ML20235X000

Forwards Response to 861107 & 1231 Ltrs on 861023 Tech Spec Change Request 148 Re Reactor Coolant Radioiodine Monitoring.Util Request Requires Rev Prior to Final Action

ML20235X000
Person / Time
Site:	Oyster Creek
Issue date:	07/20/1987
From:	Varga S Office of Nuclear Reactor Regulation
To:	Scott D NEW JERSEY, STATE OF
References
GL-85-19, NUDOCS 8707240004
Download: ML20235X000 (8)

Text

t> f G July 20,1987 Mr. David M. Scott, Chief DISTRIBUTION '

Bureau of Nuclear Engineering Docket No. 50-210 d Division of Environmental Quality NRC & Local PDRs 4 M .;

Department of Environmental Protection PDI-4 Reading RHernan State of New Jersey S. Varga CN 411 B. Boger Trenton, New Jersey 08625 A. Dromerick S. Norris

SUBJECT:

Oyster Creek Nuclear Generating Station Technical Specification Change Request (TSCR) No. 148

References:

1. Fiedler, P.B., GPUN, letter to J. A. Zwolinski, NRC, "0yster Creek Nuclear Generating Station, Docket No. %-219, Technical Specification Change Request (TSCR) No. 148," October 23, 1986.

2. NUREG-0822, " Integrated Plant Safety Assessment, Systematic Evaluation Program, Oyster Creek Nuclear Generating Station," January 1983,

3. NUREG-0123, " Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)," April 1978.

4. Generic Letter 85-19 " Reporting Requirements on Primary Coolant Iodine Spikes," September 27, 1985.

This letter provides the Nuclear Regulatory Commission (NRC) staff's response to your November 7,1986 and December 31, 1986 letters on the subject TSCR, We appreciate your interest in this matter and hope you will find the infonna-tion provided herein responsive to your questions.

As you know, General Public Utilities Nuclear (GPUN) submitted TSCR No.148 on October 23,1986 (Ref.1) as requested in tha NRC's Oyster Creek Integrated Plant Safety Assessment Report (Ref. 2). The intent of our request was to up-grade the Oyster Creet Technical Specifications in the area of reactor coolant radiciodine monitoring. The model to be used for this upgrade is the Standard Technical Specifications for General Electric Boiling Water Reacters (Ref. 3) as modified by NRC r,ereric letter No. 85-19 (Ref. 4). The NRC staff has ccm-pleted its preliminary review of GPUN's submittal and will require a revision to that submittal prior to final action on TSCR No.148. The deficiencies in the October 23 submittal have been discussed with representatives of GPUN in a public meeting on June 30, 1987 and are discussed in the enclosure to this letter.

Because the explanation of the staff's positions on concerns expressed in your-December 31, 1986 letter are rather detailed, we have made them an enclosure to j this letter. The enclosure includes a discussion of the background of the present staff position on coolant iodine activity and responses to your l

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2 specific questions. Should you have any questions on the enclosed response, please contact Mr. Alexander W. Dromerick, the NRC Project Manager assigned to Oyster Creek. Mr. Dromerick's telephone number is (301) 492 7563.

1 Sincerely,

/6 Steven /A. Varga, Director Divisi6n of Reactor Projects I/II Office of Nuclear Reactor Regulation Enclosu re:

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specific questions. This response has been prepared by the same person that prepared Generic Letter No. 85-19 dealing with primary coolant iodine spikes.

Should you have any questions on the enclosed response, please contact Mr. Alexander W. Dromerick, the NRC Troject Manager assigned to Oyster Creek.

Mr. Dromerick's telephone number is (301) 492 7563.

Sincerely, Stephen A. Varga, Director Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation

Enclosure:

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Enclosura Oyster Creek Nuclear Generating Station Technical Specification Change Request (TSCR) No. 148 NRC Response to Questions from the State of New Jersey l

Background '

A systematic evaluation of the Oyster Creek Station was conducted by the NRC in the late 1970's and early 1980's as part of its Systematic Evaluation Pro-gram (SEP). The purpose of the SEP was to review the designs of eleven older operating nuclear reactor plants in order to reconfirm and document their safety. The SEP review of Oyster Creek compared current technical regulatory positions on safety issues with those that existed in 1969, when Oyster Creek received its operating license. One of the SEP objectives was to make inte-grated and balanced decisions with respect to backfitting, including upgrading l of Technical Specifications (TS). The Oyster Creek SEP review included 83 topics and resulted in the NRC requesting 49 actions by General Public Utili-ties Nuclear (GPUN), including five TS revisions. The results of this review are documented in NUREG-0822, which was issued in January 1983 by the NRC. In Section 4.36 of NUREG-0822, the NRC established the position that adaptation of the more stringent BWR Standard TS limits for reactor coolant activity is neces-sary and sufficient to ensure that the radiological consequences of a small coolant leak are acceptably low. This position is the reason that GPUN sub-mitted TSCR No. 148.

