ML20246E091

Responds to NRC 881020 Ltr Re Violations Noted in App R Program Audit Insp Rept 50-344/88-34.Corrective Actions: Deenergization of Stated Buses at 12.47 Kv Bus Will Be Deleted from Procedure

ML20246E091
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	04/28/1989
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NUDOCS 8905110143
Download: ML20246E091 (33)

Text

r '. .

s .

a.  :~ . ..

Portland General ElectricCorpsiy 4 (Q ly e;l.

RC s vgg Ot/ y David W. Cockfield Vice President, Nuclear $

00 Wv April 28,'"l48 A// ,a g7 Trojan Nuclear Plant Docket 50-344 i License NPF-1 Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek CA 94596-5368

Dear Mr,

Martin:

Portland General Electric Company (PCE) Responses To Title 10 to the Code of Federal Regulations, Part 50 (10 CFR 50). Appendix R Audit Items The purpose of this letter is to provide Portland General Electric's

.(PGE) resolutions of Nuclear Regulatory Commission (NRC) items identified during the August 1988 Appendix R Program audit. The NRC's findings are I documented in Inspection Report 50-344/88-34, which was transmitted to PCE by letter, dated October 20, 1988. A total of 14 items were identi-fled, 11 of which are related to PCE's normal and alternative safe shutdown methodology and are addressed in Attachment A to this letter.

Attachment B provides a comparison of the current versus the revised alternative shutdown methodology. The remaining Items, namely 344/88-34-12, 344/88-34-13, and 344/88-34-14, will be addressed in separate correspondence or by followup NRC inspection.

Sincerely, Attachments e

40 g

<- o c: U.S. Nuclear Regulatory Commission ATTN: Document Control Desk M

C00 Washington DC c)x Mr. William T. Dixon

$ State of Oregon Q Department of Energy SQ N'

Mr. R. C. Barr NRC Resident. Inspector hh Trojan Nuclear Plant y

m sw saun sem ama o<egon 97204 \

m l l

' Trojrn Nucloir PIEnt Jshn B. Mirtin

• I' Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 1 of 25 PORTLAND GENERAL ELECTRIC COMPANY (PCE) RESPONSES TO NUCLEAR REGULATORY COMMISSION (NRC)

TITLE 10 TO THE CODE OF FEDERAL REGULATIONS, l PART 50 (10 CFR 50). APPENDIX R AUDIT ITEMS UNRESOLVED ITEM NO. 344/88-34 Alternste Shutdown Capability

" Based on the inspection team's review, the team determined that the licensee's alternate shutdown capability appears to be inconsistent with .

the requirements of Appendix R as follows:" )

1. "The methodology that has been implemented may not have been fully I recognized by the NRC staff during its 1985 review. The licensee's method requires an intentional loss of all AC power for approximately I 40 minutes in order to mitigate the adverse consequences of fire damage to safe shutdown circuits located in the control room or cable spreading room. Although the shedding of offsite power to preclude certain spurious activations has been a methodology that has been accepted by the staff, the concurrent unavailability of onsite AC power for 40 minutes may be excessive. Section III.L.3 of Appendix R requires that onsite AC power be available when offsite power is available, and when offsite power is not available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

"A weakness in the methodology appears to be that if fire damage to 2.

certain safe shutdown circuits occurs prior to the licensee's manual operator actions being performed, in accordance with Procedure EFP-1, the alternate safe shutdown capability may be defeated by the fire damage (i.e. fire damage to Service Water System control cables). In the event of a concurrent loss of offsite power and a loss of service water to the EDGs, damage could occur to both EDCs within three to five minutes. This appears to be contrary to section III.L.4 of Appendix R.Section III.L.4 of Appendix R requires that if such equipment and systems will not be capable of being powered by both onsite and offsite power because of fire damage, an independent onsite power system shall be provided.

3. "By implementing Alternate Shutdown Procedure EFP-1, it appears that plant operators are placed in a position to intentionally disable all AC power, perhaps prematurely, upon receipt of a cable spreading room fire alarm, in order to avoid sustaining fire damage to safe shutdown  ;

circuits prior to implementing the alternate shutdown methodology. '

This appears to involve a voluntary entry into Technical Specifi- {

cation 3.0.3. The licensee stated that the criteria for declaring a l fire an emergency event is consistent with the site Emergency Plan. f However, the licensee emphasized that the site Emergency Plan criteria allows for shift supervisor discretion under such circum-stances. This statement by the licensee appears to support the inspection team's position that the provisions of 10 CFR 50.54(x) '

i l

1 l

' Trojan Nuclozr Plant John B. M2rtin Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 2 of 25 were used as a basis for writing the procedure and placing the plant in a condition outside of those conditions analyzed in Chapter 15 of the FSAR, when the procedure is implemented.

"Once the decision is made to enter the procedure (control room evacuation), there are no additional decision points or related symptoms provided prior to the deenergization of all AC power sup-plies. The licensee indicated that this approach was intended to address worst case fire scenarios. However, a less severe fire situation could result in shift supervisor decisions to enter the

< procedure and cause an unwarranted disabling of all AC power sup- l plies. The decision point et which the shift supervisor is directed to evacuate the control room and enter the procedure is defined as "a fire lasting more than 15 minutes and affecting redundant safe shut-down trains". If the decision to evacuate the control room and enter the procedure is made 15 minutes after a verified fire, the time threshold for carrying out certain steps in the procedure may have elapsed (i.e. diesel generators must be stopped within three minutes  !

and pressurizer PORVs must be closed within five minutes).  !

4. "After one hour has elapsed, the licensee relies on operators who are assigned fire brigade duties to return and assist in implementing the methodology. This reliance appears to be based the assumption that a fire occurrence will not last for more than one hour or, additional fire fighting assistance will arrive before one hour has elapsed.  ;

i "The above items appear to be inconsistent with Sections II.A, III.G.A

[ sic) and III.2.a [ sic) of Appendix R and is considered an Unresolved Item (344/88-34-01) pending further NRC review." l PGE RESPONSE i i

1. Emergency Fire Procedure (EFP)-1, " Alternative Shutdown for Evacuation of Control Room Caused by Fire", includes a step to deenergize breakers at the 12.47 kV buses supplying offsite power to various loads onsite. This is done to limit spurious operation of equipment during shutdown. This action is taken in conjunction with opening breakers for the 12.47 kV power feed at the 4.16 kV Engineered Safety Feature (ESF) switchgears.  !

Reevaluation of EFP-1 has shown that deenergization of all offsite l power will not be necessary. The action to deenergize all ESF and non-ESF buses at the 12.47 kV bus will be deleted from EFP-1 by 4 September 29, 1989. Deenergization of ESF buses will be completed at l the 4.16-kV switchgears and 480-V load centers and motor control centers.

Spurious operation of non-safe shutdown equipment fed from non-ESF 1 buses which may adversely impact safe shutdown will be mitigated by ,

deenergizing the loads of concern at their branch feeder breaker. The l 1

I 1

J

E 1.

• trojan Nuclser plcnt John B. Martin

/' Docket 50-344 April 28, 1989

' License NPF-1 Attachment A '

l page 3 of 25 remaining non-ESF buses provide power to lighting buses and other

l. equipment which although not required for shutdown could facilitate shutdown in the event that offsite power is available.

An attempt to credit operation of the charging pumps and other safe l shutdown equipment via offsite power feed to the Train B switchgear L could complicate the alternative shutdown procedures. Since breaker l coordination for non-ESF power supplies has not'been ensured, a fire l could cause loss of offsite power due to common power supply associ-ated circuits. Therefore, the shutdown crediting offsite power to the Train B switchgear could be interrupted by a fire-induced loss of -

offsite power. This would complicate the shutdown process since.

reinitiation of the fire procedure would be required at a potentially critical point in the operation of safe shutdown equipment.

Deenergization of power to 4.16-kV ESF buses will be necessary to prevent spurious operation of safe shutdown equipment. The Train A switchgear must be deenergized to prevent spurious operation of equip-ment not utilized for alternative shutdown. The Train B switchgear must be deenergized to prevent spurious operation of safe shutdown and nonsafe shutdown equipment. Controlled reenergization of the Train B switchgear will be initiated by closure of the Emergency Diesel Generator (EDG) feeder breaker to the Train B switchgear subsequent to a manual start of the EDG. The time between deenergization of offsite power feed to the Train B switchgear and closure of the switchgear EDG breaker will be minimized by prioritization of the associated operator actions. This time frame has been reduced to approximately 4.5 minutes.

