ML20207F679

Forwards Technical Review of Util Tech Spec Change Request 148 Re Issues Including SEP Topic XV-16, Radiological Consequences of Failure of Small Lines Carrying.... NRC Response Will Provide Basis for State Concurrence

ML20207F679
Person / Time
Site:	Oyster Creek
Issue date:	12/31/1986
From:	Tosch K NEW JERSEY, STATE OF
To:	Donohew J Office of Nuclear Reactor Regulation
References
RTR-NUREG-0123, RTR-NUREG-123, TASK-15-16, TASK-RR GL-85-19, TAC-M62860, NUDOCS 8701060183
Download: ML20207F679 (6)

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9 tate of Netu 3erseg DEPARTMENT OF ENVIRONMENTAL PROTECTION DIVISION OF ENVIRONMENTAL QUALITY BUREAU OF NUCLEAR ENGINEERING CN 411 TRENTON, NEW JERSEY 08625 (609) 530-4022 December 31, 1986 Mr. Jack Donohew, Project Manager Division of BWR Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

TECHNICAL SPECIFICATION CHANGE REQUEST NO. 148 Re: Oyster Creek Nuclear Generating Station The Technical Specification Change Request (TSCR) No. 69, Revision 1,

Subject:

Radiological Effluent Technical Specifications (RETS) was issued on October 22, 1986. This TS authorizes changes to Appendix A required by 10CFR50 Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable."

While the majority of the document is a significant step towards more restrictive limits for effluent releases, one component, Basis 3.6.A, which refers to SEP Topic XV-16:

" Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Activity Outside Containment,"

presents several questions.

We understood this topic would be addressed separately and included in TAC #M62860: " Postulated High Energy Line Break within Emergency Condenser Penetrations," (Scott to Zwolinski November 7, 1986). Subsequent to that correspondence, the licensee opted to create a separate TSCR, No. 148, for this issue.

TSCR No. 148 was published in the Federal Register on November 19, 1986. The BNE responded to TSCR No. 148 through the normal "Sholly" process and presented several questions in three separate categories, (see "BNE Technical Review", attached).

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We look forward to your written response, as this will provide a basis for State concurrence on TSCR No. 148. If you have any questions, please contact me at (609) 530-4022.

Thank you.

Very truly yours, q) -

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/ /ose x Kent Tesch Bureau of Nuclear Engineering c: David Scott, BNE John Zwolinski, NRR

i BNE TECHNICAL REVIEW TSCR NO. 148 The Technical Specification Change Request (TSCR) No. 148 has presented several significant questions relating to the consequence of a small pipe break outside of containment.

The licensee-in TSCR No. 148 has not complied with several.

critical aspects of NUREG-0123. Oystor Creek's ability to mitigate the small pipe break accident in sufficient time without exceeding the 30 REM maximum off-site dose limit is in question. Because of the-unique design of Oyster Creek, the T.S. limit of 0.2 uCi/ gram D.E. I-131 has not been demonstrated to reduce the offsite dose from a small pipe break to the'30 REM limit of the Standard Review Plan. When coupled with less restrictive Limiting Conditions of Operations and reporting requirements, the effect on the margin of safety for Oyster Creek is unknown. The following outlines specific categories and questions as they relate to this TSCR.

Cateaorv 1 Basis 3.6.A: Oyster Creek is inconsistent or incomplete in their Iodine reactor coolant activity level determination and it's associated offsite consequence.

I. In T.S. 3.6.A the licensee is adopting a 0.2 uCi/ gram Dose Equivalent (D.E.) for I-131 reactor coolant specific activity limit to replace the existing 8 uCi/ gram total Iodine T.S. limit. The licensee states in TSCR No.148 that the proposed basis "will not involve a significant reduction in the margin of safety because more restrictive limits for the primary coolant radio-iodine activity will increase the margin of safety."

Question 1: What is the relationship between uci/ gram total Iodine and uCi/ gram D.E. I-131?

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

II. In the current Basis 3.6.A the primary coolant radioactivity concentration limit was calculated based on a steam-line break accident which isolates in 10.5 seconds. In the revised Basis the primary coolant radioactivity concentration limit was calculated based on the failure of small lines carrying primary coolant outside containment.

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

Question 4: If the licensee recognizes that the small line break has a greater radiological probability and/or consequence than the r

large steam line break, then why doesnft 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 coolant level to meet the off-site 30 REM maximum limit?

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-800 states "the NRC staff will pursue alternatives with the applicant to reduce the doses to within the guideline values." What alternatives are being discussed?

Cateaorv 2 LCO: Two aspects of the licensee's LCO for reactor coolant activity are less restrictive than the LCO in the Standard Technical Specifications of General Electric Boiling Water Reactors (NUREG-0123).

I. The licensee LCO states:

3.6.A.1: "Whenever an isotopic analysis shows reactor coolant activity exceeds 0.2 uCi/ gram Dose Equivalent I-131, additional analyses shall be done at least 6 times within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />..."

The Standard Technical Specification's (NUREG-0123) states:

3.4.5 ACTION b: "

...with the specific activity of the primary coolant greater than 0.2 uCi/ gram Dose Equivalent I-131... perform the sampling and analyses requirements of Item 4a of Table 4.4.5-1..." which is at least once per four hours (12 time within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />).

Comparing these LCOs shows that NUREG-0123 requests the licensee to perform twice as many analyses in the first 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

l Question 6: Why is the licensee not complying with the 4 ,

hour analysis interval? '

II. The licensee's LCO states:

3.6.A.3: "If an initial sample of the reactor coolant activity is greater than 4 uCi/ gram D.E.

I-131 a second sample shall be taken and analyzed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If the second sample indicates the reactor coolant activity is greater than 4 uCi/gm D.E. I-131, be in at least HOT SHUTDOWN with the mainsteam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

NUREG-0123 as amended by NRC Generic Letter No. 85-19 states:

3.4.5 ACTION a.1: " ...

with the specific activity of the primary coolant ... greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the mainsteam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

Comparing these LCOs shows that the licensee is allowing operation for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the reactor coolant activity is in excess of 4 uCi/gm D.E. I-131 where NUREG-0123 does not.

Question 7: Why does the licensee have an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period to make a HOT SHUTDOWN determination?

Question 8: Why doesn't the licensee have a peak limiting value in their LCO for HOT SHUTDOWN?

Cateaorv 3 Reporting requirement: Reporting requirements for primary coolant activity in excess of LCO's have been eliminated from the Special Report and i Licensee Event Report system and included in the Annual Reporting system.

The NRC's Generic Letter No. 85-19, " Reporting Requirements on Primary Coolant Iodine Spikes," states that the NRC has a continuing program to delete unnecessary reporting requirements. They have determined that the reporting requirements for iodine spiking can be reduced from a short-term report (within a 31 day Special Report or Licensee Event Report) to an item which can be included in the Annual l Report.

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The NRC states in 85-19, "The quality of nuclear fuel has .

been greatly improved over the past decade with the result l that normal coolant iodine activity (i.e. in the absence of '

iodine spiking) is well below the limit...Therefore, this Technical Specification limit is no longer considered necessary on the basis that proper fuel management by

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licensees and' existing reporting requirements should preclude ever approaching the limit."

The'NRC also states in 85-19, " Licensees are expected to continue to monitor iodine activity in the primary coolant and take responsible action to maintain it at a reasonably low level in accordance with the surveillance requirements in the Standard Technical Specification's (NUREG-0123)."

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

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

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