ML13191B346

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Charlissa Smith Pre-filed Hearing Exhibit CCS-118 - Catawba Nuclear Station - NRC Operator License Examination Report 05000413-08-301 and 05000414-08-301
ML13191B346
Person / Time
Site: Catawba, 05523694  Duke Energy icon.png
Issue date: 02/05/2009
From: Widmann M
Division of Reactor Safety II
To: Morris J
Duke Energy Carolinas, Duke Power Co
SECY RAS
References
55-23694-SP, ASLBP 13-925-01-SP-BD01, RAS 24803
Download: ML13191B346 (24)


See also: IR 05000413/2008301

Text

UNITED STATES

CCS-118

NUCLEAR REGULATORY COMMISSION

REGION II

SAM NUNN ATLANTA FEDERAL CENTER

61 FORSYTH STREET, SW, SUITE 23T85

ATLANTA, GEORGIA 30303-8931

February 5, 2009

Mr. J. R. Morris

Site Vice President

Duke Power Company, LLC

d/b/a Duke Energy Carolinas, LLC

Catawba Site

4800 Concord Road

York, SC 29745-9635

SUBJECT: CATAWBA NUCLEAR STATION - NRC OPERATOR LICENSE EXAMINATION

REPORT 05000413/2008301 AND 05000414/2008301

Dear Mr. Morris:

During the period December 1 - 4, 2008 the Nuclear Regulatory Commission (NRC)

administered operating tests to employees of your company who had applied for licenses to

operate the Catawba Nuclear Station. At the conclusion of the tests, the examiners discussed

preliminary findings related to the operating tests with those members of your staff identified in

the enclosed report. The written examination was administered by your staff on December 10,

2008.

One Reactor Operator (RO) and three Senior Reactor Operator (SRO) applicants passed both

the operating test and written examination. One RO and four SRO applicants failed the written

examination. There were ten, (10) post-administration comments concerning the written

examination. These comments, and the NRC resolution of these comments, are summarized in

Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.

The draft written examination submitted by your staff failed to meet the guidelines for quality

contained in NUREG-1021, Operator Licensing Examination Standards for Power Reactors,

Revision 9, Supplement 1, as described in the enclosed report.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its

enclosures will be available electronically for public inspection in the NRC Public Document

Room or from the Publicly Available Records (PARS) component of the NRCs document

system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-

rm.adams.html (the Public Electronic Reading Room).

DPC, LLC 2

If you have any questions concerning this letter, please contact me at (404) 562-4550.

Sincerely,

/RA/

Malcolm T. Widmann, Chief

Operations Branch

Division of Reactor Safety

Docket Nos.: 50-413, 50-414

License Nos.: NPF-35, NPF-52

Enclosures: 1. Report Details

2. Facility Comments and NRC Resolution

3. Simulator Fidelity Report

cc w/encl: (See page 3)

DPC, LLC 3

cc w/encl: Susan E. Jenkins

Randy D. Hart Director, Division of Waste Management

Regulatory Compliance Manager Bureau of Land and Waste Management

Duke Power Company, LLC d/b/a Duke S.C. Department of Health and

Energy Carolinas, LLC Environmental Control

Electronic Mail Distribution Electronic Mail Distribution

R. L. Gill, Jr. Manager R. Mike Gandy

Nuclear Regulatory Issues & Industry Affairs Division of Radioactive Waste Mgmt.