2. Loss of service water to the EDG is discussed in Item 3 below.

3. NRC guidance does not exist for post-fire shutdown procedure entrance criteria. This issue is of concern primarily for control room and cable spreading room fires Where a control room evacuation may be necessary. For these fires, initiation of the post-fire shutdown ]

procedure is particularly difficult since control room evacuation is j considered a last resort by the shift supervisor. Detailed criteria are difficult to generate since it could limit the flexibility avail-able to the shift supervisor. The final decision to evacuate must always be made by the shift sapervisor based on his evaluation of the situation utilizing his extensive operating experience.

After the shift supervisor makes the decision to evacuate the control .i' room and declare the situation an emergency event, the requirements of Technical Specification 3.0.3 are not applicable. Because the Appendix R fire scenarios are outside the conditions analyzed in Chapter 15 of the Final Safety Analysis Report (FSAR), the EFPs are consistent with the criteria of Appendix R.

l l

l

irojan Nuclotr Plent John B. Martin Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 4 of 25 1

The EFP in place at the time of the 10 CFR 50, Appendix R audit provided specific criteria for control room evacuation and dispatched an operator to the EDG to trip the EDG in case of a spurious start without service water. This action was initiated prior to implemen-tation of the alternate shutdown procedure and associated control room evacuation. To address the NRG's concern with the short time frame available for operator response to spurious EDG start or pressurizer power operated relief valve (PORV) opening, a "evision to the proce-dure was implemented, as an interim measure, to provide actions to trip the diesel and close the PORV block valves upon receipt of a fire alarm in the control room or cable spreading room.

Reevaluation of spurious pressurizer FORV operation has shown the need to modify the pressurizer PORV control circuitry. The modification to the pressurizer PORV control circuitry will enable doenergization of the valves via a double pole switch in the control room. Based on this modification, EFP-0, " Procedure in the Event of a Fire", will be revised to reflect control rcom closure of the pressurizer PORVs if the PORVs spuriously open upon receiving a fire alarm in fire areas where cables for the pressurizer PORVs are routed. The actions associated with closure of the pressurizer PORV block valves will be deleted. For a fire in the control room and cable spreading room, a high-priority backup action to deenergize the pressurizer PORVs from the electrical distribution panels will be taken to provide further assurance of closure of the pressurizer PORVs. The modification to the pressurizer PORV control circuitry will be completed during the 1990 refueling outage. EFP-0 will be revised by September 29, 1989.

Spurious EDG operation concurrent with the loss of service water cooling to the diesel has also been reevaluated. It has been deter-mined that the probability of such an event occurring, before action is taken by an operator to isolate the circuit, is low enough not to warrant design consideration. However, in order to assure that EDG isolation occurs in a timely manner, EFP-0 will be revised by September 29, 1989 to require operator action to trip the EDG in the control room upon control room indication of an EDG start in conjunction with a fire alarm in fire areas where cables of both trains of service water may be damaged. In addition, for a fire in the control room and cable spreading room, a high-priority backup action to decouple and trip the EDG at the local control panel will be taken.

Purther discussions on the EDG and the pressurizer PORV are included in the response to Unresolved Items 344/88-34-03 and 344/88-34-04, respectively.

4. The current alternative shutdown methodology described in Topical Report PCE-1012 Volume II, " Fire Protection Plan, 10 CFR 50, Appendix R Review", assumed additional operators will be available

~

. . Trojtn Nuc133r Pitnt Jchn B. Martin Docket 50-344 April 28, 1989 4

License NPF-1 Attachment A Page 5 of 25 k

after one hour to assist in actions required to achieve cold shut-down. The alternative shutdown time line was reevaluated to determin'e if the minimum Technical Specification staffing level of five oper- i ators was adequate to perform manual actions required for cold shutdown. The revised time line showed that five operators are cap-able.of completing the manual operations required for hot' shutdown, and for achieving and maintaining cold shutdown. Although additional operators, excluding operators assigned to fire brigade duties, will likely be available, the revised time line does not credit additional operators to assist in completing manual operations for cold shutdown.

Volume II of PGE-1012 will be revised by December 31, 1989 to delete reference to additional operators credited after one hour to assist in completing manual actions.

FAD /3033W

JTrojan Nuclear Plent John B. Martin

'- ~'

Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 6 of 25 l

l UNRESOLVED ITEM No. 344/P8-34 Reactor Coolant Pump Seal Injection I "The inspection team identified a concern that a loss of coolant flow to the seals can result in Reactor Coolant Pump (RCP) seal degradation and subsequent excessive leakage of primary coolant during the approximate one-hour period that the seals are without seal injection during imple-mentation of the alternate safe shutdown methodology. According to the i time lines in the procedure for implementing the methodology, it will I take approximately 40 minutes to start the EDGs and another 20 minutes for the charging pumps to restore seal cooling. The licensee's position

{

was that only normal leakage would occur during this period based on a

{

Westinghouse analysis (WCAP 10541). The inspection team's review of the j referenced analysis disclosed inconsistencies between assumptions in the l analysis and interpretations of the analysis by the licensee. The analysis assumes that a seal leakage rate of 21-gpm/ pump will occur immediately upon loss of seal injection at reactor temperature and pressure and a possible leakage rate of 480 gpm/ pump can eventually result.  !

" Based on the team's review, the team determined that the inconsistencies between the Westinghouse analysis (WCAP 10541) and the assumption that a loss of seal injection for a period up to one hour will have no adverse affect on seal integrity by the licensee represent a significant level of uncertainty about the integrity of the primary reactor coolant system  !

boundary due to the potential for RCP seal damage and subsequent leakage '

during the period that seal injection is lost.

"This appears to be inconsistent with Sections III.L.1 and III.L.2 of Appendix R and is considered an Unresolved Item (344/88-34-02) pending further NRC review."

PGE RESPONSE Revision 2 of WCAP-10541, " Reactor Coolant Pump Seal Performance Following a Loss of All AC Power", shows that the normal RCP seal leakage rate of 3 gpm/ pump will occur for the first 10 minutes j subsequent to a loss of saal cooling. After 10 minutes, the RCP seal leakage este will increase to a value in excess of 21 gpm/ pump for a short period of time, and then rapidly decline to a leakage rate of 21 gpm/ pump or less. The increase in leakage rate occurs during the transient heat-up phase and thermal equilibrium phase. During the cooldown and depressurization phase, the reduction in RCS pressure will cause the differential pressure across the No. I and No. 2 RCP seals to decrease, thereby reducing the RCP seal leakage rate towards the normal leakage rate of 3 gpm/ pump.

A new analysis for RCS makeup requirements was performed using the leakage rates stated in WCAP-10541. This analysis also determined the RCS makeup requirements based on one pressurizer power-operated

6

$ trojan Nuclear Plent John B. Martin 1 -

• Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 7 of 25 relief valve (PORV) spuriously opening for a period of three minutes, two minutes, or one minute in conjunction with RCP seal leakage.

Based on a limiting assumption of one pressurizer PORV spuriously open for three minutes, and an RCP seal leakage rate of 3 gpm/ pump for the first 10 minutes and increase to 21 gpm/ pump thereafter until

l. seal injection charging is initiated, RCS makeup would be required l within 13 minutes.

The alternative shutdown time line for operator actions was revised utilizing the new time constraint for RCS makeup. The revised alternative shutdown timeline has shown that RCS makeup can be re-established within 13 minutes using the Train B centrifugal charging pump (CCP). PCE determined that the positive displacement ,

charging pump (PDP) could not be credited for alternate charging as previously stated in our letter of June 8, 1988. The inability to credit the PDP is attributed to a previously overlooked interlock that prevents the establishing of cooling water flow to the PDP seals and lube oil cooler. To ensure the availability of charging for a control room or cable spreading room fire, procedures will be revised to trip the centrifugal charging pumps in the control room prior to evacuation. Attachment C provides a detailed discussion of this issue and PGE's corrective actions.

Additional discussion on the pressurizer PORV spurious operation is included in the response to Unresolved Item 344/88-34-04.

FAD /3033W l

i

_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ . - _ _ __ .)

