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Safety Insp Repts 50-317/89-08 & 50-318/89-08 on 890410-14. No Violations Noted.Major Areas Inspected:Licensee Actions in Response to Expeditious Actions Described in Generic Ltr 88-17, Loss of DHR, During Nonpower Operation
	ML20244E272
Person / Time
Site:	Calvert Cliffs
Issue date:	05/30/1989
From:	Eapen P, Moy D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20244E269	List: ML20244E267; ML20244E272;
References
	50-317-89-08, 50-317-89-8, 50-318-89-08, 50-318-89-8, GL-87-12, GL-88-17, IEIN-88-036, IEIN-88-36, NUDOCS 8906200241
	Download: ML20244E272 (9)
	v • d • e

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U.S. NUCLEAR REGULATORY COMMISSION

REGION

Report Nos. 50-317/89-08

50-318/89-08

Docket Nos. 50-317

50-318

License Nos. DPR-53

DPR-69

Licensee: Balttnore Gas and Electric Company

Calvert Cliffs Nuclear Power Plant

MD Routes 2 &3

Lusbv. Maryland 20657

Facility Name: Calvert Cliffs Units 1&2

Inspection At: Lusi v. Maryland

Inspection Conducted: Anffl 10-14. 1989

Inspector: N8k '

bN

D. T . M'oy , fRe " tor Engineer Date

Approved By: Q' K "

Dr. P.

&J

K. Eapen,' Chief, Special

IAD /f(9

' ate /

Test Programs Section, DRS

Inspection Summary: Routine unannounced safety inspection on

April 10-14, 1989. (Inspection Report Nos. 50-317/89-08 and

50-318/89-08.)

Areas Inspected: Review of licensee actions in response to the

" expeditious actions" described in Generic Letter 88-17, " Loss of

Decay Heat Removal" during non-power operation. The inspection

reviewed instrumentation, training, procedures and plant staff

awareness as related to mid-loop operation.

Results: All " expeditious actions" described in Generic Letter 88-17 were implemented at the Calvert Cliffs Nuclear Station

prior to drain down to mid-loop operation. The management

involvement, training and staff awareness to problems related to

mid-loop operation were adequate. Procedures, instrumentation,

and systems required to support mid-loop operation were found to

be consistent with the licensees response to Generic Letter 88-

17. No violations were identified.

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8906200241 890607

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DETAILS

1.0 Persons Contacted

1.1 Baltimore Gas and Electric Company

Charlie Cruse, Manager, Nuclear Eng. Service Dept.

J. E. Gilbert, Supervisor, Procedure Development and

Modification Acceptance Unit

John R. Hill, Supervisor, Operation Training

P. E. Katz, General Supervisor, Design Engineering

'L. S. Larragoite, Engineer, Licensing Unit

Keven Nietmann, Supervisor, Nuclear Training !

Lee Russell, Manager, Calvert Cliffs

Don Shaw, Licensing Engineer

Alan Thornton, Project Engineer

J. E. Thorp, Senior Engineer

Don Ward, Principal Engineer

Raymond Wenderlich, General Supervisor, Nuclear.Oper.

i- 1.2 U.S. Nuclear Reaulatory Commission

H. Eichenholz, Senior Resident Inspector, Calvert

Cliffs

V. Pritchett, Resident Inspector, Calvert Cliffs

Denotes those attending the exit meeting on 4/13/89. The

inspectors also contacted other administrative and-

technical personnel during the inspection.

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l 2.0 Review of Licensee Action in Response to Generic Letter No.

88-17. Loss of Decay Heat Removal (TI2515/101)

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Loss of decay heat removal (DHR) during non-power operation

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and the consequences of such a loss have been of increasing

concern to the NRC. Many events of loss of DHR have

occurred while the reactor coolant system has been drained

down for mid-loop activities such as steam generator

inspection or repair of reactor coolant pumps. The

possibility exists to bypass two fission product barriers as

the ~9 actor coolant system and containment may both be open

while the.mid loop activities are in progress.

