licensing and design basis of electrical separation: (1) BGE's self-identified backlog of modifications awaiting closeout; (2) the licensing basis criteria for electrical separation relative to redundant cables passing through free air was not clear; and (3) BGE's ongoing review to determine which cable separation barrier modification project records should be filed and maintained as quality record E8 Miscellaneous Engineering Issues (92902)

E8.1 (Closed) Unresolved item 50-317&318/93-25-01: failure to promptly perform a reportability evaluation. On August 5,1992, BGE determined that a fire in the Unit 1 cable chase could potentially cause a loss of off-site power to both 4160 V emergency busses, resulting in the loss of both trains of control room ventilation. A BGE engineer wrote an issue report documenting the finding and provided the report for supervisory review. Neither the supervisor or the initiator considered the issue NRC reportable. The issue report was then reviewed by the issues assessment unit which included members from both operations and nuclear regulatory matters (NRM-licensing), with no reportability concern. However, the inspector determined that as was BGE practice at the time, an NRM engineer had independently reviewed the issue for reportability and had concluded that no report was required. This conclusion was not formally documented until a more rigorous analysis was performed several weeks later, with the same conclusion. The inspector reviewed the formal analysis and concluded that BGE's evaluation and conclusions were reasonable, and that the compensatory actions taken were appropriate. The inspector concluded that BGE had reviewed and determined reportability of the issue in accordance with BGE practices. However, documentation of the initial review was weak. Since that time, BGE has upgraded their reportability proces This item is close .

,

,

E8.2 (Closed) Unresolved item 50-317 and 318/96-04-01 Inocerable LPSI oumo circuit breaker due to bent trio-caddle oroble The item involved the BGE corrective actions in response to a circuit breaker failure on June 14,1996. The failure was due to a bent trip-paddle in the General Electric 4160 Volt Magne-Blast circuit breaker. There were two issues in this unresolved item. The first issue pertained to the bending of circuit breaker linkages by the technicians during the preventive maintenance inspection. This issue was closed by Section E6.4 of Inspection Report 50-317&318/96-0 The second issue involved the BGE root cause analysis (RCA) for the breaker failure and bent trip-paddle. The issue was updated in Section E1.1, part b, of NRC Inspection Report (IR) 50-317 & 318/96-0 The inspectors reviewed the completed root cause analysis. The analysis [

considered a General Electric evaluation that included information concerning testing of the failed breaker. BGE concluded that the root cause for the bent trip paddle was a weak trip shaft reset spring that would allow the trip-paddle to contact the breaker frame with excessive force. General Electric concluded that the weak spring was probably an isolated incident due manufacturing defect or installation damage. BGE believed that it was also possible that the springs relaxed due to i age-related degradation. In accordance with 10 CFR Part 21, BGE submitted a i notification of these problems to the NRC by letter dated January 15,1997. The long term corrective actions were to replace the springs during scheduled breaker overhauls or to replace the 4160 volt breakers. The actions were to be completed in the next two or three years. BGE expected that the ongoing modification of the trip paddle would preclude further breaker failures. The inspectors concluded that i BGE's review, analysis, and corrective actions for these issues were extensive and l appropriate. The item is close IV. Plant Support R4 Staff Knowledge and Performance in Radiation Protection and Chemistry On January 3, the inspector observed are auxiliary building operator complete routine log taking and operations duties. During the rounds, the inspector observed l that the operator routinely used good radiological work practices that included verification of radiation and contamination levels using radiological contols postings prior to entry into radiation areas. Also, the operator verified the eC tence of low radiation leveis in areas subject to high radiation by both checking local radiation monitor readings and by completing spot radiation level checks using an alarmed dosimeter. The operator also contacted radiological controls personnel prior to each entry into a potentially contaminated area. The inspector considered the actions of the operator to be a very good demonstration of sound ALARA and radiation controls practice {

, _ . . _ -

_ . _ _ _ _ _ _ . - _ _ _ _ . . _ _ _ _ . .

-

. . _ - . . . _ - . _ _ _ _ . . _ . . _ _ _ _ _ _ _ . . _ . . _ _ _ _ .

