ML071550160

IR 05000395-07-502, on 08/15/2006 - 05/01/2007, for Virgil C. Summer Nuclear Station

ML071550160
Person / Time
Site:	Summer
Issue date:	06/04/2007
From:	James Shea Division of Reactor Safety II
To:	Archie J South Carolina Electric & Gas Co
References
EA-07-079 IR-07-502
Download: ML071550160 (19)

See also: IR 05000395/2007502

Text

June 4, 2007

EA-07-079

South Carolina Electric & Gas Company

ATTN:

Mr. Jeffrey B. Archie

Vice President, Nuclear Operations

Virgil C. Summer Nuclear Station

P. O. Box 88

Jenkinsville, SC 29065

SUBJECT:

VIRGIL C. SUMMER NUCLEAR STATION - NRC EMERGENCY

PREPAREDNESS INSPECTION REPORT 05000395/2007502

Dear Mr. Archie:

On March 28, 2007, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection

at your Virgil C. Summer Nuclear Station. The enclosed report documents the inspection

results, which were discussed via teleconference on May 14, 2007, with you and other

members of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed

personnel. The inspection also included a review of the Emergency Plan changes that

occurred between October 1980 and July 28, 2006.

Based on the results of this inspection, one apparent violation was identified and is being

considered for escalated enforcement action in accordance with the NRC Enforcement Policy.

The current Enforcement Policy is included on the NRCs Web site at http://www.nrc.gov/about-

nrc/regulatory/enforcement/enforc-pol.pdf. Based on a review of Summers Emergency Plan

changes, the staff determined there were Emergency Action Level (EAL) changes made that

decreased the effectiveness of the emergency plan and failed to maintain a standard

emergency classification scheme. This is a performance deficiency and an apparent violation

associated with emergency preparedness planning standard 10 CFR 50.47(b)(4),

10 CFR 50.54(q), and the requirements of Section IV.B of Appendix E to 10 CFR Part 50 to

obtain NRC approval prior to implementation of a revision to an EAL that changes EAL

schemes, uses alternate methods for complying with the regulations or decreases the

effectiveness of the emergency plan.

This finding was assessed using traditional enforcement. NRC Manual Chapter 0609,

Appendix BProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix B" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Section 2.2(e) states in part, Findings that potentially impede the regulatory

process (i.e., violations that impact the NRCs ability to oversee licensees activities) are not to

be evaluated through the SDP [Significance Determination Process]. Noncompliances may be

significant because they may challenge the regulatory envelope within which certain activities

SCE&G

2

were licensed. These types of violations include failures to receive prior NRC approval for

changes that result in a decrease in effectiveness of the Plan (10 CFR 50.54(q) issues). Such

violations are to be evaluated in accordance with the guidance in Section IV of the Enforcement

Policy (traditional enforcement). Additional details associated with this determination are

discussed in Section 1EP4 of the enclosed inspection report. This finding is also determined to

potentially have greater significance because the finding represents a failure to meet planning

standard 10 CFR 50.47(b)(4) and the requirements of Section IV.B of Appendix E to 10 CFR Part 50.

Regional inspectors and program office staff reviewed Summer's EALs and determined there

were additional EAL observations that support the apparent violation. The observations are

provided in Attachment 2 and should be included in the extent of condition during your review of

the four examples given in the apparent violation. You should consider if any corrective action

for the observations is warranted and include such in your corrective action program.

Before the NRC makes its enforcement decision, we are providing you an opportunity to either:

(1) respond to the apparent violation addressed in this inspection report within 30 days of the

date of this letter or (2) request a predecisional enforcement conference. If a conference is

held, it will be open for public observation. The NRC will also issue a press release to

announce the conference. Please contact Brian R. Bonser at 404-562-4653 within 7 days of

the date of this letter to notify the NRC of your intended response.

If you choose to provide a written response, it should be clearly marked as a "Response to an

Apparent Violation in Inspection Report No. 5000395/2007502; EA-07-079" and

should include: (1) the reason for the apparent violation, or, if contested, the basis for disputing

the apparent violation; (2) the corrective steps that have been taken and the results achieved;

(3) the corrective steps that will be taken to avoid further violations; and (4) the date when full

compliance will be achieved. Your response may reference or include previously docketed

correspondence, if the correspondence adequately addresses the required response. If an

adequate response is not received within the time specified or an extension of time has not

been granted by the NRC, the NRC will proceed with its enforcement decision or schedule a

predecisional enforcement conference.