Duke Power Company, LLC d/b/a Duke S.C. Department of Health and

Energy Carolinas, LLC Environmental Control

Electronic Mail Distribution Electronic Mail Distribution

Dhiaa M. Jamil Beverly O. Hall

Group Executive and Chief Nuclear Officer Chief, Radiation Protection Section

Duke Energy Carolinas, LLC Department of Environmental Health

Electronic Mail Distribution N.C. Department of Environmental

Commerce & Natural Resources

Kathryn B. Nolan, Senior Counsel Electronic Mail Distribution

Duke Energy Corporation

526 South Church Street-EC07H Elizabeth McMahon

Charlotte, NC 28202 Assistant Attorney General

S.C. Attorney General's Office

Lisa F. Vaughn P.O. Box 11549

Associate General Counsel Columbia, SC 29211

Duke Energy Corporation

526 South Church Street-EC07H Vanessa Quinn

Charlotte, NC 28202 Federal Emergency Management Agency

500 C Street, SW

Senior Resident Inspector Room 840

Duke Energy Corporation Washington, DC 20472

Catawba Nuclear Station

U.S. NRC Steve Weatherman, Operations Analyst

4830 Concord Road North Carolina Electric Membership

York, SC 29745 Corporation

Electronic Mail Distribution

David A. Repka

Winston Strawn LLP Peggy Force

Electronic Mail Distribution Assistant Attorney General

State of North Carolina

North Carolina MPA-1 P.O. Box 629

Suite 600 Raleigh, NC 27602

P.O. Box 29513

Raleigh, NC 27525-0513 Duke Energy Corporation, LLC

ATTN: Mr. Ronald Weatherford

County Manager of York County Training Manager

York County Courthouse Catawba Nuclear Station

York, SC 29745 4800 Concord Road

York, SC 29745-9635

Piedmont Municipal Power Agency

Electronic Mail Distribution

_________________________ X SUNSI REVIEW COMPLETE

OFFICE RII:DRS RII:DRS RII:DRS RII:DRS RII:DRP RII:DRS

SIGNATURE RA RA RA RA RA RA

NAME LASKA EHRHARDT KONTZ MEEKS BARTLEY WIDMANN

DATE 2/4/2009 2/4/2009 1/30/09 2/4/09 2/2/09 2/4/09

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO

U.S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket No.: 50-413, 50-414

License No.: NPF-35, NPF-52

Report No.: 05000413/2008301, 05000414/2008301

Licensee: Duke Energy Corporation (DEC)

Facility: Catawba Nuclear Station, Units 1 & 2

Location: 4800 Concord Road

York S.C. 29745

Dates: Operating Test - December 1 - 4, 2008

Written Examination - December 10, 2008

Examiners: Gerard Laska, Chief Examiner, Senior Operations Examiner

Frank Ehrhardt, Senior Operations Engineer

Craig Kontz, Operations Engineer

Michael Meeks, Operations Engineer (In-Training)

Approved by: Malcolm T. Widmann, Chief

Operations Branch

Division of Reactor Safety

Enclosure 1

SUMMARY OF FINDINGS

ER 05000413/2008301, 05000414/2008301, 12/01-04/2008 & 12/10/2008; Catawba Nuclear

Station; Operator License Examinations.

Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in

accordance with the guidelines in Revision 9, Supplement 1, of NUREG-1021, "Operator

Licensing Examination Standards for Power Reactors." This examination implemented the

operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45, as applicable.

Members of Catawba Nuclear Station staff developed both the operating tests and the written

examination. The final written examination submittal was considered to be outside the

acceptable range because it did not meet the quality guidelines contained in NUREG-1021.

The NRC administered the operating tests during the period December 1 - 4, 2008. Members of

the Catawba Nuclear Station training staff administered the written examination on

December 10, 2008. One Reactor Operator (RO) and three Senior Reactor Operator (SRO)

applicants passed both the operating test and written examination. One RO applicant and four

SRO applicants failed the written examination. Two applicants (one RO and one SRO) were

issued licenses. Two SRO applicants received pass letters pending results of any appeals.

There were ten (10) post-examination comments.

Enclosure 1

REPORT DETAILS

4. OTHER ACTIVITIES

4OA5 Operator Licensing Examinations

a. Inspection Scope

Members of the Catawba Nuclear Station staff developed both the operating tests and

the written examination. All examination material was developed in accordance with the

guidelines contained in Revision 9, Supplement 1, of NUREG-1021, "Operator Licensing

Examination Standards for Power Reactors." The NRC examination team reviewed the

proposed examination. Examination changes agreed upon between the NRC and the

licensee were made per NUREG-1021 and incorporated into the final version of the

examination materials.

The NRC reviewed the licensees examination security measures while preparing and

administering the examinations in order to ensure compliance with 10 CFR Part 55.49,

Integrity of examinations and tests.

The NRC examiners evaluated two Reactor Operator (RO) and seven Senior Reactor

Operator (SRO) applicants using the guidelines contained in NUREG-1021. The

examiners administered the operating tests during the period December 1 - 4, 2008.

Members of the Catawba Nuclear Station training staff administered the written

examination on December 10, 2008. Evaluations of applicants and reviews of

associated documentation were performed to determine if the applicants, who applied

for licenses to operate the Catawba Nuclear Station, met the requirements specified in

10 CFR Part 55, Operators Licenses.

b. Findings

The NRC determined that the details provided by the licensee for the walkthrough and

simulator tests were within the range of acceptability expected for a proposed

examination.

The NRC determined that the licensees original written examination submittal was

within the range of acceptable quality specified by NUREG-1021. However, based on

post exam comments whereby five (5) additional questions were determined to be

unsatisfactory the final written examination was determined to be outside of the range of

acceptable quality. More than 20% (24 of 100) of questions sampled for review

contained unacceptable flaws. Individual questions were evaluated as unsatisfactory for

the following reasons:

  • 10 questions failed to meet the K/A statement contained in the examination

outline.

  • 2 questions contained two or more implausible distractors.
  • 9 questions on the SRO examination were not written at the SRO license level.
  • 3 questions contained other unacceptable psychometric flaws.
  • 4 questions contained multiple unacceptable flaws.

Enclosure 1

4

Future examination submittals need to incorporate lessons learned.

One RO applicant and three SRO applicants passed both the operating test and written

examination. Two applicants (one RO and one SRO) were issued licenses. One RO

applicant and four SRO applicants passed the operating test but did not pass the written

examination.

