ML20236E900

Insp Rept 50-361/87-24 on 870903,04 & 14-18 & 1001 Telcon.No Violations Noted.Major Areas Inspected:Licensee Response to 870831 Event Involving Packing Ejection from Unisolable Shutdown Cooling Sys Valve & Resultant Primary Coolant Leak

ML20236E900
Person / Time
Site:	San Onofre
Issue date:	10/14/1987
From:	Russell J, Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20236E893	List: ML20236E891 ML20236E900
References
50-361-87-24, NUDOCS 8710300055
Download: ML20236E900 (6)

See also: IR 05000361/1987024

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o U. S.- NUCLEAR'RE'GULATORY COMMISSION-

REGION V.

l Report-No. 50-361/87-24

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. Docket No. 50-361-

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~ License No; NPF-10-

. Licensee: . Southern' California Edison Company

2244= Walnut Grove Avenue

Rosemead, California 91770.

{ Facility Name: San Onofre Nuclear Generating Station - Unit 2

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Inspection'ati San Onofre Nucle Gene. ng Station

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Inspector: M

M Russell, Radiation 1 pecialist

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Date Signed

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' Approved'by: hbb

'G. P.

. s, Chief _

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Date Sign ~ed

Facili Radiological Protection-Section

Summary:

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Inspection on September 3,~4 and'14 through 18, and a telephone conversation

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on October 1, 1987 (Report No. 50-361/87-24)-

~ Areas Inspected: This was a special, unannounced inspection to evaluate the-

-licensee's-response to an event on. August 31,.1987, involving packing ejection

from.an unisolable Shutdown Cooling-(SDC) System valve and resultant primary

l coolant leak.' Inspection procedures 30703, 90712 and 93702 were. addressed.

Results: No items of noncompliance were identified in the areas examined.

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DETAILS

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.1' Persons Contacted

E ' Licensee Personnel l

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W._ Moody, Deputy Site Manager

J.1Reilly, Technical Manager _

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P..Knapp,-Health Physics..(HP) Manager ,

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R. Krieger,: Operations Manager '

DE Shull, Maintenance Manager

.c M.3 Wharton,' Assistant Technical Manager i

W. Marsh,-Units;2/3 Superintendent: ;j

R.1 Warnock, HP. Engineering Supervisor-  !

..K.. Helm,: Effluent Engineering Supervisor

J. Madigan, Units 2/3 HP Supervisor

C.' Couser,. Comp 1iance Engineer

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All-of the'above noted individuals.were present at the exit ~ interview on

. September'18,.1987. In' addition to the individuals. identified, the:

inspector met and held discussions with'other members of the licensee's

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2. . Initial Licensee Report

San Onofre' Nuclear. Generating Station (SONGS) made.a report in accordance .j

with 10 CFR 50.72 at 2030, PDT, August 31, 1987, that the Unit 2 SDC i

suction isolation valve 2HV-9378 had developed a packing leak of 50 toi

.100 gpm at 1900, PDT, August' 31, 1987. The leak began while operators  !

were attempting to manually manipulate.the valve following failure of the

valve motor operator. The leak was not' isolable from the primary. system.

.The' unit was in Mode 5, having been shutdown on August 28, and was at 350

psi,and 125 F. The valve is located inside containment at deck height on

the:30' elevation approximately 120 feet from the pl. ant equipment hatch i

which was open at the time of the leak. The hatch was closed at 2017, l

PDT. The' leakage drained into the containment sump. .j

Licensee evaluation of the event in accordance with plant procedure

5023-VIII-1, Recognition and Classification of Emergencies, considered

that it was of insufficient magnitude.to be classifiable as an Unusual

Event in that containment purge monitor, 2RE-7828, indicated effluent J

radioactivity concentrations significantly below the Unusual Event

indication level and other event criteria applicable specifically for

modes 1-4 were not considered applicable as the plant was in mode 5.

Region V resident inspector personnel responded to the site at 2120, PDT, j

-August 31, to monitor licensee actions. On September 1 the licensee  ;

reported that five maintenance personnel had been contaminated during i

efforts to stop the leakage and that approximately 326 Curies, primarily '

noble gas, had been released from the site.

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3.- NRC On-site Followup.