In addition, please be advised that the number and characterization of apparent violations

described in the enclosed inspection report may change as a result of further NRC review. You

will be advised by separate correspondence of the results of our deliberations on this matter.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its

enclosure, and your response, if you choose to provide one, will be made available

electronically for public inspection in the NRC Public Document Room or from the NRCs

SCE&G

3

document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/

reading-rm/adams.html. To the extent possible, your response should not include any personal

privacy, proprietary, or safeguards information so that it can be made available to the Public

without redaction.

Sincerely,

/RA/

Joseph W. Shea, Director

Division of Reactor Safety

Docket No. 0-395

License No. NPF-12

Enclosure: NRC Inspection Report No. 05000395/2007502

w/Attachments: 1. Supplemental Information

2. Listing of additional EAL Observations

cc w/encl:

R. J. White

Nuclear Coordinator Mail Code 802

S.C. Public Service Authority

Virgil C. Summer Nuclear Station

Electronic Mail Distribution

Kathryn M. Sutton, Esq.

Morgan, Lewis & Bockius LLP

Electronic Mail Distribution

Henry J. Porter, Director

Div. of Radioactive Waste Mgmt.

Dept. of Health and Environmental

Control

Electronic Mail Distribution

R. Mike Gandy

Division of Radioactive Waste Mgmt.

S. C. Department of Health and

Environmental Control

Electronic Mail Distribution

(cc w/encl contd - See page 4)

SCE&G

4

(cc w/encl contd)

Bruce L. Thompson, Manager

Nuclear Licensing (Mail Code 830)

South Carolina Electric & Gas Company

Virgil C. Summer Nuclear Station

Electronic Mail Distribution

Robert M. Fowlkes, General Manager

Engineering Services

South Carolina Electric & Gas Company

Virgil C. Summer Nuclear Station

Electronic Mail Distribution

Thomas D. Gatlin, General Manager

Nuclear Plant Operations (Mail Code 303)

South Carolina Electric & Gas Company

Virgil C. Summer Nuclear Station

Electronic Mail Distribution

David A. Lavigne, General Manager

Organization Development

South Carolina Electric & Gas Company

Vigil C. Summer Nuclear Station

Electronic Mail Distribution

Distribution w/encl:

R. Martin, NRR

C. Evans, RII EICS

L. Slack, RII EICS

RIDSNRRDIRS

PUBLIC

OFFICE

RII:DRS

RII:DRS

RII:DRP

RII:DRS

RII:EICS

SIGNATURE

RA

RA

RA

RA

RA

NAME

KREH

MILLER

E.GUTHRIE

BONSER

EVANS

DATE

4/26/2007

4/26/2007

4/30/2007

4/30/2007

4/27/2007

E-MAIL COPY?

YES

NO YES

NO YES

NO YES

NO YES

NO YES

NO YES

NO

Enclosure

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket No.:

50-395

License No.:

NPF-12

Report No:

05000395/2007502

Licensee:

South Carolina Electric and Gas

Facility:

Virgil C. Summer Nuclear Station

Location:

576 Stairway Road

Jenkinsville, SC 29065

Dates:

August 15, 2006 - May 1, 2007

Inspectors:

Lee Miller, Senior Emergency Preparedness Inspector

James Kreh, Emergency Preparedness Inspector

Approved by:

Brian R. Bonser, Chief

Plant Support Branch 1

Division of Reactor Safety

Enclosure

Summary of Findings

IR 05000395/2007-502; 08/15/2006-05/01/2007; Virgil C. Summer Nuclear Station; Emergency

Action Level (EAL) and Emergency Plan Changes

The report covered an announced inspection by two emergency preparedness inspectors. One

apparent violation (AV) was identified. The NRCs program for overseeing the safe operation of

commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 4, dated December 2006.

A.

NRC-Identified and Self-Revealing Findings

Cornerstone: Emergency Preparedness

TBD. The NRC identified an apparent violation (AV) related to the licensees

implementation of EAL changes which decreased the effectiveness of the Emergency

Plan, and related to the licensees failure to maintain a standard emergency

classification scheme. The AV is associated with 10 CFR 50.54(q), emergency

preparedness planning standard 10 CFR 50.47(b)(4), and the requirements of Section

IV.B of 10 CFR 50, Appendix E.