One SRO applicant passed the operating test, but passed the written examination with

an overall score between 80% and 82% and the SRO-only portion with a score between

70 and 74 %. One SRO applicant passed the operating test, but passed the SRO-only

portion of the written examination with a score between 70% and 74%. Each of these

applicants were issued a letter stating that they passed the examination and issuance of

their license has been delayed pending any written examination appeals that may

impact the licensing decision for their application.

Copies of all individual examination reports were sent to the facility Training Manager for

evaluation of weaknesses and determination of appropriate remedial training.

The licensee submitted ten (10) post-examination comments concerning the written

examination. A copy of the final SRO and RO written examination and answer key, with

all changes incorporated, and the licensees post-examination comments may be

accessed in the ADAMS system (ADAMS Accession Number(s) ML090230071,

ML090230075, ML090230038, and ML090230060).

4OA6 Meetings, Including Exit

Exit Meeting Summary

On December 4, 2008, the NRC examination team discussed generic issues associated

with the operating test with Mr. James R. Morris, CNS Site Vice President, and

members of the Catawba Nuclear Station staff. The examiners asked the licensee if

any of the examination material was proprietary. No proprietary information was

identified.

Enclosure 1

5

KEY POINTS OF CONTACT

Licensee personnel

J. Morris, CNS Site Vice President

R. Hart, CNS Manager Regulatory Compliance

R. Weatherford, Training Manager

J. McConnell, Shift Operations Manager

S. Coy, Operations Training Manager

H. Dameron, Operations Initial Training Supervisor

G. Hamilton, Operations Training

J. Suptela, Operations Training

T. Garrision, Operations Training

NRC personnel

A. Sabisch, SRI

R. Cureton, RI

Enclosure 1

FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS

A complete text of the licensees post-examination comments can be found in ADAMS under

Accession Number ML090230060.

Written Examination - Question # 5

Licensees Comment:

Operations administrative procedure OPM 1-7 (Emergency/Abnormal Procedure

Implementation Guidelines) has a section 7.6 (Deviation From Approved Procedures) and 7.7

(Situations Not Covered by Procedure). The question developers considered this condition to fit

into a situation not covered by procedure; therefore, OMP 1-7 section 7.7 would apply and

paragraph B which states, The planned course of action shall be reviewed and approved by a

second SRO would require one additional SRO to approve the desired action.

The applicants who chose answer C believed that OMP 1-7 section 7.6 (Deviation From

Approved Procedures) applied because the stem of the question stated that the OSM

determined an immediate need to take action. Section 7.6 paragraph C.3 states that actions

outside approved procedures can be taken when, Actions are needed to minimize immediate

personnel hazard/injury or damage to plant equipment. Section 7.6 paragraph D. 2 states that

only one SRO must approve the action.

If the OSMs chosen actions are taken from various procedures unrelated to the current

condition, then section 7.6 would apply. If the OSMs chosen actions arent described in any

procedure then section 7.7 would apply. The question didnt provide enough information for the

applicants to know whether to apply section 7.6 or 7.7. Therefore, we request that both

answers C and D be accepted as correct answers.

NRC Response:

The NRC agrees that the stem of the question did not provide enough information for the

applicants to determine which section of OMP1-7 to apply, and that there is a basis for two

possible correct answers given. However, the two answers contain conflicting information as

described in NUREG 1021 Revision 9 Supplement 1, 403 D.1.c. Therefore, this question will be

deleted from the examination.

Enclosure 2

Written Examination - Question # 19

Licensees Comment:

Unwarranted continuous rod movement is an entry condition for procedure AP/1/A/5500/015

Rod Control Malfunction Case II. The immediate actions of the AP are to place the rod bank

select switch in manual, verify rod motion stops and trip the reactor if the rods continue to move.

The question developers considered strict procedural compliance when developing the

question.

The intent of the step C.1 is to remove the CRD Bank Select switch from Auto. Any rod

movement with the switch in any position other than Auto indicates a fault in the rod control

system, and a reactor trip is warranted.

If the CRD Bank Select switch is in any position other than AUTO the rods can only be moved

manually. The applicants who selected answer B applied NSD 705 allowance of intent met, and

understood that any position other than AUTO is a position that only supports manual control of

the rods; therefore, the intent of step C 1 was already met and the only required action was an

immediate trip of the reactor per step C.2 RNO.

The applicants who selected answer D considered that strict procedural compliance required

the rod bank select switch to be placed in the MAN position. Per strict procedural compliance

answer D is correct.

Using the allowance of intent met, the first required action is to trip the reactor, and answer B is

correct. Considering strict procedure compliance the first required action is to place the CRD

Bank Select switch to manual, so the first require action is to place the switch in the MAN

position, and answer D is correct. Therefore, we request that both answers B and D be

accepted as correct.

