ML19322A152

Forwards Responses to NRC 781129 Ltr Re Practice of Containment Purging During Normal Plant Operations. Justification for Unlimited Purging Is Provided W/Evaluation of Safety Actuation Signal Circuits

ML19322A152
Person / Time
Site:	Fort Calhoun
Issue date:	12/28/1978
From:	Short T OMAHA PUBLIC POWER DISTRICT
To:	Reid R Office of Nuclear Reactor Regulation
References
NUDOCS 7901030058
Download: ML19322A152 (7)

Text

.

Omaha Public Power District -

1023 HARNEY a OMAHA. NE8RASKA 68102 a TELEPHONE 536 4000 AREA CODE 402 December- 28, 1978 Director of Nuclear Reactor Regulation ATTN: Mr. Robert W. Reid, Chief Operating Reactors Branch No. 4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Reference:

Docket No. 50-285 Gentlemen:

The Omaha Public Power District received a letter from the Com-mission, dated Ncrember 29, 1978, concerning the practice of contain-ment purging during normal plant operation. Specifically, the letter requested that a commitment be made to cease all containment purges during operation of the Fort Calhoun Station, or that a justification be provided for continued purging. In addition, the letter requested that the design of all safety actuation signal circuits which incor-porate a manual override feature be reviewed. Accordingly, the Dis-trict offers the attached discussions which justify unlimited purging operations and provide an evaluation of the aforementioned safety actuation signal circuits.

Pending completion of the NRC staff review of the justification for continued purging, the District commits to limit purging to an absolute minimum not to exceed 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year; however, this does not preclude venting through the 2" vent line to keep containment pressures within Technical Specification limits. It must be emphasized that a timely review of this document is required in order to assure Fort Calhoun Station availability. This is because efficient opera-tion of the unit depends upon the ability to purge containment during operation, as more fully explained in the attachment.

Sincerely, n )

1. Shor Assistant General Manager TES/KJM/BJH:jmm

' Attach.

l cc: LeBoeuf, Lamb, Leiby & MacRae

! 1757 "N" Street, N. W. j (

Washington, D. C. 20036 Y,

i

\

i 7901030c n p

r a

e ,

1. BASIS F0F CONTINU2D CONTAIU!EIT PURGING General System Description The containment purge system is a separate part of the containment air cooling and ventilation system (ACVS) as described in subsection 9 10.2.3 of the Fort Calhoun FSAR. The ACVS consists of two other separate systems besides the containment purge system; i.e. , the containment air recirculation cooling and iodine removal system and the nuclear. detector well cooling system. The containment purge system in designed to perform the following functions:

! a. Provides means for the reduction of concentrations to radio-active materials, including noble gases, in the containment.

Noble gases cannot be reduced by the filtration equipment in the containment air recirculation, cooling, and iodine removal system; and

b. Ventilates the containment building to provide a suitable en-vironment during personnel access.

Operating Characteristics Operating experience at Fort Calhoun Station has indicated a need to frequently purge the centainment to control the airborne acti-vity levels to permit personnel access during various modes of opera-tion. Fort Calhoun Station is also equipped with a 2" containment pressure relief line (very small size as compared to 8" lines pro-vided for new plants) to alleviate containment pressure increases which may be caused by temperature mad humidity transients during startups or air leakage from pneumatic controllers during reactor operating modes. It should be emphasized that, for these reasons, all new plant designs include the capability to purge and vent the containment continuously through about eight inch, in diameter, purge / pressure relier lines.

In addition to the above-described reasons for containment purging, sampling and inspection of certain equipment located in the contain-ment is performed during reactor operation in accordance with the Technical Specifications. This requires personnel entry in the containment which, in most cases, necessitates the operation of the containment purge system in order to reduce the atmospheric activity in the containment within the occupational MPC levels.

Also, sometimes it becomes necessary to purge the containment building, while the reactor is in hot standby or hot shutdown condition, to allow personnel access to the containment for per-forming any maintenance on the safety-related equipment / components.

It should also he pointed out that subsection 9.10.h.3 of the FSAR allows the use of containment purge system during power operation, prior to and during personnel access to the containment.

• l l

1

Response to Issues Relating to Purcinc in SRP 6.h.2 and BTP CSB 6 h (Includes evaluation of impact of purging during operation of ECCS performance and evaluation of radiological consequences of DBA re-quiring containment isolation during purge operations. )

The design of the containment purge system is in compliance with the applicable sections of Acceptance Criteria under Standard Re-view Plan , Section 6. 4.2, Revision 1. As the Branch Technical Position CSB 6-4 provides guidance on the design and use of con-tainment purge systems which may be used during the normal plant operating modes, therefore, it is appropriate to address all the items discussed under Section B of the Branch Technical Position.