The licensee's implementation of EAL changes that resulted in a decrease in

effectiveness of the Emergency Plan and a failure to maintain a standard emergency

classification scheme are performance deficiencies. This finding is greater than minor

because it is associated with the Emergency Preparedness Cornerstone and affects the

cornerstone objective to ensure that the licensee is capable of implementing adequate

measures to protect the health and safety of the public in the event of a radiological

emergency. The finding is an identified weakness that demonstrates a level of

performance that could preclude effective implementation of the Emergency Plan in an

actual emergency. This finding is also determined to potentially have greater

significance because the finding represents a failure to meet planning standard 10 CFR 50.47(b)(4) and the requirements of Section IV.B of 10 CFR 50, Appendix E to obtain

NRC approval prior to implementation of a revision to an EAL that changes EAL

schemes, uses alternate methods for complying with the regulations or decreases the

effectiveness of the emergency plan. (Section 1EP4)

B.

Licensee-Identified Violations.

None

Enclosure

REPORT DETAILS

1.

REACTOR SAFETY

Cornerstone: Emergency Preparedness

1EP4 Emergency Action Level (EAL) and Emergency Plan Changes

a.

Inspection Scope

Between August 15, 2006 and May 1, 2007, an in-office review of Summers EALs was

conducted by the emergency preparedness program office staff and region based

emergency preparedness inspectors. The program office staff determined that Revision

5 of the Emergency Plan was the last NRC-approved revision. Revision 53 of the

Emergency Plan was the current revision at the time of the last Emergency

Preparedness Program inspection for Summer. The licensees available documentation

for Revisions 5 through 53 was reviewed to determine if the changes made to the EALs

had decreased the effectiveness of the emergency plan. The EAL revisions were also

reviewed to determine if they were consistent with the guidance in NUREG-0654,

Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and

Preparedness in Support of Nuclear Power Plants, Appendix 1, Emergency Action

Level Guidelines for Nuclear Power Plants.

The in-office review was conducted in accordance with NRC Inspection Procedure 71114, Attachment 04, Emergency Action Level and Emergency Plan Changes. The

applicable planning standard, 10 CFR 50.47(b)(4), and its related 10 CFR 50,

Appendix E requirements were used as reference criteria. The criteria contained in

NUREG-0654 and Regulatory Guide 1.101, Emergency Planning and Preparedness

For Nuclear Power Reactors, Revision 4, were also used as references.

The inspectors reviewed multiple change packages for the licensees emergency plan,

EP-100, Radiation Emergency Plan, Revision 5 through Revision 53.

b.

Findings

Introduction. The NRC identified an apparent violation (AV) related to the licensees

implementation of EAL changes which decreased the effectiveness of the Emergency

Plan and related to the licensees failure to maintain a standard emergency classification

scheme.

Description. NRC Inspection Report 05000395/2006012, dated September 1, 2006,

identified an Unresolved Item (URI), URI-05000395/2006012-01, Blending of EAL

Schemes. The URI was opened to review the Summer EALs to determine if the

changes made between October 1981 (Revision 5) and April 2006 (Revision 53)

resulted in decreases in effectiveness or a blending of EAL schemes.

4

Enclosure

Summers EALs are based on NUREG-0654. Revision 5 to the Summer EALs was the

last NRC-approved set of EALs for Virgil C. Summer Nuclear Station. Revision 53 was

the revision the inspectors used to compare to the last NRC-approved EAL revision.

The standard approved EAL schemes are based on NUREG-0654, NUMARC/NESP-

007, or NEI 99-01. Revisions to the EALs that contain combinations of two or more of

the standard EAL schemes are termed blending of EALs and must be approved by the

Commission prior to implementation. Revisions to the EALs that are not consistent with

the standard EAL schemes are termed non-standard EALs and must be approved by

the Commission prior to implementation.

The staff review identified two General Emergency and two Site Area Emergency EAL

changes that decreased the effectiveness of the emergency plan and/or resulted in a

failure to maintain a standard EAL scheme. The EALs are from EP-100, Table 4-1. The

four examples, documented below, compare the NRC approved revision, Revision 5, to

Revision 53, the current revision at the time of the inspection.

Attachment 2 provides additional observations that support these findings.

1) General Emergency (GE) EAL Number 411

The Initiating Condition in Revision 5 stated, Transient initiated by loss of feedwater

and condensate systems (principle heat removal system) followed by failure of

emergency feedwater system for extended period. Core melting possible in several

hours. Ultimate failure of Reactor Building possible if core melts.