A regionally. based Radiation Specialist reported to the site at 1130, i

PDT, September 3, 1987,' to review the radiation safety aspects of the ]

event. Based on a review of containment entry logs, Radiation Exposure s

Permits (REPS), surveys and discussions with licensee Operations, f

Maintenance and HP personnel, the following was revealed. A Shift

Superintendent Accelerated Maintenance (SSAM) action had been implemented

in accordance with procedure 50123-0-21, Equipment Status Control, to

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investigate valve'2HV-9378, which would not open. The valve isolates a

ten inch 500 line from the Reactor Coolant System (RCS) and is used in

parallel with a sixteen inch line during forced cooldown. The isolation j

valves in the sixteen inch line had opened properly with their motor

operators and cooldown was being provided through that portion of the

system. Valve 2HV-9378 had failed to open with its. motor operator. The q

SSAM and followup Maintenance Order (MO) #87082460 were initiated and <

several attempts were made to manually operate the valve. REP #81010, ,

Maintenance Support Activities U-2 Ctmt, provided radiological controls  !

for the attempt at approximately 1845 on August 31. The radiological

controls appeared appropriate for the manual operation of the valve.

however, while a worker was using a pipe wrench as a " cheater" on the

valve handwheel, the follower plate studs were apparently broken, the

valve packing ejected and an estimated 100 gpm leak began. I

At the time of the event the primary system was at approximately 350 psi

and 130 F. Primary coolant activity levels were approximately 1.6 pCi/ml ,

noble gas, 0.56 pCi/ml particulate activation products, and 0.43 pCi/ml

radioiodine; due to fission product spiking after shutdown and a crud

burst initiated by plant Chemistry's addition of hydrogen peroxide.

There was also a potential for irradiated fuel particles in the primary

coolant due to fuel degradation. The equipment hatch was open, the

personnel hatch had both doors open with their interlocks defeated and

containment purge was in progress.

The workers and the HP technician providing coverage immediately exited

the containment and reported the leak. None of these personnel were

reported as being contaminated and their whole-body exposures were low.

Plant pressure and temperature remained essentially constant although an

initial decrease in pressurizer level was noted. Containment evacuation

was begun. After a siting of the area by Operations, Maintenance and HP,

the containment hatch was ordered closed and plant depressurization and

pressurizer cooldown were initiated at 1925.

Operations initiated an effort under a SSAM to stop the leak by sending

five workers and two HP technicians into containment at 2030, August 31.

The workers signed in under REP #81012, Refueling Activities in U-2 '

Ctmt-Zone III, as a matter of convenience. The controls of the REP were

not specifically appropriate to the actions to be taken. The work was to

be performed under an exception in HP procedure 50123-VII-9.9, Radiation

Exposure Permit Program, which allows deviations from the normal control

program in " Operations Shift Supervisor-declared emergency conditions."

The REP contained the radiological controls which Operations and HP felt

would most closely fit the situation, i.e. , a full set of

anti-contamination clothing, white plastic suit, plastic cap (used but

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not specified on the REP), a particulate respirator, a whole-body.

thermoluminescent dosimeter (TLD), two ring TLDs (one for each hand), and

continuous HP coverage. Gamma radiation levels in the vicinity of the

valve had previously been measured at approximately 10 mr/hr and were not

expected to have_ changed. No specific pre-job surveys were taken. Beta

levels around the valve from the mist of primary coolant had been

measured at 800 mrad /hr. Airborne radioactivity levels around the valve

were not known as samples taken at the initiation of the event had not

then been analyzed. However, the HP supervisor stated the he expected

that they would be high and that significant radiciodine levels would be

present. Particulate respirators were worn because respirators which

could remove radiciodine were not readily available and no air supply

manifold had yet been set up in containment.

Efforts to force the valve packing back into the valve and stop the flow  ;

of water were not successful and the HP foreman stopped the job at 2130

because of this and because the workers were getting soaked. All five of

the workers were found to have varying levels of skin contamination, up

to 30,000 cpm as measured with a handheld frisker. All five were  !

subsequently given whole-body counts and four showed radioiodine intakes

less than the action point of the counter, i.e., less than 5% of a

maximum permissible body burden, and one showed more than nine times the

action point although he was still externally contaminated with iodine.

As of September 18, a representative of the HP engineering group informed I

the inspector that the most recent estimate of the worker's intake from

the event was approximately 51 MPC-hrs for all isotopes. External

whole-body and extremity gamma doses and skin doses from contamination

were low and less than regulatory and plant administrative limits. The

contaminations and radioiodine uptakes appeared to be due to the choice

of anti-contamination clothing and respiratory protective equipment used  :

during the event. I

The Unit 2 containment purge monitor, 2RT-7829, was in service and

provided a maximum reading of 5.34E-4 pCi/cc dt. ring the release with its

alert setpoint at 3.80E-3 pCi/cc and the unusual event criteria, as

specified in 5023-VIII-1, being 6.6E-2 pCi/cc for 1 hr; all values are

for noble gas. Monitor 2RT-7828 provides control room indication only 1

for noble gas but also collects a particulate and an iodine sample which i

were later evaluated. These indicated an average effluent radiciodine

concentration of 6.5E-8 pCi/cc and an average particulate concentration

of 2.7E-9 pCi/cc predominately cobalt and cesium. The effluent

radioiodine and noble gas concentrations corresponded well with grab

samples taken in containment during the event. No Technical

Specification (TS) dose or dose rate limits were exceeded during the

release.

