ML17059B100

and 50.73 Reportability

3.3. 1

and 50.73 Reportability

October

1993

.

3.3.2

and 50.73 Reportability - March 1995

.

3.4

Procedural

Compliance

.

. '.

.

4

5

6

7

8

4.0

REVIEW OF

LER 95-05,

"BUILDING BLOWOUT PANELS OUTSIDE THE

BASIS BECAUSE OF CONSTRUCTION

ERROR,"

DESIGN

9

5.0

SAFETY SIGNIFICANCE

.

5. 1

Review of the Unit

1 Individual

5.2

Review of the

UFSAR Commitments

5.3

Safety Assessment

and

Summary

.

6.0

MANAGEMENT MEETINGS

.

~

Plant Examination

.

~

9

10ll

12

0

DETAILS

1. 0

INTRODUCTION

1.1

Purpose of Inspection

t

The purpose of this special

inspection

was to review concerns identified by

.

the

NRC staff during the review of the Nine Mile Point Unit

1 (Unit 1)

Licensee

Event Report

(LER) 95-05, "Building Blowout Panels

Outside the Design

Basis

Because of Construction Error," dated

November 30,

1995.

1.2

Event Overview

On October 25,

1993, with the reactor at

IOOX power, Unit

1 engineering staff

determined that the safety-related

blowout panels in the reactor

and turbine

buildings would not blowout at the design relief pressure

of 45 pounds per

square foot (psf).

The purpose of the blowout panels is to provide pressure

relief to prevent collapse of the superstructure

due to a break of an

emergency cooling system,

or other primary coolant system line in the reactor

building,

and

a steam line break in the turbine building.

Unit

1 found that

the existing shear bolts

on the blowout panels

were larger than those

identified on the design drawings.

This was documented

in a Deviation/Event

Report

(DER 1-93-2526).

The initial engineering

evaluation indicated that the

turbine

and reactor building blowout panels

would relieve at

a pressure

greater

than the design basis value,

but less

than the design basis fail

pressure for the reactor

and turbine buildings.

The

1993 engineering

evaluation

recommendation

and

NMPC resolution

was to leave the as-found

condition in place.

On March 27,

1995, during the completion of the recommended

actions

included

in the

DER,

an engineering

review determined that the blowout panels

would

actually not relieve until pressures

in excess of the structural

design of 80

psf for both buildings.

At this time, Niagara

Mohawk Power Corporation

(NMPC)

completed

a design

change to bring the relief pressures

back in conformance

with the

During subsequent

reviews of the issue,

NMPC determined that

the condition was reportable

under Title 10 of the

Code of Federal

Regulations

Part 50.73,

(10 CFR 50.73)

"Licensee

Event Report System,"

and issued

LER

95-05.

1.3

Scope of Inspection

During this inspection,

the inspectors

reviewed

LER 95-05, applicable

Technical Specifications

(TSs),

Updated Final Safety Analyses

Report

(UFSAR)

and Individual Plant Examination

(IPE) sections,

portions of related

calculations,

procedures,

DERs, Station Operations

Review Committee

(SORC)

meeting minutes

and other licensee

documentation.

The inspectors

also

conducted interviews with various

members of the

NMPC staff and management,

and conducted

walkdowns of the applicable

areas of the facility.

The

inspectors

focused their review on the following aspects

of the issue:

~ Engineering

Support/Design

Control

~ Reportability/LER Adequacy

~ Safety Review and Assessment

~ Proper

Procedure

Implementation

0

Additionally, the inspector

assessed

the accuracy of the applicable

UFSAR

sections.

2.0

DETAIL DESCRIPTION

OF THE EVENT

Attachment

1 to this report is

a time line of events for this review.

During NMPC's effort to resolve contradictions identified in the

UFSAR

regarding the blowout panel relief pressure,

Calculation

S7-RX340-WOl, dated

August 23,

1993, was'enerated.

Due to a lack of documentation

regarding the

material properties of the bolts identified on the design drawings,

NMPC

determined to test

a sample of the installed bolts to obtain actual material

properties.

After initial calculations

were performed,

a number of bolts were

replaced with new 3/16" diameter American Society for. Testing

and Materials

(ASTM) A-307 bolts.

The previously installed bolts were tested to determine

their strength.

Upon receipt of the test results,

the structural

engineer

identified that the bolts were 1/4",

and not 3/16" as specified

on plant

drawings

(C-18713-C).

