ML18038A116

Forwards Response to W Butler 861115 Ltr Re Conformance to Reg Guide 1.97.Encl Provides Info Necessary to Close Confirmatory Item 10

ML18038A116
Person / Time
Site:	Nine Mile Point
Issue date:	01/20/1986
From:	Mangan C NIAGARA MOHAWK POWER CORP.
To:	Adensam E Office of Nuclear Reactor Regulation
References
CON-IIT07-464A-91, CON-IIT07-464B-91, CON-IIT7-464-91A, CON-IIT7-464A-91, CON-IIT7-464B-91, RTR-REGGD-01.097, RTR-REGGD-1.097 (NMP2L-0589), (NMP2L-589), NUREG-1455, NUDOCS 8601240044
Download: ML18038A116 (16)

Text

NIAGARA MOHAWK POWER CORPORATIOk/300 ERIE BOULEVARD3VES~SYRACUSE. N.Y. 13202/TELEPHONE (315) 474-1511 1

January 20, 1986 (NMP2L 0589)

Ms. Elinor G. Adensam, Director BWR Project Directorate No. 3 U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Washington, DC 20555

Dear Hs. Adensam:

Re: Nine Mile Point Unit 2 Docket No. 50-410 Attached is the Nine Mile Point Unit 2 response to the letter from W. Butler (NRC) to B. G. Hooten (NHPC), dated November 15, 1986 concerning conformance to Regulatory Guide 1.97.

This material provides the information necessary to close Confirmatory Item 10.

Very truly yours,

. V. Hanga Senior Vice President TRL:]a Attachment xc: R. A. Gramm, NRC Resident Inspector Project File (2)

ATTACHMENT

1. The licensee should identify plant-specific Type A variables and verify that the instrumentation for them is Category I.

~Res onse Table 421.36-1 of the Nine Hile Point Unit 2 Final Safety Analysis Report will be revised to identify the following Type A variables.

1. Containment hydrogen concentration (2) (same as variables Gila and b)

2. Containment oxygen concentration (2) (same as variables C12a and b)

3. Reactor vessel pressure (2) (same as variables B6a and b)

4. Reactor vessel level (4) (same as variables B4a, b, c and d)

5. Suppression pool water temperature (8) (same as variables D6a and b)

6. Drywell atmosphere temperature (18) (same as variables D7a and b)

7. Drywell atmosphere pressure (2) (same as variables D4a and b)

2. Neutron flux The applicant should provide redundant Class lE power sources for this instrumentation; the applicant should show that the source and intermediate ranges have sufficient overlap.

~Res onse The source of power (see Nine Mile Point Unit 2 Final Safety Analysis Report figure 8.3-10> for ~Sour e Ran e Monitors ( > and Intermediate Range Mon'Itors (IRMs> ri >nates f~ reliable normal dc sources. Normal

~supp y originates from stub buses (2NJS-US6 and USS> to 24/48-V dc distribution panels (2BWS-PNL300A and 2BWS-PNL300B). A normal power source feeds two battery chargers per division that service the 24/48-V dc distribution panel(s) and maintain the charge on two 24-V dc batteries. The batteries are available to service the"24/48-V dc distribution panels on loss of normal power.

The power source for Low Power Range Honitor (LPRH) groups and Average Power Range Monitor (APRH) channels is from the RPS/UPS channelized Divisions 1 through 4. This power is fed to RPS buses by means of a OPS which has normal, alternate and battery backup sources. Power sources are channelized RPS Divisions 1 through 4. The power distribution system for this instrumentation is described in detail in Section 8.3. 1. 1.3 of the Final Safety Analysis Report.

It is Niagara Hohawk's determination that this design provides reliable power sources.

Although the power supplies are classified as nonsafety-related, this h~

does not impair the ability of the neutron monitor1nng nstYomentation to perform its requ1red detection and trip funct1ons. The trip funct1on is configured to trip and 1nit1ate a scram on loss o ower The instrumentation is seismically and environmen a ly qualifi so the trip f n loss of power.

The operating ranges of the source range (SRM) and intermed1ate range (IRH) devices are as follows:

SRM 1 x 103 to 1.5 x 109 nv IRM 1 x 108 to 1.5 x 10l3 nv The overlap of the ranges is 1 x 108 to 1.5 x l09 nv.