NRC Response:

The NRC does not agree with accepting both answers B and D as correct. In this case placing

the rod control select switch in manual may have stopped rod motion. Shutdown banks do not

move out when the rod control select switch is in the manual position. Rod speed in the SBB

position is 64 steps per minute (spm), and rod speed in the manual position is 48 spm. The

path that the rod out impulse takes is different in SBB and manual positions. Therefore, it is

important to take the rod control select switch to manual in this case. The first Immediate Action

of AP/1/A/5500/015 Rod Control Malfunction Case II, specifically states to place the switch in

Manual.

Also noteworthy was the fact that the reference stated above that the applicants applied

allowing them to assume that the intent of the step was met was titled NSD 705 Instructions for

the Verification and Validation of Technical Procedures. This NSD did not contain directions on

procedure use. Furthermore it was discovered that NSD 704 Technical Procedure Use and

Adherence, which was the NSD that did contain direction on when the intent of a step was met,

was not applicable to abnormal and emergency procedures.

Therefore, answer D is the only answer that will be accepted as correct.

Enclosure 2

Written Examination - Question # 23

Licensees Comment:

The question developer considered the EMFs 71-74 to be correct because their location on the

steam lines makes them the first monitors to detect the change in secondary contamination.

The applicants who chose answer C selected 1EMF 33 because it will be the first EMF to

generate an alarm.

The question asks for the, best indication (most sensitive and most timely). The candidates

selected different answers due to making different assumptions about what indication is being

observed. Normally the operators infrequently monitor the EMF readings but are frequently

monitoring the EMFs alarm state. AP/1/A/5500/003 (Load Rejection), which would have been

implemented due to the runback, does not require the operators to monitor the EMF readings.

AP/1/A/5500/010 (NC System Leakage), which would be entered once a tube leak greater than

5 gpd is detected, requires monitoring of EMF readings every 15 minutes but only if the SG leak

rate is greater than 40 gpd. Given the situation described in the question the operators would

be monitoring the EMF alarm state not the EMF readings.

In accordance with NSD 513 (see attached) EMFs 71-74 are set to alarm at 5 GPD. 1EMF-33

readings input to a calculation that runs continuously on the Operator Aid Computer (OAC). Per

NSD 513 that calculation is set to produce an OAC alarm at 5 gpd. 1EMF-33 will produce an

alarm on the annunciator panel based upon a predetermined increase in count rate above the

background. Consequently, EMF-33 produces an annunciator due to increasing count rate

before an OAC alarm based upon the calculated leak rate.

EMFs 71-74 are located on the steam line coming from each of the SGs. EMF-33 is monitoring

the offgas from the condenser air ejectors. Due to their locations, EMFs 71-74 will be the first to

detect an increase in secondary activity due to a tube leak.

This scenario was performed on the simulator at 100% power and again after a runback on loss

of a CF pump. A 12 gpd leak in 1A SG was inserted, and in both cases 1EMF-71 count rate

was the first EMF to increase, but 1EMF-33 was the first EMF to produce an alarm.

Based upon observing the EMF alarm status EMF-33 will be the timeliest indicator, which would

make answer C correct.

Based upon monitoring the EMF readings EMFs 71-74 will be the timeliest because they are the

first monitors to be exposed to the increase in secondary activity which makes answer D

correct.

Since the question didnt clearly ask if the operators were monitoring the EMF readings or alarm

state, we request that both answers C and D be accepted as correct.

NRC Response:

The NRC does not agree with accepting answers C and D. The stem of the question asked for,

Which one of the following indicators will provide the best indication (most sensitive and timely)

that the S/G tube leak has increased. The question did not ask which one would alarm first.

The applicants who chose distractor C assumed that the indication would have to cause an

alarm first to alert the control room. NUREG 1021 appendix E, part B (7) states:

Enclosure 2

If you have any questions concerning the intent or the initial conditions of a question,

do not hesitate to ask them before answering the question. Note that questions

asked during the examination are taken into consideration during the grading process

and when reviewing applicant appeals. Ask questions of the NRC examiner

or the designated facility instructor only. A dictionary is available if you need it.

When answering a question, do 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. For example, you should not assume that any

alarm has activated unless the question so states.

Therefore, answer D is the only correct answer.

Enclosure 2

Written Examination - Question # 42

Licensees Comment:

The question developer considered the level required to support all ECCS and NS pumps taking

suction on the containment sump. The crew enters EP/ES-1.3 when the FWST level decreases

to 37%. The ND pump suctions automatically align to the containment sump, and the operators

will align the remaining ECCS pumps suctions from the FWST to the ND pump discharge per

ES-1.3. When FWST level decreases to 11% the operators will align the NS pumps suction to

the containment sump per ES-1.3.

The stem states that the crew has just entered EP/ES-1.3; therefore, at that point in time the

only pumps with their suction aligned to the sump are the ND pumps and all other pumps are

still aligned to the FWST. EP/ES-1.3 step 2 checks for a sump level > 3.3 feet. If it isnt, the

RNO verifies sump level > 2.5 feet at step 2.f. If level is > 2.5 feet then the NV and NI pumps

suctions can be aligned to the containment sump. In this situation a level of 2.5 feet will support

the operations all ECCS pumps while the NS pumps are still aligned to the FWST.