Furthermore,

the test results indicated that the

strength of the bolts to be higher than that used in Calculation

S7-RX340-W01.

Therefore,

the structural

engineer initiated Revision

1 to the calculation,

which indicated that the relief pressure for the reactor building to be

53

psf,

and

60 psf for the turbine building.

On October 28,

1993,

DER 1-93-2526

was written to address

the difference

between the size of the installed blowout panel bolts and the size indicated

on the plant drawings.

When the

DER was reviewed

by the

on November

1,

1993,

an operability determination

was attached that indicated the relief

pressures

of blowout panels for both the reactor

and turbine buildings would

exceed the value described

in the

Because

the calculated relief

pressures

were less

than the buildings internal failure pressure,

engineering

recommended

that the blowout panels still be considered

operable.

The

accepted this recommendation.

Additionally, the

SSS did not consider the

condition to be reportable.

As part of DER 1-93-2526 Action Plan, Unit

1 was to complete

a calculation to

identify exactly which bolts were required to be replaced to restore the

blowout panels

in. conformance with the

UFSAR relief pressure

of 45 psf.

This

calculation

was scheduled

to be completed

by June 30,

1995.

The calculation

was completed

on March 27,

1995, during the Unit

1 refueling outage

13.

Based

upon the results of this calculation,

the licensee

determined different relief

pressures

for the as-installed

configurations.

These

new relief pressures

were in excess of the fail pressure for both the reactor

and turbine buildings

as stated in the

According to the engineering

supervisor,

there

was

an error in the assumptions

used during the October

1993 calculation that caused

the previous incorrect

results.

Particularly, loading of the panels

was

assumed to be equally

distributed in both the horizontal

and vertical directions.

Therefore,

the

engineer incorrectly concluded that

a failure of the sheet

metal at the top of

the panels

would be sufficient to relieve pressure

as required.

To provide

'ufficient

pressure relief,

a failure of the bolts connecting the sides

of. the

panels to the supports

would have

been required.

Based

on the

new

3

calculations

the correct relief pressure for the reactor building was

91 psf,

and 89 psf for the turbine building.

No DER was written at this time to

address

the

human performance

issues

associated

with the design control

deficiencies

inherent in the calculation error and the independent

design

review (this area is described further in Section 3.3).

During the re-

disposition of the

DER in March 1995,

NMPC indicated that the event

was not

- reportable

under

10 CFR 50.72 or 50.73 for the following reasons:

~

UFSAR Section XVI.D.2.0 states

that Unit

1 was designed prior to 10 CFR 50 Appendix A, General

Design Criterion 4,

and that it was not designed

for the dynamic effects of a double-ended guillotine pipe rupture,

and

that the probability of this kind of occurrence

is extremely low.

The above

assumption

was substantiated

by a 1984 leak-before-break

analysis.

This analysis

concluded that

a full double-ended

pipe break

need not,be postulated

as

a design

basis for defining loads at Unit 1.

The results of this study were

used to define the Unit

1 design basis

for masonry walls at Unit 1,

and were submitted to the

NRC via letter

dated

June 8,

1984.

The leak-before-break

analysis eliminates the need

for the blowout panels,

because

the high energy line break

(HELB) event

would be preventable

by detection of the leak,

and timely shutdown would

follow.

Therefore,

the condition was not outside the Unit

1 design

bases.

During the

1995 refueling outage,

Unit

1 evaluated

the situation

and decided

to remove every other bolt used to hold each of the blowout panels

in place

due to the higher relief pressures.

This would provide

a relief pressur e of

approximately

45 psf as per the

DER 1-93-2526

and associated

oper ability determinations

were updated

and the blowout panels

were declared

inoperable,

on March 27,

1995.

Modification Nl-95-001

LG329 was initiated,

and the bolts were removed prior to plant restart

from the refueling outage in

1995.

The Structure

Engineering Supervisor determined that because

NMPC was

completing the design

change to place the relief pressure

back in accordance

with the

UFSAR,

no additional

analyses

were needed;

such

as analyses

to

determine the actual

internal fail pressures

of the reactor

and tur bine

buildings, the events that would exceed

these fail pressures

or the subsequent

consequences

of exceeding

these failure pressures.