Additionally, there is a typographical error in Table 421.36-1. The lower end of the IRH range should be 4.0 x 10-5 percent power. This will be corrected in the next Final Safety Analys1s Report amendment.

3. Reactor coolant system soluble boron concentration The applicant should identify the range of the instrumentation being supplied for this variable.

~Res onse The range is 50 to 2,000 ppm boron in solution. This information will be incorporated in the next F1nal Safety Analysis Report update.

4. 'oolant level in reactor - The applicant should identify the remainder of this instrumentation in accordance with Section 6.2 of NUREG-0737, Supplement No. l, identify any deviations, and )ustify those deviations identified.

~Res onse Conformance to Regulatory Guide 1.97, Revis1on 3, 1s accomplished by the use of two transmitters per d1vision, one fuel zone and one w1de range.

As noted in Section 3.3.3 of your November 15, 1985 letter, the wide range transmitters were omitted from the table and will be added in the next Final Safety Analys1s Report amendment.

The wide range transmitters (2ISC'LT9C and D) are calibrated to monitor the 375.70 in. level, which is inside the fuel zone transmitter range, to the 585.70 in. level which is 62.3 in. below the centerline of the main steam lines at 648 in.

This range is considered to meet the 1ntent of the regulatory guide which is to restore and maintain reactor pressure vessel water level to ensure adequate core cooling.

5. Drywell pressure The appl1cant should provide independent Class lE power sources for these instrument channels.

~Res onse The two instrument channels are powered from separate Class lE sources.

Table 421.36-1 is incorrect and will be corrected in a future Final Safety Analysis Report amendment.

6. Drywell sump level The applicant should provide instrumentation for this variable.

~Res onse See Item 7.

7. Drywell drain sumps level The applicant should provide instrumentation for this variable.

~Res onse <Items 6 and 7>

The drywell sump level instruments provide indication of identified and unidentified leakage during normal operating conditions. The instrumentation, which is nonsafety grade (Category 3), is located on drain tanks in the secondary containment, outside. of the drywell. Under accident conditions, these drain tanks are automatically isolated from the primary containment to prevent the escape of any post-accident reactor fluid from the drywell. In this situation, the drywell sump level 1ndication is no longer meaningful and thus serves no post-accident

-safety function.

Other instrumentation is available to identify leakage into the drywell.

This includes drywell pressure, drywell temperature, and containment radiation. These instruments meet the Category I requirements of Regulatory Guide 1.97.

8. Radiation level in c1rculating primary coolant The applicant should supply the recommended instrumentation and the information required by Sect1on 6.2 of NUREG-0737, Supplement No. l, ident1fy any deviations from the regulatory guide, and ]ustify those deviations.

~Res onse This instrumentation is not provided at Nine M1le Point Unit 2.

Justification The usefulness of information obtained by monitoring the radiation level in the circulating primary coolant, 1n terms of helping the operator in his efforts to prevent and mitigate accidents has not been substantiated. The particular planned operator action to be taken based on mon1toring this variable 1s not specified in the current draft of the Emergency Pr'ocedures. The critical actions taken to prevent and mitigate a gross breach of fuel cladding are to shutdown the reactor and maintain water level. Monitoring primary coolant radioactivity has no influence

. on either of these actions. The purpose of this monitor falls in the cateogry of "information that the barriers to release of radioactive material are being challenged" and

~ t "identic cation of degraded conditions and their magnitude, so the 1

operator can take actions that are available to mit1gate the consequences." Additional operator actions to mitigate the consequences of fuel barriers be1ng challenged, other than those based on Type A and B var1ables, have not been tdentif1ed.

Regulatory Guide 1.97 spec1fies measurement of the radioactivity of the circulating pr1mary coolant as the key variable 1n monitor1ng fuel cladding status during 1solation of the nuclear steam supply system (NSSS). The words "circulating pr1mary coolant" are interpreted to mean coolant, or a representative sample of such coolant, that flows past the core. A bas1c criterion for a valid measurement of the specified variable is that the coolant being monitored is coolant that is in act1ve contact with the fuel, i.e., flowing past the failed fuel. Monitoring the active coolant (or a sample thereof) 1s the dominant considerat1on.

