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IR 05000352-05-302, 05000353-05-302, 05000277-05-302 and 05000278-05-302; June 13-15, 2005; Limerickand Peach Bottom Stations; Initial Limited Senior Operator Licensing (Lsro) Examination. Three of Three Applicants Passed the Examinations
	ML052160317
Person / Time
Site:	Peach Bottom, Limerick
Issue date:	08/04/2005
From:	Conte R; Division of Reactor Safety I
To:	Crane C; Exelon Generation Co, Exelon Nuclear
Shared Package
ML042440144	List: IR 05000277/2005302; IR 05000443/2005301;
References
	IR-05-302
	Download: ML052160317 (34)
	v • d • e

August 4, 2005 Mr. Christopher President and CNO Exelon Nuclear Exelon Generation Company, LLC 200 Exelon Way KSA 3-E Kennett Square, PA 19348 SUBJECT: LIMERICK GENERATING STATION UNIT 1 AND UNIT 2 AND PEACH BOTTOM ATOMIC POWER STATION UNIT 2 AND UNIT 3 LIMITED SENIOR REACTOR OPERATOR INITIAL EXAMINATION REPORT NOS. 05000352/2005302, 0500353/2005302, 05000277/2005302 AND 05000278/2005302

Dear Mr. Crane:

This report transmits the results of the Limited Senior Reactor Operator (LSRO) licensing examinations conducted by the NRC during the period of June 13-15, 2005. The examinations addressed areas important to public health and safety and were developed and administered using the guidelines of the Examination Standards for Power Reactors (NUREG-1021, Revision 9).

Based on the results of the examination, all three applicants passed all portions of the examination. The applicants included three LSROs. Examination results indicated that the applicants were well prepared for the examination. Mr. J. Caruso, Chief Examiner, discussed performance insights observed during the examination with Mr. C. Rich on June 15, 2005. On July 21, 2005, final examination results were given during a telephone call between Mr.

J. Caruso and Mr. C. Rich of your training organization. The issuance of license numbers will be delayed pending receipt of written certification from Exelon to the U.S. Nuclear Regulatory Commission stating that the applicants have acquired all experience for which they were previously granted waivers.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). These records include the final examination and are available in ADAMS (Master File - Accession Number ML042440215; LSRO Written - Accession Number ML052060104; LSRO Operating Sections A and B- Accession Number ML0052060114. Note for LSRO examinations there is no Operating Section C. ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Mr. Christopher Should you have any questions regarding this examination, please contact me at (610) 337-5183, or by E-mail at RJC@NRC.GOV.

Sincerely,

/RA Richard J. Conte, Chief Operations Branch Division of Reactor Safety Docket Nos. 50-352, 50-353, 50-277, 50-278 License Nos. NPF-39, NPF-85, DPR-44, DPR-56 Enclosure: Initial Examination Report Nos. 05000352/2005302, 05000353/2005302, 05000277/2005302 and 05000278/2005302 cc w/encl:

Chief Operating Officer, Exelon Generation Company, LLC Site Vice President, Peach Bottom Atomic Power Station Plant Manager, Peach Bottom Atomic Power Station Regulatory Assurance Manager - Peach Bottom Senior Vice President, Mid-Atlantic Operations Vice President - Operations Support Vice President - Licensing and Regulatory Affairs Director, Licensing and Regulatory Affairs, Exelon Generation Company, LLC Vice-President, General Counsel and Secretary J. Bradley Fewell, Assistant General Counsel, Exelon Nuclear D. Quinlan, Manager, Financial Control, PSEG R. McLean, Power Plant and Environmental Review Division (MD)

Director, Nuclear Training, Peach Bottom Manager, Operations Training, Peach Bottom Correspondence Control Desk Site Vice President - Limerick Generating Station Plant Manager, Limerick Generating Station Regulatory Assurance Manager, Limerick Senior Vice President - Nuclear Services Vice President - Mid-Atlantic Operations Manager, Licensing - Limerick Generating Station Director, Training, Limerick Generating Station Chairman, Board of Supervisors of Limerick Township D. Allard, Director, Pennsylvania Bureau of Radiation Protection (SLO, PA)