'Trojtn Nucle:r Plant John B. Martin Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 8 of 25 UNRESOLVED ITEM NO. 344/88-34 Mitigating Damage to Both Emergency Diesel Generators "The inspection team identified the concern that damage could occur to both emergency diesel generators due to a loss of cooling water during implementation of the alternate safe shutdown methodology. The team is concerned that in the event of a concurrent loss of offsite power with a loss of service water to the diesels, damage could occur to both EDGs q within three to five minutes after a diesel automatic start is initiated  !

upon loss of offsite power. The time lines examined for this occurrence I by the team from PGE-1012 and the licensee's internal memorandum dated June 1, 1988 (D. J. Harvey to R. L. Russell), entitled ' Validation for EFP-1,' indicates an elapsed time of more than three minutes for an operator to arrive at the EDG rooms to shut the EDGs down before this damage occurs. During the walkdown of the procedure by the licensee's staff and the inspection team on August 25, 1988, the team also deter-mined that more than three minutes could elapse before an operator completes these actions. Based on this assessment, the team determined that the marginal time frame permitted for an operator to complete required actions before damage occurs to the EDGs is of questionable adequacy.

"This appears to be inconsistent with Sections III.G.3 and III.L.4 of Appendix R and is considered an Unresolved Item (344/88-34-03) pending further NRC review."

PGE RESPONSE The fire areas in which cables for both trains of the service water system are routed through are the control room (C11), cable spreading room (C7), service water pump room (II), Manholes 3 and 4 (M3 and H4) and the Auxiliary Building general area (A4). A fire in these fire areat could cause loss of service water cooling to the emergency diesel generator (EDG) which could damage the EDG if it spuriously starts upon loss of offsite power. All other fire areas will have at least one train of service water available for cooling the EDG jacket water.

It is PGE's position that the probability of a fire causing an unacceptable spurious signal, in conjunction with loss of offsite power before action is taken by an operator to isolate the circuit, is low enough not to warrant design change consideration. Because of the low probability of this occurrence, the alternative shutdown methodology will not consider the spurious operation of the EDG in conjunction with loss of service water prior to (a) tripping the EDG j from the control room and (b) decoupling and/or tripping the EDG at l the local control panel.

V )

l **

' Trojan Nuclear Plant John B. Martin

. Docket 50-344 April 28,.1989 License NPF-1 ' Attachment A Page 9 of 25 i

I The current shutdown methodology requires an operator to locally monitor the EDG jacket water if the EDG is running. The operator will then trip the EDG when the jacket water reaches a temperature greater than 190*F_without service water available. This action is completed for a fire in all fire areas except the control room and cable. spreading room. The current methodology for fire in the control room or cable spreading room requires an operator to trip the i EDGs at the local control panel.

The Emergency Fire Procedures will be revised-by. September 29 1989 to instruct an operator to trip the EDG in the control room upon control room indication of a diesel generator start in conjunction with a fire indication in those fire areas where both trains of service water are routed.

FAD /3033W l

l I

l

<Trojen ruelser Pirnt John B. Mirtin  ;

Docket 50-344 April 28, 1989 l License NPF-1 Attachment A i Page 10 of 25 l UNRESDLVED ITEM No. 344/88-34 Mitigating Spurious Operation of ,

Pressurizer Power Operated Relief )

Valves (PORVs) i l

"The inspection team raised the concern that an operator is directed to '

open the DC supply breakers for the pressurizer's power operated relief valves (PORVs) to prevent or mitigate spurious operation of the valves in the event of a CR/CSR fire. The licensee's Procedure EFP-1 cautions operators that this action must be accomplished within five minutes.

According to the licensee, the five-minute time frame is based on sustaining spurious operation of one pressurizer PORV and the loss of primary system inventory through the open valve. If the PORV is closed within five minutes, the licensee's position is that fuel damage will not occur and the loss of primary system inventory can te recovered When normal charging is restored approximately 40 minutes later.

"The inspection team expressed concern to the licensee regarding the narrow time line established to accomplish this action. The team's position is that if fire damage occurs to the control circuits to the valves and certain control room instrumentation prior to receipt of a fire alarm, time 0.00 for closing the valve starts when the damage occurred. In this case, by the time operators respond to the receipt of a fire alarm, evacuate the control room, implement the procedures for disabling all AC power and closing the valve, more than five minutes could have elapsed. The team recognized, however, that operators would probably attempt mitigating actions for this condition prior to control room evacuation. During the team and licensee's walkdown of Proce-dure EFP-1 on August 25, 1988, the team observed that the operator assigned responsibility for opening the DC supply breakers for the PORV's appeared to accomplish this action in excess of five minutes, When time 0.00 was established as the time that the shift supervisor decided to evacuate the control room because of a postulated verified cable spreading room fire of significant magnitude.

"This appears to be inconsistent with Sections III.C.3 an III.L.1 of Appendix R and is considered an Unresolved Item (344/88-34-04) pending further NRC review."

PGE RESPONSE The design of the Trojan pressurizer PORVs (PCV-455A and PCV-456) could result in the opening of the PORVs as a result of a hot short to the valve control circuitry. The valves would remain open for the duration of the hoh shart but would close When the hot short no longer affects the valve control circuits.

A double pole switch configuration in the valve's main control switch j is provided for PCV-456. This design configuration provides the i

capability to manually close the valve in the control room. This

'Trojen Nuclear Plent John B. Martin Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 11 of 25 switch, when manually actuated by placing the switch in the CLOSE position and pulling the switch to lock it in the CLOSE position, i prevents spurious opening of the valve by a single hot short. Two  !

proper polarity shorts are required to spuriously open the valve subsequent to actuation of this switch. The double pole switch configuration has not been implemented on PCV-455A. The current configuration requires a single hot short on the PCV-455A control circuit to spuriously open the pressurizer PORV. To enhanco the design of PCV-455A, a modification to add a double-pole switch configuration, similar to PCV-456, will be completed during the 1990 refueling outage such that two proper polarity shorts will be required to open the pressurizer PORV.

NRC guidance contained in Generic Letter 86-10 requires that the effect of two proper polarity DC faults be evaluated for high/ low pressure interface boundary components. The pressurizer PORVs are considered high/ low pressure interface boundary components. In an October 15, 1985 Safety Evaluation Report (SER) for the Trojan Nuclear Plant's alternative shutdown capability, the NRC concurred with the PGE position that the PORVs would not spuriously operate upon isolation of DC power. This SER approved the PGE position that two proper polarity faults on the valve control circuits are not credible due to the low probability of such faults.

Emergency Fire Procedures, EFP-0 and EFP-1, will be revised by September 29, 1989 to require manual action to close the pressurizer PORVs and pull the switch to lock the pressurizer PORVS closed from the control room in response to any spurious PORV operation in conjunction with an indication of a fire in fire areas where PORV cables are routed. These fire areas are the Train B electrical switchgear room (C2), 61-foot elevation mechanical and telephone rooms (C6), cable spreading room (C7), 77-foot elevation mechanical room (C8), control room (Cll), electrical penetration area and main steam support structure (E1), and the Containment (Rl). These fire areas are provided with fire detection systems which alarm in the control room to ensure early warning to control room personnel.

EFP-0 will also be revised by September 29, 1989 to include a high- l priority manual action to deenergize the PORVs at the 125-V DC distribution panel for a fire in the control room and cable spreading room.

Spurious PORV operation prior to manual action to close the valves in the control room would require a sustained hot short for a total period of three minutes in order to potentially affect safe shut-down. Spurious operation of pressurizer PORVs for a total period of three minutes is not considered credible due to the low probability of a hot short sustained for this period of time.

FAD /3033W

.. 1

~

. Trojen Nuclear Pltnt John B. M1rtin i

Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 12 of 25 UNRESOLVED ITEM NO. 344/88-34 Availability of Process Instrumentation "The inspection team identified the concern that for a period of i 40 minutes while power is not available from the EDGs, diagnostic instru-mentation for Thot, Teold, and source range neutron flux monitoring is not available on the remote shutdown (C-160) panel. When power is avail-able from the EDGs, this instrumentation will be available at the Bailey l process display console adjacent to the C-160 panel. The team's concern i is that the lack of capability to read these process variables during this period was not recognized by the NRC staff during its 1985 review of this methodology.

"This appears to be inconsistent with Section III.L,2 of Appendix R and is considered an Unresolved Item (344/88-34-05) pending further NRC review."