The Detection Method in Revision 5 stated, Reactor trip on low feedwater flow; and

Decreasing wide-range steam generator levels toward off-scale low on all steam

generators; and 1) Emergency feedwater flow indicators indicate zero flow 2 min.

after required; or 2) Status lamps indicate emergency feedwater pumps not running

2 min. after required and Emergency feedwater cannot be restored within 30 min.

The Initiating Condition for EAL number 411 remained essentially the same between

Revisions 5 and 53. However, significant changes were made to the detection

methods between Revisions 5 and 53.

The Detection Method in Revision 53 stated, ALL of the following exists (1 AND 2):

1. Inability to Establish Bleed and Feed Cooling when required per EOP-15.0 - AND

- 2. Core Exit Temperatures > 700°F.

The revised EAL applied more restrictive criteria to when the EAL would be met and

could reduce the number of classifiable events or could delay the GE declaration.

As a result, the EAL changes appear to have resulted in a decrease in the

effectiveness of the emergency plan. Additionally, the revised detection methods

were not consistent with the standard EAL schemes, resulting in a non-standard

EAL.

5

Enclosure

2) General Emergency (GE) EAL Number 401

The Initiating Condition in Revision 5 stated, Small or large LOCA with failure of

ECCS to perform leading to severe core degradation or melt. Ultimate failure of

Reactor Building possible for meltdown sequences.

The Detection Method in Revision 5 stated, Safety injection signal with reactor trip;

and 1) Status lamps indicate safety injection system and RHR pumps not running :

or 2) Flow indicators for Safety Injection Systems read zero; and RMG-5, RMG-7,

RMG-18, high alarms; and RM-A2 high alarm.

The Initiating Condition for EAL number 401 remained essentially the same between

Revisions 5 and 53. However, significant changes were made to the detection

methods between Revisions 5 and 53.

The Detection Method in Revision 53 stated, Failure of BOTH of the following after

depressurizing the RCS to < 140 psig per EOP-14.0. Failure of (1 AND 2): 1. High

Head Injection Flow AND 2. Low Head Injection Flow

The revised EAL applied more restrictive criteria to when the EAL would be met, and

could reduce the number of classifiable events or could delay the GE declaration.

As a result, the EAL changes appear to have resulted in a decrease in the

effectiveness of the emergency plan. Additionally, the revised detection methods

were not consistent with the standard EAL schemes resulting in a non-standard

EAL.

3) Site Area Emergency (SAE) EAL Number 301

The Initiating Condition in Revision 5 stated, Known Loss of Coolant Accident

(LOCA) greater than charging pump capacity.

The Detection Method in Revision 5 stated, Pressurizer low pressure reactor trip;

and Pressurizer low pressure safety injection signal, and RM-A2 high alarm; and

High Reactor Building sump level; and High Reactor Building humidity; and High

Reactor Building pressure.

The Initiating Condition for EAL number 301 remained essentially the same between

Revisions 5 and 53. However, significant changes were made to the detection

methods between Revisions 5 and 53.

The Detection Method in Revision 53 stated, ANY of the following indications (1 OR

2 OR 3 OR 4):

1. Evaluate the following indications to determine if a LOCA condition exists

(similar to EOP-1.0):

a. Pressurizer low pressure reactor trip.

b. Pressurizer low pressure safety injection.

6

Enclosure

c. Reactor Building pressure > 1.5 psig,

d. Abnormal Reactor Building sump level,

e. RBCU Drain Flow High,

f. Abnormal radiation levels on RM-A2 or RM-G7, or RM-G18. - OR -

2. Direct Entry into EOP-2.0 from EOP-1.0 due to the RCS NOT Being Intact. -

OR -

3. Stuck Open and Unisolable Pressurizer PORV or Safety Valve Leading to

Pressurizer Relief Tank Rupture - OR -

4. Initiating Bleed and Feed per EOP-15.0. (Refer to Initiating Condition 411

for possible escalation.)

The changes to the EAL may increase the number of classifiable SAE events. An

unwarranted SAE declaration is a non-conservative action which may place

members of the public at risk during an unnecessary evacuation process. The

Revision 5 detection methods were definitive criteria that had no procedural delay in

reaching a determination for the SAE declaration. The detection methods which are

reliant on an EOP transition point or entry point could result in a delay in making the

SAE declaration. As a result, the EAL changes appear to have resulted in a

decrease in the effectiveness of the emergency plan. Additionally, the revised

detection methods were not consistent with the standard EAL schemes resulting in a

non-standard EAL.