During the closeout

SORC review of DER 1-93-2526,

on June

22,

1995, the

SORC

questioned

the reportability of the issue,

and requested

that engineering re-

evaluate

the reportability of the events.

Nuclear Engineering confirmed the

bases for the earlier decision not to report the condition.

This information

was documented

in

NMPC Memorandum

ESB1-S95-0039 to file, and presented

to

SORC

on July 6,

1995.

On October 31,

1995,

DER 1-95-3012

was written to prompt another evaluation of

the condition for reportability.

This

DER was generated

as

a result of a non-

required review of original condition performed

by NMPC personnel.

As a

result,

NMPC determined to report the condition under

10 CFR 50.73(a)(2)(ii).

LER 95-05 was submitted to the

NRC on November 28,

1995.

According to

LER 95-

05, the use of engineering, judgement

was improperly credited in the previously

0

concluding that this condition was not reportable

and that the

UFSAR inferred

that the relief panels

were credited with functioning for certain events.

DER 1-95-3012 identified the failure to properly report the condition to be

caused

by inadequate

engineering

reviews,

and inadequate

investigation of the

reportability requirements

when the error was identified.

Based

on these

causes,

NNPC initiated the following corrective actions:

~

Emphasis to be given to the Structural

Engineering

Group to perform an

adequate

review of documentation

and verification of assumptions

used in

calculation before final issue.

(Completed shortly after the end of

Refueling Outage

13.)

~

Engineering to write a lessons

learned transmittal to address

the

responsibility of engineering staff to promptly inform operations

when

, new information is identified that could affect reportability.

(Scheduled for completion October 1996.)

~

Provide

a training session

to certain engineering

and plant personnel

regarding reportability (NUREG-1022,

and

10 CFR 50.72,

and 50.73),

and

include .this training in the

NHPC continuous training cycle.

(Scheduled

for completion October 1996.)

3.0

ASSESSMENT

OF THE EVENT

Based

on the inspectors'eview,

concerns

were identified in the following

areas:

Design Control;

50.59 safety evaluations;

Reportability;

Procedure

compliance;

and

Root Cause

and Corrective actions

as described

in the

LER.

3.1

Design Control

The inspectors

reviewed the following facets of design control:

~

the root cause for the installation of the oversized bolts during

construction;

~

the October

1993 calculational

err or;

and

~

the corrective actions to address

the calculational error.

3. 1.1 Root Cause of the Installation of Oversized Bolts during Construction

The cause of the oversized bolts used to install the reactor

and turbine

building blowout panels

was documented

in LER 95-05 to be inadequate

quality

control measures

in place during construction.

The corrective action credited

in the

LER is additional quality control

and quality assurance

requirements

that have

been

im'plemented for the design

and construction activities since

r

'

initial construction,

which should prevent similar. deficiencies.

NMPC

reported that, Unit

1 was designed

and constructed prior to the implementation

of 10 CFR 50 Appendix 6, "guality Assurance Criteria for Nuclear Power Plants

and

Fuel Reprocessing

Plants."

The inspectors

acknowledged

NMPC statement

ia

this regard,

but did not focus

on the performance

problems in the construction

period.

Those

same quality assurance

requirements

referenced

to by

NMPC were

in effect during the performance

problems since

1993.

The inspectors

focused

their review on the time period from 1993.

3.1.2 October

1993 Calculational

Error

In October

1993, after the oversize bolts were identified, the engineer

made

an error in his assumptions

leading to the determinations of the. incorrect

relief pressure

fo} the safety-related

reactor

and turbine building blowout

panels.

Specifically, the assumption

that the loading of the panels

was

equally distributed in both the horizontal

and vertical directions resulted

in

the incorrect determination that the reactor

building blowout panels

would

relieve at 60 psf and the turbine building blowout panels

would relieve at 53

psf.

This error was not caught

by the checker or the approver of the

calculation

as part of design review.

As determined

in March 1995, the actual

relief pressures

would be

91 psf for the reactor building and 89 psf for .the

turbine building.

According to the

UFSAR, the reactor building and turbine

building blowout panels

were designed to relieve at 45 psf to prevent failure

of the building superstructures

at pressures

in excess of 80 psf.

This is an apparent violation of 10 CFR 50 Appendix 8 Criterion III, "Design

Control," in that Calculation

S7-RX340-W01 incorrectly determined

the blowout

panel relief pressures

to be less than the 80 psf failure pressure

of the

reactor

and turbine buildings,

and the calculation. was inadequately

design

reviewed.