PCE RESPONSE Alternative shutdown control and instrumentation is provided at the C-160 panel and the Bailey Net-90 distributed control system located i in the Remote Shutdown Station. Instrumentations and controls on the C-160 panel are powered from the Engineered Safety Features (ESF) power distribution system and are therefore battery backed. All required safe shutdown instrumentation is available at the C-160 panel with the exception of Reactor Coolant System (RCS) temperatures and neutron source range flux indications. Prior to installation of the new Remote Shutdown Station (RSS), RCS temperature and source range flux indication were to be obtained in the penetration area via local temperature indicators and local installation of a source range flux drawer. This methodology was reviewed and approved by the NRC as stated in the October 15, 1985 Safety Evaluation Report (SER).

The installation of the new RSS enhanced the design for the source range flux indication since it will no longer require local wiring of the spare source range flux drawer thereby making source range flux indication available in a shorter time frame.

The need for RCS temperature and source range flux indication is associated with restoration of auxiliary feedwater. Source range flux indication is required When a potential exists for increase in l reactivity. This would come about at a time when boron dilution is occurring or When RCS cooldown is initiated. Boron dilution will not occur since charging is via the refueling water storage tank (RWST)  ;

and RCS leakage paths are isolated for inventory control. RCS cool-down will not occur prior to initiation of auxiliary feedwater flow.

RCS temperatures are required to verify natural circulation and monitor RCS cooldown. These functions are required subsequent to

~

- 'Trojtn Nucisar Plcnt Jchn B. MIrtin i' Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 13 of 25 establishment of auxiliary feedwater flow. Since Emergency Diesel Generator (EDG) power to the ESF power distribution system becomes available prior to the establishment of auxiliary feedwater flow, the lack of battery backing to ensure continuous readout of these instruments will not impact the availability of the instruments for use in monitoring RCS parameters during shutdown. A revised alternative shutdown timeline analysis has shown that power to the Bailey Net-90 distributed control system will be restored in approximately nine minutes.

FAD /3033W i

1 t

I

f l

l

. ' Trojen NuclSer Plant John B. Mirtin j i

J Docket 50-344 April 28, 1989 I License NPF-1 Attachment A Page 14 of 25 UNRESOLVED ITEM NO. 344/88-34 Safe Shutdown Procedures "The inspection team and the licensee's staff performed a walkdown of the procedure for implementing alternate CR/CSR safe shutdown on August 25, 1988 at 11:00 a.m. During the walkdown, the inspection team assessed emergency lighting, communications effectiveness, operater familiarity with the procedure and equipment locations, and the availability of required tools, equipment, and other instructions. Procedural defi-ciencies were identified during the walkdown of Alternate Safe Shutdown Procedure EFP-1 as follows:"

1. "The NRC concern for the design basis parameters for safety-related valves and actuators, fail-safe positions and movement times with respect to a rapid or gradual loss of air was identified to the licensee in Region V Inspection Report 50-344/87-31. NRC Generic Letter 88-14 Information Notice 87-28 and NUREG 1275, provided the licensee with information on instrument and service air system problems and recommendations for action.

" Paragraph 2.B of Region V Inspection Report 50-344/87-31 documents the NRC concern that a gradual loss of instrument air test at Trojan has never been performed and safety-related backup accumulator testing has been accomplished on only a few selected accumulators.

This brings into question whether the backup accumulators at Trojan will perform their intended function.

"During the implementation of EFP-1, operators will be out of the control room with the minimum required process instrumentation provided on the (C-160) remote shutdown panel. This instrumentation does not include instrument air monitors or provisions for monitoring the f ail-safe positions of air-supplied safety-related or non-safety-related instruments. Since the reliability of the instrument air system design and backup safety-reinted accumulators is not supported by testing, and the licensee did not provide the inspection team with a supporting analysis, disabling the plant instrument air system in accordance with Step "B" of Attachment 3 to Procedure EFP-1 appears to potentially place the plant in a condition outside of those analyzed in Chapter 15 of the FSAR.

2. "The procedure was not updated to reficet new licensee identified changes.

3. " Operators were apparently unsure of some steps in the procedure (ie, the operator implementing attachment 1 of the procedure was unsure of Steps K and 0).

4. "An operator did not have a key to access the RHR pump room.

1. Trojen Nucisar Plant

, Jchn B. M!rtin Docket 50-344 April 28, 1989

[ License NPF-1 Attachment A l Page 15 of 25

5. "A sound-powered phone was missing in one location where the operator implementing Attachment I had to restore the nitrogen supply to the pressurizer [ sic] PORVs and make pressure adjustments in accordance with the shift supervisor instructions.

6. " Sound-powered phones are the designated means of communications, but they were not stored in secured cabinets.

7. "The emergency fire procedures stored in the "B" diesel generator room were found to be one revision behind the current revision.

8. "In keeping track of the status of the shutdown, it was necessary for the shift supervisor to work between the body of the procedure and the attachments. The team took the position that this process could be enhanced by providing tabs for the attachments and referencing the steps in the body of the procedure to the steps in the attachments."

The above items are considered an Unresolved Item (344/88-34-06) pending further license action and NRC review".

PGE RESPONSE

1. As described in a letter dated February 8,1989 from PCE to the NRC, a detailed action plan has been developed to address all actions requested in Generic Letter 88-14, " Instrument Air Supply System Problems Affecting Safety-Related Equipment". The detailed action plan identifies individual responsibilities and action completion dates to assure that the specific actions required by Generic Letter 88-14 are completed in a timely manner.

The design of the safety related portion of the instrument air system will be verified to be in accordance with its intended function by assessing instrument air system test data and assessing existing inservice test (IST) data and maintenance test data. The tests will be conducted during the 1989 refueling out-age. The scheduled testing includes sampling of the instrument air system to determine actual air quality being supplied to components. A rapid and gradual loss of instrument air test will also be conducted on portions of the safety-related section of the instrument air system. The rapid loss of air data will be obtained using two methods. One set of data will be obtained by testing the safety-related accumulators and associated check valves for: (1) four usin steam isolation valves, and (2) four  ;

auxiliary feedwater pump steam inlet valves, and (3) two pressur-ized power-operated relief valves. The other rapid loss of instrument air test data will be obtained using test results from the IST Program and a Maintenance Procedure. The gradual loss of

u

'Trojen Nuclear Plant John B. Martin J Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 16 of 25 instrument air tests will be limited to testing the same accumu- i lators and associated check valves included in the rapid loss of instrument air test.

Verification of the adequacy of maintenance practices, emergency fire procedures and training on the instrument air system, to ensure that safety-related equipment will function as intended on loss of instrument air, is being conducted and is scheduled to be completed by August 1989.

PGE will develop a program for maintaining proper instrument air quality when all design and procedure review and system testing has been completed, and after all significant modifications, if any, have bc.en completed.

For alternative shutdown, air-operated valves can be deenergized at their respective distribution panel. EFP-1 currently requires the entire instrument air system to be isolated, in addition to deenergizing the valves at the distribution panel. This action to isolate the instrument air system will not be necessary since air is vented from the valves when deenergized. EFP-1 will be revised by September 29, 1989 to eliminate the operator action to isolate the instrument air system.

2. Identified changes to EFP-1 have been completed.

3. Operator training on EFP-1 has been conducted, and enhancements made to the Operator Training Program. Refer to Unresolved Item 344/88-34-07 for further discussion.

4. To assure that the Shift Supervisor has the keys to the Train B RHR Pump Room when implementing the alternative shutdown method-ology outside the control room, EFP-1 will be revised to require the Shift Supervisor to obtain the set of keys to the Train B RHR Pump Room prior to evacuating the control room. This revision to EFP-1 will be completed by September 29, 1989.

5. A sound-powered phone jack has been installed and a sound-powered phone is stored at the nitrogen supply for the steam generator PORVs.

6. Cabinets have been provided for the storage of sound-powered phones, except for the Remote Shutdown Station. A Maintenance Request has been initiated to provide a storage cabinet at this location.

7. The most current revision of the EFPs are stored at safe shutdown stations.

8. The need to enhance the EFPs by addition of tabs and referencing steps in the body of the procedure to the steps in the attach-ments, will be evaluated as part of a future update to the EFPs.