Reactor coolant activity is monitored primarily by measuring the radioactivity of several isotopes of iodine, which are products of the fission prccess, and i mathematically converting those measurements to a " dose equivalent iodine-131" l

_ (D.E. 1-131) value. This value is an indicator of how much leakage exists I

between the reactor fuel and the reactor coolant system as the result of fuel i cladding failures. The historical fuel failure rate for nuclear reactors is i typically very low (approx. 0.02%) resulting in D.E. I-131 equilibrium activity levels in the range o' O.0001 to 0.) microcuries per gram (pCi/g). The equili-brium value of radio 1/ine activity is reached only after several days of opers-tion at a given reactar power level. Each power level has a unique value of D. E. 1-131. At Oyster Creek, the equilibrium iodine activity is typically about 1

0.0003 pCi/g. Approximately 25% of the operating reactors have experienced a transient increase in coolant iodine levels called " iodine spiking." These spikes arc normally the result of a significant change in plant operating conditions, such as a reactor trip or large changes in power level, and sub-side af ter several hours allowing D.E.1-131 activity to return to pre-spike concentrations. The temporary iodine concentrations can be as high as 1000 times the normal equilibrium concentrations. In order to discern between these temporary iodine ? pikes and gross fuel element failures, the Standard Technical Specifications require more frequent monitoring of coolant activity (every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) once D.E. 1-131 values reach a predetermined value. This value is 0.2 pCi/g for BWR's and 1.0 pCi/g for PWR's. If the D.E. 1-131 levels remain above 0.2 pCi/g in a BWR for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or reach 4.0 pCi/g at any time, major i

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fuel degradation is indicated and the reactor must be shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, These limits have been established by the NRC in the Standard Technical Spe-cifications and are based upon assuring-compliance with the offsite exposure guidelines of 10 CFR 100. In establishing the D.E. 1-131 limits, the NRC as- I sumed that a small loss of coolant accident (or failure of a small reactor cool-ant line) occurs simultaneous with a very large iodine spike oc degraded fuel l condition and that radiciodines wculd account for a small fraction (10%) of the off-site dose. The Standard TS assumptions are therefore very conservative since the existence of high D.E. I-131 and a small break are virtually coinci-dental (i.e. one would not normally cause the other) and is very unlikely.

Every BWR nuclear power plant licensed to operate since 1977 has.the Standard TS requirements for reactor coolant activity. The 22 BWR's licensed before then have a variety of limits pertaining to iodine activity ranging from 0.2 to 25.0 pCi/g. The present Oyster Creek TE limit is 8.0 pCi/g. measured as total iodine, not D.E. 1-131. For a number of years, the NRC required utilities experiencing high iodine activities, usually due to iodine spiking as discussed above, to report pertinent information via a special report on a case basis.

. This requirement was changed by Generic Letter 85-19 to allow annual reporting of iodine activity ievels in excess of the TS limit. In a number of plants, iodine spikes are relatively predictable following certain plant transients and NRC notification each time is net necessary.from a regulatory standpoint unless release of significant quantities of fission products may be involved.

In cases where an increase in D.E. 1-131 activity level is unexpected, the licensee is required by 10 CFR 50.73 to submit a Licensee Event Peport (LER) to the NRC. In extreme cases involving potential for releasing significant quantities of fission products or if a reactor shutdown is required by the TS, the licensee would also be required to provide immediate notification to the NRC under the provisions of 10 CFR 10.72. It has been the NRC's experiente over the past several years that licensees are very aware of the quality and performance of their fuel long before fuel degradation becomes a regulatory concern. A number of utilities have noted fuel degradation and shut down the j affected reactor to resolve the problem even though D.E. 1-131 had not even 1 approached the TS value which would have required more frequent sampling (but not necessarily require plant shutdown).

NRC Response to State of New Jersey Questions Question 1: What is the relationship between pCi/ gram total Iodine and pCi/ gram D.E. I-131?

Response

Since D.E. I-131 is a number which is calculated using activity levels of several different isotopes of iodine, there is no fixed conversion factor between total iodine and D.E. 1-131. Each isotope has a different decay half-life and radioa::tive emission energy level. As stated in TSCR No. 148, the dose conversion factors listed in NRC's Regulatory Guide 1.109 are used for this calculation (once an activity level for each iodine isotope is measured).

The relative concentration of each iodine isotope is dependent on the power history of the reactor. If the reactor has been operating steadily at a high power level for several weeks, an equilibrium conditicn would have been estab-lished and the approximate conversion from total iodine to D.E. I-131 would be about 5 to 1. In other words, 1.0 pCi/ gram total iodine would be equivalent to about 0.2 pCi/ gram D.E. I-131. The total iodine value would always be larger.

than the D.E. 1-131 value.

Question 2: Is the D.E. I-131 of 8.0 uCi/ gram total Iodine greater than 0.2 pCi/ gram D.E. I-131?