GL 87-12, " Loss of Residual Heat Removal (RHR) while the

Reactor Coolant System (RCS) is partially filled" was issued

to all licensees of operating PWR's and holders of

construction permits on July 9, 1987. Respcnses indicated

that the licensees did not understand the identified

problems, and the problem continued as evidenced by events

at Waterford on May 12, 1988 and Sequoyah on May 23, 1988.

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The seriousness and continuation of this problem has

resulted in the issuance of GL 88-17. In addition, the

Director of NRR has written to the CEO of each licensee

operating a PWR, in which he said, "We consider this issue

to be of high priority and request that you assure that your

organization addresses it accordingly." He also wrote to

each licensed operator at all PWR plants on " Operator

Diligence while in Shutdown Conditions," and enclosed a copy

of Generic Letter 88-17.

GL 88-17 requires the recipients to respond with two plans

of actions:

a. A short-term program entitled " Expeditious Actions,"

and

b. A long-term program entitled " Programmed Enhancements."

This inspection assessed the short-term licensee actions as

outlined in the " Expeditious Actions" Section of GL 88-17.

2.1 Review of Licensee Response to Generic Letter 88.17.

NRC reviewed the licensee's responses dated January 3,

1989 to Generic Letter 88-17. This review concluded

that the licensee responses met the intent of the

generic letter with respect to expeditious actions,

even though the responses to some items were brief.

The NRC staff reviewed the licensee's mid-loop

operations in 1988 as detailed in the NRC inspection

reports 50-317/88-16 and 50-318/88-16.

The inspector reviewed the licensee response dated

January 3, 1989 to Generic Letter 88-17 to understand

the actions committed to by the licensee. The licensee

response provided a description of action taken to

address the eight recommended expeditious actions

identified in the Generic Letter. The inspector

verified that the licensee actions were implemented

during mid-loop operation in accordance with the NRC

guidance in Generic Letter 88-17 as detailed below.

2.2 Core Exit Temperature Indications

The inspector verified that for mid-loop operation, the

licensee has taken adequate administrative and

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procedural steps to provide at least two independent,

continuous coolant temperature indicators that are

representative of the core exit conditions. The

licensee monitors the core exit temperature using

thermocouple. At least two thermocouple are

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maintained operable to monitor core exit temperature

during mid-loop operation. A third thermocouple is

used as standby. The inspector reviewed step (A) in

Appendix (5) of Operating Procedure (OP)-5 " Plant

Shutdown from Hot Standby to Cold Shutdown." This step

requires as an initial condition that at least two RCS

temperature indications be provided in the control

room.

The inspector verified that there are at least two

independent core exit temperature indications in the

control room. The process computer alarm is 20 degrees

F above the current Core Exit Thermocouple (CET)

temperature. Periodic logging of CET temperature is

required whenever CET temperature or process computer

temperature alarm is not functional.

The inspector also verified that the core exit

thermocouple temperature recorders (2TR-131A for CET's

T13 and #T33 and 2TR-1323 for CET #T31) were

functioning adequately. The indicated temperatures

were within the acceptance limits specified in OP-5.

Based on the above, the inspector concluded that the

temperature indication system is consistent with the

expeditious actions of Generic Letter 88-17.

2.3 RCS Water Level Indication

The inspector verified that the licensee has procedures

and administrative controls to provide at least two

independent continuous RCS water level indications

whenever the RCS is in a reduced inventory condition.

Water level indication for mid-loop operation is

provided by a permanently installed pressure

transmitter which senses pressure at the bottom of the

piping between steam generator No. 11 and the reactor

vessel. Indication is provided in the control room by

a digital meter which is calibrated before each.use.

The meter has operator adjustable high and low level

alarms. The setpoints are controlled by the operating

procedure for a reduced inventory condition.

The inspector reviewed step (F) in Appendix (5) of

l Operating Procedure (OP)-5 " Plant Shutdown from Hot

Standby to Cold Shutdown" for RCS water level

indication. The water level alarm is set at +0.2 feet

of the refueling level. This setpoint was chosen based

on the Calvert Cliffs past operating experience.

In addition to the water level indicator in the control

l room a tygon tube provides local indication at the 27

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ft. elevation in the containment. This level

indication is also taken from the bottom of the piping

between steam generator No. 11 and the reactor vessel.