3. 1.3 Corrective Actions to Address the Calculational

Error

The inspectors

reviewed Modification Nl-95-001

LG329, which initiated the

removal of every other blowout panel bolt,

and identified no immediate

concerns.

The inspectors verified that the applicable drawings

and

calculations

wer e updated.

Additionally, the inspector s also

"walked down"

the blowout panels for the reactor building and verified through sampling that

the installed configuration was consistent with the plant drawings.

However,

the

NRC did not complete

a detail review of the calculations

as of the close

of the inspection period.

Region I is performing

a confirmatory independent

review of the calculations.

This is considered

an unresolved

item pending the

completion of NRC staff's review.

(URI 50-220/96-05-01)

3.2

10 CFR 50.59 Safety Evaluations

The inspectors

evaluated

the licensee's

implementation of 10 CFR 50.59

"Changes,

Tests

and Experiments," for the relief pressures

exceeding

the

values stated

in the

UFSAR identified in October

1993,

and for the

modification to remove every other bolt used to install the blowout panels

completed in March 1995.

0

'

3.2.1

10 CFR 50.59 Safety Evaluation - October

1993

In October

1993, during NMPC's evaluation of the installed oversized bolts,

NHPC determin'ed

the blowout panels to be operable.

With respect to the

original relief pressure of 45 psf and the new calculated relief pressures

of

approximately

60 psf both being less

than the structural

design value

(80

psf), clearly the safety margin was reduced in these facts.

Unit

1 decided to

leave the oversized bolts installed until the completion of their corrective

actions,

scheduled

to be completed

June 30,

1995.

When this decision

was

made,

no

10 CFR 50.59 evaluation

was completed.

A delay or partial correction

of a condition adverse to safety or quality for a structure,

system,

or

component described

in the

UFSAR is considered

by the

NRC staff to be

a change

in the facility, which is subject to a 10 CFR 50.59 review.

Additionally, the

above facts indicated

a reduced safety margin that needed to be evaluated

in

accordance

with 10 CFR 50.59.

h

The inspectors

considered

the failure to complete

a

10 CFR 50.59 evaluation,

to allow for the approx'imately

one

and

a half year delay in resolving the

differences

between

the

UFSAR stated

design relief pressures

and the

installed/calculated relief pressures

for the reactor

and turbine building

blowout panels,

an apparent violation on

10 CFR 50.59.

3.2.2

10 CFR 50.59 Safety Evaluation

March 1995

During the evaluation of Modification N1-.95-001

LG329,

an applicability review

was completed

by the Unit

1 staff indicating no need for the conduct of a

10 CFR 50.59 safety evaluation.

The inspectors

noted that the

NHPC documented

basis in the applicability review for not completing

a

10 CFR 50.59 safety

evaluation

was because

the proposed

change

would bring the, facility back onto

compliance with the

The inspector s acknowledged this basis in the

applicability review,

and also verified that the size

and spacing of the

blowout panel bolting were not described

in the

However,

changes

in

~

the facility as described

in the

UFSAR are considered

by the

NRC staff

(NRC

Manual Chapter Part

9900) to pertain to any changes

in the facility which

alter the design,

function, or method of performing the function of a

component,

system

or structure described

in the

Accordingly, the

NHPC

Modification Nl-95-001

LG329 made in March 1995, which consisted of the

removal of every other blowout panel bolt, is considered

by the

NRC staff to

be

an .alteration to the design of a structure described

in the

Therefore,

the failure of NHPC to perform

a safety evaluation for the subject

modification is considered

another

example of an apparent violation of 10 CFR 50.59.

3.3

10 CFR 50.72 and 50.73 Reportability

The inspectors

evaluated

the licensee's

implementation of 10 CFR 50.72

and

50.73,

"Immediate notification requirements for operating

nuclear

power

reactors,"

and "Licensee. event report system," for the relief pressures

exceeding

the design values stated

in the

UFSAR identified in October

1993,

and for the relief pressures

exceeding

the structural

design pressures

of the

reactor

and turbine buildings values stated

in the

UFSAR identified in March.

1995.

"

7

The inspectors

reviewed the applicable revisions of the licensee's

procedure

regarding reportability and determined it to provided appropriate, requirements

to ensure that

and 50.73 reportability regulations related to

conditions that are outside of the design basis of the plant.