Commonwealth of Pennsylvania (c/o R. Janati, Chief, Division of Nuclear Safety, Pennsylvania Bureau of Radiation Protection)

R. Fletcher, Maryland Department of Environment T. Snyder, Director, Air and Radiation Management Administration, Maryland Department of the Environment (SLO, MD)

Mr. Christopher

SUMMARY OF FINDINGS

IR 05000352/2005302, 05000353/2005302, 05000277/2005302 and 05000278/2005302;

June 13-15, 2005; Limerick Generating Station and Peach Bottom Atomic Power Station; Initial Limited Senior Operator Licensing (LSRO) Examination. Three of three applicants passed the examination (3 LSROs).

The written examinations were administered by the facility and the operating tests were administered by 2 NRC region-based examiners. There were no inspection findings of significance associated with the examinations.

A. Inspector Identified Findings No findings of significance were identified.

B. Licensee Identified Findings No findings of significance were identified.

ii

REPORT DETAILS

REACTOR SAFETY

Mitigating Systems - Limited Senior Reactor Operator (LSRO) Initial License Examination

a. Scope

of Review The facility developed the written and operating initial examination and together with NRC Region I examiner staff verified or ensured, as applicable, the following:

The examination was prepared and developed in accordance with the guidelines of Revision 9 of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors. A review was conducted both in the Region I office and at the Limerick plant and training facility. Final resolution of comments and incorporation of test revisions were conducted during and following the onsite preparation week.

A test item analysis was completed on the written examination for feedback into the systems approach to training program.

Examination security requirements were met.

The NRC examiners administered the operating portion of the examination to all applicants from June 13-14, 2005. Limerick training staff administered the written examination on June 15, 2005.

b. Findings

Grading and Results Three of three applicants (3 LSROs) passed all portions of the initial licensing examination.

The facility had two post-examination comments. These are detailed in Attachment 2.

One procedure enhancement issue was identified during the administration of the Operating Test at Limerick. This was communicated to the licensee.

Examination Administration and Performance No findings of significance were identified.

4OA6 Exit Meeting Summary

On July 21, 2005, the NRC provided conclusions and examination results to a Limerick management representative via the telephone. License numbers for the applicants were not provided at this time pending receipt of written certification from Exelon to the U.S. Nuclear Regulatory Commission stating that the applicants have acquired all experience for which they were previously granted waivers. The NRC expressed appreciation for the cooperation and assistance provided by the licensees training staff during the preparation and administration of the examination.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

J. White, Director of Training

C. Rich, Manager, Operator Training

C. Fritz, Facility Exam Development Team

C. Goff, Facility Exam Development Team

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

ITEM NUMBER TYPE DESCRIPTION

NONE

ATTACHMENT 2

Licensees Post Written Examination Comments Publically Available

ADAMS Accession No. ML052060249

The licensees post exam comments regarding Questions 26 and 37 were received by the NRC

on June 17, 2005. After initial review, the NRC provided comments to and requested additional

information from the licensee on June 20, 2005. After further review by the NRC staff and

discussions with the licensee, the licensee submitted revised comments for both LSRO

questions #26 and #37, dated and received in Region I on June 22, 2005 (see below). The

NRC provided the licensee further comments on July 5, 2005 regarding question #37 and

certain stated assumptions made by the applicants and the fact that under the conditions stated

in the stem that there was no technical basis for assuming that Fuel Pool Cleanup was either

NOT running at the onset, nor for assuming it was lost following the event. The licensee

submitted a third and final revision for question #37 on July 11, 2005. During the exam there

were no questions from the applicants regarding either LSRO question #26 or #37. The NRCs

resolution for these two post exam comments is based on the independent reviews that were

conducted by both NRC examiners assigned to the exam team (plus a third independent

examiner for the final submittal on question #37) as well as the Branch Chief. There was some

additional input from a DRP Engineer who was a former Peach Bottom SRO.