The licensee had provided accurate elevation marks

(from 27 ft. elevation) so that the temporary level

indicating scale for the tygon tubing could be properly

placed. According to step (F) in Appendix (5) of

Operating Procedure (OP)-5, the indication between this

tygon tubing water level and centrol room water level

should agree within +0.50 ft. reading. The indicated

water levels were within the acceptance range specified

in OP-5.

The inspectors observed that the water level monitor

cabinet in the control room was not secured. This

unsecured cabinet has the potential to damage other

safety related equipments in the control room in the

event of Design Base Earthquake (DBE). The inspector

identified this matter to the licensee and the licensee

issued Non-Conformance Report (NCR) No. 7882 to

evaluate this problem and take corrective and

preventive action, as necessary.

On April 14, 1989, the licensee issued revision 2 to

the " Electrical and Controls Section Standard Practice

No. 36, Storage of Transient / Semi-portable Equipment

in Safety-Related Areas," to include the control room

as one of the safety-related areas to be covered by the

procedure. The inspector reviewed a copy of tnis

revision to standard practice No. 36 and had no further

questions in this regard.

Based on the above, the inspector concluded that the

licensee's level indicating system was consistent with

the expeditious actions described in Generic Letter 88-

17.

2.4 RCS Inventory Control

The inspector verified that the licensee has procedures

and administrative controls to provide at least two

means of adding inventory to the RCS, in addition to

pumps that are a part of the normal DHR systems. One

source of inventory makeup is from the High Pressure

Safety Injection (HPSI) pump. The second independent

source of makeup comes from the Containment Spray (CS)

pumps. Containment spray pumps are aligned to pump

water from the Refueling Water Tank (RWT) to the

Reactor Coolant System (RCS). Each injection flow path

will provide sufficient flow to keep the core covered.

The inspector verified that the water flow path for

these systems does not bypass the reactor vessel before

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exiting any opening in the RCS. Based on the above,

the inspector concluded that the licensee actions in

this regard are consistent with expeditious action

described in Generic Letter 88-17 and had no further

questions concerning this issue.

2.5 RCS Perturbations

The inspector verified that the licensee has

implemented procedures and administrative controls to

avoid operations that cause level perturbations in the

RCS. Based on interviews and system walkdowns with the

reactor operators, the inspector concluded that the

operators are adequately trained to preclude

unnecessary RCS/SDC perturbations. Furthermore, step

(D) in Appendix (5) of OP-5 procedures, requires the

reactor operator to log each perturbation in the

control room log. The Control Room Supervisor (CRS) is

required to evaluate each such logged perturbation.

Inspector verified that the level of detail for each of

the RCS perturbation log entries is adequate for

tracking and evaluating RCS perturbation problems.

Based on the above, the inspector concluded that the

licensee procedures and controls were consistent with

the requirements described in the Generic Letter

concerning this issue.

2.6 Steam Generator Nozzle Dams

The inspector verified that the licensee has

implemented procedures and administrative controls to

assure that all hot legs are not blocked simultaneously

by nozzle dams unless a large enough vent path to

prevent pressurization of the upper plenum of the

reactor vessel is provided. The licensee uses steam

generator nozzle dams during mid-loop operation. The

recommendation of NRC Information Notice 88-36 had been

incorporated into the licensee's maintenance procedure

SG-19, " Installation Use and Removal of Steam Generator

Primary Nozzle Dams." This procedure provides the

proper sequence of steam generator manway and nozzle

dams installation and removal along with appropriate

caution notes and the reasons for the required

sequence. The steam generator nozzle dam maintenance

procedure requires an adequate reactor vessel head vent

prior to installing all steam generator hot leg nozzle

dams.

Based on the above, the inspector concluded that the

licensee established adequate measures to control

activities related to nozzle dam installation and

removal during mid loop operations.

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2.7 Loon Ston Valves

Loop stop valves are not part of the Calvert Cliffs

Units'1&2 system design.

2.8 Containment Closure

The inspector verified that the licensee has prepared

procedures and administrative controls to reasonably

assure containment closure before core uncovery

following'a loss of DHR event.