The post-accident sampling system (PASS) provides a representative sample which can be monitored.

The concern of Regulatory Guide 1.97 assumes a s1tuation in which the NSSS is isolated and the reactor is shutdown. This assumption is

]ustified because the monitors 1n the off-gas system and main steam tunnel provide reliable and accurate information on the status of fuel cladding when the plant is not isolated. Further, the PASS, once activated, prov1des an accurate status of coolant radioactivity and hence, cladding status. In the interim between NSSS isolation and operation of the PASS, monitoring of the pr1mary containment rad1ation a'nd hydrogen levels provides information on the status of the fuel cladding.

Present emergency procedures provide that once initial core damage 1s estimated using 1nformation obtained from the analysis of PASS samples, the estimate is confirmed using conta1nment hydrogen analysis, containment high-range radiation monitoring, water level indications, and Sr, Ba, La, and Ru analyses. Therefore, no Type C Category I instrumentation is provided to measure the subject variable.

The Niagara Mohawk position agrees with the BWR Owners Group position on this variable.

9. Analys1s of primary coolant The applicant should identify the range of the instrumentation being supplied for this variable.

~Res onse The Instrument range is 10-6 to 10l Ci/gm and will be incorporated 1n the next Final Safety Analysis Report update.

10. Radiation exposure rate The applicant should show that the ranges encompass the expected radiation levels in the1r locations.

~Res onse Access is not required in any area of secondary containment to service safety-related equipment in a post-acc1dent situation. When access1bility is reestabl1shed in the long term, 1t will be done by a combination of portable radiation survey instruments and post-acc1dent sampling of the secondary containment atmosphere.

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Area monitors provided 1n areas outs1de secondary containment where access may be required post-accident have ranges that envelop the dose rates expected 1n these areas at the time access is required.

ll. Residual heat removal heat exchanger outlet temperature Env1ronmental qualif1cat1on should be addressed in accordance with 10CFR50.49.

~Res onse The instrumentation for this variable is Category I and environmentally qualified to 10CFR50.49. It is in compliance w1th the requirement for Regulatory Guide 1.97, Category 2 instrumentation. Table 421.36-1 of the Final Safety Analysis Report is incorrect and w111 be corrected in a future amendment.

12. Cooling water temperature to engineered safety-feature system components - The applicant should )ustify the deviat1on in range.

~Res onse The temperature range of 2SWP*TE31A and B compl1es with the intent of Nuclear Regulatory Commission Regulatory Guide 1.97, Revision 3. The intent of the regulatory guide is to ensure that instrument ranges are selected so that the instrument will always be on scale. 2SWP*TE3lA and B are located on the service water supply header and are used to mon1tor service water supply temperature. The temperature range of service water instrument is based upon the range for Lake Ontario, which normally varies from 38 F to 77'F and which is well within the instrument range of 32'F to 130'F.

13. Secondary containment area radiation The applicant should supply the information required by Section 6.2 of NUREG-0737, Supplement No. 1, identify any dev1ations, and ]ustify those deviations; environmental qualification should be addressed in accordance with 10CFR50.49.

~Res onse Environmentally qualified area monitors with ranges of 10-l to 104 R/hr are not provided in secondary containment at Nine Mile Point Unit 2.

Justification The use of local area radiation monitors to detect breach or leakage through primary containment penetrat1ons is inappropriate. In general, radiation levels in the secondary containment will be largely a function of radioactivity in primary containment and in the fluids flowing in emergency core cooling system (ECCS) piping. Localized hot spots due to piping> sources and primary containment penetrations and hatches will provide ambiguous indications. Breach of primary containment will be detected by the reactor building exhaust gaseous effluent monitor prior to the reactor building isolation and the noble gas channel of the main stack gaseous effluent monitor following the isolation of the reactor building. Therefore, these area monitors are not necessary and are not 1mplemented at Nine Mile Point Unit 2.

In the long term, accessibility to the secondary containment will be reestablished using a combination of portable radiation survey instruments and post-accident sampling of the secondary containment atmosphere.