Original LSRO Question 37:

PBAPS Unit 2 plant conditions are as follows:

- Mode 5

- Core Shuffle Part 1 has just begun

- RBCCW is backing up TBCCW

- The CRD system is in service

- The RWCU system is in service in a normal lineup dumping 60 GPM to the Main

Condenser

- A fire header break in the RBCCW system room has caused both RBCCW

pumps to trip

WHICH ONE of the following describes the operational implications of this condition?

A. Higher than normal plant dose rates.

B. Loss of Instrument Air to the Refueling Bridge

C. Reactor Cavity and fuel pool visibility will degrade

D. Reactor Cavity and Fuel Pool water level will begin to lower.

Originally designated correct answer D Submitted answer explanation (original):

Justification

a. Incorrect - RWCU will not isolate on high temperature. The isolation temperature is

200 degrees F but the Reactor Cavity temperature will be between 110 F and 130 F

during refueling operations. Generally, the temperature is maintained well below 110 F;

around 90 F.

b. Incorrect - The Refueling Bridge at PBAPS has an air compressor mounted on the

bridge and is, therefore, independent of station air systems.

c. Incorrect- RWCU will not isolate on high temperature. The clarity of the Reactor Cavity

will not change due to this event.

d. Correct- With RBCCW supplying TBCCW loads, RBCCW is supplying cooling water to

the CRD pump lube oil coolers and thrust bearings. The CRD pumps will trip after loss

of RBCCW. The loss of 60 GPM from the CRD system into the reactor cavity will

cause Reactor Cavity and Fuel Pool Levels to slowly lower. RWCU Dump Flow is still

in service and would lower cavity level at the rate of 60 GPM. This question tests

differences between LGS and PBAPS.

Licensees Original Justification for Change

Question Number:

(Missed by all three candidates)

Facility Regrade Request:

Change the correct answer to c

Justification:

It also provides that Control Rod Drive System (CRD) is in service, which provides a normal

flow of approximately 60 gpm to the reactor vessel. With RWCU dump flow out of the reactor

compensating for CRD flow into the reactor, reactor cavity/spent fuel pool level will be stable.

The question then states that both RBCCW pumps trip. The answer key indicates that the

operational implication of this would be that Reactor Cavity and Fuel Pool water level will begin

to lower, which is answer d. The justification for this on the answer key states that The CRD

pumps will trip after a loss of RBCCW. If this were the case and the running RWCU pump

remains running, then reactor cavity/spent fuel pool level would lower.

However, there is no direct trip of the CRD pump due to a loss of RBCCW flow. The PBAPS

Initial Licensed Operator Training lesson plan for Control Rod Drive Hydraulic System, PLOT-

5003A states on page 18 of 24, under Interlocks, that the pump will trip on low suction

pressure and various electrical malfunctions. This is also supported by the Annunciator

Response Card for CRD WATER PUMP TRIP (ARC-211, F-1 and G-1) that lists only Low

suction pressure and Motor overcurrent as the automatic trips of the CRD pumps. The

lesson plan also states (on page 20 of 24) that A loss of TBCCW and RBCCW will cause the

CRD pump to overheat. Therefore, the CRD pump will remain running even with a loss of

TBCCW and RBCCW.

Since RBCCW also cools the RWCU pump motor coolers (see Design Basis Documents P-S-

for RBCCW and P-S-36 for RWCU) a loss of RBCCW will result in an automatic trip of the

RWCU pump due to high temperature in the RWCU pump motor windings at a setpoint of 149

deg.

