ML20117L055

Forwards Response to NRC Request for Addl Info Re Conformance to Reg Guide 1.97 (SER Open Item 6) on post-accident Instrumentation.Info Will Be Included in Amend 11 to FSAR

ML20117L055
Person / Time
Site:	Hope Creek
Issue date:	05/14/1985
From:	Mittl R Public Service Enterprise Group
To:	Butler W Office of Nuclear Reactor Regulation
References
RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 8505160064
Download: ML20117L055 (5)

Text

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O Putsc Servico

)#S E Ektnc arn1 Gas R Company 80 Par k Plata, Newar k, NJ 07101/ 201430 8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation May 14, 1985 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bothosda, MD 20814 Attention: Mr. Walter Butler, Chief Licensing Branch 2 Division of Licensing Gentlement SER OPEN ITEM 6 - ADDITIONAL INFORMATION, POST-ACCIDENT MONITORING INSTRUMENTATION .

l IlOPE CREEK GENERATING STATION DOCKET NO. 50-354 Pursuant to llopo Crook Gonorating Station Safoty Evaluation Report (SER) Outstanding Issuo No. 6 and in rosponso to the NRC request for additional information regarding conformanco to Rogulatory Guido 1.97 (letter f rom A. Schwoncor, NRC, to R. L. Mitti, PSE&G, dated March 26, 1985), Public Sorvico Electric and Gas Company hereby submits the attachod additional information for review.

The enclosed responsos correspond to the four items listed in Section 4; Conclusions, to the EG&G Idaho report, .

"Conformance to Regulatory Guido 1.97 (March 1985)," submit- l tod with the March 26, 1985, lotter. l l

The typographical error correction on FSAR Tablo 7.5-1, discussed in the responso to the coro spray flow variablo '

(1.0. Item 3), and the proposed revision to FSAR Soction 1.8.1.97.4.9, discussed in the responso,to the reactor building or secondary containment area radiation variablo '

(i.e. Item 4), will bo includod in Amondment 11 to the llCGS i FSAR. l 1

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Director of Nuclear Reactor Regulation 2 5/14/85 Should you have any questions in this regard, please contact t us.

Very truly yours, f

Enclosure f

C D. H. Wagner l USNRC Licensing Project Manager A. R. Blough USNRC Senior Resident Inspector s

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l-MP 12 01/02 I

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ENCLOSURE HOPE CREEK GENERATING STATION RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAI) -

CONFORMANCE TO REGULATORY GUIDE 1.97 (SER OPEN ITEM 6)

ITEM 1 - Neutron Flux (Variable B1): Hope Creek Generating Station will follow industry development of this equipment.

ITEM 2 - High Pressure Coolant Injection (HPCI) Flow (Variablo D14): Per GE design specification data sheet 22A6237AB, rated HPCI flow is 5600 GPM, not 6350 GPM. The installed flow monitoring instrumentation provides 0-6000 GPM indication which corresponds to 0-107% of rated flow. The Regulatory Guido 1.97 requirement is 0-110% of rated flow.

ITEM 3 - Core Spray Flow (Variable D15): Coro spray flow monitoring instrumentation provides an indication of 0-10,000 GPM which fully complies with the Regulatory Guido 1.97 requiremont. The 0-1000 GPM referenced by the RAI was a typographical error on FSAR Tablo 7.5-1 and has boon corrected. This correction will be included in Amendmont 11 to the HCGS FSAR.

ITEM 4 - Reactor Building or Secondary Containment Area Radiation (Variablo E2): Further justification for not implomonting Variablo E2 has boon provided by a proposed revision to FSAR Section 1.F 1.97.4.9 (soo Attachment 1). This rovision will be included in Amendment 11 to the HCGS FSAR.

No attempt han boon made to correlate the monitor-ing rangos provided by the noble gas monitors of the south plant vont and filtration, recirculation and vontilation system vont radiation monitoring nystems to the range specified in Regulatory Guido 1.97 for area radiation monitors. A comparison of thin typo would not be foanible. The aron radia-tion monitorn are not nonsitivo to beta radiation becauno of their design and conntruction. Further-more, the gonoric gamma officiency of the fission product noble gason is much lower than the bota officiency and the inotopo mix changos rapidly duo to docay.