!^  !

i i

• .

!

17 L

V. Moneaement Meetinas X1 Exit Meeting Summary  :;

.

During this inspection, periodic meetings were held with station management to T!

[ discuss inspection observations and findings. On January 27,1997, an exit ,

i meeting was held to summarize the conclusions of the inspecticn. BGF '

management in attendance acknowledged the findings presente "

!

,

i

t I

,

f i

?

k f

v

!

. . . . .- . . . . . . - - . . - - . - - . .

. - _ - . - - - _ . -

-

I

,

.

ATTACHMENT 1

,

PARTIAL LIST OF PERSONS CONTACTED

!

D_QE P. Katz, Plant General Manager i K. Cellers, Superintendent, Nuclear Maintenance i K. Neitmann, Superintendent, Nuclear Operations *

P. Chabot, Manager, Nuclear Engineering T. Camilleri, Director, Nuclear Regulatory Matters 8. Watson, General Supervisor, Radiation Safety C. Earls, General Supervisor, Chemistry I'

T. Sydnor, General Supervisor, Plant Engineering

!

!

INSPECTION PROCEDURES USED f IP 62707: Maintenance Observation IP 71707: Plant Operations IP 93702: Prompt Onsite Response to Events at Operating Power Reactors IP 61726: Surveillance Observations IP 37550: Engineering IP 37551: Onsite Engineering IP 71750: Plant Support Activities IP 83750: Occupational Exposure IP 92700: Onsite Follow-up of Written Reports of Non-routine Events at Power Reactor Facilities IP 92002: Follow up - Engineering i IP 82701: Operational Status of the Emergency Preparedness Program I l

l

!

l

l'

!

l

!.

Attachment 1 2 ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-317&318\96010 01 VIO Failure of BGE to establish procedures that assure that in a fuel handling event, released gases would be filtered through charcoal adsorbers and the failure to follow the briefing and procedure control procedures 50-317&318/97010-02 VIO Failure to develop documentation to support dry fuel storage cask unloading 50-317&318/96010-03 URI Old design issues identified during the BGE UFSAR review 50-317&318/96010-04 VIO Incomplete corrective actions for electrical separation barriers Closed 50-318/96004 LER Missed Surveillance Due to Less than Adequate Review of Surveillance Test Procedure 50-317&318/93025-01 URI Failure to Promptly Perform Reportability Evaluation 50-317&318/96007-01 IFl Engineering for IFSF1-02, Cask Unloading 50-317&318/96008-01 URI Cable Separation issues 50-317&318/96004-01 URI Inoperable circuit breaker due to bent trip-paddle i LIST OF ACRONYMS USED AFW Auxiliary Feedwater ALARA As Low As Reasonably Achievable  !

BGE Baltimore Gas & Electric CFR Code of Federal Regulations l ECCS Emergency Core Cooling System EDG Emergency Diesel Generator DSC Dry Shielded Canister FCR Facility Change Request HPSI High Pressure Safety injection

'

IFl Inspector Followup Item IR Inspection Report or issue Report (non-conformances)

ISFSI Independent Spent Fuel Storage Installation LER Licensee Event Report LPSI Low Pressure Safety injection MO Maintenance Order NCR Nonconformance Report or issue Report l NDE Non-destructive Examination NRC Nuclear Regulatory Commission NRM BGE Nuclear Regulatory Matters Group NUHOMS Nutech Horizontal Modular Storage PDR Public Document Room 1 l

. . . _ = . . -. . . . . - . _ _ _ _ _ . . . . _ . ._ . . _ _ _ _. . ..

..N'

l

.

i

Attachment 1 3 i PSIG Pounds per Square Inch-Gauge RAS Recirculation Actuation Signal  :

RCA Root Cause Analysis "

RWT Refueling Water Tank SFP Spent Fuel Pool  :

TCB Trip Circuit Breakers '

TS Technical Specification l

! UFSAR Updated Final Safety Analysis Report i URI Unresolved item  !

VIO Violation  ;

i f

i e

!  :

'

l

!

.

'

)

l 1 ,

i

' l

i s

!

• J i

,

i i

- . _ _