F. This is supported by ARC-215, A-2 and B-2, which are provided. In addition, the

RWCU System Manager at PBAPS, Luis Feliu (717-456-3634) indicated that this trip would

occur fairly soon after a loss of RBCCW at rated conditions. He also indicated that with the

reactor shutdown and cooled down to a temperature typically seen during a refueling outage,

the high temperature trip setpoint may take longer to reach due to the absence of heat

conduction input from the system, but would still reach the high temperature trip setpoint due to

the heat generated due to the motor winding current. This was also confirmed by the alternate

RWCU System Manager at PBAPS, who was the previous RWCU System Manager, as well as

engineering personnel at LGS, which has similar RWCU pumps. The System Managers also

indicated even with the RPV flooded up to normal level for refueling operations, no dump flow

would be expected after RWCU pump trip, due to the lack of RPV pressure and the high

headloss of the circuitous RWCU dump flowpath (RPV pressure will be approximately 0 psig

since the reactor is in Mode 5 with Core Shuffle Part 1 in progress). When the author of this

question was asked why RWCU was assumed to remain in service, he responded that he

overlooked the high motor winding trip for the RWCU pumps.

The candidates were interviewed to determine why answer c was chosen. The candidates

indicated that they assumed the Fuel Pool Cleanup System was not in service because it was

not stated in the question stem. The candidates also indicated that they were aware of

operational configurations where the Fuel Pool Cleanup System is not in service. The Fuel

Pool Cleanup System is routinely removed from service for other outage related activities (i.e.,

Service Water System maintenance). When the Fuel Pool Cleanup System is removed from

service, the Residual Heat Removal (RHR) System must be placed in service to remove the

decay heat from the reactor cavity and spent fuel pool, as specified in P-S-09, Residual Heat

Removal System. When the RHR System is being used to remove decay heat from the reactor

cavity and spent fuel pool, hot water from the reactor core rises to the surface of the reactor

cavity, flows into the reactor cavity weir opening and through the transfer canal into the fuel

pool, and ultimately flows into the skimmer surge tanks. The water that entered the skimmer

surge tanks is cooled and returned to either the reactor vessel or the spent fuel pool. The

natural circulation currents cause particulate from the reactor cavity to be transported into the

spent fuel pool, degrading overall spent fuel pool water quality. Since none of the particulate is

being removed via the RWCU System, corrosion products will continue to accumulate. These

particles diminish visibility due to the light from the installed spent fuel pool and reactor cavity

lights reflecting off the particles, resulting in the appearance of a haze in the water. The effect

worsens as the particle concentration in the reactor vessel increases.

A loss of RBCCW or RWCU does not require suspension of fuel handling activities, and fuel

movements would continue, causing more corrosion products to be dispersed in the water.

Since the concentration of corrosion product particles will rise in the reactor cavity and spent

fuel pool water when RWCU is out of service, and RWCU pumps will trip, visibility will degrade

over time without RWCU in service. Per OP-PB-112-101-1011, Fuel Handling Director Shift

Turnover Checklist, the LSRO is required to verify once per shift that conditions support error-

free operations, including water clarity. This is required due to the possibility that visibility will

degrade slowly over a period of time, and requires an LSRO to assess conditions at least once

per shift to determine if fuel handling activities may safely continue.

Since the running RWCU pump will trip on high motor winding temperature, and the CRD pump

will not trip, reactor cavity/spent fuel pool level will actually rise very slowly. This makes answer

d incorrect. Also, since the running RWCU pump trips, the clarity of the water in the reactor

cavity/spent fuel pool will degrade, making c the correct answer.

Therefore, the Facility Licensee requests the correct answer for Question 37 be changed to c.