When the FWST level decreases to 11% ES-1.3 directs aligning the NS pumps to the

containment sump using enclosure 2. Step 2 checks for a sump level of > 3.3 feet. If it isnt

then the NS pump suction isnt aligned to the containment sump. Therefore, after FWST level

has decreased to 11%, 3.3 feet in the containment sump is required to support operation of all

ECCS pumps.

The stem didnt provide the applicants information concerning the FWST level. That information

is needed to determine which pumps are supposed to be aligned to the containment sump. If

FWST level is <37% and > 11%, then answer B is correct. If the FWST level is < 11% then

answer C is correct.

Since the question didnt have enough information for the applicants to know the point in time

they are required to evaluate the question, we request that both answers B and C be accepted

as correct.

NRC Response:

The NRC does not agree with accepting both B and C as correct. After a review of the

procedure and the construction of the stem of the question, What is the minimum containment

sump level that will support operation of all ECCS pumps and the NS pumps, it is clear that the

question is asking for the containment sump level as specified in ES-1.3 (Transfer to Cold Leg

Recirculation) that would be sufficient to provide a net positive suction for all ECCS pumps and

NS pumps. In accordance with ES-1.3 (Transfer to Cold Leg Recirculation), a containment

sump level of greater than 3.3 feet is required for all pumps to take a suction on the containment

sump.

Therefore, answer C is the only correct answer.

Enclosure 2

Written Examination - Question # 55

Licensees Comment:

Valve VQ-10 gets a close signal at 0 psig. The fans are large enough to reduce containment

pressure below the Tech Spec limit (See Attached). Therefore, basis for closing the valve at

that 0 psig is to prevent the VQ fans from reducing containment pressure to the minimum tech

spec value.

The 6 applicants who chose answer D rejected answer C because the wording of the answer

implied that the minimum tech spec value had been reached when the valve closed which is

incorrect since the minimum tech spec value is -0.1 psig. At 0 psig the plant is in compliance

with Tech Specs; therefore, the answer is not technically correct. Had the answer stated, To

prevent non-compliance then the answer would have been correct.

The VQ fans are sized small enough to prevent them from opening the ice condenser doors;

therefore answer D is wrong. (See attached.)

We recommend that this question be deleted from the exam since there is no technically correct

answer.

NRC Response:

The NRC agrees with the licensee in that there is not a technically correct answer, and the

question will be deleted from the examination.

Enclosure 2

Written Examination - Question # 76

Licensees Comment:

The stem told the applicant that the ND suction relief was leaking. The applicant was required

to know that the ND relief valves discharge to the PRT. The applicant was also required to

know that AP/27 will transition the operator to AP/19 if PRT level is increasing without indication

that the input is from the NC system pressurizer.

The symptoms of this event would be pressurizer level and pressure decreasing and PRT level

increasing. These symptoms match the entry conditions for AP/19 rather than AP/27. (See

attached.) Therefore, entry into AP/27 was an incorrect diagnosis of the event. In the event of

a leak step 3 of AP/27 will stop any ND pump taking suction on an NC system loop to protect

the ND pump from damage.

AP/27 step 4 looks for PRT level increasing without indication of safety valve input. The intent

of this step is to rule out input to the PRT from the pressurizer safety. If the safety valve is not

discharging to the PRT, then the procedure assumes the input is from the ND system and the

operator is directed to transition to AP/19. The only indication available for the operator to

determine if the PRT input is from a pressurizer safety is safety valve tailpipe temperature and

acoustic flow monitors. The question did not provide the applicant the status of those indicators.

Additionally, the question didnt provide the applicant with information about the status of PRT

level before or after the actions of AP/27 were performed. AP/27 step 4 doesnt specifically

state pressurizer safety. The ND relief valve is a safety valve which discharges to the PRT.

The background document for that step doesnt clarify that the step applies to pressurizer safety

valves. The stem stated that the ND relief was open can be interpreted as indication that a

safety is discharging to the PRT. Procedure change request number CNS-2008-5216 has been

submitted to revise AP/27 step 4a to state pressurizer safety valve. All of the applicants

correctly answered part 2 of the question. However, the applicants were not given information

about pressurizer safety valve status and PRT level response which was needed adequately

determine the proper procedure flowpath.

Given the ambiguity of AP/27 step 4, and the lack of information to properly evaluate the status

of the pressurizer safeties and PRT level we request that question 76 be deleted.

NRC Response:

The NRC does not agree with deleting the question. After reviewing the entry conditions for

AP/27, and AP/19, it appears that either procedure could be entered for the above conditions.

Furthermore, the question stated that AP/27 was entered.