3.3. 1 10 CFR 50.72 and 50.73 Reportability October

1993

The inspectors

discussed

with the licensee their reasoning for not declaring

the event reportable

under

10 CFR 50.72,

50.73 in October

1993,

and was

informed, that even though the relief pressure

of the blowout panels

exceeded

the design basis

values stated in the

UFSAR, the structural

design basis

pressure of the reactor

and turbine buildings would not be exceeded.

Additionally,

NMPC reviewed their design basis

and determined that

HELBs were

outside their design basis that there

was

no credible postulated

event that

would cause

pressures

to challenge

the failure pressure of the reactor or

turbine buildings.

Based

on these

reasons,

NHPC determined that they were not

outside the design basis; therefore,

the condition was not reportable.

The inspectors

evaluated

the reportability decision

made

by NHPC.

Based

on

the definition of Design Basis

as provided in 10 CFR 50.2,

"Design bases

means that information which identifies specific functions to be performed

by

a structure,

system,

or component of a facility, and the specific values or

ranges of values

chosen for controlling parameters

as reference

bounds for

design...."

Since the specific values

chosen for controlling the relief

pressures

of the reactor

and turbine building blowout panels

were exceeded

as

stated in the

UFSAR, the inspectors

considered

Unit

1 to be in a condition

outside their design basis.

The inspectors

concluded that the Unit

1 failure to complete the reports,

as

required in October

1993, is an apparent violation on

and 50.73.

3.3.2

and 50.73 Reportability Narch

1995

During the period of March through July 1995,

NHPC evaluated

the reportability

of the blowout panel relief pressures

on several

occasions.

Through

discussions

with NHPC staff and management,

the inspectors

ascertained

that

NHPC understood that they were outside their design basis relief pressures

of

45 psf as stated

in the

Furthermore,

they understood that the

calculated relief pressures

exceeded

the 80 psf failure pressure for the

reactor

and turbine buildings,

as stated in UFSAR Sections

VI-C.1.2 and III-

A.l.2, respectively.

However,

NHPC decided that the condition was not outside

the Unit

1 design

bases,

since the

UFSAR Sections

XVI.D.2.0 describes

that

HELB events

are not part of the Unit

1 design basis;

and, therefore,

no

postulated

pressure-related

event would challenge

the reactor building blowout

panels.

The inspectors

evaluated

the reportability decision

made

by

NHPC for this time

period.

Unit

1

UFSAR Sections

VI-C.1.2,

"Pressure

Relief Design," describes

specifically, that the reactor building pressure relief is provided to prevent

collapse of the superstructure

due to a break of an emergency cooling system,

or other primary coolant system line in the reactor building.

Further,

the

pressure relief function is provided by the blowout panels,

that were designed

0

to fail at

an internal

pressure

of approximately .45 psf, to prevent excessive

internal pressure

on the superstructure

walls, roof, and their supports,

which

would fail at

an internal pressure

in excess of 80 psf.

A similar description

is provided for the turbine building in UFSAR Section III-A.1.-2, specifying

the initiating event

as

a steam line break.

Since the specific values

chosen for controlling the relief pressures

of the

reactor

and turbine building blowout panels

as stated in the

UFSAR were

exceeded,

and the specific values

chosen for the reactor

and turbine building

failure pressure

as stated in the

UFSAR were exceeded,

the inspectors

considered

Unit

1 to be in a condition'utside their design basis.

Furthermore,

the detailed description of the emergency cooling system line

break

and the steam line break provided in the

UFSAR as the events for which

the. blowout panels

were installed to protect against,

indicated that there

are

analyzed pressure-related

events

would challenge

the reactor

and turbine

building blowout panels.

The inspectors

concluded that Unit

1 failure to complete the reports,

as

required in March 1995, is an apparent violation on

and 50.73.

3.4

Procedural

Compliance

The inspectors

reviewed the applicable revisions of the licensee's

DER

procedure to verify the completion of DER 1-93-2526

was in accordance

with the

procedure.

Although the inspectors

found the

DER completed

as required

by the

procedure,

the inspectors

noted that the length of time taken

between the

identification of the problem until the review of the

was longer than

expected

considering

the potential significance of the condition.

Particularly, the condition was identified on October 25,

1993, the

DER was

initiated on October 28,

1993,

and signed

by the