NRC Initial Response

For answer c to be correct, your explanation does not address or explain how the

loss of RWCU (proposed as an explanation for "D" being incorrect) would affect the Fuel Pool

visibility as long as the Fuel Pool Cooling/Cleanup remains in service. Information in the stem

is consistent with Fuel Pool Cooling/Cleanup in service and remaining in service for the duration

of the event. The "normal" configuration during Core Shuffle Part 1 would be to have Fuel Pool

Cooling system in service This system is needed both for decay heat removal and to

establish/maintain clarity in the Fuel Pool. The use of RHR Fuel Pool Cooling assist would NOT

be the "normal" configuration during Core Shuffle Part 1 since it has a tendency to introduce

flow disturbances in the core area. It was also confirmed by Limerick Training that any

scheduled outage of Fuel Pool Cooling would NOT normally be done during Core Shuffle Part

1. Rather it would be done during Core Shuffle Part 2 when there is lower decay heat load.

Excerpts from the System Description for the "Fuel Pool Cooling and Cleanup System", P-S-52,

Revision 5 (provided by the facility licensee) include the following:

1. The Fuel Pool Cooling and Cleanup System has sufficient cooling capacity to maintain the

Spent Fuel Pool water at a temperature at or below 150F for normal decay heat load with two

pumps and two heat exchangers operating.

2. If an abnormally large heat load is placed in the spent fuel pool, a cooling train of the RHR

System consisting of an RHR pump and heat exchanger, is substituted for the Fuel Pool

Cooling pumps and Heat exchangers.for cooling the pool water. The conditions under which

cooling of the Spent Fuel Pool water by the RHR System alone would be required include the

unloading of a full core of irradiated fuel into the pool.

3. Additional heat removal capability may be needed when full core off-loading occurs and less

than three Fuel Pool Cooling Pumps/heat exchangers are available.

4. The service water system shall support operation of the Fuel Pool Cooling System by

providing cooling flow at a rate of 800 GPM.

5. Design basis... minimize corrosion product buildup and control water clarity through filtration

and demineralization.

6. The Fuel Pool Cooling and Cleanup System performs the decay heat removal function

whenever spent fuel is stored in the Spent Fuel Pool, including refueling.

As specified in NUREG -1021, Appendix E. The applicant is directed to ask questions "if you

have any questions concerning the intent or the initial conditions of a question". No applicant

asked any questions regarding question 37 during the exam. The NUREG goes on to specify

that the applicant should "NOT make assumptions regarding conditions that are not specified in

the question unless they occur as a consequence of other conditions that are stated in the

question".

Fuel Pool Cleanup would NOT be lost as a consequence of the loss of RBCCW. It would

appear that the RHR System lineup would only be required if there was a Fuel Pool Cooling

System partial or full outage. There would be normally NO planned outage of Fuel Pool

Cooling during Core Shuffle Part 1. No outage of Fuel Pool Cooling equipment was specified in

the stem. Nor was there any mention of a "full core off-load". Therefore, the applicant would

have no technical basis for assuming that Fuel Pool Cleanup was either NOT running at the

onset, nor for assuming it was lost following the event. Therefore, as presented in your

submittal, there appears to be no basis for concluding that the stem conditions would result in

degrading Fuel Pool visibility.

Revised and Final Facility Regrade Request:

Change the correct answer to c.

Justification:

The question provides that Reactor Building Closed Cooling Water System (RBCCW) is

backing up Turbine Building Closed Cooling Water System (TBCCW), and Reactor Water

Cleanup (RWCU) is dumping 60 gpm to the main condenser. It also provides that Control Rod

Drive System (CRD) is in service, which provides a normal flow of approximately 60 gpm to the

reactor vessel. With RWCU dump flow out of the reactor compensating for CRD flow into the

reactor, reactor cavity/spent fuel pool level will be stable.

The question then states that both RBCCW pumps trip. The original answer key indicates that

the operational implication of this would be that Reactor Cavity and Fuel Pool water level will

begin to lower, which is answer d. The justification for this on the answer key states that The

CRD pumps will trip after a loss of RBCCW. If this were the case and the running RWCU

pump remains running, then reactor cavity/spent fuel pool level would lower.