Step 4 of AP/27 States: Verify Leak is on ND:

a. Plant alarms and indications - INDICATE LEAK OUTSIDE CONTAINMENT

OR

PRT Level - INCREASING WITHOUT INDICATION OF SAFETY VALVE INPUT.

b. GO TO AP/1/A/5500/019 (Loss of Residual Heat Removal System).

Enclosure 2

The stem of the question states that the leak was from the ND via one of the ND suction relief

valves that had lifted and HAD FAILED TO RESEAT. With the information given in the stem,

and not making any new assumptions, the applicant had enough information to determine/verify

that the leak is on the ND system and that a transition to AP/19 is clearly warranted.

Therefore, answer D is the only correct answer.

Enclosure 2

Written Examination - Question # 77

Licensees Comment:

The developer considered the basis for Tech Spec 3.8.1 which states either off site or on site

power is available, and in this scenario off site power is maintained. Therefore, entry into 3.0.3

was considered to be the time that the design criteria were no longer met. Tech Spec 3.8.1

action B2 requires declaring 2B NI inoperable 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after 2B DG was declare inoperable.

Thus, at 0700 2B NI is still considered operable; so, the ECCS design criteria for a large break

LOCA was met.

All of the applicants that selected answers C and D understood that the 2B NI didnt have to be

declared inoperable until 0900. Those who selected answer D considered design criteria to be

separate from the declaration of inoperability. Declaration of inoperability is an administrative

function. The Regulatory Compliance department was asked to interpret this scenario.

Regulator Compliance contacted Excel Services who writes our Tech Specs. The following is

their reply:

From: Dan Williamson [dan.williamson@excelservices.com]

Sent: Monday, December 15, 2008 8:32 PM

To: pwrog@excelservices.com

Subject: RE: Initial License Exam TS Question

Few comments:

>> If not yet adopted, consult TSTF-273 for intent clarifications related to this situation.

>> The ECCS design criteria for a large LOCA is different than loss of safety function

typically used in TSpecs / SFDP. The design criteria was not met when the first 1A SI was inop --

> loss of single failure protection.

>> The example is a bit confusing when the ending question mentions when 2A DG becomes

inoperable -- prior to this, 2A DG was not at issue (?) Seems a typo of some kind.

>> The [A-SI + B-DG] is still is not a loss of safety function (see TSTF-273). The directed

declaration of B-SI inop at 4 hrs due to B-DG inop (and one can wait the full 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to make this

declaration) can be argued to be the first time that a loss of safety function exists --> both A & B

SI inop.

Dan Williamson

EXCEL Services Corporation

Main Offc/Cell: (904) 272-5300

Given that the design criteria were not met when 2A NI pump was declared inoperable, we

request that the correct answer be changed to D.

NRC Response:

NRC agrees with the licensees explanation. It is clear that the thought process involved in the

question development was equipment operability and not actual ECCS design criteria. The

exam key will be changed to make D the correct answer.

Enclosure 2

Written Examination - Question # 83

Licensees Comment:

When plant control is aligned to the control room and a VCT Lo-Lo Level (4.3%) is detected the

suction valves from the FWST open and the suction valves from the VCT close. The Design

Basis Document for Loss of Control room states that all automatic NV functions are disabled

when control is transferred to the Auxiliary Shutdown Complex (ASC). The DBD also states that

the suction valves from the VCT open upon transfer to the ASC and are blocked from closing on

Lo-Lo Level. The Loss of Control Room lesson plan states the same information found in the

DBD. The DBD for the NV system doesnt discuss how the suction valves from the FWST are

affected by swapping control to the ASC. Additionally, AP/1/A/5500/017 (Loss of Control Room)

Enclosure 1 page 12 directs manual alignment of the NV pump suction to the FWST if VCT level

is < 23%. The background document for the procedure states that, All automatic transfer of the

NV pump suction to the FWST on low VCT level is lost when control is transferred to the ASP.

Based upon controlled information available to the question developers they determined that the

automatic swap of the NV pump suction to the FWST on lo-lo VCT level would not occur.

During the exam review the applicants stated that they were taught that the swap to the FWST

will occur automatically. The instructor who teaches the Loss of Control Room had determined

that the suction valves from the FWST are unaffected by the swap to the ASC, and had included

that information in the notes section of the Power Point presentation used to teach the lesson.

A copy of the Power Point presentation had been provided to the applicants. The notes section

of a single slide of the presentation includes the statement, NV-252A & NV-253B will auto open

on Lo-Lo VCT level, but NV-188A & NV 189B will not close. Brian Woolweber (Senior

Engineer) and Nick Burgess (Engineer III) reviewed the electrical drawings and confirmed that

the FWST suction valves are unaffected by a swap to the ASC and will in fact open on a VCT

Lo-Lo- Level signal. (See attached note from engineering.)

Answer B is technically correct because if the suction of the NV pumps isnt manually aligned to

the FWST when VCT level is < 23%, then the valves will automatically open on Lo-Lo VCT level

and primary side makeup would be assured.