FAD /3033W

i l

i

. Trojan Nuclear Plant John B. M!rtin

. Docket 50-344 April 28, 1989

. License NPF-1 Attachment A Page 17 of 25 UNRESOLVED ITEM NO. 344/88-34 Operator Training l

"The inspection team reviewed operator training records and interviewed a 4 l

selected sample of plant operators to determine the extent that operators had been trained on the safe shutdown modifications and procedures ,

implemented by the licensee for compliance with Appendix R. In general, i the inspection team found the operators to be knowledgeable of the safe shutdown method. Appropriate classroom sessions had been conducted.

Ilowever, deficiencies in operator training were identified as follows:"

1. " Training did not include periodic scheduled walkdowns of the emergency fire procedures. Only one shift of plant operators had performed a walkdown of the procedures.

2. " Operators appeared to be unaware that they would be without certain diagnostic instrumentation (Thot, Teold, and source range) for approximately 40 minutes while implementing the alternative safe shutdown methodology.

3. "The operators interviewed indicated that they had a lack of confi-dence in the use of the Bailey computer system for alternate safe shutdown. They also indicated that they had no knowledge of the mounting brackets provided for portable emergency lights that the method relies on to accomplish certain required manual operator actions.

"During the inspection, the licensee indicated that the operator training program would be reviewed and immediate corrective actions taken where required.

"The above appears to be inconsistent with Section III.L.1 of Appendix R and is considered an Unresolved Item (344/88-34-07) pending further NRC review."

PGE RESPONSE

1. The requirement for all operating crew members to complete an annual walk-through of the Emergency Fire Procedures (EFPs) has been established and communicated to the Operations Department.

This training has been completed for the 1989 training cycle.

Future scheduling of this activity will be conducted by the Operations Department and, upon receipt of completion documen-tation, it will be tracked by the Training Department.

L i

. *Trajcn Nuclocr Plcnt' John B. Martin

,. Docket 50-344 April 28, 1989 License NPF-1 Attachment A I Page 18 of 25 l

2. Information pertaining to the availability of diagnostic instru-mentation while implementing alternative shutdown outside the control room has been included in the Remote Shutdown Station (RSS) and EFP lesson plans. All licensed personnel attended these lectures during the 1988/89 Retraining Sessions and have received this information.

3. Completion of the issued Job Performance Measures concerning the operation of the Bailey Net-90 distributed control system (DCS) for miscellaneous tasks, EFP-1 performance, and Off-Normal Instruction (ONI)-17, " Control Room Inaccessibility" portions is expected to increase operator experience and confidence using this system. An updated revision of these materials to more closely reflect actual conditions is expected to be completed in the first half of 1989. Full operational performance using the simulator equipment will address inadequacies identified by the Operations Department.

Information concerning emergency battery light mounting brackets has been included in lesson plan materials. Training has been enhanced to instruct the operators on use of the portable light-ing units and where the lighting' units and their mounting brackets are located.

FAD /3033W L - -- - - - _ _ _ _ -

L..

- ' * ~ Trojan Nucisar. Pitnt John B. Martin Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 19 of 25 OPEN ITEM NO. 344/88-34 Use of Fire Pumps As a Backup to the Service Water System l "For the case where the. fire pumps are used as a backup to the Service-Water System, the licensee indicated that the diesel fire pump may be scheduled to be out-of-service for maintenance. Therefore, during this period, it will not be available for this purpose. During-the inspec .

tion, the licensee was evaluating what (if any) compensatory measures would be implemented during this period.

"The licensee's internal Memorandum No. RQR-Q15-88, dated August 31, 1988 .which was forwarded to the NRC, Region V, concluded that for a fire in Fire Areas M3 (Manhole 3), M4 (Manhole 4) and II (Service Water Pump Room), both fire pumps would be needed to supply the 4450 gallons per minute (spm) that would be needed as a backup water supply for the Service Water System. Therefore, during the period that either fire pump is out-of-service, the licensee's internal memorandum indicates that fire patrols will be established to detect and mitigate a fire in any of the three fire areas to prevent the loss of all three service water pumps 'and the subsequent need to use the fire pumps as a backup to the service water pumps. The memorandum indicates that these measures will continue until an in-depth review is completed or relief is granted by the NRC.

"This is considered an open item (344/88-34-08) pending further licensee and NRC action".

PCE RESPONSE:

PGE has reviewed the compensatory measures implemented for fire pump outages and has determined that hourly fire patrols will continue to be established and maintained until the out-of-service fire pump is returned to service. The use of hourly fire patrols to detect and mitigate fires in fire areas M3 (Manhole M3), M4 (Manhole M4), and 11 (service water pumps room) is the most feasible means of ensuring the availability of service water during fire pump outages.

FDP/3033W l

> Trojan Nuclect Plant John B. Mirtin Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 20 of 25 UNRESOLVED ITEM NO. 344/88-34 Associated Circuits

1. "High-Impedance Fault Concern - The high-impedance fault concern arises in the case where multiple fire-induced faults exist as loads on a power supply required to achieve safe shutdown. Such faults are postulated to be of a value which is just below the trip point of the individual load circuit protective devices. The occurrence of a i sufficient number of such faults within a given fire area may result in a trip of the power supply feeder breaker and therefore may caure the loss of the required power source.

"In response to this concern, the licensee states in Section 3.5.5.2 j of the Trojan Fire Protection Plan (PGE-1012) that 'due to the low probability of occurrence, the coordination for multiple high-impedance circuit faults was not considered within the scope of its review'.

"To further highlight this concern to the licensee, by letter dated January 24, 1985, the NRC staff requested that the licensee provide additional information regarding this concern. The staff requested that the licensee provide either, 'A quantification of the argument which demonstrates that if all the load side circuits in a common fire area were to receive a fault, the coordination ensures that the power source is not lost'; Or, 'A description of how a tripped breaker on the power supply resulting from multiple faults would be identified to the operators and the corrective action that would be taken. Also, verify that these actions can be taken prior to the plant entering a nonrecoverable state.' This positicn was repeated j by the staff and given to all licenseo's as guidelines for addressing i this concern in Section 5.3.8 of Generic Letter 86-10. )

l "By letter dated March 6, 1985, the licensee responded to the staff's request and basically reiterated their previous position that due to the low probability of occurrence, further consideration of multiple high impedance faults was unwarranted. Based on these statements by the licensee, the NRC staff concluded in SER Supplement 5, dated j I

October 15, 1985, that operatorn could identify the location of a fault and the status of the breaker where an appropriate action could be taken to restore power.

"The inspection team's review disclosed that the licensee had not  !

performed an analysis for this concern and did not have procedures in j place to direct operator actions to mitigate the potential affects of such occurrences. The NRC staff stated in Supplement 5 to the SER that the licensee would rely on operator actions to restore the l affected buses. However, due to the lack of procedural guidance, it )

was not clear to the inspection team how operators would identify

4 l

1 l

l t

s Trojan NucisEr'Plcnt John B. Mirtin

.. Docket 50-344 April 28, 1989 i License NPF-1 Attachment A Page 21 of 25 l

those power sources which may be affected by the occurrence of high-impedance faults or how operators would be aware of which loads are required to be reloaded onto the bus in a timely manner during bus restoration."

2. " Circuit Breaker and Relay Maintenance - The inspection team's review ,

disclosed that the licensee currently performs circuit breaker maintenance on a rotating outage basis such that all breakers are tested at a frequency not to exceed six years for 480-V MCCs and five i years for 120-V ac and 125-V de breakers.

]

"The NRC staff position stated in the clarification letter to Generic Letter 81-12 states that, 'In order to take credit for the coordinate protection of molded-case circuit breakers, the devices must be periodically exercised and inspected for case of operation, and on a refueling outage basis, a sample of these breakers must be tested to verify the drift is within the allowed design limits.'

"While the licensee currently has an established maintenance program for molded-case circuit breakers, the inspection team identified the concern that this program is relatively new, having been implemented by the licensee just prior to the 1988 refueling outage. Apparently, due to limitations of scheduling and manpower since the implemen-tation of the program, all molded-case circuit breakers of concern have not yet been subjected to the program's maintenance and testing requirements. Therefore, the maintenance activities described in the NRC staff's clarification letter to Generic Letter 81-12 have not been performed for all molded-case circuit breakers relied on to provide coordinate circuit protection for fire-induced faults.

" Based on the inspection team's review, the licensee's apparent lack of adequate procedures to address required operator actions during bus restoration from potential damage due to multiple fire-induced high-impedance faults and the lack of verification of molded-case circuit breaker operability, places the adequacy of the licensee's resolution to the common bus concern in question.