Response

The conversion of 8.0 pCi/ gram total iodine to D.E. 1-131 would always be a number larger than 0.2 pCi/ gram. As discussed in the respense to Question 1, the exact conversion depends en a number of variables and is therefore not constant. Iodine-131 is normally the dominant isotope from the stanc' point of thyroid dose and therefore has the largest conversion factor. The other isotopes have a small contribution to thyroid dose and therefore have smaller conversion factors.

Question 3: Why was the accident sequence changed from a large pipe break to a small pipe break accident.

Response: 1 The NRC's basis for the iodine limits in the Standard Technical Specifications for General Electric BWR's is a small loss of coolant accident concurrent with a very large iodine spike, as discussed above under " Background." The NRC evaluated radiological consequences _of small line breaks and main steam line failures at Oyster Creek as part of the SEP. These evaluations are documented in sections 4.36 and 4.37, respectively, of NUREG-0822. Our conclusion was that, because the small-line failure is more limiting than the main steam line failure from a radiological consequences standpoint, adaptation of the Standard TS limit will result in acceptably low consequences from either accident.

Question 4: If the licensee recognizes that the small line break has a greater radiological probability and/or consequence than the large steam line break, then why doesn't the licensee.take action by either limiting the flow of discharge through the line, e.g. in-line flow restrictors, or lower the allowable radio-iodine reactor ecolant level to meet the off-site 30 REM maximum limit?

Response

As discussed above under " Background," the " normal" D.E.1-131 coolant activity level at Oyster Creek is about 0.0003 pCi/ gram. The NRC calculations of off-site dose consequences assume a D.E.1-131 level of 4 pCi/ gram during a sinall

line. break in order to approach a 30 REM thyroid dose (or 10% of the 10 CFR 100 guideline of 300 REM thyroid dose) to a member of the public. In other words, Oyster Creek coolant activity would have to be over 10,000 times more radio-active than normal at the same time (coincidentally) that a small line rupture occurs.in order to approach ;t0% of the statutory guideline of 10 CFR 100. The NRC staff concluded in NUREG-0822 that backfitting of flow-restricting devices in these lines at Oyster Creek is ng1 justified by the small potential reduction in risk. GPUN can certainly opt to install such devices at Oyster Creek by its own initiative as long as such modifications did not introduce the potential for new unanalyzed risks, such an, increasing the probability of failure in one of the small lines in question.

Question 5: If the dose calculations are not within exposure guidelines of the Standard Review Plan NUREG-0800, 15.6.2: " Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment" NUREG-0800 states "the NRC staff will pursue alternatives with the apolicant to reduce the doses to within the guideline valuer,." What alternatives are being discussed?

. Response:

The NRC Stan4rd Review Plan was adapted in its present form in July 1981 to provide guidance to the NRC ctaff in the review of license applications for plants which had not yet been licensed to operate. The Plan did not exist

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at the time Oyster Creek was licensed in 1969. The NRC initiated the SEP specifically to re-evaluate some of the older operating reactors in terms of

- licensing criteria of the early 1980's. NUREG-0822 is the result of the NRC's SEP review of Oyster Creek. Therefore, no further discussion of al-ternatives is appropriate unless new information is presented.

Question 6: Why is the licensee not complying with the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> analysis interval?

Response

The licensee was informed during a public meeting on June 30, 1987, that a resubmittal of TSCR No.148 would be necessary which complies with the intent ^

of the Standard TS sampling frequency before a license amendment would be issued.

Question 7: Why does the licensee have an B hour period to make a HOT SHUTDOWN determination?

Response

i See the response above to Question 6. The NRC will require the reactor to be

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l in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> V initially determining that coolant activity exceeds 4 pCi/ gram D.E. 1-131. GPUN must resubmit TSCR No. 148 to be consis- l tent with the Standard TS.

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I' Question 8: Why doesn't the licensee have a peak limiting value in their l LC0 for HOT $HUTDOWN?

Response: 1 The peak value is 4 pCi/ gram, which requires shutdown of the reactor.

Question 9: How will the NRC or any other outsida agency know whether proper fuel management is being maintained-at Oyster Creek on a monthly basis?

Response

As discussed above under " Background," the requirements of 10 CFR 50.72 and 10 CFR 50.73 provide the NRC with adequate notification in those cases involv-  !

ing reactor safety regulatory concerns. In fact, Oyster Creek experienced a i minor fuel degradation in 1986 and prepared a Licensee Event Report (LER No.86-016) to inform the NRC. The maximum D.E. 1-131 level measured during I . the period that followed~was approximately 0.005 pCi/ gram.

Question 10: Why is Oyster Creek allowed to maintain less restrictive LC0's for reactor coolant activity when 85-19 states that the li-censee is expected to continue utilizing NUREG-0123 surveillance requirements?

I Response:

• The NRC will amend the Oyster Creek license in response to TSCR No. 148 only after it is assured that the intent of NUREG-0822 is satisfied and the appro-priate provisions of Generic Letter 85-19 have been adapted. l E--------_---_---- - - - --- -