4) Site Area Emergency (SAE) EAL number 397

The inspectors noted that there was not an equivalent EAL in Revision 5 for EAL

number 397. This EAL was not included in the NUREG-0654 EAL scheme nor was

it required by NRC regulations. The licensee's Revision 23 to their EALs stated that

the EAL was added per Generic Letter (GL) 87-12, Loss of Residual Heat Removal

While The Reactor Coolant System is Partially Filled, but provided no evaluation or

technical basis to support the addition of this EAL. NRC review of GL 87-12

determined that there was no requirement to add an EAL.

The Initiating Condition in Revision 53 stated, Loss of Residual Heat Removal flow

for more than 40 minutes during half-pipe operations with vessel head installed and

High Head Safety Injection/charging unavailable.

The Detection Method in Revision 53 stated, ALL of the following (1 THROUGH 5)

For a period greater than 40 minutes:

1. Both RHR Loop A FLO LO AND RHR Loop B FLO LO annunciators in alarm, -

AND-

2. NEITHER RHR pump is running, - AND -

3. Core exit thermocouple temperatures increasing or at saturation temperature for

the current RCS pressure - AND -

4. Reactor Vessel Head is in place and RCS loops are drained to 434-7.43 or less.

- AND -

5. NEITHER train of Charging/SI is available.

7

Enclosure

This EAL was not consistent with the standard EAL schemes resulting in a non-standard

EAL.

Analysis. A performance deficiency was identified for the licensees implementation of

EAL changes which decreased the effectiveness of the Emergency Plan and for the

licensees failure to maintain a standard emergency classification scheme. This finding

is greater than minor because it is associated with the Emergency Preparedness

Cornerstone and affects the cornerstone objective to ensure that the licensee is capable

of implementing adequate measures to protect the health and safety of the public in the

event of a radiological emergency. The finding is an identified weakness that

demonstrates a level of performance that could preclude effective implementation of the

Emergency Plan in an actual emergency. This finding is also determined to potentially

have greater significance because the finding represents a failure to meet the

requirements of 10 CFR 50.54(q), the risk significant planning standard 10 CFR 50.47(b)(4), and the requirements of Section IV.B of 10 CFR 50, Appendix E.

MC 0609, Appendix B, § 2.2.e states, in part: "Findings that potentially impede the

regulatory process (i.e., violations that impact the NRCs ability to oversee licensees

activities) are not to be evaluated through the SDP. Noncompliances may be significant

because they may challenge the regulatory envelope within which certain activities were

licensed. These types of violations include failures to receive prior NRC approval for

changes that result in a decrease in effectiveness of the Plan (10 CFR 50.54(q) issues).

Such violations are to be evaluated in accordance with the guidance in Section IV of the

Enforcement Policy (traditional enforcement)."

Enforcement. 10 CFR 50.54(q) states, in part, The nuclear power reactor licensee may

make changes to these plans without Commission approval only if the changes do not

decrease the effectiveness of the plans and the plans, as changed, continue to meet the

standards of § 50.47(b) and the requirements of Appendix E to this part. Planning

standard 10 CFR 50.47(b)(4) states, A standard emergency classification and action

level scheme, the bases of which include facility system and effluent parameters, is in

use by the nuclear facility licensee, and State and local response plans call for reliance

on information provided by facility licensees for determinations of minimum initial offsite

response measures. Section IV.B of 10 CFR 50, Appendix E states, in part: A revision

to an EAL must be approved by the NRC before implementation if: (1) the licensee is

changing from one EAL scheme to another EAL scheme (e.g., a change from an EAL

scheme based on NUREG-0654 to a scheme based upon NUMARC/NESP-007 or

NEI 99-01); (2) the licensee is proposing an alternate method for complying with the

regulations; or (3) the EAL revision decreases the effectiveness of the emergency plan.

8

Enclosure

Contrary to 10 CFR 50, Appendix E, between October 1980 and April 2006, the licensee

made changes to its Emergency Plan which decreased the effectiveness of the Plan

and were not consistent with the NUREG-0654 EAL scheme. These changes were not

submitted to the NRC for approval prior to implementation.