Answer D is technically correct because the suction supply valves from the FWST are manually

opened per the requirements of procedure AP/17 to ensure primary side makeup is assured.

We request that both answers B and D be accepted as correct.

NRC Response:

The NRC does not agree with accepting B and D as being correct. Per AP/1/A/5500/017 (Loss

of Control Room), primary side inventory is assured by manually swapping NV pump suction to

the FWST.

Step 17 of AP/1/A/5500/017 (Loss of Control Room) Enclosure 1 ASP actions directs the

operator to:

17. Control VCT level as follows:

a. Ensure charging and letdown flow -

ADJUSTED TO MAINTAIN PZR LEVEL

AT 25%.

Enclosure 2

b. IF NV pump suction aligned to the VCT,

THEN maintain VCT level using one of the following:

Normal boration:

1) Start a Boric Acid Transfer pump.

2) Open 1NV-186A (B/A Blender

Otlt To VCT Otlt).

3) Open 1NV-238A (B/A Xfer Pmp

To Blender Ctrl).

OR

Emergency boration:

1) Start a Boric Acid Transfer pump.

2) Open 1NV-236B (Boric Acid To

NV Pumps Suct).

c. Verify VCT level - GREATER THAN c. IF VCT level decreasing,

23%. THEN notify

Unit 1 Aux Bldg operator to align

Unit 1 NV pump suction to FWST.

REFER TO Enclosure 4 (Aux Bldg

Operator Actions), Step 5.

d. IF AT ANY TIME VCT level decreases

to less than 23%, THEN perform 17.c

RNO.

The procedure directs the operator to manually align FWST suction to the NV pump if VCT

level decreases to <23 %, this manual alignment also includes manual closing 1NV-188A and

1NV-189B. A note prior to step 14 of enclosure 1 states:

CAUTION With NV pump suction valves from the VCT (1NV-188A and 1NV-189B)

and FWST (1NV-252A or 1NV-253B) open, suction supply may be lost

when VCT level drops to 0% due to the H2 pressure maintained in the

VCT.

The automatic opening of 1NV-252A or 1NV-253B when VCT level decreases to less than

4.3%, does not assure a continued NV pump suction because signal this does not close 1NV-

188A and 1NV-189B, as stated in the Licensees description above. Based on the above note

this is required to assure NV pump suction. Therefore answer D is the only correct answer.

Enclosure 2

Written Examination - Question # 87

Licensees Comment:

During the first stage of a LOCA the ice condenser is the major heat sink for cooling the

containment atmosphere. After the ice has melted then NS becomes the major heat sink. RN

flow rate to the NS heat exchangers is a constant value; therefore, the temperature of NS is

directly related to RN temperature. Once the ice has melted containment pressure will be

related to the NS temperature, and if NS temperature is higher, then containment pressure will

be higher. The higher NS temperature would have little to no affect on containment pressure

before the ice melts because the ice is the major heat sink, but pressure would be affected after

the ice was melted. (See attached excerpt from Tech Spec 3.7.9 bases.) The developer

included the word significant in the second part of the answer because the difference in NS

temperature will be observable to the operator in the control room.

If the Lake Wylie temperature reaches the SLC limit, the remedial action is to align at least one

train of RN to the Standby Nuclear Service Water Pond (SNSWP). The Tech Spec basis for the

(SNSWP) states, NSWS (Nuclear Service Water System) temperature influences containment

pressure following a Loss of Coolant Accident and offsite dose following a Main Steam Line

Break. The containment peak pressure analysis can accommodate NSWS temperatures up to

100oF. Since the Lake Wylie temperature, thus NSWS temperature had not exceeded 100oF

the applicants who chose answer D determined that the elevated RN temperature would not

have a significant affect on containment temperature. Therefore, they rejected answer C and

selected answer D as the most correct for the given conditions.

Question 2 asked the applicant to compare the affect of the higher lake temperature, but it

doesnt ask which higher temperature to use, the last observed or the SLC limit, or what

temperature it should be compared to. In reference to answer C for question 2, the first part of

the answer is correct, but statement that the affect would be significant cannot be supported

since question didnt imply how big a temperature difference to consider. Consequently, answer

C cannot be supported as correct.

Answer D is a correct answer since all of the temperatures given for comparison are below the

analyzed value of 100oF. Thus, the impact or consequence would be minimal throughout the

entire sequence of the accident.

We request that the correct answer be changed to D.

NRC Response:

The NRC does not agree that answer D is a correct answer. SLC16.9.4 states that the water

temperature of Lake Wylie shall be 95.5 ºF. If temperature is greater than 95.5 ºF the SLC

directs the operator to align at least one NSWS loop to the Standby Nuclear Service Water

Pond (SNSWP). Therefore actions must be taken to prevent exceeding any accident analysis

assumptions. TS 3.7.9 Standby Nuclear Service Water Pond basis document discusses the

effects of elevated NSWS temperatures. The basis document states in part:

The peak containment pressure occurs when energy addition to containment (core

decay heat) is balanced by energy removal from the Containment Spray and Component

Cooling Water heat exchangers. This balance is reached after the transition from

injection to cold leg recirculation and after ice melt. Because of the effectiveness of the

Enclosure 2

ice bed in condensing the steam which passes through it, containment pressure is

insensitive to small variations in containment spray temperature prior to ice meltout.

