ML20214D446

Forwards Supplemental Responses to SER Questions 12a,b,j,k & 13e Re NUREG-0737,Item I.C.1 Concerning Emergency Operating Procedures Upgrade Program & B&W Owners Group Response.Epri NSAC-101 Re Steam Generator Tube Ruptures Also Encl

ML20214D446
Person / Time
Site:	Three Mile Island
Issue date:	11/18/1986
From:	Hukill H GENERAL PUBLIC UTILITIES CORP.
To:	Stolz J Office of Nuclear Reactor Regulation
Shared Package
ML20214D448	List: ML20214D446 ML20214D465 ML20214D503
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.1, TASK-TM 5211-86-2194, NUDOCS 8611240101
Download: ML20214D446 (5)

Text

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GPU Nuclear Corporation NggIgf Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057-0191 717 944 7621 TELEX 84 2386 November 18, ggr's Direct Dial Number:

5211-86-2194 Office of Nuclear Reactor Regulation Attn: J. F. Stolz, Director PWR Projects Directorate hq. 6 U.S. Nuclear Regulatory Comission Washington, DC 20555

Dear Mr. Stolz:

Three Mile Island Nuclear Station, Unit 1 (THI-1)

Operating License No. DPR-50 Docket No. 50-289 Emergency Operating Procedures (EOP) Upgrade Program Response to SER Questions (NUREG 0737, I.C.1)

Your Safety Evaluation Report (SER) for NUREG 0737, I.C.1 dated March 28, 1984 concluded that the TMI-l Procedures Generation Package (PGP) was found to be acceptable but required confirmatory documentation. GPUN's letter of June 29, 1984 provided a response to the NRC's questions. Part of our response needed the results of analyses being performed by the Electric Power Research Institute (EPRI).

EPRI's report, NSAC-101, which includes the results of their analyses was received by GPUN on October 19, 1986 and has been used in the preparation of our response to the remaining SER questions. Attachment 1 provides the

, additional confirmation in response to SER questions 12a and k. In addition, i we are providing updated responses to questions 12b, j and 13e.

1 TDR-517 was transmitted to the NRC as part of GPUN's PGP. This document is l being revised to incorporate the results provided in the reference documents which are enclosed as described in Attachment 1. The revised TDR-517 will be made available at the TMI-l site for NRC review on completion of GPUN review and approval.

Sincerely, SW22hn nab $r / l-I H. D. ki 1 Vice President & Director, TMI-l HDH/MRK/spb:0710A cc: R. Conte r 0 g\

J. Thoma i Attachment GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation l

l

ATTACMENT 1 GPUN RESPONSE TO NUREG 0737, I.C.1 (EMERGENCY OPERATING PROCEDURES)

SER QUESTIONS (SUPPLEMENT RESPONSES TO QUESTIONS 12a, b, j, k and 13e) 12a. QUESTION:

With regard to the SGTR guidelines, provide the following:

The information provided in TDR 406 to support the intention to steam the damaged 0TSG for the duration of the SGTR event is largely qualitative and has not adequately demonstrated the need for continuous steaming of the damaged 0TSG. Therefore, for both the earlier and existing methods (i.e. isolation of the damaged 0TSG at the predetermined RCS temperature versus continued steaming), provide the following systems analyses. Provide the assumed sequence of events including timing of automatic and operator action. Include the projected radiological consequences (whole body and thyroid): (1) Design basis SGTR accident with offsite power (and condenser) available; and (2) SGTR accident with offsite power unavailable. The analyses should assume that the accident begins with the primary coolant iodine concentration at the technical specification limit of 1.0 uci/gm and that an iodine spike occurs as a result of the primary system depressurization. Provide the following information from these analyses: RCS (pressurizer) and OTSG (both) pressure, temperature, mass and level, break flow, atmospheric dump and main steam safety valve flows, and safety injection flow. These analyses should include the period of RCS cooldown to decay heat removal (DHR) system initiation.

Compare the offsite doses in the above calculations to the FSAR assumption and results. Use these analyses to show the need for using the techniques of TDR 406 for either limiting offsite dose or avoiding more serious events (i.e. filling the faulted 0TSG, lifting the damaged 0TSG safety valves, etc.).

RESPONSE

Since the promulgation of the guidelines in TDR 406, the B&W Owners Group has incorporated them into the Emergency Operating Procedures Technical Basis Document which forms the generic basis for symptom oriented procedures. Questions regarding the offsite doses associated with the use of the TBD guidelines were subsequently directed to the Operator Support Committee of the B&W Owners Group. An assessment of doses for the continuous steaming versus isolated 0TSG cases was provided to the Staff on June 17, 1986 by a letter from Mr. Mark A. Linn, Chairman Operator Support Committee. This report (Reference 1, enclosed), entitled " Response to NRC Questions on the Steam Generator Tube Rupture Chapter of the B&W Owners Group Emergency Operating Procedures Technical Bases Document #47-1164369-00)"

demonstrates acceptably small doses using eithcr the continuous steaming or the isolated steam generator approach.

The thermal hydraulic response of the B&W NSSS design to tube rupture during both continuous steaming and isolated steam generators is reported in an Electric Power Research Institute document (Reference 2, enclosed), entitled " Response of a B&W Plant to Steam Generator Tube Ruptures, NSAC-101, September 1986". The report analyzes both single and multiple steam generator tube ruptures. This report provides all of the thernal hydraulic information requested by the staff. Thermal hydraulic response, while acceptable for both the steaming and isolated conditions, is shown to be simpler for the continuous steaming cases.