"This appears to be inconsistent with Sections III.G.3 and III.6.7 of Appendix R, and is considered an Unresolved Item (344/88-34-09) pending further NRC review."

PGE RESPONSE

1. The position taken by the Trojan Nuclear Plant, as doeuraented in Section 3.5.5.2 of PGE-1012 (Volume II) and accepted by the NRC in the October 15, 1985 SER for the alternative shutdown capability, is that coordination for multiple high-impedance circuit faults does not need to be considered because of the low probability of such an occurrence.

1

.?.

• Trojan Nuclear Plant John B. Martin Docket 50-344 April 28, 1989 License NPF-1 Attachment A l Page 22 of 25 This position is supported by detailed analyses performed by other Plants. For example, Bechtel Power Corporation completed a High-Impedance Fault Calculation, Bechtel Calculation E4C-079, for San Onofre Nuclear Generating Site, Units 2 and 3. This calculation showed that a multiple high-impedance fault causing the main supply breakers on a bus to trip and remove the bus from service is not credible.

Irrespective of the above, cautionary guidance will be provided in the Emergency Fire Procedures by September 29, 1989 to make operators aware that a high-impedance fault could cause the loss of safe shutdown power supplies, and that operators will need to strip the non-safe shutdown loads and reclose the feeder breaker. ,

2. The NRC's recommendation for testing of molded-case circuit breakers was issued in a clarification letter to NRC Generic Letter 81-12.

The maintenance and functional tests of safety-related molded-case circuit breakers are completed on a rotating refueling outage basis consistent with the NRC recommendation. The periodic testing to date has shown a failure rate of less than 2 percent.

In addition to periodic testing, breakers undergo a pre-installation verification test to compare the manufacturer data with the actual data obtained from the tested breaker. Pre-installation testing has also shown a failure rate of less than 2 percent.

It should also be noted that an extensive program is underway, in response to NRC Bulletin 88-10. " Nonconforming Molded-Case Circuit Breakers", to assure the traceability of safety-related molded-case circuit breakers to the circuit breaker manufacturer. This program is described in a letter dated April 14, 1989 from PGE to the NRC.

Based on the breaker testing programs in-place and those efforts underway in response to NRC Bulletin 88-10, PGE does not consider any additional baseline testing to be necessary.

FAD /3033W 2

. Trojan Nucisar Plant Jchn B. M:rtin Dockst 50-344 April 28, 1989 l License NPF-1 Attachment A Page 23 of 25 I

l l

UNRESOLVED ITEM NO. 344/88-34 High-Low Pressure Interface Concern "The high-low pressure interf ace concern exists where there is the poten-tial for a single fire to induce the spurious operation of redundant motor-operated isolation valves. Such an event may result in a fire-initiated unisolable loss of coolant accident (LOCA) outside of the primary containment due to the inability of a low-pressure system to withstand the substantially higher primary system pressure.

"The licensee's high-low pressure interface analysis did not take credit for the normally open motor-operated PORV block valves [M08000A and M0800B(sic)] which are connected in series and located upstream of the PORVs. The primary purpose for the licensee's position is that following a CR/CSR fire and control room evacuation, the licensee's procedures require the shedding of offsite power and the disabling of both emergency diesel generators. Therefore, AC power would not be available to operate these valves.

" Based on the inspection team's review of applicable schematic and con-trol wiring diagrams, it appears that the occurrence of a single hot short within either the CR/CSR could result in the spurious operation of either of the two pressurizer PORVs. To prevent or mitigate this con-dition, the licensee's Procedure EFP-1 directs operators to open the DC supply breakers for the PORVs. The procedures state that this action must be accomplished within five minutes. This preventive or mitigating action appears to be intended to address the case of a large exposure fire requiring control room evacuation. However, the team determined that a small fire postulated to' occur in the CR/CSR that initiates opening of the PORV prior to propagating into a large exposure fire and an immediate fire alarm, has the potential to significantly impact the five minute time line established by the licensee for this condition.

Furthermore, the occurrence of multiple fire induced hot shorts on the POP.V cables was not considered in the licensee's analysis.

"This appears to be inconsistent with Sections III.C.3 and III.L.7 of Appendix R."

PCE RESPONSE The response to this item is included in the response to Unresolved Item 344/88-34-04.

FAD /3033W

.Trojen Nuclear Plant John B. Martin Docket 50-344 April 28, 1989.

License NPF-1 Attachment A Page 24 of 25 VIOLATION NO. 344/88-34 Emergency Lighting "Section III.J of Appendix R to 10 CFR 50 required the licensee to install emergency lighting units with at least an eight-hour battery power supply in all areas needed for operation of safe shutdown equipment and in access and egress areas thereto. Open Item 344/83-18-11 identi-fled several NRC concerns regarding the adequacy of the licensee's lighting provided to support safe shutdown in 1983.

"The concern that no emergency lighting was provided in the access and egress route between the Turbine Building and Intake Structure was iden-tified in Open Item 344/83-18-11. As of this inspection, this condition continues to exist. In addition, a subsequent analysis by the licensee disclosed that for a fire in Fire Areas II, M3, and Me, equipment needed for safe shutdown is located near the south wall and on the west side of the Intake Structure. However, the required emergency lighting was not provided in support of this safe shutdown method. Additionally, the required emergency lighting was not provided in the circulating water pump pit at the pump discharge and suction valves and at the fire hydrant east of the circulating water pump pit where other operator actions are required for safe shutdown.

"Instead of the required fix eight-hour battery power supply emergency lighting, the licensee elected to provide four six-hour hand-held port-able lighting units for operator use in these areas. Two of the hand- )

held portable lighting units are stored in the control room and two are stored in a radiologically controlled access area. The portable units stored in the control room may be left there during a control room evacu-ation because there is no procedure requiring their use upon evacuation.

These portable units are electrically charged by a non-vital electrical receptacle which would de-energize upon a loss of AC power. In this case, the portable lighting units automatically provide emergency DC lighting for the control room. During this period, their six-hour capacity would be reduced. In the event that these portable lighting units are used by operators in support of safe shutdown, they are rather bulky, so the licensee installed mounts for them at the locations where they are expected to be used in support of safe shutdown.

"During interviews with operators, the inspection team learned that the j operators interviewed had no knowledge of the mounts for the portable ['

emergency lighting units. The mounts are constructed with unistrut channels, and when the inspection team requested licensee personnel to demonstrate installing the units into the mounts, the individuals encountered difficulty in completing this task due to the construction of the mounts. This request was made by the inspection team during daylight hours. If this task had to be performed in the dark, the inspection team questioned whether or not it could be achieved.

O____ -_ -_ ___ _

' Trojtn Nuclear Plant John B. Mirtin Docket 50-344 April 28, 1989 License NPF-1 Attachment A Page 25 of 25 "Although the use of portable lighting units is an alternate approach that the staff considers viable on a case-by-case basis, the licensee did not request and was not granted an exemption from the NRC for the lack of fixed eight-hour battery-powered emergency lighting units in support of safe shutdown. In addition, the inspection team questioned the adequacy of installed emergency lighting in the 'A' switchgear room, Turbine Building, and the 45-foot and 61-foot elevations of the Auxiliary Build-ing during the walkdown of alternative safe shutdown Procedure EFP-1.

"The licensee's failure to install eight-hour battery-powered amergency lighting units in support of safe shutdown is considered an apparent violation of Section III.J of Appendix R to 10 CFR 50 (344/88-34-11)."

PGE RESPONSE An exemption request for use of portable six-hour battery-powered emergency lighting units was submitted to the NRC in a letter dated December 30, 1988 in response to the Notice of Violation dated October 20, 1988. Training has been enhanced to instruct the oper-ators on use of the portable lighting units and where the lighting units and their mounting brackets are located.

FAD /3033W.0489 1

l

E r.

TrojanNuclearPlant John B. Martin

,; Dockst 50-344 Aprl1 28, 1989

' License NPF-1 Attachment B Page 1 of 2 OVERVIEW OF THE REVISED ALTERNATIVE SHUTD0lm METHODOLOGY

{ ,

Current Methodology Revised Methodology Conments

1. Control room actions prior control room actions prior to The Train B CCP will be available to evacuation: evacuation: for alternative shutdown. See response in Attachment A for

• Trip the reactor.