14. Noble gas, radwaste vent The applicant should either provide the recommended range or )ustify the use of the lesser range.-

~Res onse See Item 15.

15. Noble gas, common plant vent - The applicant should either provide the recommended range or )ustify the use of the lesser range.

~Res onse (Items 14 and 15>

The noble gas channels of the gaseous effluent monitors for the radwaste/reactor building vent and the main stack release points have design ranges of 10-6 to 104 uCi/cc, which meet Regulatory Guide 1.97 requirements.

16. Plant and environs radiation The applicant should identify the ranges of this instrumentation and show that the ranges are adequate.

~Res onse Two types of portable radiation detection and instrumentation 'are provided to monitor the plant and environs. An ion chamber detector with a range of 10-3 to 50 R/hr is used for low-level gamma and beta radiation monitoring. A Geiger-Muller Teletector type detector with a range of 10-2 to 10> R/hr is used for high level gamma radiation monitoring. With a combined range of 10-3 to 103 R/hr, these instruments have adequate range to envelop the dose rates expected outside the plant buildings after an accident.

17. Accident sampling (primary coolant, containment air and sump) The applicant should provide the information required by Section 6.2 of NUREG-0737, Supplement No. 1, identify any deviations from the regulatory guide, and justify those deviations.

~Res onse

l. Instrument range Analysis range is given in Table II.B.3-2 (See Section 1.10, Table II.B.3-2 of Nine Mile Point Unit 2 Final Safety Analysis Report). The ranges meet or exceed requirements of Regulatory Guide 1.97, within instrument limitations, with the exception of the dissolved gas sample analysis. The ranges given for dissolved gas analysis were approved by the Nuclear Regulatory Commission in a letter to General Electric (letter from W. Johnston

[Nuclear Regulatory Commission3 to G. Sherwood [General Electric]

dated July 17, 1984).

2, 3, 4, 6. Environmental qualification, seismic qualification, quality assurance, and power supply These have been addressed in Table 421.36-1 and meet the requirements of'egulatory Guide 1.97.

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5. Redundancy and sensor locations - This is not applicable, s1nce analys1s 1s done 1n a chem1stry laboratory on grab samples.

Regulatory Guide 1.97 has no specific provis1on.

7. D1splay location - Th1s 1s not applicable, since analysis is done in a chemistry laboratory, the d1splay 1s on each individual 1nstrument. This meets the requirements of Regulatory Guide 1.97.

l8. Primary containment isolation valve position - The applicant should provide Justification for the exemption of instrumentation for the travers1ng incore probe system isolation valves.

~Res onse The traversing incore probe (TIP) system isolation valves consist of ball valves, operated when the probe is out of the guide tube, and shear valves manually operated if the probe is in the gu1de tube.

The TIPs are normally withdrawn and the ball valves are closed. If event occurs while the TIP is 1nserted into the core and the TIP should an

'ail to retract, the shear valve can be operated manually to provide the necessary containment isolation.

These valves are classified nonessential and are provided with non-Class lE automatic isolation signals and power and as such, cannot meet Regulatory Guide 1.97, Category l or 2 requirements. (For further explanation, refer to the Nine Mile Point Unit 2 F1nal Safety Analysis Report Section 1.10, TMI act1on item II.E.4.2 concerning the Containment Isolation Dependability.)

Additionally, any leakage through this line has been incorporated in the radiological LOCA analysis of Chapter 15.6.5 of the Nine Mile Point Unit 2 Final Safety Analysis Report.

19. Please verify that Category I 1nstrumentation 1s or will be provided for neutron flux instrumentat1on (from H. R. Butler letter of October l5, 1985). (Table 421.36-1 of Final Safety Analysis Report commits to environmentally and seismically qual1fied equ1pment.)

~Res onse Neutron flux measurement devices located 1n harsh environment areas are environmentally and mTsmicall; qualified for the antic1pated environments. The env ronmental qualif1cat1on under harsh environment conditions 1s for a limited t1me, but the time 1s sufficient to perform the detection, mitigation, and monitoring functions required of the instrumentation. Instrumentation located in mild environment areas is seismica11y quaiified.

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