First of all, it needs to be pointed out that the design of the Fort Calhoun Station does not have the provisions for an "on-line purge system" which could be used independe'nt of the purge system during the cold shutdown and refueling modes. The following para-graphs delineate responses to all the acceptance criteria listed under Section B of the Branch Technical Position.

B.1 Acceptance Criteria for the Purge System Design The containment purge system design is in conformance with all the requirements under this item with the exception of paragraph 1.c which is not considered a potential discrepancy, especially considering the fact that the radiological conse-quences for the existing containment purge system would not be inimical to the health and safety of the public as dis-cussed further under paragraph B.5.a. Further detailed dis-cussions on all the criteria addressed under this item can be found in Sections 5.9, 7.3, and 9 10 of the FSAR.

B.2 Purge System Reliance for Temperature and Humidity Control The containment purge system is not designed to control the temperature and humidity in the containment. These parameters, along with pressure increases due to leakage from air lines and valve operators inside the containment, are controlled by the 2" containment pressure relief line. See also the above discussion under " Operating Characteristics".

B.3 Provisions for Containment Atmosphere Cleanup Systems The design of Fort Calhoun Stction does have the provisions for a separate containment atmosphere cleanup system designated as the containment air recirculation, cooling, and iodine re-moval system. This system, however, cannot reduce the noble gases concentration in the containment prior to personnel entry during the reactor operating modes. Therefore, on these occasions, the containment purge system is relied upon for diluting the containment atmosphere.

2

- B. h Provisions for Testing During Reactor Operation Provisions for testing the availability of the valves in question are made as follows:

a. The valves are placed in the non-accident position, an

. accident- signal is then generated, and the valves' move-

. ' ment to their accident position is verified.

b. The leakage rate of the purge valves on line is performed, one penetration at a time, in accordance with the test metho'ds outlined in Technical Specification 3.5(3).

B.5 Analyses Regarding the Purge System Design

a. The containment purge system isolation valves are designed to close, including instrumentation delays, within five secor.ds . This design capability is provided in order to mirimize the release of containment atmosphere to the environs, to mitigate the offsite radiological conse-quences and to assure that emergency core cooling sys-tem (ECCS) effectiveness is not degraded by a reduction in the containment backpressure (see also response to paragraph B. 5. c ) .

Following a LOCA, the purge valves vill receive a signal to close within three seconds and will be closed within two additional seconds (see FSAR, subsection T.3.2.6).

In this time, it is assumed that the contents of the reactor coolant system are emptied into the containment vessel, and fission products contained in the water are available to pass through the purge valves. This is a conservative assumption since the blowdown is not com-plete at this time. The resulting activity releases are estimated to be:

1. 31 dose equivalent curies of I-131; and f

2. 590 dose equivalent curies of XE-133.

Applying a five-percentile atmospheric dispersion factor of 6.30 E-04 sec/m3 for ground level releases, the two-hour doses at the exclusion area boundary are:

1. 10.0 rems to the thyroid; and

2. 0.0h rems to the whole body.

Therefore, the containment purge system isolation design assures that, in the event of a LOCA, the radiological ,

consequences would not exceed 10 CFR Part 100 limits.

3

(

{

b. There are no safety-related fans, filters, and ductuork located in the vicinity of the containment purge system, especially beyond the purge system isolation valves (FSAR, Sections 6.0 and 9 10). Therefore, the loss of function or inoperability of the containment purge system, due to the environment created by escaping air / steam, will not affect the performance of any safety-related equipment required to function following a LOCA.

c. The criterion regarding the containment purge valves closure time within five seconds was considered during the ECCS evaluation. It was concluded the ECCS effective-ness would not be degraded due to any partial loss of containment pressure during and/or following a LOCA.

d. The leakage rate criteria for the purge and vent isola-tion valves is determined by subsection 3.5(3)b. of the Technical Specifications in which the combined leakage rate for all contributing valves is insured to be less than 0.6 La-(where La means the maximum allowable leak-age rate at the DBA pressure equivalent to 60 psig).

Each individual valve's leakage rate is reviewed for its leakage contribution and maintainability to insure that the combined leakage of 0.6 La is not exceeded.