• Trip the reactor. Unresolved items 344/88-34-01, l 344/88-34-03, 344/88-34-04. See l also Attachment C.

' Trip the Main Steam *Close the pressurizer power-Isolation Valves. operated relief valves (PORVs).

l

• Trip the centrifugal charging pumps (CCPs).

I I

• Trip the emergency diesel generators (EDGs).

• Trip the Main Steam Isolation Valves.

2. Intentional loss of ac AC Power to Engineered Safety See response in Af . Nnent A for power for 40 minutes to Feature (ESF) loads will be de- Unresolved item 344/8b-34-01.

all 12.47-kV and 4.16-kV energized at the 4.16-kV level buses, for approximately five minutes.

AC power to non-ESF loads wil'. not be intentionally deenergized except to certain loads of concern (such as reactor coolant pumps, pressur-Izer heaters and steam generator blowdown isolation valves).

3. Additional operators will No credit taken for additional See response in Attachment A for be available to assist operators. Unresolved Item 344/B0-34-01.

completion of manual actions after one hour.

4. Reactor Coolant System RCS makeup requirements revised See response in Attachment A for (RCS) makeup restored based on (a) seal leakage rate of Unresolved Item 344/88-34-02.

within one hour based on 3 gpm/ pump for first 10 minutes, normal seal leakage rate then increase to 21 gpm/ pump of 3 gpm/ pump and assum- thereafteruntilsealinjection ing a pressurizer PORV charging is reestablished, and j

. spuriously opens for five (b) the pressurizer PORV spuri. ]

minutes. ously opens for 3 minutes. Time l constraint for RCS r:akeup reduced i from 60 to 13 minutes.

l

TrojanNucigarPlant .lihn B. Martin Docket 50-344 April 28, 1989 License NPF-1 Attachment B Page 2 of 2 i DVERVIEW OF THE REVISED ALTERNATIVE SHUTDOW METHODOLOGY Current Methodology Revised Methodology Conments

5. Operator sent to EDG rooms Operator will trip the Train B EDG See response in Attachment A for tomonitorjacketwatertem- in the control room if it spuri- Unresolved Item 344/88-34-03. 3 perature and trip / decouple ously starts prior to evacuation, j theEDGswhenthejacket A high-priority backup action will i water temperature reaches also be taken to trip the Train B 190'F. EDG at the local control panel in the EDG room.

6. Pressurizer PORVs are de- Operator will close the pressurizer A double-pole switch has been in-energized at the 125-V DC PORVs in the control room prior to stalled for PCV-456. An equiva-distribution panel. As an evacuation. A high-priority back- lent switch will be installed interim measure, pressuri- up action will also be taken at for PCV-455A. See response in zer PORV block valves are the 125-V DC distribution panel. Attachment A for Unresolved closed in the control Items 344/88-34-04 and room upon fire alarm 344/88-34-10. l Indication.

7. RCS Thot and Tcold AC power will be restored to the Unavailability of RCS temperature indicators and source range Bailey Net-90 computer system indicators and source range range neutron flux monitors within 10 minutes. Therefore, neutron flux monitors was linked on the Bailey Net-90 com- RCS temperature indicators and with the loss of AC power for puter system will not be source-range neutron flux moni- 40 minutes. See response in available for 40 minutes, tors will be available within Attachment A for Unresolved 10 minutes. Item 344/88-34-05.

8. The NRC's October 15, A caution statement will be added High Impedance Fault Analysis for 1985 Safety Evaluation to the Emergency Fire Procedures San Onofre Units 2 and 3 showed Report states that high- stating that a high-impedance fault high-impedance faultt are un-Impedance faults are un- could cause the loss of safe shut- likely. See response in Attach-likely and that operators down power supplies, and that opera- ment A for Unresolved Item are capable of identifying tors will need to strip the nonsafe 344/88-34-09.

the location of a fault shutdown loads and reclose the and take appropriate feeder breaker.

action to restore power.

KLB/rc 3033W.0489

.. Trojen Nuclear Plcnt John B. M1rtin I s Dockst 50-344 April 28, 1989 License NPF-1 Attachment C 1 Page'1 of 5 DESCRIPTION OF THE COOLING REQUIREMENTS OF THE POSITIVE DISPLACEMENT CHARGING PUMP AND CORRECTIVE ACTION PLAN AND SCHEDULE

Background

By letter dated June 8, 1988, Portland General Electric Company (PGE) described two scenarios involving spurious closure of the outlet valves for the Volume control Tank (VCT), MO-112B and MO-1120, which could .j result in the loss of both Centrifugal Charging Pumps (CCPs), Also ]

described in the referenced letter was a scenario involving fire-induced j spurious closure of the CCP miniflow isolation valves, MO-8110 and I MO-8111, concurrent with the loss of offsite power which could result in damage to both CCPs. Nine fire areas were identified in which both CCPs l j

could be damaged. For a fire in these areas, the positive displacement

]

chargi g pump (PDF), P-217, was to be used as an alternate charging j source to provide Reactor Coolant System (RCS) makeup.and Reactor Coolant 1 Pump (RCP) seal cooling. Table 1 summarizes in matrix format, for each relevant fire area, how this problem will be corrected.

During the reevaluation effort of the alternative shutdown methodology..a problem was discovered with the PDP that severely restricts its use as an alternative charging source. This problem and its proposed resolution is described in detail below.

Description The PDP requires cooling water for the lube oil cooler and the hydraulic coupler. The cooling capability is only provided from the Train 'A loop of the Component Cooling Water-(CCW) System. The CCW system is a closed cycle system which transfers its heat to the service water system (SWS) via the CCW heat exchanger. Fire areas A4, C7, Cll, T1 and T8, will only have the Train B CCW and SWS systems available. For these fire areas, credit was taken for aligning the CCW and SWS swing pumps (P-2100 and P-108C, respectively) to the Train A CCW and SWS loop. Both the swing pumps and the normal Train B CCW and SWS pumps were to be powered from the Train B'4.16-kV switchgear, A2. Manual operation of locked close valves was required to align the CCW and SWS swing pumps.

During the reevaluation effort of the alternative shutdown methodology, the configuration of the PDP and its cooling requirement was reviewed.

This review disclosed that an electrical interlock exists in the Train B switchgear, A2, between the swing pump and the normal pump. A similar electrical interlock exists in the Train A switchgear, A1. The elec-trical interlock prevents both the swing pump and the normal pump from operating concurrently on the same switchgear. Because of the electrical interlock, CCW and SWS cooling requirements for the PDP will not be available for fire areas A4, C7, Cll, T1, and T8.

n l

,, ' Trojan Nucicer Plent John B. Mr.rtin

" Docket 50-344 April 28, 1989 License NPF-1 Attachment C Page 2 of 5 A preliminary review was performed to determine the impact on safe i shutdown should a modification be implemented to the circuitry of the j Train B switchgear to bypass the electrical interlock. The calculation l on RCS makeup requirements was revised to address the concerns raised during the Nuclear Regulatory Commission (NRC) audit on PGE's Fire Protection Program (see Attachment B, Unresolved Item 344/88-34-02). The time constraint for RCS makeup was reduced to 13 minutes, assuming a pressurizer power-operated relief valve (PORV) spuriously opens for a ,

period of three minutes in conjunction with normal RCP seal leakage rates. The alternative shutdown timeJine was reviewed to determine if i the operators are capable of manually aligning both the CCW and SWS swing pumps and manually racking-in the PDP breaker in the switchgear within the 13-minute time constraint. As a result of the discovery of the pump interlock, it was determined that the required manual operations to utilize the PDP for alternate shutdown would exceed the 13-minute time requirement for RCS makeup. The results of the preliminary review show {

that the most feasible way of providing charging capability is to protect j the Train B CCP, P-205B, by tripping P-205B in the control room prior to evacuation. The probability of fire damage to the circuits for the VCT outlet valves and/or the CCP miniflow valves causing an unacceptable spurious signal, before action could be taken to trip the CCPs, is extremely low. This action ensures that P-205B will be available for charging.