This finding is identified as Apparent Violation (AV) 50-395/2007501-01, EAL Changes

Resulted in Decreases in Effectiveness and a Non-Standard EAL Scheme. This issue

has not yet been entered into the licensee's corrective action system.

4.

OTHER ACTIVITIES

4OA6 Meetings, including Exit

On May 14, 2007, the inspectors conducted a telephonic exit to discuss the results of

the inspection with Mr. J. Archie, Vice President, Nuclear Operations, and other

members of his staff. The inspectors confirmed that no proprietary information was

received by the inspectors during the inspection.

Attachments: 1. Supplemental Information

2. Listing of Additional EAL Observations

Attachment 1

1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

J. Archie, Vice President, Nuclear Plant Operations

D. Gatlin, General Manager, Nuclear Plant Operations

S. Zarandi, General Manager, Nuclear Support Services

G. Lippard, Manager, Operations

R. McCauley, Quality Assurance

F. Miller, Supervisor, Quality Control

B. Thompson, Manager, Nuclear Licensing

A. Cribb, Supervisor, Nuclear Licensing

R. Williamson, Supervisor, Emergency Services

NRC

J. Zeiler, Senior Resident Inspector

B. Bonser, Chief, Plant Support Branch 1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000395/2007502-01

AV

EAL Changes Resulted in

Decreases in Effectiveness and a

Non-Standard EAL Scheme (Section

1EP4)

Opened and Closed

None

Closed

05000395/2006012-01

URI

Blending of EAL Schemes.

(Section 1EP4)

Discussed

None

Attachment 1

2

LIST OF ACRONYMS

AV

Apparent Violation

CFR

Code of Federal Regulations

EAL

Emergency Action Level

ECCS

Emergency Core Cooling System

EOP

Emergency Operating Procedures

ERO

Emergency Response Organization

FLO LO

Flow Low

GE

General Emergency

GL

Generic Letter

IMC

Inspection Manual Chapter

LOCA

Loss of Coolant Accident

NEI

Nuclear Energy Institute

NUMARC/NESP

Nuclear Management and Resources Council/National Environmental

Studies Project

PORV

Power Operated Relief Valve

RBCU

Reactor Building Cooling Unit

RCS

Reactor Coolant System

RHR

Residual Heat Removal

SAE

Site Area Emergency

SDP

Significance Determination Process

SI

Safety Injection

TBD

To Be Determined

URI

Unresolved Item

Attachment 2

1

LISTING OF ADDITIONAL EAL OBSERVATIONS

Regional inspectors and program office staff reviewed Summer's EALs and determined

that an apparent violation for failure to maintain a standard EAL scheme and for

decreases in effectiveness of the licensee's emergency plan had occurred. Four EALs

received an in-depth review and were used as examples for Apparent Violation (AV)05000395/2007502-01, EAL Changes Resulted in Decreases in Effectiveness and a

Non-standard EAL Scheme (Section 1EP4). Additional observations were noted with

other EALs. The observations are provided below and should be included in the

licensee's extent of condition during their review of the four examples provided in the

AV. The licensee should consider if any corrective action for the observations is

warranted and include such in their corrective action program.

The observations are listed by EAL number. The listing is sequenced in the following

order: General Emergency, Site Area Emergency, Alert, and Notification of Unusual

Event.

General Emergency

402 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase Pressurizer low pressure reactor trip and Pressurizer low pressure

safety injection signal and added the phrase loss of primary or secondary

coolant in progress.

403 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase LOCA as identified in Site Emergency and containment status panel

indicates incomplete isolation; and added the phrase a. dose equivalent I-131

activity > 1 ci/gm in primary coolant. OR b. core exit temperature >700EF.

431 - Detection method changes between Revisions 5 and 53 resulted in an additional

requirement for verification of primary coolant dose equivalent I-131 activity >

300ci/gm for RM-L1 alarm.

441 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase Undervoltage alarms on 1DA and 1DB buses for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and

added the phrase Either a OR b a) Steam driven Emergency Feedwater pump

fails to start AND is unavailable for one hour OR b) Core exit temperature

>700EF

493 -

Revision 53 has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

Site Area Emergency

302 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrases pressurizer low pressure alarm and reactor trip, or pressurizer level

rapidly decreasing; and Pressurizer low-level alarm; and undervoltage alarms on

Attachment 2

2

1DA and 1DB and steam generator water level rapidly increasing in one or more

steam generators, falling in the others, and ... The changes added the

condition Entry into EOP-4.0.