Evaluation of Purge and Vent Isolation Valves Capability to Close Against Dynamic Forces of DBA The .large h2-inch containment purge valves PCV-Th2 A, B, C, D and the smaller 2-inch vent line valves HCV-Th6 A, B are capable of closing against the dynamic forces of a design basis loss-of-coolant accident as described in the FSAR. Section 7.3 2.6 of the FSAR states that, in a DBA, the purge valves will receive a signal within three seconds and will be closed within another two seconds. In addition to the evidence presented in the FSAR, the District has received verification from the valve manufacturers that valve closure can be accomplished against a DBA pressure of 60 psig.

2. REVIEW OF SAFETY ACTUATION SIGNAL CIRCUITS WHICH INCORPORATE A MANUAL OVERHIDE FEATURE The District has reviewed all safety actuation signal circuits which incorporate a manual override feature to ensure that over-riding of one safety actuation signal does not also cause the by-pass of any other safety actuation signal, that sufficient physical features are provided to facilitate adequate administrative controls, and that the use of each such manual override is annunciated at the system level for every system impacted.- l The valve control circuits for all valves actuated by the follow- .

[ ing safeguards actuation signals were inspected, both on an individual

[ basis as well as'a system basis: PPLS, CPHS, CRHS, CSAS, STLS, L

4 ,

1-j- .

1 -

-p s SGLS , SIAS , RAS , CIAS , and VI AS.

• In addition, the RPS/ safeguards

- signal generation and actuation circuitry including sequencer and diesel generator actuation circuitry were also inspected. Major

! areas of interest in the inspection were:

a. Whether the valve control circuit contained a safeguard actu-ation signal override feature.

b. Whether or not there was a safeguards actuation signal over-ride "at the system level".

c. Whether or not the overriding feature described in a. and b.

above was annunciated at the system level.

d. Whether or not the overriding feature described in a. and b.

above caused the bypass of any other safeguards actuation signals.

With the exception of the control circuitry for the PCV-2909, 2929, 2949, and 2969 valves (safety injection leakage cooler valves which are referred to later in this letter), all circuits / designs and related administrative procedures were found- to be in conformance with the functional requirements of IEEE 279, " Criteria for Pro-tection Systems for Nuclear Power Generating Stations", as described in Section 7 3.1 of the FSAR. In particular, a review and evalua-tion of containment purge isolation equipment has related that the only override associated with the PCV-Tk2 A, B, C, and D valves is via the containment isolation valve test switches, 01/AI 43A and 01/AI-43B located on panel AI-43 Placing the 01/AI-43A switch in TEST will override the CIAS safeguards actuation contacts con-tained in the control circuit to PCV-Th2 A/C. Placing the 01/AI h3B switch in TEST will override the CIAS safeguards actuation contacts contained in the control circuit of PCV-Th2 B, D. It should, how-ever, be pointed out that the surveillance test procedures preclude the simultaneous placement of 01/AI-43A and 01/AI h3B switches in the " TEST MODES" during various operating modes. Also, the placing of either of these test switches in TEST, however, is annunciated and will in no way adversely affect the operation of any other safe-guards actuation signal in the purge valve's control circuitry.

• PPLS Pressurizer Pressure Low Signal SGLS Steam Generator Low Signal CPHS Containment Pressure High Signal SIAS Safety Injection Acutation Signal CRHS Containment Radiation High Signal RAS Recirculation Actuation Signal CSAS Containment Spray Actuation Signal CIAS Containment Isolation Actuation STLS Safety Injection Tank Low Signal Signal C

VIAS ~ Ventilation Isolation Actuation Signal ,

5 9

p.

.-*.4- j

-Listed as the lone exception vere the safety injection leakage i cooler valves. The control switch for these valves (PCV-2909, 2929, 29h9, -mnd 2969) incorporates an " override" position. When set to this. override position, the SIAS actuation signal to these valves is overridden / bypassed. Presently, there is no annunciation of the override condition of these valves.

As a result of these investigations, all safety actuation si6 nal

' circuits which incorporate a manual override feature have been found to be adequate to the extent that operation of a bypass will affect no safety functions other than those analyzed and discussed on the Fort Calhoun docket. The safety injection leakage cooler valves control circuitry vill be modified to provide annunciation i of override, or procedural control vill be established to preclude bypassing the circuits during power operation, hot standby, or hot shutdown. These controls will be established by the next refueling outage at the Fort Calhoun Station.

t E

6

.