REFERENCES:

1. White, J.R., Santee, G.E., Jensen, R.T., Response of a 53W Plant to Steam Generator Tube Ruptures. NSAC-101, September 1986. Nuclear Safety Analysis Center, Palo Alto, CA.

2. B&W Owners Group Operator Support Committee, " Response to NRC Questions on the Steam Generator Tube Rupture Chapter of the B&W Owners Group Emergency Operating Procedures Technical Bases Document." Doc. No. 47-1164369-00. Babcock and Wilcox Co.,

Lynchburg, VA. (this report was transmitted to GPUN by letter dated June 17,1986. A revised page 12 of the report was transmitted to GPUN by letter dated June 30, 1986 from B&W).

12b. QUESTION:

Recent test data at TMI-l has indicated that if the energency feedwater system (EFWS) is utilized instead of the main feedwater system (MFWS) to maintain secondary water inventory, a rapid cooldown rate would result in exceeding the allowable tube /shell delta T of 70*F.* Since the EFWS would be initiated for a variety of conditions, provide the maximum cooldown rate for these conditions that would maintain the tube /shell delta T 70*F. If different from the cooldown rate assumed in Question 12a, calculate the projected radiological consequences assuming the TDR 406 guidelines or show that this cooldown rate would not be more limiting than that used in the analyses to address Question 12a.

• A 90*F/ hour cooldown rate resulted in a tube /shell delta T of ll2*F in OTSG "A" and 99'F in OTSG "B" (Reference - TDR 488).

RESPONSE

The tests performed at TMI-1 were performeo without any steaming of the OTSGs due to the very low core decay heat. The combination of low level and zero steam flow allowed the shell/ tube delta T to reach 100F*.

During expected plant operating conditions, a cooldown at low OTSG level

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(i.e. RC pumps running) using EFW would only occur due to a loss of the main feedwater pumps (FW-Pl A/B). The electric driven condensate booster pumps have a shut-off head of about 675 psig, and if available would be adequate to control shell/ tube delta T below 70F*. With offsite power available, the condenser would normally be available. Offsite doses will be considerably reduced due to iodine removal in the condenser.

Therefore, even if cooldown were delayed due to high shell/ tube delta T, offsite doses would be much smaller (by a factor of 100 - 10,000) than for a case in which steaming is directly to the atmosphere. Results for steaming to the atmosphere are reported in Reference 2 in response to Question 12a.

With a loss of offsite power, EFW will be controlled at the natural circulation setpoint of 50%; the combination of OTSG level and steaming of the OTSG will limit the differential temperature below 70F*.

12j. QUESTION:

Clarify the TDR 406 assessment (Section 2.2.2) of the control of the damaged and intact OTSG 1evels, as function of HPI pumps, EFW pumps and number of broken tubes. The discussion regarding raising the OTSG 1evel of 95% " tempered with the need to control the RCS cooldown rate" is particularly unclear. The discussion regarding simulator experience (Item 4.2.1) is also not clear (Was it not possible to raise OTSG level

. to 95% with full HPI on, while steaming the OTSG and maintaining a 100*F/hr cooldown?).

RESPONSE

Based on the simulator experience in June 1983, it was very difficult with a saturated RCS for the operators to raise OTSG level to 95% while HPI was on full flow. At the simulator, all three pumps were running, providing about 1800 gpm. GPUN has conducted in-house small break LOCA analyses using the RELAPS computer code and the TMI-l HPI capacity as limited by cavitating venturies. Those analyses deraonstrate that the RCS cooldown rate can be controlled while still maintaining a continuous, minimum flow to the OTSGs.

12k. QUESTION:

Section 2.3 indicates that a large portion of the analytical effort required to resolve the issues identified in TDR 406 is still ongoing.

This includes simulation of single and multi-tube ruptures under various conditions, including unavailability of RCPs, unavailability of condenser, high radiation releases, steam if ne flooding, loss of SCM, and PORY availability versus unavailability. Describe how these aspects are being evaluated and provide a schedule for their resolution.

RESPONSE

NSAC-101, " Response of a B&W Plant to Steam Generator Tube Ruptures", is being provided as the GPUN response to the questions raised here. Cases analyzed specifically include:

1. Single tube rupture

2. Multiple tube ruptures in one OTSG

3. Multiple ruptures in both OTSGs

4. RCPs running

5. RCPs tripped

6. Subcooling available

7. Subcooling lost The NSAC analyses were not carried out to the RCS conditions which allowed decay heat system initiation. Rather, the transient was terminated when RCS pressure decreased below 1000 psig and RCS temperature was decreasing. This plant state was reached in less than one hour for all of the transients analyzed. Once this condition is reached, then the steam generator safety valves would not re-open, and the leaking OTSG can remain isolated.

It should be noted that the RELAPS model conservatively modeled the upper vessel head region in predicting the creation of a head bubble by restricting flow into the upper-upper head (see Figure 3-8 of the NSAC report) by modeling a single flow path to the uppermost head region.

This assures a bounding prediction of the head bubble effect in delaying RCS depressurization. The analyses have demonstrated that a head bubble cannot prevent RCS depressurization below the main steam safety valve lift pressure.

13e. QUESTION:

The " Summary" table is not clear. Clarify the wording of the table headings and the numbering of parenthetical reference entries.

, RESPONSE:

This question addresses clarity of the Summary Table in TDR-406 and not TDR-517. TDR 517 provices the Plant Specific Guidelines for TMI-1 and provicks the source document fron which plant procedures are written.

If TDR 406 is revised in t:a future, the table headings will be clarified to be consistent with TDR 517.

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