Long term equipment qualification of safety related components required to mitigate the

accident is based on a continuous, maximum NSWS supply temperature of 100oF or

less.

To ensure that the NSWS initial temperature assumptions in the limiting analysis are

met, Lake Wylie temperature is also monitored. During periods of time while Lake Wylie

temperature is greater than 95.5°F, the emergency procedure for transfer of Emergency

Core Cooling System (ECCS) flow paths to cold leg recirculation directs the operator to

align both trains of containment spray to be cooled by loops of NSWS which are aligned

to the SNSWP. Swapover to the SNSWP is required at 95.5°F rather than 95°F

because Lake Wylie is not subject to subsequent heatup due to recirculation, as is the

SNSWP. Therefore, the 100°F design basis maximum temperature is not approached.

ES-1.3 Transfer to Cold Leg Recirculation Enclosure 2 Step 13 Directs the operator to

verify adequate heat sink by determining if the RN (NSWS) system is aligned to Lake

Wylie, and if it is, to verify Lake Wylie temperature is less than 93°F. If RN is aligned to

Lake Wylie, and temperature is greater than 93°F the operator is directed to align both

trains of RN to the SNSWP.

The actions required by SLC16.9.4 and the statement in the basis document requiring the

operator when swapping to cold leg recirculation and the actions listed in ES-1.3 indicates that

the rising temperature will have an effect after the ice has been depleted.

These actions also indicate that there is more than a minimal impact during the entire accident

sequence. However, it is difficult to determine if the effects of the increased Lake Temperature

will be significant. Because the word significant is subjective in nature the NRC has

determined that there is not a correct answer and this question will be deleted.

Enclosure 2

Written Examination - Question # 89

Licensees Comment:

The developer was considering that to isolate a ruptured S/G (RSG) a level of > 11% is a

precondition that must be satisfied. However, to completely isolate a RSG steps 3 - 6 within

EP/E-3 must be performed, and step 6 which completes the isolation can only be performed if

RSG level is > 11%.

The question didnt differentiate between initiating the isolation of an RSG and completely

isolating an RSG.

Answer A is correct because, once steps 3, 4, & 5 are reached; the operator is required to

perform these actions as soon as the RSG is identified. There are no preconditions to

performing these steps.

Answer B is correct because it is part of the guidance which completes the isolation of the RSG

by isolating the auxiliary feedwater supply when level is > 11%.

If the question had asked for the guidance to completely isolate the RSG, then there would be

no correct answer to the question; however, the question asked for the procedural guidance

regarding isolation which is found in both answers A & B, so both answers A and B are correct.

We request that both answers A and B be accepted as correct.

NRC Response:

The NRC does not agree with accepting both A and B as correct answers. The term isolation is

not defined in the procedures, and was not defined in the stem. The NRC does agree that the

stem of the question did not ask for when to start the isolation, or when the isolation would be

considered complete. Answer A is not correct because the steam generator would not be totally

isolated, however answer B is not correct either because the NC system cooldown could begin

even if the affected steam generator was not isolated, and its level was less than 11% narrow

range level. Therefore, none of the answers are correct, and the question will be deleted.

Enclosure 2

SIMULATOR FIDELITY REPORT

Facility Licensee: Catawba Nuclear Station

Facility Docket No.: 05000413, & 05000414

Operating Test Administered: December, 01- 04, 2008

This form is to be used only to report observations. These observations do not constitute audit

or inspection findings and, without further verification and review in accordance with Inspection

Procedure 71111.11 are not indicative of noncompliance with 10 CFR 55.46. No licensee

action is required in response to these observations.

While conducting the simulator portion of the operating test, examiners observed the following:

Item Description

Reactivity Response During performance of scenario # 3, (EOL boron concentration of

differences between 215 ppm) the simulator exhibited different responses to the

scenarios. amount of dilution. One crew diluted only 400 gallons of water to

get the required temperature increase to allow the crew to

commence an increase in power, and two other crews did not

receive the same temperature response after diluting over 2700

gallons of water.

Simulator ANSI limits Several times during the performance of JPMs the simulator gave

exceeded the indications that it was outside of its ANSI limits. Simulator

operators had to override the issue to allow the JPMs to be

completed. Simulator Work Request SGB-009 submitted.

Critical Scenario/Simulator The simulator parameter data collection required to be saved

Data not captured. during the administration of the operating test scenarios was not

saved during the first scenario, either due to personnel error or

equipment malfunction.

Enclosure 3