The revision to PSAR Section 1.8.1.97.4.9 npocific-ally identifion that the noblo gas monitoring inntrumontation provided (with the rangon linted in P8AR Table 11.5-1) in a much more rollablo moann of obtaining the information required by variablo E2 of Rogulatory Guido 1.97.

t ATT ACH M EN!T 1 HCGS FSAR 8/84

~

Note that HCGS has implemented variable D29 (condenser cooling water flow) by monitoring the circulating water temperature rise across the condenser as a positive AT across the condenser i coupled with no decrease in condenser vacuum is an adequate

indication of condenser cooling water flow.

1.8.1.97.4.9 ISSUE 9 - VARIABLE E2 l Regulatory Guide 1.97 specifies that " Reactor building or secondary containment area radiation" (variable E2) should be

, monitored over the range of 10-8 to 10' R/h for Mark I and II containments, and over the range of 1 to 107 R/hr for Mark III containments. The classification for Hope Creek is Category 2; for Mark III, the classification is Category 1.

As discussed in the variable C14 position statement (Issue 6),

reactor building area radiation is an inappropriate parameter to use to detect or assess primary containment leakage.

F iNSER.T /N 4 Therefore, it is the position of HCGS that the specified reactor building area radiation monitors are not required for HCGS.

i 1.8.1.97.4.10 ISSUE 10 - VARIABLE E3 l Regulatory Guide 1.97 specifies in Table 1, variable E3, that radiation exposure rate (inside buildings or areas where access is required to service equipment important to safety) be i monitored over the range of 10-8 to 10' R/hr for detection of significant releases, for release assessment, and for long-term surveillance.

4 In general, access is not required to any area of the reactor building in order to service safety-related equipment in a post-accident situation. When accessibility is reestablished in the long term, it will be done by a combination of portable radiation survey instruments and post-accident sampling of the reactor building atmosphere. The existing lower-range (typically 3 decades lower than the Regulatory Guide 1.97 range) area radiation monitors would be used only in those instances in which anticipated radiation levels were within measurable instrument ,

ranges.

1.8-89 Amendment 7 ,

e T INSERT A l

The reactor building exhaust and refueling floor area exhaust are continuously monitored by their respective radiation monitoring system as described in Sections 11.5.2.1.3 and 11.5.2.1.2. Any concentration of airborne radioactivity in excess of preset limits as detected by either of these systems (possibly indicating a leak from the primary containment) will initiate the Filtration, Recirculation, and Ventilation System (FRVS) and the FRVS Vent (FRSV) System and will also provide '

signals to the primary containment isolation system to initiate primary containment isolation to the extent described in Section 7.3.1.1.5.

The reactor building exhaust and refueling floor area exhaust are normally routed to the south plant vent where the effluent air flow is monitored by the south plant vent radiation monitoring system as described in Section 11.5.2.2.2. The south plant vent radiation monitoring system instrumentation ranges and sensitivities are listed in Table 11.5-1.

If the FRVS and FRVSV system are initiated (either manually or automatically by the reactor building exhaust or refueling floor area exhaust radiation monitoring systems), the reactor building exhaust and refueling floor area exhausts are automatically shifted to the FRVSV system. The FRVSV effluent air is monitored by the FRVSV radiation monitoring system as described in Section 11.5.2.2.3. The FRVSV radiation monitoring system instrumentation ranges and minimum sensitivities are listed in Table 11.5-1.

It is the Hope Creek position that the monitoring functions performed by the south plant vent radia-tion monitoring system and the FRVSV radiation monitoring system with the ranges and sensitivities listed in Table 11.5-1 provide a much more reliable means of detection of significant releases, release assessment, and long term surveillance than could be provided by reactor building area radiation monitors.

RSCivw 5/14/85 MP 12 3/4

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