A particulate and iodine sample was also taken at the open equipment

hatch between 1927 and 2015 and indicated no iodine and 1.3E-11 pCi/cc

particulate. The plant vent stack monitor, 2RT-7865, also showed an

increase from 2E-6 to 4E-5 pCi/cc during the event. These indicate some

activity was escaping through the personnel hatch but that only

insignificant amounts escaped through the equipment hatch. Environmental

particulate and iodine samples, which are drawn continuously at locations

as indicated in the TS, were evaluated by the licensee's contractor for

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the period of the event. These. indicated no detectable iodine

concentrations off-site and a maximum of 0.038 pCi/ cubic meter gross

beta. The lower limit of detectability for radioiodine for these samples

is 0.04 pCi/ cubic meter and each sample covers a period of one week at a  ;

sampling rate of 1 cfm.

By 0700, September 1, primary pressure ~had been reduced to 5 psi and a

second effort to stop the valve leakage had been initiated. For this

entry, prejob surveys were taken, an ALARA job review was performed and .h'

a specific REP was written in accordance with plant HP procedures. By

1045, September 1, leakage had been reduced to less than 5 gph.

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No violations or deviations were identified.

4. Inspector Observations

NRC regulation 10 CFR 50.54 (x) prescribes specific conditions under

which the licensee may be allowed to take action outside Technical

Specification limits in order to protect the public health and safety.

Licensee procedure 50123-0-1, Shift Superintendent's Authority,

Responsibilities & Duties, in paragraph 6.1.9 specifies:

"The Shift Superintendent shall approve reasonable action that

departs from a LC0 (Limiting Condition of Operation) or a Technical

Specification in an emergency when, in his opinion, this action is ',

immediately needed to protect the public health and safety and no

action consistent with License Conditions and Technical

Specifications that can provide adequate or equivalent protection is

immediately apparent."  !

The licensee representative stated that the provisions of 10 CFR 50.54 i

(x) and paragraph 6.1.9 of S0123-0-1, were not implemented in this case.

Technical Specification 6.11, Radiation Protection Program, states:

" Procedures for personnel radiation protection shall be prepared

consistent with the requirements of 10 CFR'Part 20 and shall be

approved, maintained and adhered to for all operations involving

personnel radiation exposure." l

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Licensee procedure 50123-VII-9.9, Radiation Exposure Permit Program,

allows deviations from the normal control program for " Operations Shift

Supervisor - declared emergency conditions. . . ."

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In this case, the Operations Shift Supervisor declared a " plant

emergency" which allowed deviation from the normal radiological control

program. As described earlier, the licensee did not consider the

provisions of S0123-0-1 to be in effect. Also, as noted earlier, the

event was not classifiable into any of the emergency classifications

identified in 5023-VIII-1, Recognition and Classification of Emergencies.

It is also worth noting that neither S0123-VII-9.9 or 50123-0-1 provide

any definitions for the terms " emergency," " declared emergency

conditions" or " plant emergency." From the inspector's review of the

event and discussions with licensee personnel, no clear nexus could be

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discerned between the context of an " emergency" as used in 50123-VII-9.9,

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S0123-0-1 and 5023-VIII-1. Of particular concern is the provision of

S0123-VII-9.9 that allows deviation from the established radiological

control program. Such action, without concurrent implementation of

paragraph 6.1.9 of 50123-0-1 could result in a violation of T.S. 6.11 or

10 CFR 20, Standards for Protection Against Radiation. Also of. concern

are the provisions of S0123-VIII-9.9 and S0123-0-1 that recognize an

emergency and authorized departure from established requirements without-

necessarily being in an emergency classification as defined by

5023-VIII-1. Without further review, it is not clear whether this aspect'

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represents an inconsistency between the procedures. The issues.as

discussed above will be reviewed in a subsequent inspection.

(50-361/87-24-01).

5. Exit Interview  !

The inspector met with the licensee representatives, denoted in paragraph

1, at the conclusion of the inspection on September 18, 1987. The scope

and findings of the inspection were summarized.

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