Fire areas A4, T1 and T8 utilize normal shutdown means (as opposed to alternate shutdown). Because the circuits for the pressurizer PORVs are not routed in these fire areas, spurious opening of the pressurizer PORVs does not need to be postulated. The revised calculation on RCS makeup requirements determined that RCS makeup will not be required for 38 minutes, if a pressurizer PORV does not spuriously open. Thus, more time will be available for operators to align both the CCW and SWS swing pumps and rack-in the breaker to the PDP. The preliminary review has shown that operators will be capable of completing the required manual operations within 38 minutes. However, cables for the Train B CCP, P-205B are not routed in fire areas A4, T1 and T8 (see Table 1). There-fore, only a loss of offsite power would start P-205B. The probability of the VCT suction valves and/or the CCP miniflow valves spuriously closing (due to fire damage to the circuit, concurrent with a loss of offsite power or fire damage to the VCT and RWST suction valve interlock circuit), prior to an operator action to trip the CCP in the control room, is low enough to credit an action to trip P-205B in the control room.

Corrective Action

1. As a result of reevaluating the alternative shutdown methodology, an l operator action to trip the Train B CCP, P-205B, in the control room I prior to evacuation will be credited in order to ensure the avail-ability of the pumps. A high-priority backup action will also be 1

l

l .'

l , ., * *Trojen Nuclser Plant John B. Martin .

l' *' Docket 50-344 April 28, 1989

. j License NPF-1 Attachment C j Page 3 of 5 l taken at the 4.16-kV switchgear to trip the pumps. Emergency Fire l Procedure, (EFP)-0, will be revised to incorporate the control room action to trip the Train B CCP and the high-priority backup action to trip the Train B CCP at the 4.16-kV switchgear.

2. Further review is necessary to evaluate the safety impact of over-riding the electrical interlock in the Train B switchgear. An interim operator action for fire areas A4, T1 and T8 will be ,

implemented to trip the Train B CCP, P-205B, in the control room.  !

EFP-0 will be revised to incorporate the interim action. l,

3. The feasibility of supplying B-Train cooling to the PDP will be evaluated. Depending on the outcome of this evaluation, modifica-tions may be implemented to provide B-Train cooling, thus enhancing the availability of charging using the PDP.

Corrective Action Schedulo Corrective action schedules for the deficiencies described above are proposed as follows:

1. Prior to September 29, 1989:

Revise EFP-0 to include an operator action to trip the Train B CCP, P-205B, in the control room for a fire in fire areas A4, C7, Cll, T1 and T8. EFP-0 will also be revised to include a backup operator action to trip P-205B at the Train B switchgear, A2, for a fire in fire areas C7 and C11.

2. Prior to October 1. 1989: l Complete feasibility evaluation of (a) overriding electrical inter-locks in the Train B 4.16-kV ESF switchgear, and (b) supplying Train 1 B cooling to the PDP.

FAD /mr 3033W.0489

e e e n r r r r e6 r

pM r s a a . i i i n in ,i) u q

u q

u q

i n .

o f o l 1 f oa C i t i A e e e i d rt e x r r r rt A u a e u a u a ovf A e e e e o v a h e r n n n h ee

- l 5f A o o o - l r 1 E 2 o( N N N 1 E A p .

a . r . A) r 8E o e y . t W 2 ef Bl nl m r n n r1 rt 5 b ip o A oe o i1 oa) . 0 a a p n o f d i c - r9 C 2l r u ir n ye .

n t D D aA - i T s i S u . eM p( d P a ( 8l al p a P l At d e n v r 5 o r p e4 M o 5 a r n s a a r a 7 e 0 r T pd A U s 0 c oa e o 1 w 2 t u n P e 2 of e r8 e 2 o - n es i a R - l ,d a2 A b - p P o d e G P e s Ci 1 5 P p

c i r 7 r N rt v e1 0 l B v e 1 A I e s r u n r i d o r -

iD 2

l i

e s r ie rh ow r o 2

- f G U o c aP f M P w U o T t P p P o R

A H

C eo L l t A b .

G

. ad4 . . .

U n n c eA n n n F i o i o  ; ad t

i o i o i o

I R m t t d c n t t t T e c c e oa c c c N

l u u vl u u u E b s s l e3 s s s C o o rA r P P s P P P N P C C eC s C C C O C C r2 a C C C 1 e f f m1 r f f f T

C o o le - A o o o A D P s s b M e s s s M s s o r s s s I o L o rpt oi F L o L o

L o

L N

O 1 ' '

1 I 1

T 1 (

E A B X X X X X L R -

B E D A P M T O 0 .

1 1 S 1 A e U 8 X X X O aB I

R e U r P E X X X X X A S n R 2 ii O

1 1

D X X X pa F - a X D r wp I M C X X X X R e T

A r M B X X X i e f k N r O ) u I

T B ob U 7 h L

1 n -

O 2 i 3i S - a X E P r X X h o R T t s

( i

/

M wr E df L X X e B 5 B O 0 2

d i e R - v r P P A X X oip r

pq t

e yr e s l nf n t t g

g

, o ot n g i 7 i e e n n n n n t 7 t e r r ;

o i t i t i a v h af v u y i d et d e d t

p l e e l e l l e g l e7 cl i f e i f i el u el 7 ai u f u u uE o uE A o e r B5 B5 B m F r F d 5 r c 7 o / ,h / , n ys a 0 s A s yn7 yn yo ys t 2 e e r o r o r R r g r g eD aih ai a a int a int Pb r i t g i t i A i e i e i l a u l a l 5 l d e l d e r F i v o i v i 0 i l f i l f o x e r x e x2 xi xi f l ul h ul u - u u 5 u u1 A E t A E A P A B 4 A B 6 e l

b a

o. a C C N l 2 3 4 4 A A A A A ) )

1

( (

l

e e r e r r r r pM i r i i i i m u i u u u u o q f . q q q q C e 2 e e e e R rC r r r r u

e oa e e e e n h e n n n n o - r o o o o N 1 A N N N N A) y s t . . .

n

. e inl r n m m m s n p A oe o o o a e o a p r u n f d n

n o i r i R n o n o i r n o i r f

5 it u

T s

( ialyep . B l B l B l B l r

e P l r r p e2 5 o 5 o 5 o 5 o w M o 7 e T pd C 0 r 0 r 0 r 0 r o U s 1 w u n 2 t 2 t 2 t 2 t p P e 2 o

- p e si a - n - n - n - n R d e P o P o P o P o l G P i r7 r p

c p C

p c p c o N B v e1 A r e o w2 e e i e e t I

G s r U o P r o -

pP o f irh T t irh T t rh T t irh T t n

o R c A

H d C d d d d d n L n n n n n a A a a a a a G . n U n n n n n n e o o o o o o p f

I i i i i o R m t t t it it t T e c c c c c c y N

l u u u u u u l b s s s s s s l E

C o r P a

P P P P P m N P C C C C C C r O C C . C . C . C . C . o w w w w w n f f o f o f o f o f o T

C o o l f

ol f

ol f

ol f

ol f i s

A s s i si si si si P s s n s n s n s n s n 7 M o o i oi oi oi oi 1 I

L L m L m L m L m L m 2 N

- O I

T A

1 1

1 8

X X X P

f r

o R -

E O r P M e O 0 k S

1 a U

1 X X X X e 8 r O B I

R U E X X X X P n S o R 2 i

- F O

1 1

D X X X X X t a

r X D e p

M C X X X X R

I l

o T

A r

M B X X X X X e k

a N e O ) r I

T B b l 7 l* 1 n d n 0 2 i i l io 5

E R

- a P r T

X X

X ot s a r

_ (

/

M rop e

E f o

_ L

_ B 5 B X X X 0 d p e m

_ O 2

_ R - r u

_ P P A X X X X i p

_ u

_ q s eu r

m o

touior

_ g m r o r np

_ n n o ,a ,R , ,t ,a s

- o i o g e g g g e g e  ;

_ i d R n g n g n n et n g yt t l ih i n i f e ih l n id n e p iC d c di d e d c

_ ai u l t l d l n u l 5 f l t ov

_ e r B d ii i a ic i4 ii e

_ r c n u w u e u o u 3 u w s r e p

_ A s ya B S B r B R B n9 B S e r p o l eD aB l B l S l l eih eA b o

_ r i 2 o o oo inta gu n n

at c

l 1 r n r e r r i

b v o ibi m d F i x -

uD 1 t n ia m o r o o

t l nb o a t t n n o o r e r ul h r a o u r o l

ov e

_ A M C T R C C C C T E t T T R r o t m n e

. o r o 1 C N 9 A

2 C

7 C

1 C

1 T

8 T )

2

(

_