The licensee's estimate for containment closure time in

the event of a loss of shutdown cooling event are:

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a. If openings totaling less than one square inch are

present in the RCS cold leg or reactor coolant

pump (RCP), closure will be within 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

b. If openings totaling greater than one square inch \

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are present in the RCS cold leg or RCP, closure

will be within 45 minutes.

c. If openings totaling greater than one square inch

are present in the RCS cold leg or RCP and a

sufficient vent path is provided for the reactor

vessel head to prevent a loss of inventory out the 4

maintenance opening, closure will be accomplished-

within 2.0 hours0 days 0 hours 0 weeks 0 months .

The inspector reviewed step (E) in Appendix (5) of OP-5

procedure for containment closure. Containment

integrity is established per containment integrity

verification procedure STP No. 055A-2. Any closure or

penetration which deviate from the containment j

integrity verification procedure will be logged and

maintained by the shift supervisor in the control room.

Based on the review of STP No. 055A-2 procedure and the

closure deviation log, the inspector concluded that the

licensee has adequatae measures to isolate the

containment prior to core uncovery. l

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2.9 Trainina

The inspector verified that the licensee training made j

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" the licensee personnel adequately aware of the risks

associated with mid-loop operation.

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The inspector reviewed training program documentation,

held discussions with the training supervisor and

senior reactor operators regarding reduced inventory

operation at Calvert Cliffs Stations. Both the

supervisor and senior reactor operators were

knowledgeable of the Loss of Decay Heat Removal event

at Diablo Canyon Unit 2 in 1987 and the lessons learned

from it. In addition, there is adequate procedural

information in the control room to guide the operators

during a loss of shutdown cooling event. The inspector

reviewed the following:

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Loss of shutdown cooling with the RCS in reduced

inventory condition for licensed operator

requalification training program (lesson plan LOR-

202-3B-89 & 203-5-89).

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Pre-shift briefing (GSO instruction 88-5).

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Loss of component cooling water / shutdown cooling

(lesson plan LOR-202-3B-1002-89).

The inspector interviewed parsonnel responsible for

installing temporary core exit thermocouple jumpers,

containment closure and steam generator nozzle dam

installation. All these personnel were knowledgeable

of their inspection responsibilities. Based on the

above, the inspector concluded that the licensee has

established and implemented an effective training

program to provide guidance to personnel during a loss

of shutdown cooling event.

2.10 Summary

Licensee management and supervision were actively

involved in the conduct of reduced inventory operation.

This involvement was particularly evident in the

establishment of procedures for mid loop operations.

These procedures thoroughly addressed the concerns

discussed in Generic Letter 88-17. The personnel

involved in mid-loop operation were knowledgeable and

adequately trained in their responsibilities. The

involvement of supervision was particularly evident in

the review of containment closure deviations and core

exit temperature deviations by control room

supervisors. The licensee was responsive to the

concerns addressed in Generic Letter 88-17.

During the inspection, the inspectors raised a concern

regarding the unsecured level instrumentation cabinet

in the control room for mid loop operation. The

licensee promptly issued an LER to assess this concern.

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Prior to the conclusion of the inspection, the licensee

revised an existing' procedure for the control of

movable objects in the vital areas to include the

control room.

Licensee's QA involvement for mid-loop operation was

not very apparent. For example, the inspector observed

no QA/QC coverage for the mid loop operation.

Additionally, the applicable procedures did not have

" hold points" for verification by the line organization

or QA.

3.0 Plant Tours

The inspector made several tours of the plant including

the control room, auxiliary building and turbine

building.to observe work in progress, housekeeping and

cleanliness. No unacceptable conditions were noted.

4.0 Exit Interview

On April 13, 1989, an exit interview was conducted with

the License's senior site representatives (denoted in

Section 1) to summarize the observations and

conclusions of this inspection. At no time during this

inspection was written material provide to the licensee

by the inspector. Based on the NRC Region I review of

this report and the discussions held with licensee

representatives during this inspection, it was

determined that this report does not contain

information subject to 10 CFR 2.790 restrictions.

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ML20244E272

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