303 - Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase RM-A2 high alarm with no justification and added the requirement

for verification of dose equivalent I-131 activity > 300 ci/gm.

321-

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase Thot and Tcold rapidly diverging (T rapidly increasing) or no T

across core and added the phrase no indication of forced or natural circulation.

The requirement for verification of failed fuel monitor offscale (>106 cpm) with

determination of dose equivalent I-131 activity > 300 ci/gm was added.

322 - Detection method changes between Revisions 5 and 53 resulted in deletion of

RM-A2" with no justification and addition of RM-G5.

341-

Revision 53 has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

342 -

Revision 53 has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

361 -

Revision 53 has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

362 - Detection method changes between Revisions 5 and 53 resulted in addition of

the phrase radiation monitoring teams measure thyroid dose rates (equivalent I-

131concentrations) at one mile or greater from the plant. The initiating condition

specifies at the exclusion area boundary which the emergency plan defines as

within one mile.

371 - Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase major fire that defeats redundant safety system trains or functions

and added the phrase fire that renders both trains of a safety system or function

inoperable per the Technical Specifications.

392b - Detection method changes between Revisions 5 and 53 resulted in direction to

the user to Initiating Condition 394, which contains an error.

392c - Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase as detected by portable instrumentation AND which renders a train

of safety related system inoperable.

394 - Detection method changes between Revisions 5 and 53 resulted in deletion of

the OR between methods 2 and 3. The revisions added the phrase AND

between methods 2 and 3.

Attachment 2

3

396 -

Revision 53 has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

Alert

201 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the and/or between the first and second detection methods in Revision 5 and

added a third detection method IPC CHGNET.

202 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase pressurizer low pressure alarm and reactor trip; and pressurizer low

level alarm; and RM-A9 high alarm and pressurizer safety injection signal and

undervoltage on 1DA and 1DB buses; and RM-L3, RM-L7, and RM-L10 high

alarm, and added the phrase all of the following: Primary to Secondary

Leakage exceeds 10 gpm as determined per AOP-112.2 AND safety injection is

NOT required per AOP-112.2 AND a loss of offsite power has led to the loss of

condenser vacuum.

203 -

Detection method changes between Revisions 5 and 53 resulted in deletion of all

the Revision 5 methods and added the phrase Entry into EOP-4.0.

204 -

Detection method changes between Revisions 5 and 53 resulted in deletion of all

the Revision 5 methods and added the phrase EOP network has determined a

faulted steam generator exists and primary to secondary leakage exceeds 10

gpm ...

221 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phase laboratory analysis which indicates an increase in failed fuel of 1% in

30 minutes or a total failed fuel of 5%.

222 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

RM-A2" with no justification and the addition of RM-G5.

262 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

RM-L3" and the addition of RM-A3 (Gas and Iodine) and RM-A4 (Gas).

Several of the radiation monitor alarm setpoint levels were changed. The

reference to setpoints established in the discharge permit or while steam

generator blowdown is directed to the blowdown system for RM-L5,

RM-L7, and RM-L9 was also deleted.

271 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase fire potentially affecting safety systems and added the phrase fire

that has the potential for renders both trains of a safety system inoperable per

the Technical Specifications.

296 -

Detection method changes between Revisions 5 and 53 did not incorporate the

loss of indication in the control room or compensating non-alarming indications

unavailable with a significant transient in progress.

Attachment 2

4

297 -

Revisions 53 has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

Notification of Unusual Event

101 -

Revision 53 has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

104 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase reduced RCS temperature and pressure and added the phrase see

Initiating Condition 102."

106 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrases Dose equivalent I-131 activity concentration greater than the limit in

figure 3.4-1 of Technical Specifications and laboratory analysis which indicates

an increase in failed fuel of 0.1% in 30 minutes. The revision added the phrase

Primary coolant dose equivalent I-131 activity > 300 ci/gm.

107 -

Revision 53 has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

108 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

RM-A9, RM-A10, and RM-A13" and deletion of the phase in valid alarm mode

for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and added specific values above background which had to

be maintained for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the radiation monitors .

109 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

RM-L10 and RM-L3." No 50.54(q) documentation was available for review.

112 -

Revision EAL has a detection method that appeared to be inconsistent with the

NUREG-0654 scheme.

115 -

Detection method changes between Revisions 5 and 53 resulted in deletion of

the phrase significant loss of vital assessment.